B12000, Proposed Tech Spec Revs,Expanding Storage Capacity of Spent Fuel Pool by Storing Consolidated Spent Fuel in Fuel Storage Racks
| ML20195D719 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/31/1986 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20042C972 | List: |
| References | |
| B12000, TAC-61658, TAC-65274, NUDOCS 8606050028 | |
| Download: ML20195D719 (193) | |
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1 Docket No. 50-336 B12000 i
Attachment 1 Millstone Nuclear Power Station, tinit No. 2 Storage of Consolidated Spent Fuel Technical Specification Revisions-a 1
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DEFINITIONS VENTING 1.35 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during venting. Vent, used in system names, does not imply a VENTING process.
MEMBER (S) OF THE PUBLIC 1.36 MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.
The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.
SITE BOUNDARY 1.37 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.
UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industria!, commercial institutional and/or recreational purposes.
STORAGE PATTERN 1.39 The Region II spent fuel racks contain a cell blocking device in every 4th rack location for criticality control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and adjacent and diagonal Region II cell locations surrounding the blocked location.
MILLSTONE - UNIT 2 1-8
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REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 1800 pounds, with the exception of the consolidated fuel storage box, shall be prohibited from travel over irradiated fuel assemblies in the storage pool.
APPLICABILITY: DURING ALL CRANE OPERATION.
ACTION:
With the requirements of the above specification not satisfied, place load in a safe condition.
SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks and/or physical stops which prevent crane travel with loads in excess of 1800 pounds over irradiated fuel assemblies shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to initiation of irradiated fuel handling operations and at least once per 7 days during irradiated fuel handling operations.
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MILLSTONE - UNIT 2 3/49-7 i
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REFUELING OPERATIONS MOVEMENT OF FUEL OVER REGION II RACKS LIMITING CONDITION FOR OPERATION 3.9.17 Prior to movement of a fuel assembly, or a consolidated fuel storage box, over a Region 11 rack in the spent fuel pool, the boron concentration of the pool shall be maintained uniform and sufficient to maintain a boron concentration of greater than or equal to 800 ppm.
APPLICABILITY: Whenever a fuel assembly, or a consolidated fuel storage box, is moved over the Region Il racks in the spent fuel pool.
ACTION:
With the boron concentration less than 800 ppm, suspend the movement of all fuel over Region 11 racks.
SURVEILLANCE REQUIREMENTS 4.9.17 Verify that the boron concentration is greater than or equal to 800 ppm within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to any movement of a fuel assembly, or a consolidated fuel storage box, over a Region II rack in the spent fuel pool and every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter.
e MILLSTONE - UNIT 2 3/4 9-21 m
REFUELING OPERATIONS SPENT FUEL POOL LIMITING CONDITION FOR OPERATION 3.9.18 The Reactivity Condition of the spent fuel pool shall be such that K eff si less-than-or-equal-to 0.95 at all times.
APPLICABILITY: Whenever fuel is in the spent fuel pool.
ACTION:
Borate until a K eff ess-than-or-equal-to l 0.95 is reached.
SURVEILLANCE REQUIREMENTS 4.9.18.1 Ensure that all fuel assemblies to be placed in Region II (as shown in Figure 3.9-2) of the spent fuel pool are within the enrichment and burn-up limits of Figure 3.9.1 by checking the assembly's design and burn-up documentation.
4.9.18.2 Ensure that the contents of each consolidated fuel storage box to be placed in Region II (as shown in Figure 3.9-2) of the spent fuel pool are within the enrichment and burn-up limits of Figure 3.9-3 by checking the design and burn-up documentation for storage box contents.
MILLSTONE - UNIT 2 3/4 9-22
REFUELING OPERATIONS SPENT FUEL POOL LIMITING CONDITION FOR OPERATION 3.9.19 Prior to consolidation of spent fuel assemblies, the - candidate fuel assemblies must have decayed for at least 5 years.
APPLICABILITY: During all consolidation operations.
ACTION:
With the requirements of the above specification not satisfied, replace candidate assembly with an appropriate substitute or suspend all consolidation activities.
SURVEILLANCE REQUIREMENTS
~4.9.19 The decay time of all candidate fuel assemblies for consolidation shall be determined to be greater than or equal to five years within 7 days prior to moving the fuel assembly into the consolidation work station.
MILLSTONE - UNIT 2 3/4 9-23
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1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 FUEL ASSEMBLY INITIAL ENRICHMENT, WT. % U-235 FIGURE 3.9-3 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 AS CONSOLIDATED FUEL
REFUELING OPERATION SPENT FUEL POOL LIMITING CONDITION FOR OPERATION f
3.9.20 Each STORAGE PATTERN of the Region II spent fuel pool racks shall meet either:
(1) ' A cell blocking device is installed in those cell locations shown in Figure 3.9-2, or (2) If a cell blocking device has been removed, all cells of the STORAGE PATTERN must have consolidated fuel in them, including the formerly blocked location, or (3) Meet both (a) and (b):
(a) If a cell blocking device has been removed, all cells of the STORAGE PATTERN must have consolidated fuel in them except the formerly blocked -
location.
(b) The formerly blocked location is vacant and a consolidated fuel box or cell blocking device is immediately being placed into the formerly blocked cell.
APPLICABILITY: Fuel in the Spent Fuel Pool ACTION:
Take immediate action to comply with either 3.9.20(1), (2) or (3).
SURVEILLANCE REQUIREMENTS 4.9.20 Verify that 3.9.20 is satisfied at the following times.
(1) Prior to removing a cell blocking device.
(2) Prior to removing a consolidated fuel storage box from its Region 11 storage location.
MILLSTONE - UNIT 2 3/4 9-26
REFUELING OPERATIONS BASES 3/4.9.6 CRANE OPERABILITY - CONTAINMENT BUILDING The OPERABILITY requirements of the cranes used for movement of fuel assemblies ensures that: 1) each crane has sufficient load capacity to lift a fuel element, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lif ting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly and CEA over irradiated fuel assemblies ensures that no more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident. Specific analysis has been performed for the drop of a consolidated fuel storage box on an intact fuel assembly. This assumption is consistent with the activity release assumed in the accident analyses.
3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 1400F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.
The requirement to have two shutdown cooling loops OPERABLE when the refuel pool is unavailable as a heat sink ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel water level at or above the vessel flange, the reactor vessel pit seal installed, and a combined available volume of water in the refueling pool and refueling water storage tank in excess of 370,000 gallons, a large heat sink is readily available for core cooling. Adequate time is thus available to initiate emergency procedures to provide core cooling in the event of a failure of the operating shutdown cooling loop.
3/4.9.9 and 3/4.9.10 CONTAINMENT RADIATION MONITORING AND CONTAINMENT PURGE VALVLOSOLATION SYSTEM The OPERABILITY of these systems ensures that the containment purge valves will be automatically isolated upon detection of high radiation levels >
, within the containment. The OPERABILITY of these systems is required to restrict the release of radioactive material from the containment atmosphere to the environment.
3/4.9.11 and 3/4.9.12 WATER LEVEL-REACTOR VESSEL AND STORAGE POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
MILLSTONE - UNIT 2 B 3/4 9-2
REFUELING OPERATIONS BASES 3/4.9.13 STORAGE POOL RADIATION MONITORING The OPERABILITY of the storage pool radiation monitors ensures that sufficient radiation monitoring capability is available to detect excessive radiation levels resulting from 1) the inadvertent lowering of the storage pool water level or 2) the release of activity from an irradiated fuel assembly.
3/4.9.14 & 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM The limitations on the storage pool area ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are cons!=ter.: sith the assumptions of the accident analyses.
3/4.9.16 SHIELDED CASK The limitations of this specification ensure that in an event of a cask tilt accident 1) the doses from ruptured fuel assemblies will be within the assumptions of the safety analyses, and 2) Keff will remain less-than-or-equal-to 0.95, 3/4.9.17 MOVEMENT OF FUEL OVER REGION II RACKS The limitations of this specification ensure that in the event of a fuel assembly, or a consolidated fuel storage box, drop accident into a Region 11 rack location completing a 4-out-of-4 fuel assembly geometry, Keff will remain less-than-or-equal-to 0.95.
3/4.9.18 SPENT FUEL POOL The limitations described by Figure 3.9-1 ensure that the reactivity of fuel assemblies, and consolidated fuel storage boxes, introduced into the Region II spent fuel racks are conservatively within the assumptions of the safety analysis.
MILLSTONE - UNIT 2 B 3/4 9-3
REFUELING OPERATIONS BASES i
3/4.9.19 SPENT FUEL POOL The limitations of these specifications ensure that the decay heat rates and radioactive inventory of the candidate fuel assemblies for consolidation are conservatively within the assumptions of the safety analysis.
3/4.9.20 SPENT FUEL POOL The limitations of this specification ensure that the reactivity conditions of the Region 11 storage racks and spent fuel pool Keff will remain less than or equal to 0.95.
The Cell Blocking Devices in the 4th location of the Region II storage racks are designed to prevent inadvertent placement and/or storage of fuel assemblies in the blocked locations. The blocked location remains empty to provide the flux trap to maintain reactivity control for fuel assembly storage in any adjacent locations. Only loaded consolidated fuel storage boxes may be placed and/or stored in the 4th location, completing the STORAGE PATTERN, after a 1 adjacent, and diagonal, locations are occupied by loaded consolidated fuel storage boxes.
MILLSTONE - UNIT 2 B 3/4 9-4
DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,060 + 700/-0 cubic feet.
5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.6 FUEL STORAGE CRITICALITY 5.6.1 a) The new fuel (dry) storage racks are designed and shall be maintained with sufficient center to center distance between assemblies to i l
ensure a K eff ess-than-or-equal-to 0.95. The maximum fuel enrichment to be I stored in these racks is 3.70 weight percent of U-235.
b) Region 1 of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center to center distance between storage locations to ensure a Keff less-than-or-equal-to 0.95 with the storage pool filled with unborated water. Fuel assemblies stored in this region may have a maximum fuel enrichment of 4.5 weight percent of U-235. Consolidated fuel storage boxes may also be stored in this region.
c) Region 11 of the spent fuel storage pool is designed and shall be maintained with a 9.0 inch center to center distance between storage locations to ensure a Keff less-than-or-equal-to 0.95 with the storage pool filled with unborated water. Fuel assemblies stored in this region must comply with Figure 3.9-1 to ensure that at least 85% of the design burn-up has been sustained. The contents of consolidated fuel storage voxes to be stored in this region must comply with Figure 3.9-3.
d) Region 11 of the spent fuel storage pool is designea to permit storage of consolidated fuel in the 4th location of the storage rack and ensure a Keir less-than-ar-equal-to 0.95.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 22'6".
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 384 storage locations in Region I and 962 storage locations in Region II for a total of 1346 storage locations.
MILLSTONE - UNIT 2 5-5
DESIGN FEATURES 5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I Items in Section 5.1.1 of the FSAR shall be designed and maintained to the original design provisions contained in Section 5.8 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.
5.9 SHORELINE PROTECTION 5.9.1 The provisions for shoreline protection described in Amendments 34, 35 and 36 to the FSAR shall be completed by June 15,1976.
MILLSTONE - UNIT 2 5-6 9
Docket No. 50-336 B12000 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Storage of Consolidated Spent Fuel Safety Analysis Report May,1986
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- TABLE OF CONTENTS Page
1.0 INTRODUCTION
1-1 1.1 License Amendment Requested 1-1 1.2 Current Status 1-1 1.3 Summary of Report 1-5 1.4 Conclusions 1-6 2.0
SUMMARY
OF SPENT FUEL RACK DESIGN 2-1 3.0 NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS 3-1 3.1 Neutron Multiplication Factor 3-1 3.1.1 Normal Storage 3-1 3.1.2 Postulated Accidents 3-2 3.1.3 Criticality Methods Development for Consolidated Fuel 3-3 3.1.4 Design Calculation Methods 3-4 3.1.5 Acceptance Criterion for Criticality 3-9 3.2 Cooling Considerations 3-10 3.2.1 Spent Fuel Fool Cooling and Cleanup System 3-10 General Description 3.2.2 Spent Fuel Pool Cooling System Performance 3-10 3.2.3 Thermal Hydraulic Methods Development for Consolidated Fuel 3-14 3.2.4 Fuel Cooling Analyses 3-15 3.2.5 Fuel Element Heat Transfer 3-16 3.2.6 Spent Fuel Pool Chemistry Control 3-17 i
3.3 Potential Fuel and Load Handling Accidents 3-25 3.4 References 3-26 4.0 MECHANICAL, MATERIAL, AND STRUCTURAL 4-1 CONSIDERATIONS 4.1 Description of Structure 4-1 4.1.1 Description of the Auxiliary Building 4-1 4.1.2 Description of Spent Fuel Racks and Consolidated Fuel Storage Boxes 4-1 4.2 Applicable Codes, Standards, and Specifications 4-6 4.3 Seismic and Impact Loads 4-8 4.4 Loads and Load Combinations and Structural Acceptance Criteria 4-3 4.4.1 Pool / Auxiliary Building Analysis 4-11 4.5 Design and Analysis Procedures 4-25 4.5.1 Seismic Methods Development for Consolidated Fuel 4-25 4.5.2 Spent Fuel Storage Rack Design Analyses 4-26 4.5.3 Computer Code Descriptions 4-38 4.5.4 Consolidated Fuel Storage Box Analyses 4-41 4.6 Materials, Quality Control, and Special Construction Techi.iq aes 4-41 4.6.1 Materials 4-41 4.6.2 Quality Control 4-43 4.7 Testing and In-Service Surveillance 4-44 4.3 References 4-55 Section 4 Appendix A - Spent fuel Pool Evaluation Section 4 Appendix B - Reference 4-2 li
5.0 NEED ASSESSMENT AND ENVIRONMENTAL IMPACT 5-1 5.1 Need Assessment 5-1 5.2 Environmental Effects 5-1 5.2.1 Heat Dissipation Effects 5-1 5.2.2 Radiological Considerations 5-1 5.2.3 Chemical Discharges 5-1 6 5.3 References 5-6 lii
f LIST OF TABLES i- Table -Title Page 3.2-1 - Spent Fuel Pool Cooling System Heat Loads and 3-18
- Operating Temperatures 3.2-2 Thermal-Hydraulic Design Parameters 3-19 L
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LIST OF FIGURES Figure Title Page 3-1 Fuel Pool Arrangement: Two Region 3-20 3-2 - Spent Fuel Rack Storage Configuration for 3-21 Region 11 3-3 Effective Multiplication Factor as a function of Burnup 3-22 for Fuel Assemblies in Region 11 3-4 Effective Multiplication Factor as a function of Burnup for Consolidated Fuelin Region 11 3-23 3-5 Initial Enrichment vs. Minimum Allowable Burnup 3-24 for Consolidated Fuel in Region 11 4-1 Typical Spent Fuel Storage Rack Module / Poison Box 4-46 (Region 1) 4-2 Typical Spent Fuel Rack Module 4-47 (Region II) 4-3 Adjustable Foot 4-48 4-4 Poison Box 4-49 4-4a Poison Box Section View 4-50 4-4b Poison Box Section Detail 4-51 4-5 Consolidated Fuel Storage Box 4-52 4-6 Consolidated Box Cover Assembly Installation 4-53 4-7 CESHOCK model of 7 x 9 Spent Fuel Rack Module 4-54 y
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1.0 INTRODUCrf0N 1.1 1.ICENSE AMENDMENT REQUESTED
/
Inf 3anuary, l986 the NRC approved Northeast Nuclear Energy Company's (NNECO) license amendment request for the instaliation cf new' rack; to be i placed in the Millstone 2 spent fuel pool. The new racks increue the 1
amount of spent fuel that can be stored in the existing spent fuel pool. A further increase in capacity may be achieved by storing consolidated fuel, in these. racks. .Under this approach, fuel rods are removed from their spent fuel assemblies and placed in a close-packed triangular array in specially designed storage boxes whic'h can be handled and stored in the new racks like fuel assemblies.
'This Safety Analysis Report supports a request for a license amendment to the Millstone 2 Facility Operating License, and more specifically a request for technical specification revisions required as a result of storing consolidated fuel todether with spent fuel assemblies in the spent fuel pool.
In this report,'censolidated fuel is defined as Millstone Unit 2 spent fuel with a consolidation ratio of 2:1 derived from Combustion Engineering 14x14 fuel assemblies or their equivalent fabricated by Westinghouse.
Consolidated fuel has achieved at least 85% of its designed burnup and has decayed for at least five years since removal from the reactor.
1.2 CURRENT STATUS 1
NNECO's Millstone Unit No. 2 facility received its operating license in August, 1975.- At that time there was capacity to store 301 spent fuel assemblies, or about 1.3 full cores, in the spent fuel storage pool.
in November 1976, NNECO concluded that spent fuel reprocessing facilities would not be available in the near-term. As a result, discharged fuel was filling up the pool. Therefore, a capacity gxpansion of the spent fuel pool was necessary to support the engineering practice and corporate l-1
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policy of reserving storage space in the spent fuel pool to receive an entire discharged reactor core (" full-core-offload") should it become necessary due to operational considerations.
A spent fuel pool rerack license amendment was obtained and the project was completed prior to the first refueling in the fall of 1977. The modified storage pool provided storage locations for 667 fuel assemblies. The increased spent fuel storage capacity provided for the " full-core-off-load" capability as well as the capacity needed for the spent fuel discharges through 198t.f The pool design configuration was composed of nine rack modules, each containing 63 fuel assembly storage locations in a 7 x 9 array and one rack module containing 100 fuel assembly storage locations in a 10 x 10 array. The modules stored the fuel assemblies with a nominal center-to-center spacing of 12.19 inches.
Af ter the Cycle 6 refueling outage in 1985, Millstone 2 lost the " full-core-off-load" capability that NNECO's engineering practices and corporate policy dictate, given the then existing capacity of 667. NNECO, therefore, applied in June, 1985 for a license amendment to support a second reracking of the Millstone 2 spent fuel pool. This license amendment was issued by the NRC on January 15,1986.
The design of the recently approved spent fuel storage racks reflect several recent developments within the nuclear industry. First, current circumstances in the back-end of the nuclear fuel cycle make it necessary that fuel owners establish and implement a plan for " life-of-reactor-storage" of nuclear spent fuel. Consistent with the Nuclear Waste Policy Act (NWPA) of 1932, which requires fuel owners to provide on-site spent fuel storage until a government repository is available, numerous utilities are presently planning capacity expansion projects for which interim storage of spent fuel will have to be provided over the next fif teen to twenty years.
Second, proposed NRC Regulatory Guide 1.13, Revision 2, permits credit to be taken for reactivity depletion in nuclear spent fuel. In particular, this NRC guidance document a new approach to the design of a spent fuel storage rack which will provide a substantially closer center-to-center 1-2
spacing for increased capacity. Additionally, this new approach makes possible the concept of a " region strategy" that can be employed to achieve the maximum utilization of the storage space available in the spent fuel pool.
The " region strategy", utilized by NNECO for the spent fuel pool design configuration approved in January,1986, was a two-region dual-pitch pool of both poisoned and non-poisoned spent fuel racks as shown in Figure 3-1.
Region I will contain the high-enrichment, core off-load assemblies. Fuel assemblies may be stored in every location. The region consists of poisoned spent fuel racks with a nominal center-to-center cell spacing of 9.8 inches. The five modules of Region I totaling 384 storage locations are designed to accommodate 1.7 reactor cores of high enrichment nuclear spent fuel.
The spent fuel rack design for Region I is based upon the commonly accepted physics principle of a " neutron flux trap" with the use of neutron absorber materials. The racks are designed to store Millstone 14 x14 fuel with an initial enrichment of 4.5 w/o U-235. The poison material used is Boroflex.
Region II is reserved for fuel that has sustained at least 85% of its design burn-up. Fuel assemblies are stored in a three-out-of-four logic pattern as shown on Figure 3.2. The spent fuel rack design is based on the criticality acceptance criteria specified in Revision 2 of Regulatory Guide 1.13 which allows credit for reactivity depletion in spent fuel. (Previously, the physics criteria for fuel stored in the spent fuel pool were defined by the maximum unirradiated initial enrichment of the fuel.) The fourth location of the storage configuration remains empty to provide the flux trap for reactivity control. Blocking devices are used to prevent inadvertent placing of a fuel assembly the fourth location.
Region 11 consists of fourteen modules of non-poisoned spent fuel racks whose nominal center-to-center cell spacing is 9.0 inches. The modules consist of 962 cells with storage capacity for up to 728 unconsolidated fuel assemblies that have sustained at least 85% of design burn-up.
1-3
The spent fuel rack modules in both regions of the fuel pool are designed to be free-standing and direct bearing onto the spent fuel pool floor liner.
The rack modules are fabricated from 304 stainless steel and are seismically qualified without mechanical dependence on neighboring modules or the pool walls. They are classified ANS Safety Class !!I.
A third development influencing spent fuel storage options is the consolidation of spent fuel. NNECO, in cooperation with the Electric Power Research Institute (EPRI), Baltimore Gas & Electric Co. and Combustion Engineering, is completing the development of a process which packs the fuel rods from two spent fuel assemblies in a storage box which can be handled like a fuel assembly. The storage box fits into a cell of the new Millstone 2 fuel racks. Storage of consolidated fuel would therefore, effectively double the net capacity of the new fuel racks af ter reserving
. space for the full core off load and for those assemblies undergoing 5 year decay.
3 NNECO is now applying for a license amendment to support the storage of consolidated fuel, together with spent fuel assemblies, in Regions I and II of the Millstone 2 spent fuel pool as reract<ed in 1986.
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, No structural changes to the storage rack modules will be required as they were initially designed to withstand the static and dynamic loads resulting from the storage of both intact fuel assemblies and consolidated fuel in Regions I and II.
Similarly, both regions of the spent fuel pool were initially designed to store both intact fuel assemblies and consolidated fuel in a safe, coolable, l
subcritical configuration with Keff less than 0.95 under the conditions stated in Reference 4-1.
The physical arrangement of the cells within the racks and the total l
number of racks provide for a total of 1346 storage locations. Of this number, 384 locations are contained within Region I and may be used for the storage of new, spent, or consolidated fuel. The remaining 962 locations are contained in Region 11 and 728 of these are available for the storage of spent fuel assemblies which have experienced the required
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burnup, or consolidated fuel storage boxes. The remaining 234 locations 1-4
c are blocked to prevent storage of intact spent fuel assemblies in more than a 3-out-of-4 array.
These 234 locations become available for the storage of consolidated fuel in that a blocking device may be removed once the cell containing it is completely surrounded by consolidated fuel storage boxes. Removal of a cell blocking device is only to be permitted if all cells adjacent to the blocked cell cuitain consolidated fuel.
With these storage restrictions, and the thermal load restrictions imposed by the cooling system, allowable storage capacity of the spent fuel pool becomes:
10 " spare" cells 362 intact fuel assemblies with less than 5 yr decay 217 empty cells in Region I reserved for full core offload 688 cells containing consolidated fuel (at a 2:1 consolidation ratio, this is equivalent to 1376 assemblies)
Anm'al fuel cycles are assumed, with one-third of a core replaced at each refueling.
It should be noted, however, that both the spent fuel racks and the pool / building structure have been analyzed and qualified for the conditions and maximum loadings associated with consolidated fuel stored in each of the 1346 storage locations.
1.3
SUMMARY
OF REPORT This report follows the guidance of the NRC position paper entitled
" Review and Acceptance of Spent Fuel Storage and Handling Applications",
April 1978, as amended by NRC letter dated January 18, 1979. Section 2.0 presents a summary of the spent fuel rack design. Sections 3.0 through 5.0 of this report are consistent with the section/ subsection content of the above NRC position paper, Sections ill through V.
1-5
This report contains the nuclear, thermal-hydraulic, mechanical, material, structural, and radiological design criteria for the fuel racks when loaded with both intact assemblies and consolidated fuel. The nuclear and thermal-hydraulic aspects of this report (Section 3.0) address the neutron multiplication factor, considering normal storage and handling of spent fuel as well as postulated accidents, with respect to criticality and the ability of the spent fuel pool cooling system to maintain sufficient cooling.
Mechanical, material, and structural aspects (Section 4.0) involve the capability of the fuel assemblies, storage racks, loaded consolidated fuel storage boxes and spent fuel pool system to withstand ef fects of natural phenomena and other service loading conditions.
The environmental aspects of the report (Section 5.0) concern the thermal and radiological release from the facility under normal and accident conditions. This section also addresses the occupational radiation exposures and generation of radioactive waste.
1.4 CONCLUSION
S On the basis of the evaluations and information presented in this report NNECO concludes that consolidated spent fuel can be safely stored in the Millstone Unit 2 spent fuel storage facilities as proposed, and that such storage is consistent with the facility design and operating criteria as provided in the FSAR and operating license.
The following sections provide information to assist the NRC in its review and approval of the request for an amendment to the Millstone Unit No. 2 Operating License. The amendment will help NNECO meet the intent of the Nuclear Waste Policy Act that licensees provide for interim onsite storage of spent fuel.
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F-2.0
SUMMARY
OF SPENT FUEL RACK DESIGN The spent fuel racks are designed to store both normal and consolidated fuel from Millstone Unit 2. These racks have been approved to store up to 1112 intact fuel assemblies in two regions of the spent fuel pool.
Alternatively, the racks have the capacity to store up to 688 loaded consolidated fuel storage boxes based upon 5 year decay. The maximum rack capacity is based on structural and criticality considerations; the maximum spent fuel pool storage capacity is given in Section 1.2.
Region I consists of two 8x9 modules and three 8x10 modules, i.e., 5 modules with storage cells for up to 384 intact assemblies or loaded consolidated fuel storage boxes. The rack design is based on the commonly accepted physics principle of a " neutron flux trap" with the use of neutron absorber materials. Each cell contains a poison box insert to control the reactivity of fresh fuel with a maximum enrichment of 4.5 w/o U-235. The poison material used is Boroflex. The normal center-to-center spacing of the cells is 9.8 in.
Region II consists of fourteen (14) modules of non-poisoned spent fuel racks containing 962 total cells of which up to 728 (75%) may be used for storage of intact fuel assemblies or loaded consolidated fuel storage boxes. The unused cells are neutron flux traps (to maintain the required subcriticality for intact fuel assemblies) and are provided with cell blocking devices.
When all the cells adjacent to a blocked cell become filled with consolidated fuel, the blocking device may be removed and the cell used for storage of consolidated fuel. Removal of all the cell blocking devices in this way gives the Region II racks a storage capacity of up to 962 consolidated fuel storage boxes. Region II is used to store fuel which has experienced sufficient burnup such that storage in Region Iis not required.
The nominal center-to-center spacing of the cells is 9.0 inches.
The spent fuel racks are fabricated from 304 stainless steel which is 0.135 inches thick. Each cell is formed by welding along the intersecting seams which enables the assembled cells to become a free-standing module which 2-1
meets the seismic design requirements without depending on neighboring modules or fuel pool walls for support.
s 2-2
i l
l 3.0 NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS 3.1 NEUTRON MULTIPLICATION FACTOR Criticality of both intact and consolidated fuel assemblies in the spent fuel storage rack is prevented by designing the rack to limit fuel interaction.
This is done by fixing the minimum separation between assemblies (or storage boxes) and/or inserting neutron poisons between assemblies.
The design basis for preventing criticality is, including uncertainties, a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (Keff) of the spent fuel array will be less than or equal to 0.95. This criterion is recommended in ANSI N210-1976 and in "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (April 1978), as modified (January 1979).
The following subsections describe: a) the normal and postulated accident conditions in the spent fuel pool which are assumed in calculating the effective neutron multiplication factor (keff), b) the analysis methodology, and c) the analysis results demonstrating that the design meets the acceptance criterion for criticality.
3.1.1 elormal Storage
- a. The analysis considers the most limiting storage condition. In Region I the racks are designed to store fuel assemblies containing 4.5 wt % U-235, or consolidated spent fuel, in every storage location.
In Region 11 intact spent fuel assemblies may be stored in 3 out of 4 locations with cell blocking devices in 1 out of 4 locations (Figure 3-2). Consolidated fuel may be stored in 4 out of 4 locations provided that cell blocking devices are not removed until all adjacent cells contain consolidated fuel. In the criticality analysis for Region 11, credit was taken for reactivity depletion in the spent fuel (consistent with Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis", proposed Revision 2.)
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r-F
- b. The moderator is assumed to be pure water at the normal operating temperature. The reactivity changes for off-normal temperatures are included in the uncertainties.
- c. The Region I and Region 11 arrays were assumed to be infinite in lateral extent and in length,
- d. Mechanical uncertainties (manufacturing tolerances, uncertainty of assembly position in storage racks, material tolerances, etc.) are treated by performing sensitivity studies for the various uncertainties and applying a resultant combined uncertainty to -the - Keff value. For consolidated fuel, a conservatively assumed uncertainty is applied to the Kef t value,
- e. No control element assemblies (CEAs) or contained burnable poisons are assumed to be present.
3.i.2 Postulated Accidents The double contingency principle of ANSI N16.1-1975 states that it is not necessary to assume concurrently two unlikely independent events to ensure protection against a criticality accident. This contingency principle is applied for the following postulated accidents:
- 1) Dropping a fresh fuel assembly into a blocked location of a Region !! rack, completing the 4 out of 4 geometry,
- 2) misplacement of a fresh fuel assembly into a Region 11 rack in one of the 3 open locations,
- 3) dropping a spent fuel assembly or a consolidated fuel storage box into the blocked location of a Region 11 rack, completing the 4 out of 4 geometry, 3-2
- 4) dropping a heavy load (100 ton cask)into Region II,
- 5) dropping a heavy load (2 ton gate) into Region I, and
- 6) dropping a consolidated fuel storage box into Region 1.
The bounding accident was determined to be the dropping of a fresh fuel assembly into a blocked fourth location of a Region 11 rack. To ensure that Keff remains less than or equal to 0.95 for these accidents, the boron concentration in the spent fuel pool will be greater than or equal to 800 ppm during operational conditions under which any of these accidents could occur.
3.1. 3 Criticality Methods Development for Consolidated Fuel 3.1.3.1 Consolidated Fuel Storage Box Loading Density Test The density of compacted fuel is an important parameter in determining the reactivity of the array. A box loading density test was performed to establish an achievable metal / water ratio for use in criticality calculation. Depleted fuel rods and stainless steel rods of the same dimensions were loaded on a triangular pitch into a horizantal test box. It was estabitshed that all 352 rods from two fuel assemblies can be loaded into the 8.3" square cross section of the consolidated fuel storage box.
3.1.3.2 Computer Code Evaluation Criticality experiments (Reference 3-1) were carried out on fuel rods in triangular and square arrays at the Babcock and Wilcox CX-10 f acility in 1981. The rods were clustered in 5x5 modules and both rod pitch and module spacing were varied. Moderator boron concentration was also varied. Analysis of the experiments with the Monte Carlo criticality program KENO IV, employing the 123-group XSDRN cross section set produced multiplication f actors which 3-3
1 showed no discernible dependence on water-to-fuel ratio in the module with the same nominal intermodular spacing, but a clear dependence on the intermodule spacing.
The three dimensional,123 neutron group KENO model is considered impractical for fuel rack design and licensing purposes, in which one and two dimensional transport codes are usually employed. A design model with 16 neutron energy groups and a homogenized fuel module was therefore developed, which should result in a reasonable computer running time for two dimensional calculations of differential reactivity effects with the DOT transport code. The homogenized fuel module representation in the model contained three regions in order to represent adequately the differing water fractions between the interior and the peripheral fuel pins. It was determined by comparative analyses that such a model would not adversely affect the accuracy of the KENO predicted multiplication factors for the critical experiments.
3.1.4 Design Calculation Methods The total uncertainty value to be applied to the value of Keff for intact fuel assemblies in the storage racks is obtained from the results of the following calculations.
delta Keff Change Region I Region Il Minimum center-to-center +0.0130 +0.0038 distance of storage cells Maximum Poison Box I.D. +0.0051 --
Minimum steel thickness --
+0.0018 Maximum steel thickness +0.0034 --
Off center placement of fuel +0.0001 -0.0074 assemblies in adjacent cells Temperature Change +0.0031 +0.0021 Poison Box not centered +0.0014 --
in monolith R.M.S. Value 0.0148 0.0088 3-4
r Experiments were used to determine the bias and uncertainties. The critical experiments covered a wide range of parameters, including experiments with boron between fuel clusters as in Region I of the Millstone Unit No. 2 racks. The critical experiments also included steel between fuel clusters as in Region II.
T he CEPAK lattice program is used to calculate the fission product and transuranium isotopic content in burned fuel. Fission products are calculated by the CINDER module in CEPAK. Fourteen (14) fission product chains representing the more important fission products are used instead of the sixty-nine (69) chains employed in Reference 3-4. The difference in fission product poisoning between the condensed and detailed fission product chain is treated as a lumped fission product for each fissioning nuclide.
The possible reactivity changes in the fuel subsequent to removal of the fuel from the reactor for storage in the spent fuel racks are also calculated. Fuel nuclide concentrations appropriate to the end of the 3000 hour0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> cooling period are used to calculate the microscopic cross sections for use in the DOT analyses of the effective multiplication of Region II.
The maximum reactivity change due to the axial burnup shape was determined by analyzing the two extreme axial shapes that have been observed in reactor core analysis (cosine axial shape, and a flat distribution). The two extreme distributions were picked from axial shapes taken from reactor core analysis; as expected, the cosine-like shape occurred at the end of Cycle 1. The flattest distribution occurred at the end of what is considered an equilibrium cycle. The axial average for both of these burnup shapes was calculated. The average burnup became the uniform burnup case, which did not take into consideration the axial shape. The reactivity was calculated for both the uniform burnup and the actual burnup profiles. The elfective multiplication factors w ere calculated using KENO assuming the entire Region 11 rack was full of fuel with the same 3-5
burnup shape. The effective multiplication factors obtained using the KENO code are as follows:
a) Assembly discharged at EOCl Uniform Burnup Distribution Non-Uniform Burnup Distribution 0.97575 1 000306 0.96584 1 000353 b) Assembly discharged at EOC6 (Equilibrium)
Uniform Burnup Distribution Non-Uniform Burnup Distribution 0.99720 + 0.00145 1.00834 + 0.00164 The analysis shows that the assumption of uniform burnup axially may result in underestimating the multiplication factor in some cases, and overestimating it is other cases. For the Millstone Unit No. 2 Region 11 spent fuel racks, the resulting reactivity difference for the EOC6 was 0.0114 delta K ef fective, which was conservatively determined to be equivalent to 1800 mwd /t. The Region 11 allowable (average) burnup for each initial enrichment reflects this reactivity penalty.
The consideration for axial burnup penalty associated with the fuel being consolidated was considered first with the candidate fuel assemblies as described above prior to introduction into Region 11 of the spent fuel pool and second with the consolidation analyses that utilized an allowable Keff of less than 0.90.
The B-10 areal density is a minimum, no credit was taken for the poison effect of non B-10 material, and nominal dimensions were used.
The fuel enrichment is the maximum allowed in storage. The fuel density used in the analysis is a nominal fresh stacking density - this is conservative relative to exposed fuel densities. Nominal values of other fuel assembly paraineters were used since their variation was considered statistically non-significant for this analysis.
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Some fuel assemblies, from the original core, now stored in the Millstone Unit No. 2 spent fuel pool contain burnable poisons. These burnable poison pins have essentially no B-10 since these poison pins were depleted during the initial fuel cycle. Analyses supporting the criticality calculations for the spent fuel racks treat these fuel assemblies in a highly conservative manner by assuming that fuel pins are present rather than poison pins. Thus, no credit has been assumed in the design and analysis of the spent fuel racks for any burnable poisons that might be present in the fuel inventory.
For Region I, lower pool temperatures constitute a maximum reactivity condition. For Region II, higher pool temperatures constitute a maximum reactivity condition. The design temperature used for both Regions was 900F. The difference between 900F and 680F (For Region 1), and the dif ference between 900F and 2360F (for Region 11) was assigned in the uncertainty analysis. Thus, for Region I, the resulting neutron multiplication factor (Keff) is 0.943 including all uncertainties and calculated biases.
The individual contributions were combined in a root mean squared manner to yield the values shown as the last entry in the table. When the latter values were then combined in a direct additive manner with the bias and calculational uncertainty, noted above, overa:1 uncer' utics of 0,021 and 0,015 were obtainad for Regions I and II, respectively. All components in these overall uncertainties are at least at the 95/95 confidence level. Since consolidated fuel is considerably less reactive, an assumed uncertainty of 0.05 was applied to the value of K egg for consolidated fuel in the storage racks.
NITAWL The NITAWL code reads in the cross sections in AMPX library formats then performs resonance self-shielding calculations, and 3-7
collects the data into " workable" arrangements such as the format required by the ANISN, DOT, or MORSE Codes.
XSDRNPM The XSDRNPM program performs the following functions:
- 1. Provides a one-dimensional transport calculation capability for calculating reaction rates, eigenvalues, and critical dimensions.
- 2. Allows spatial cross section weighting to be performed.
XSDRNPM is being used to collapse the original fine group library obtained from NITAWL to a broader group library for use with DOT-2W.
CEPAK The CEPAK lattice program is employed to calculate the basic broad group cross section data for the fuel assembly, spent fuel rack structure, and water. This program is a synthesis of a number of l computer codes, e.g., FORM, THERMOS, and CINDER (References 3-2, 3-3, and 3-4). These codes are interlinked in a consistent way with inputs from an extensive library of differential cross section data.
NUTEST 1
NUTEST is a two-dimensional integral transport code which employs the collision pro %bility technique to compute sub-region dependent reaction rates in an explicit geometric representation of the fuel rods and associated structure of a fuel assembly. This code is used to calculate the flux advantage factors which are applied as correction factors to the basic broad group cross sections computed by the CEPAK lattice program to account for heterogeneous lattice ef fects not represented in either the multigroup spectrum or homogenized cell spatial calculation, e.g., heterogeneous fast fission effect in fuel pellets.
3-8
DOT-2W The spatial flux solution and multiplication factor for an infinite array of individual or clusters of fuel storage cells are computed with the _ two dimensional, discrete ordinates transport code, DOT-2W (Reference 3-5). The major features of the method used in this code are:
a) Energy dependence is considered using the multigroup treatment.
b) The derivative terms and spatial dependence are approximated using a finite difference technique.
c) Dependence upon the direction variables is treated using the discrete ordinates method.
d) The scattering integral is evaluated using a discrete ordinates quadrature in . combination with a Legendre expansion of the scattering kernel to approximate anisotropic scattering.
3.1.5 Acceptance Criterion for Criticality The acceptance criterion for the neutron multiplication factor (Keff) is that it be less than or equal to 0.95, including uncertainties, under all postulated conditions. For Region I the resulting neutron multiplication factor for intact fuel assemblies is 0.943 including all uncertainties and calculational biases. For Region 11 reactivity depletion is a function of the percentage of burnup achieved, not of the initial enrichment. The resulting reactivity with intact fuel assemblies, including uncertainties and calculational biases, as a function of the average burnup for several initial enrichments is shown in Figure 3-3. The minimum allowable average burnup for a given initial enrichment is that corresponding to Keff = 0.95. When these minimum burnup values are plotted as a function of initial enrichment, regions of acceptable and unacceptable burnup are identified (see Figure 3-5).
3-9
Nominal reactivity with consolidated fuel as a function of burnup for several initial enrichments is shown in Figure 3-4. Minimum burnup values for consolidated fuel for storage in Region 11 as a function of initial enrichment are also shown in Figure 3-5. The consolidated fuel curve however is based on a nominal Keff = 0.90, conservatively allowing an uncertainty value of 0.05. It can be seen from Figure 3-5 that any intact spent fuel assembly which has sufficient burnup to be acceptable for storage in Region 11 can also be stored there af ter consolidation. For any given initial enrichment there is a range of burnup within which fuel in the intact assembly form may only be stored in Region I but in consolidated form may be stored in Region II.
3.2 COOLING CONSIDERATIONS 3.2.1 Spent Fuel Pool Cooling and Cleanup System General Description The fuel pool cooling and cleanup system consists of two circulating pumps with a capacity of 850 gpm each, and two heat exchangers with a total heat removal capacity of 11.3 x 106 BTU /hr. The system was designed to maintain the fuel pool bulk water temperature below 1200F during the normal heat load condition. With the aid of the shutdown cooling system, a temperature of 1500F can be maintained under the maximum heat load conditions. The maximum heat load was defined in the FSAR as 28.0 x 106 BTU /hr. The maximum heat load conditions for the enlarged fuel pool capacity are defined in Section 3.2.2. The disposal cartridge type fuel pool filters and a demineralizer maintain water quality. This system is described in Section 9.5 of the FSAR.
3.2.2 Spent Fuel Pool Cooling System Performance The adequacy of the cooling system has been analyzed for the storage of consolidated fuel together with intact assemblies in the spent fuel racks. Because of the substantial increase in spent fuel inventory, a comprehensive analysis of the performance of the spent fuel pool 3-10
cooling system heat exchangers was made in order to obtain a best estimate of the bulk pool water temperature at end of life. The ST-4 shell-and-tube heat exchanger computer code of Heat Transfer Research, Inc. was used (Reference 3-6). Table 3.2-1 summarizes the cooling system perfornance for the normal refueling and full core offload conditions.
The decay heat loads were calculated using the computer code ORIGEN, developed at Oak Ridge National Laboratory. ORIGEN is a point depletion code which solves the equations of radioactive buildup and decay for large numbers of isotopes.
The design basis heat loads were determined from the following refueling conditions. Annual refueling was assumed.
- 1. Normal Refueling: The maximum spent fuel capacity of 362 intact fuel assemblies, with less than 5 years' decay time, and consolidated fuel from 1376 assemblies present in the pool.
- 2. Emergency Full Core Offload: 217 fuel assemblies (full core) offloaded in the cycle following the refueling shutdown in which the normal capacity in (1) above is reached.
The projected maximum capacity of the fuel pool is 389 intact assemblies plus consolidated fuel from 1376 assemblies, however, the ten " spare" assembly spaces were assumed also to contain defective assemblies from the last cycle.
ORIGEN predicts 37.8 x 106 BTU /hr for the design basis heat load.
Since the combined cooling capability of the Spent Fuel Pool Heat Exchangers and one train of the Shutdown Cooling Heat Exchangers is 38.5 x 106 BTU /hr, the pool temperature will not increase. The size of the original fuel pool cooling system as described in the FSAR was based on infinite irradiation time for the fuel assemblies, whereas the ORIGEN heat load was based on finite irradiation time and the actual refueling schedule.
3-11
A single failure analysis of the spent fuel pool cooling system is presented in Table 3.2-2.
As indicated in Table 3.2-2, the heat load in the spent fuel pool immediately following a one-third core offload would result in spent fuel pool water temperature in excess of 1400F assuming a failure of one Spent Fuel Pool cooling train. The calculated maximum temperature is 1780F following the failure of a single train of the Spent Fuel Pool cooling system. Northeast Utilities Service Company has performed independent calculations which predict a maximum temperature of 1770F. These results assume a one-third core offload which has decayed 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> and that no corrective action has been taken to provide additional cooling.
If the most recent offload is assumed to have decayed for 13 days, the Spent Fuel Pool cooling system in its normal alignment is sufficient to maintain the pool at 1220F. In addition, alternate means of cooling could be provided. A LPSI pump could be aligned to take suction from the Spent Fuel Pool and discharge through the shutdown cooling heat exchangers into the reactor vessel. In this arrangement, coolant would flow from the vessel, through the fuel transfer tube and back into the spent fuel pool.
To satisfy a condition of the license amendment allowing a storage capacity increase to 1112 storage locations, an additional amendment request is being prepared to restrict plant operation if both spent fuel pool cooling trains are not operable during the time when a single train could not maintain the bulk pool water temperature below 1400F. Submittal of this amendment request is targeted for July, 1986.
The discussion for using the Shutdown Cooling System for cooling the spent fuel pool may be found in Section 9.5.3.2 of the Millstone Unit No. 2 FSAR titled Abnormal Operation (Spent Fuel Pool Cooling System). The section reads as follows:
"In the event that an emergency full core offload is placed in the spent fuel pool, a combination of spent 3-12
fuel pool cooling and shutdown cooling system components will be used to remove decay heat. One low pressure safety injection pump will take suction from the spent fuel pool and pump * ' cooling water through one of the shutdown heat uchangers. The shutdown cooling system is placed into service by manual initiation."
The Shutdown Cooling System has been used for cooling fuel during the last two refuel outages.
The connection between the spent fuel pool cooling system and the shutdown cooling system inlet is a spool piece located between valves 2-S1-442 and 2-RW-il. The connection between the shutdown cooling system outlet and the spent fuel pool inlet is a spool piece between valves 2-51-458 and 2-RW-15.
Valves 2-51-442 and 2-51-458 are manually operated. Valves 2-RW-il and 2-RW-15 are manually operated by two operators which are in verbal communication with an operator at Control Room Panel CO-1.
The procedure for performing the spent fuel cooling function is contained in Operating Procedure OP2310, Shutdown Cooling System.
Additional information concerning the spent fuel pool and reactor building component cooling water (RBCCW) systems heat exchangers is contained in the following table:
Spent Fuel Pool Cooling System Heat Exchangers a) Heat rate transfer coefficient 413 BTU /Hr-f t20p b) Design flow rate of RBCCWS water through exchanger 1100 GPM 3-13
RBCCW Heat Exchangers a) Heat rate transfer coefficient 347 BTU /Hr-f t2oF b) Design flow rate of service water through exchanger 11,700 GPM c) Total normal heat load per exchanger (excluding spent fuel decay heat load) 26.33(l) MBTU/HR.
d) Maximum RBCCW temperature 850F(l)
Maximum service water temperature 750F(I)
(1) Millstone Unit No. 2 FSAR Section 9.4.3.1.
3.2.3 Thermal Hydraulics Methods Development for Consolidated Fuel 3.2.3.1 Modifications to CEPOOL The CEPOOL computer code uses a flow network method to predict coolant temperature, velocity and quality in fuel storage celb. In the flow network, a row of cells is represented such that each cell has an internal resistance and is connected to its neighbors through cross flow resistance at the bottom of the cells; a constant pressure drop is maintained across the flow network by the entire pool which is much larger than the fuel region. When a box of consolidated fuelis stored in a fuel cell of a rack, a water gap exists between the cell wall and the consolidation box as shown in Figure 3-6. Thus, coolant can bypass the fuel rods within the box and flow through the gap.
The CEPOOL code was modified to account for bypass flow through the box / cell wall gap. Conservative assumptions were made in developing these modifications to avoid the increased complexity 3-14
associated with the calculation of lateral heat transfer from fuel within the consolidation box to the gap coolant. All heat generated by the fuel within the box is assumed to be removed by coolant flowing through the box. However, the coolant density change associated with this heat removal is also assumed to exist as driving pressure for bypass flow through the box / cell wall gap. This introduces conservatism into the analysis since no credit is taken for heat removal from the fuel by lateral heat transfer to the coolant in the box / cell wall gap.
Further modifications were made to CEPOOL to include the capability for modeling a mixed row of fuel cells, i.e., a row of cells containing different types of fuel. This capability was used in the Thermal Hydraulic design analyses for the Millstone Unit 2 fuel pool which can accommodate both intact fuel assemblies and consolidated fuelin the same row of storage cells.
3.2.3.2 Consolidated Fuel Heat Transfer Test An experiment, using a heated test section, was performed to confirm the conservatism of the design analyses for consolidated storage cells. The test section contained nineteen full length heated rods which were tightly packed to form triangular flow channels. The test section was placed in a water filled vessel to simulate the spent fuel rack conditions. The set-up was instrumented to measure coolant and cladding temperature at various design conditions. The experiment is described in Reference 3-7, which also gives additional information on the thermal-hydraulic methods development.
3.2.4 Fuel Cooling Analyses 3.2.4.1 Cooling of Spent Fuel in the Racks The thermal-hydraulic analyses are based on the design parameters given in Table 3.2-3. It is assumed in the analyses that the fuel pool is loaded to full capacity with intact fuel assemblies and consolidated spent fuel. The spent fuel racks are designed to adequately cool the 3-15
spent fuel during normal and accident conditions subject to the following criteria.
- 1. Bulk boiling of the entire pool must not exist during normal operation.
- 2. Maximum fuel clad temperature will not exceed 6500F during both normal operation and accident conditions.
Normal operation includes any arrangement of intact fuel assemblies and consolidated spent fuel with a bulk pool temperature of 1500F and minimum pool depth of 23 feet of water above the fuel.
Accident conditions assume that, as a result of loss of external cooling, coolant is evaporated to a minimum pool depth of 10 feet of water above the racks and that the racks are blocked by a dropped fuel consolidation cannister. The results of the thermal-hydraulic analysis confirm that the above two design criteria are met for both accident assumptions. Furthermore, analyses performed for normal operation indicate that the local coolant temperature in the racks will not reach saturation temperature; thus, localized boiling will not occur.
3.2.5 Fuel Element Heat Transfer An analysis of the maximum fuel cladding temperature was performed for the postulated case of complete loss of coolant circulation to the pool. The analysis assumed maximum anticipated heat load in the pool, with the hottest assembly located in the least cooled storage area. The maximum cladding temperature will occur at the location of maximum heat flux. Natural circulation flow rates within storage tubes will result in a heat transfer coefficient in excess of 50 BTU /hr. f 2t oF. Because the heat flux is small, very large uncertainties in the film coefficient are acceptable without causing prohibitively high clad temperatures. The design upper limit temperature for the clad in the spent fuel pool is 6500F. The fuel cladding temperature analysis verifles that this limit is maintained for the Millstone Unit No. 2 racks.
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3.2.5 _
Spent Fuel Pool Chemistry Control Water chemistry and optical clarity will be maintained by the existing spent fuel pool cleanup system. The cleanup system is a non-safety related system and has been designed to non-Seismic Category I requirements. Isolation capabilities from the Category I portion of the fuel pool cooling system have been provided by Seismic Category I isolation valves. The cleanup system consists of two refueling water purification pumps, two filters, and a demineralizer and associated valves and piping. In addition, the spent fuel pool has been provided with a skimmer pump and two filter assemblies to facilitate the removal of accumulated surface debris. The cleanup system has been designed to process water through the purification loop from the refueling pool and refueling water stnrage tank.
The radioactive contaminant levels in the pool are primarily a function of failed fuel fraction and reactor operating level and are highest during and shortly following refuelings. Since the concentration of impurities is controlled by continuous removal in the demineralizer and by natural radioactive decay, the increase in contaminant levels due to the long-term stored fuel in the form of both intact assemblies and consolidated rods is negligible when compared to the removal mechanisms. Therefore, the installed systems and equipment are deemed adequate for maintaining water chemistry and optical clarity with the expanded pool storage capacity.
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TABLE 3.2-1 SPENT FUEL POOL COOLING SYSTEM HEAT LOADS AND OPERATING TEMPERATURES Normal Maximum Heat Load: 15.2 x 106 BTU /hr(I)
Abnormal Maximum Heat Load: 37.8 x 106 BTU /hr(2)
OPERATING CONDITION POOL TEMPERATURES Design Basis Design Calculated Normal Maximum 1200F 1310F(3)
Abnormal Maximum 1500F 1200F(4)
Single Active Failure of a SFP Cooling Train Normal Maximum 2120F 1780F Abnormal Maximum 2120F 1850F Total Loss of Forced Pool Cooling Normal Maximum 2120F 9 3/4 hrs, to boiling Abnormal Maximum 2120F 4 hrs. to boiling (1) This heat load is predicted for normal refueling with the most recently unloaded one-third core having decayed for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. After 12 more days, the decay heat load will be less than 11.3 x 106 BTU /hr.
(2) This heat load is predicted for spent fuel in the pool with the entire core of floaded.
(3) This temperature is a function of using only the Spent Fuel Pool cooling heat exchangers.
(4) This temperature value is a function of using one train of the Shutdown Cooling Heat Exchangers in addition to Spent Fuel Pool Heat Exchangers.
l
Table 3.2-2 THERMAL HYDRAULIC DESIGN PARAMETERS
/
Number Assemblies Normal Reload 1/3 core Minimum Decay Time One Assembly 3 days Minimum Decay Time Full Core . 6 days Minimum Cooling Time of a Fuel Assembly 5 years before the Rods are Consolidated Water Height Above Fuel 23 feet Minimum Water Height Above Rack (Accident)* 10 feet /1.4 feet *
- Maximum Bulk Water Temperature (Normal) 1500F Maximum Water Temperature in the Rack Region (Accident)* 2350F/230oF * *
- First value is for loss of exte nal cooling accident. *
- Although it is not 4V considered a design basis acci. fat, the review of this submittal has considered the consequences of a reactor cavity seal failure accident. These values represent the water height and maximum temperature for such an accident.
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(@ 24-22- ACCEPTABLE FOR 3 20- STORAGE IN REGION 2 g ta-g . i._ g 14-ga 12-O UNACCEPTABLE FOR g 10-8- STORAGE IN REGION 2 6-4- 3 2- 0 0-I I I I I i 1 I I 2.5 3.0 3.5 4.0 4.5 5.0 1.0 1.5 2.0 FUEL ASSEMBLY INITIAL ENRICHMENT, WT. % U-235 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 AS CONSOLIDATED FUEL FIGURE 3-5 1
3.3 POTENTIAL FUEL AND LOAD HANDLING ACCIDENTS i i j 3.3.1 Fuel Handling Accident in the Spent Fuel Pool This accident has been addressed in the Millstone Unit No. 2 FSAR (Section 14.19). A complete list of assumptions is provided in FSAR l Table 14.19-1. Results of the analysis, which are well below the l limits of 10 CFR Part 100, are presented in Section 14.19.3. , I i 3.3.2 Load Handling ) Refueling operations at Millstone Unit No. 2 are conducted in $ accordance with strict procedures to prevent inadvertent dropping of heavy objects into the pool. Strict procedural controls as well as Technical Specifications prohibit the movement of heavy objects over spent fuel stored in the pool. i f 6 + 3-25 c - -
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4
3.4 REFERENCES
3-1 G. S. Hoovler et al., " Critical Experiments Supporting Underwater Storage of Tightly Packed Configuration of Spent Fuel Pins," BAW-1645-4, November 1981. j 3-2 FORM - A Fourier Transform Fast Spectrum Code for the IBM-7090, McGoff, D. J., NAA-SR-Memor 5766, September 1960. 3-3 THERMOS - A thermalization Transport Theory Code for Reactor Lattice Calculations, Honeck, H., BNL-5316, July 1961. 3-4 CINDER - A One Point Depletion and Fission Product Program, England, T. R., WAPD-TM-334, Revised June 1964. i 3-5 R. G. Soltesz, et. al., " Users Manual for DOT-2W Discrete Ordinates Transport Computer Code," WANL-TME-1982, December 1969. 3-6 Computer Manual 1, ST- Users Manual for Program ST-4, Design or Rating of Shell and Tube Heat Exchangers, Heat Transfer Research, Inc. 3-7 Fuel Consolidation Program, EPRI Research Project 2240-2, Interim Report-Thermal Hydraulics Design, December 1984. f 3-26
4.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS
4.1 DESCRIPTION
OF STRUCTURE 4.1.1 Description of the Auxiliary Building The auxiliary building is a multi-story, reinforced concrete structure with flat slabs and shear walls. Some open areas of the building are supported by structural steel columns. The portion of the building west of column line M.7 is founded on bedrock approximately 60 feet below the ground surface, while the eastern end of the building is supported by compacted structural backfill. These two portions of the building are separated from each other by an expansion joint at line M.7, to allow for differential movements. Spent fuel storage is provided between column lines 17.2 and 18.9 and column lines H.4 and L.5 at Elevation (-)2'-0". The storage area consists of a reinforced concrete pool lined with one-fourth inch thick stainless steel plate to Elevation 33'-6". Normal water level is to Elevation 36'-6".
. 4.1.2 Description of Spent Fuel Racks and Consolidated Fuel Storage Boxes 4.1.2.1 Design and Fabrication of Spent Fuel Racks The spent fuel storage racks are fabricated from 304 stainless steel having a maximum carbon content of 0.065% The racks are monolithic honeycomb structures with square fuel stora,,e cells as shown in Figure 4-1. Each storage cell is formed by welding stainless steel sections along the intersecting seams, permitting the assembly to be a load bearing structure, as well as framing the storage cell enclosures. Each module _is free standing, and seismically qualified without mechanical dependence on neighboring modules or pool walls.
This feature allows remote installation (or removal if required for pool maintenance). Reinforcing plates at the upper corners provide the required strength for handling. 4-1
Stainless steel bars, which are inserted horizontally through the rectangular slots in the lower region of the module, support the intact fuel assemblies and consolidated fuel storage boxes. The support bars are welded in place and support an entire row of fuel assemblies and boxes. The module is supported by adjustable pads to f acilitate leveling at installation. Loading of the fuel racks is facilitated via movable lead-in funnels. The openings of the funnels are symmetrical. The funnels are placed on top of the rack module. The 304 stainless steel module wall thickness is 0.135 inch. Region I is located within 5 modules and comprises a total of 384 cells. Region I is the high-enrichment, core off-load region. The fuel assemblies can be stored in every location. Region I is designed for a maximum of 334 usable cells (i.e.100% storage) for intact assemblies of fuel enriched up to and including 4.5 w/o U-235, and for consolidated fuel. Each cavity in Region I contains a poison insert for neutron absorption. These inserts are made up of Boroflex sheets enclosed, but not sealed, between stainless steel sheets and are configured such that a minimum water gap is maintained between the insert and the cell wall. The inserts lock into the storage cavity using a spring locking mechanism on tne upper end which snaps into a hole in the surrounding cell wall. A typical Region I fuel rack module and poison insert are shown in Figures 4-1 and 4-4. The neutron absorption material, Boraflex, used in the Region I spent fuel racks is manufactured by Brand Industrial Services, Inc., and fabricated in accordance with the quality assurance criteria of 10CFR50, Appendix B. Boraflex is a silicone based polymer containing fine particles of boron carbide in a homogeneous, stable matrix. Boraflex contains a minimum 10B areal density of 0.030 gm/cm2 for Region I racks. 4-2
Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments and to verify its structural integrity and suitability as a neutron absorbing material. Tests were performed at the University of Michigan exposing Boraflex to 1.03 x 101I rads gamma radiation with a substantial concurrent neutron flux in borated water. These tests indicate that Boraflex maintains its neturon attenuation capabilities before and af ter being subjected to an environment of borated water and 1.03 x 1011 rads gamma radiation. Long term borated water soak tests at high temperatures were also conducted. It was shown that Boraflex withstands a borated water immersion of 2400F for 260 days without visible distortion or softening. Boraflex maintains its functional performance characteristics and shows no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide. The actual tests verify that Boraflex maintains long-term material stability and mechanical integrity and can be safely utilized as a poison material for neutron absorption in spent fuel storage racks. The inservice poison material surveillance program is addressed in Section 4.7 of this submittal. Region 11 consists of a total of 962 cells and is reserved for fuel which has achieved at least 85% of its designed burnup. Within Region II, intact fuel assemblies may be stored in 75% of the total cells (see Figure 4-1) for an available intact assembly storage capacity of up to 728 assemblies. Cell blocking devices are used to preclude placement of intact fuel assemblies into every fourth cell, which remains empty and provides a flux trap for reactivity control. Alternatively, up to 728 loaded consolidated fuel storage boxes can be stcred in the unblocked cells, and up to an additional 234 can be stored in the normally blocked cells after removing the blocking devices. A blocking device may only be removed from its cell when all the adjacent cells contain consolidated fuel. 4-3
The foregoing capacities are subject to the overall maximum capacity of the spent fuel storage pool given in Section 1.2. Figure 3-1 shows the arrangement of Region I and Region 11 modules in the Millstone 2 pool. 4.1.2.2_ Support of Spent Fuel Racks The spent fuel racks have been designed for direct bearing onto the spent fuel pool floor. An adjustable pad (Figure 4-3) is provided under each corner of the fuel rack. Fuel rack module leveling is accomplished by adjusting each pad height to conform to the pool floor. Each foot can be raised or lowered 1/2" and can rotate 2 to 3 degrees to accommodate pool floor variations during installation. 4.1.2. 3 Design of Consolidated Fuel Storage Boxes The consolidated fuel storage boxes are open ended square tubes, each providing sufficient space to store the fuel rods from two disassembled C-E 14x14 fuel assemblies. The boxes are designed to fit the spent fuel storage rack cells. An insert floor plate in each box contains perforations to provide adequate cooling of the fuel rods by natural circulation of the spent fuel pool water. The consolidated fuel storage boxes are fabricated from type 304 stainless steel and are classified Seismic Category I. The cover assembly (see Figure 4-5) for a consolidated fuel storage box is a spring loaded self locking device that is installed on the consolidation box af ter the fuel rods have been loaded into the box. The cover has a screen across the top which allows adequate cooling flow while retaining fuel rods inside the box and preventing entry of foreign material. The cover actuator, including the ears, is dimensionally similar to the upper end fitting of the fuel assembly. This permits the consolidated fuel storage box to be lifted by the fuel handling machine. The cover assembly also contains a fuel accountability feature which provides a visual indication of whether the cover assembly has been 4-4
unlatched. The accountability feature works in the following manner. During installation of the cover assembly on the consolidation box (see Figure 4-6), the accountability indicator rests on the guide plate attached to the cover latch. The indicator is located in a slot in the cover to keep the indicator positioned during cover installation. As the latch is rotated by the cover assembly latching tool into the slots in the consolidation box, the guide plate slides under the indicator until the indicator engages the slot in the latch. This is the installed position of the cover assembly and indicator. In the latched position, the indicator is engaged with the slots in the cover and the latch. The horizontal tabs on the indicator are visible in the cutout in the cover. If the cover assembly is unlatched for removal, the rotation of the latch will deform the indicator and cause it to drop out of the cover. Evidence of cover removal is displayed by the fact that the tabs on the indicator are no longer visible in the cutout in the cover. A place is provided on the cover for an i identification number. 4.1.2.4 Fuel Handling i Storage of consolidated spent fuel will not affect the conclusions of the fuel handling accidents presented in the FSAR. The radiological dose for the postulated loaded consolidated fuel storage box drop accident is well within the 10 CFR 100 criteria. Keff remains less than or equal to 0.95. The fuel handling accidents are presented in Section 3.1.2 of this report. 1 E 4-5
4.2 APPLICABLE CODES, STANDARDS, AND SPECIFICATIONS The spent fuel racks are designed in accordance with the applicable sections of the following:
- 1. Code of Federal Regulations 10CFR Part 50:
a) Appendix A, " General Design Criteria for Nuclear Plants," Criteria 2,3,4,5,61,62,63. b) Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."
- 2. ASME Boiler and Pressure Vessel Code Section 111, Subsection NF, " Nuclear Power Plant Components."
l
- 3. ASME Boiler and Pressure Vessel Code Section IX.
4.. American Society for Testing Materials Documents: i a) ASTM - A240 - Specification for Corrosion Resisting Chromium Nickel Steel Plate, Sheet & Strip for Fusion-Welded Unfired Pressure Vessels. b) ASTM - A276 - Specification for Stainless and Heat Resist-ing Bars and Shapes
- 5. American National Standards Institute:
j a) ' ANSI - N210, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations, I' 1976. b) ANSI - N16.1, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,1975.
- j. >
l l 4-6 l
1
- 6. United States Nuclear Regulatory Commission:
i
- 1. a) Standard Review Plan, Section 9.1.2, Rev. 2 " Spent Fuel S torage."
b) Regulatory Guide 1.13, Rev. 2 Draft," Spent Fuel Storage Facility Design Basis." , c) Regulatory Guide 1.26., Rev. 3, " Quality Group Classifica-tion and Standards for Water, Steam and Radioactive Waste
- Containing Components of Nuclear Power Plants."
1 d) Regulatory Guide 1.29, Rev. 3, " Seismic Design i Classification." e) Regulatory Guide 1.31, Rev. 2," Control of Stainless Steel Welding" as modified by Branch Technical Position MTEB-51, " Interim Position on Regulatory Guide 1.31, ' Control of Stainless Steel Welding'." f) Regulatory Guide 1.122," Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Rev.1, February 1978. g) Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Report for Nuclear Power Plants, Rev. 3." l h) NRC Guidance " Review and Acceptance of Spent Fuel Storage and Handling Applications," April 1978, and i Modifications dated January 18, 1979. j 4-7
4.3 SEISMIC AND IMPACT LOADS Maximum loads transmitted to the floor by the spent fuel racks at Millstone Unit No. 2 are given below. Floor-Rack Interface Loads (Ibp/ Pad) North-South East-West Vertical (Down) OBE + Dead Weight 12,100 12,100 107,300 (4 pads in contact) SSE + Dead Weight 39,100 38,900 131,200 (2 pads in contact) SSE + Dead Weight + 39,100 38,900 133,000 Vertical Impact (4 pads in contact) The seismic analysis of the spent fuel rack includes an assessment of the maximum sliding and tipping that can be expected. The racks are installed with a nominal gap of 2 inches between modules and are 3 inches from the pool walls. The analysis has shown that the maximum motion of the racks, including tipping, sliding, and thermal expansion is less than the gap between adjacent modules. Therefore, no contact is predicted. 4.4 LOADS, LOAD COMBINATIONS AND STRUCTURAL ACCEPTANCE CRITERIA The loads, load combinations and structural acceptance criteria used in the structural analysis of the spent fuel racks are consistent with NRC guidance in " Review an Acceptance of Spent Fuel Storage and Handling Applications" (Reference 4-1). Load Combination (Elastic Analysis) Acceptance Limit D+L Normal limits of NF 3231.la D+L+E Normal limits of NF 3231.la D + L + To Lesser of 2Sy or Su stress range D + L + To + E Lesser of 2Sy or Su stress range D + L + Ta + E Lesser of 2Sy or Su stress range D + L + Ta + E' Faulted Condition Limits of NF 3231.lc The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan where each term is defined, except for Ta which is defined as the highest temperature associated with the postulated abnormal decision conditions. 4-8
Final Stress combinations are derived from R.S.S. method of each component stresses magnitude regardless of the direction (e.g., a typical element may be comprised of both tension and compression stress combined together). The component stresses assume a three directional earthquake having their peaks occurring simultaneously. The maximum stress values associated with the analyses performed , for the Millstone 11 spent fuel racks are provided below. These values are based upon the SSE load condition. Except for the adjustment screw, the stresses associated with the SSE load condition are lower than the OBE allowable stress limits and therefore are acceptable for both the OBE and SSE conditions. The stress values for the adjustment screw and their allowable stress limits are provided for both OBE and SSE condition. The design 3
, margin is defined as (allowable - 1) X 100%.
actual NOTE: In most cases the maximum stress is asscciated with SSE
- load condition, while the allowable stress is for the OBE condition.
Maximum Stress Values Stresses do not necessarily occur at the same location Design A. Monolith Maximum Stress Allowable Stress OBE Margin 4 Membrane stress = 17,560 psi 18,300 psi 4.2% Membrane plus bending = 21,760 psi 27,450 psi 26.2 % Primary plus thermal = 23,511 psi 55,000 psi 92.9 %
\
4-9
B. Support Bars Bending stress = 5,454 psi 16,500 psi 202.3 % Shear stress = 526 psi 11,000 psi 1991.3 % C. Adjustable Foot
- 1. Block Shear Stress = 2,918 psi 11,000 psi 277.0 %
Axial plus bending OBE = 13,665 psi 16,500 psi 20.8 % SSE = 19,290 psi 33,000 psi 71.1 %
- 2. Adjustment Screw Design OBE ConditionMaximum Stress OBE Allowable Stress Margin Axial stress = 11,810 psi 49,350 psi 317.9 %
Shear stress = 18,230 psi 33,500 psi 83.8 % Bending stress = 24,980 psi 50,250 psi 101.2 % Combined axial compress. plus bending = fa + fb _ .736 1 20.8 % Fa Fb Design SSE Condition Maximum Stress SSE Allowable Stress Margin Axial stress = 14,773 psi 91,000 psi sl6% Shear stress = 29,400 psi 54,600 psi 35.7 % Bending stress = 60,554 psi 91,000 psi 30.28 % Combined axial compress, plus bending = fa + fb = .828 1 20.8 % Fa Fb 4-10
I Design SSE Condition Maximum Stress SSE Allowable Stress Margin Thread shear = 6,710 psi 11,000 psi 63.9 % 4.4.1 Pool / Auxiliary Building Analysis 4.4.1.1 Introduction The Millstone Unit No. 2 spent fuel storage pool capacity expansion program requires an analysis of the spent fuel pool and auxiliary building for the increased loads caused by the increased storage of spent consolidated fuel. This section addresses the guidelines and acceptance criteria that NNECO followed to qualify the spent fuel pool and auxiliary building for the new loadings. The Millstone Unit No. 2 auxiliary building is a multi-story concrete structure. Spent fuel storage is provided between column lines 17.2 and 18.9 and column lines H.4 and L.5 at elevation (-)2'-0". The storage area consists of a reinforced concrete pool lined with a one-quarter inch thick stainless steel liner to elevation 38'-6". Normal water level is to elevation 36'-6". A leak chase system consisting of channels ernbedded behind the liner at all seams and connected to a collector system is used to monitor and control any possible leak from the pool. Construction materials used in the construction of the pool include ASTM A-240, Type 304 stainless steel, ASTM A-615 Grade 60 reinforcing steel, and 3,000 psi 28-day strength concrete. The analysis of the spent fuel pool and associated components of the auxiliary building to accommodate the loadings associated with increased storage capacity (consolidated fuel) was accomplished with the use of a large finite element model. The following sections provide the details of the spent fuel pool / auxiliary building structural analysis. Accompanying figures and tables are presented in Appendix A to this section. 4-11
4.4.1.2 Pool / Auxiliary Model Data The models were derived based upon information supplied from NUSCO-Millstone Unit No. 2 drawings. The spent fuel pool and associated auxiliary building components model contain over 9,600 degress of freedom. 4.4.1.3 Mathematical Model The extent of the structural model includes the pool walls, cask laydown and fuel transfer canal area walls (excluding the gates), pool floor slab and fuel transfer canal floor slab and the foundation walls directly beneath this portion of the auxiliary building. All walls directly adjacent the pool (including the fuct transfer canal inside wall and cask laydown area walls) and the pool floor slab are modeled with two layers of eight node solid elements to permit proper application of thermal gradients and to provide good definition of stress variations through the wall thickness. Four node membrane elements of negligible thickness were used on the inside, middle, and outside surfaces of the wall or floor to obtain stress values at the solid elements faces as well as at the solid element centroids. In this manner, five integration points through the walls and floors were obtained. The outer walls and floor slab of the fuel transfer canal area were modeled with a single layer of solid elements since these components were only included for their stif fness properties and were not evaluated according to stress criteria. The portions of the foundation which were modeled include the south, west, north, inner west, inner south and east foundation walls. These components were modeled with only one layer of solid elements with membrane elements on the inside and outside surfaces since there is no thermal gradient through the walls of the compartments at this elevation. The other structural components modeled in the foundation were the pier (solid elements) and the extensions of the inner west and east foundation walls (which were modeled with membrane elements to represent their in-plane stiffness). 4-12
Since rotations at the node points of the three-dimensional solid elements are not defined, all rotational degrees of freedom in the model were restrained. Stiffnesses of the walls and floors framing into the pool model were represented using direct matrix additions. The matrix coupling terms were computed assuming that, due to cracking, one-half of the wall or floor panel stiffness is available. The nodes at the base of the foundation which are remote from the structural areas of interest in the pool were completely restrained. The liner plate was modeled such that all weld seams and anchor locations were coincident with node lines or node watons. Global and local coordinate systems were specified psd, trat they were coincident with the pool floor slab elements in the SAP 6 finite element model. All rotations and displacements normal tothe plate were restrained for the non-thermal case. For the thermal load case, an additional modeling refinement was made by including the horizontal portion of the 900 bent plate which forms the corner piece at the intersection of the floor slab with all walls. Rotations were permitted at the overlap of this plate which forms the corner piece at the intersection of the floor slab with all walls. Rotations were permitted at the overlap of this plate and the liner plate. Lateral degrees of freedom are unrestrained for all nodes except weld seams and anchor locations, which were identified as boundary degrees and restrained by nonlinear load / displacement relationships that represent test-measured anchor characteristics. The results of the finite element model were examined to insure that realistic deflections and stresses existed for each individual load case. Classical solutions were also prepared for selected components for comparison to the finite element model results. Gross force and moment reactions were calculated and resulting stresses were compared to those in the computer model. The general behavior of the model under the loads was determined to be reasonable by viewing deformed geometry plots and screening stresses at key locations. 4-13
The material properties used in the mathematical model were obtained from design criteria specifications or by NUSCO Engineering. Concrete Material Properties Concrete Compressive Strength 4,800 lb/in2 Reinforcing Yield Strength 60,000 lb/in2 Reinforcing Elastic Modulus 29.0 x 106 lb/in2 Concrete Elastic Modulus 3.95 x 106 lb/in2 Concrete Poisson Ratio 0.17 Concrete Thermal Expansion Coefficient 5.5 x 10-6 in/in/0F Concrete Weight Density 8.68 x 10-2 lb/in3 (150 lb/f t3) Liner Plate Material and Anchor Properties Plate Aiaterial 304 Stainless Steel Plate Thickness 0.25 inches Plate Thickness Tolerance 16 % Poissons Ratio 0.24 Coefficient of Thermal Expansion 8.82 x 10-6 in/in0F Yield Strength 30 ksi Weld Electrode E308-16 Electrode Tensile Strength 90 ksi 4.4.1.4 Load Cases To provide flexibility for formulation of the load combinations, a static analyses was performed for the loads described in this section with the appropriate factors and permutations applied to these loads for formulation of the SRP load combinations. The load applied to the mathematical model of the spent fuel pool and liner were derived based on a 2:1 consolidated fuel load. 4-14
4.4.1.4.1 Structural Individual Load Cases The twelve individual loads applied to the finite element model are described in Table 3.2-2. Loads which were excluded from this evaluation include fuel cask drop, crane load, rack impact and accident flood load. Fuel cask drop has been previously addressed and therefore is not considered in this analysis. The loads from the fuel handling crane were excluded since the effect on the overall pool structure was considered beneficial when considered in combination with other loads. This assumption is based upon the fact that the relatively small compressive vertical load exerted on the pool walls, due to the crane weight, aids the concrete section's ability to carry shear forces as well as other axial and moment loadings. Impacting . of the rack pads due to tipping was considered a local ef fect and was addressed as a separate item. Accident flood load has also been eliminated from consideration since the flood gates protect the auxiliary building to tt e maximum probable flood height. Dead weight of the pool structure was defined as a 1.0g vertical acceleration. Hydrostatic loading of the structure was analyzed for a pool water depth of 38'-6". The hydrostatic forces are applied to the wetted surface of the pool by computing nodal forces in the three directions as the product of the pressure at the nodal elevations by an area vector (A xe A y> Az) which is computed from adjacent element areas. Membrane elements (only for the purpose of load application) were used to represent the gates in the fuel transfer canal and cask laydown areas so that the hydrostatic forces on the gates were accounted for. A resultant force was computed for this load verifying application of the load and additionally, confirming correct orientation of the elements since the nodal area vectors are based on the local coordinate systems of the membrane elements. Individual load cases 3,4, and 9 through 12 are nominal 1,000 pounds per square foot loads applied to the pool floor slab in the negative global z (vertical), x and y directions. These unit load cases were used to later formulate vertical (z) racks loads and lateral (x-y) loads. Application of the load in each direction was subdivided into two load cases to provide for the dif ferential fuel rack configurations in 4-15
regions 1 and 2 of the pool. Load cases 5 and 6 are operating and accident thermal loads, corresponding to pool water tmperatures of 1500F and 2120F, respectively. The ambient (or stress free) temperature for all compartments outside the pool (including the cask laydown and fuel transfer canal areas) was defined as 550F. These loads were applied by defining nodal temperatures for all nodes in the model based on linear interpolation of temperatures between adjacent compartments. The accident pool temperature of 2120F is justified since the pool water free surface is at atmospheric pressure. The pool bulk temperature will also be fairly uniform as a result of convection currents caused by heating of the water at lower elevations resulting in the movement of this lower density water toward the top of the pool. Building seismic effects and the associated hydrodynamic forces due to lateral earthquake loads are included in load cases 7 and 8. The horizontal earthquake acceleration applied for these loads was calculated by taking the average of the floor zero period accelerations, determined from the auxiliary building seismic analysis for the various levels over the pool height, and applying this acceleration to the structural mass of the model. Using the peak acceleration value from the various floor elevations over the pool height, the average peak horizontal acceleration value was found to be 0.21 g's for the 0.09 g (OBE) building base excitation. To facilitate load combinations, this seismic acceleration was expressed in terms of a nominal 2.34 g peak acceleration at the spent fuel pool elevation. This nominal 1.0 g base excitation and resulting 2.34 g fuel pool acceleration is indicated in Table 3.2-2 for individual load cases 7 and 8. 4-16
Earthquake response of the pool water was based on the methodology outlined in TID-7024, " Nuclear Reactors and Earthquakes," which provided a basis for computing pool wall and floor pressures which i result from eathquake-induced pool fluid motion. Hydrodynamic forces were calculated as the product of the pressure profiles over the wetted surfaces of the pool and their associated area vectors, similar to the application of the hydrostatic forces described previously. Gross hydrodynamic forces and moments were computed from these nodal forces, with verification by comparison to forces and moments calculated from formulas in TID-7024. These hydrodynamic responses were also normalized to a 1.0 g earthquake to facilitate load combinations. Vertical earthquake loads were not included as individual load cases, since acceleration of the pool water mass and concrete mass are equivalent to applying appropriate load factors to their respective static load cases to account for dynamic amplification of the seismic motion. Table 3.2-3 summarizes the load definition parameters used i.. evaluating the concrete structure. 4.4.1.4.2 Composite Load Cases The twelve individual loads just described were combined to formulate the composite load cases applicable to this evaluation. The composite loads are shown in Table 3.2-4a and 3.2-4b and include dead load (D), live load (L), operating and accident thermal (To and Ta), and SSE and OBE earthquake. The tables also define the relationship between individual loads and composite loads. The Standard Review Plan load combinations which are described later in this section are formulated from these load cases. Table 3.2-4a is based on fuel rack loads developed as part of the initial analyses of the fuel racks. The fuel pool was initially analyzed 4-17
in accordance with the composite loads shown in Table 3.2-4a. Further analysis of the fuel racks resulted in changed fuel rack loads, which are reflected in Table 3.2-4b. The effect of these revised fuel rack loads is limited to the fuel pool slab, slab liner plate, and foundation since structural response of the pool walls and wall liner plate is dominated by thermal and hydrostatic loads. The evaluation presented in Appendix A to Section 4, pages 28 through 53, appropriately uses the Table 3.2-4a and Table 3.2-4b load combinations. < 4.4.1.4.3 Dead Loads Dead load includes dead weight of the concrete structure, hydrostatic pressure and weight of the fuel rack modules excluding their fuel complements. The rack module dead weight was 365 pounds per cell. Since the individual load cases for rack loads were based on nominal 1,000 psi vertical loads over Regions 1 and 2 of the pool floor slab, individual load cases 3 and 4 are factored by 0.374 and 0.607. As discussed above, fuel rack dead loads were revised to 384 psf in region 1 and 216 psf in region 2, based on the sum of individual submerged rack weights. 4.4.1.4.4 Live Loads s Live load consists entirely of the submerged weight of the consolidated fuel and storage box. The weight of these two items is 2,500 pounds per cell. Region I supports 384 cells and region 2 supports 962 cells. Basd on these values, the floor slab vertical load was computed as 2,560 psf over region I and 4,155 psf over region 2. Subsequently, the region 1 and 2 boundaries were redefined with region I having an area of 290 ft2 and region 2 having an area of 639 ft 2. The cell weight and number of cells per region remained as defined above to yield a region I load of 3,310 psf and a region 2 load of 3,770 psf. 4-18
_ _ .~ . . . . - - - _ . - 4.4.1.4.5 Thermal Loads Operating and accident thermal composite loads were taken directly 4- as their individual load cases with factors of 1.0. i -4.4.1.4.6 - Earthquake Loads b Operating basis earthquake (E) was specified as 0.09 g horizontal and 0.06 g vertical ZPA levels measured at the base of the foundation. Since amplification of the base motion acceleration levels was ! accounted for in the individual load cases, a coefficient of 0.09 was applied to the horizontal response loads (load cases 7 and 8). Similarly, the response to vertical earthquake is constant over the , pool height as specified in the plant design manual, so a factor of 0.06 on the dead weight load was used for this load case. SSE horizontal and vertical reactions for the submerged racks were
- specified as 3,500 pounds per cell and 1,000 pounds per cell, respectively. OBE loads are calculated as 56 percent of the SSE loads. Based on these cell reactions, the OBE vertical loads are >69 psf over Region 1 and 923. psf over Region 2. The resulting OBE horizontal loads are 1,992 psf over Region 1 and 3,232 psf over Region 2. These loads were subsequently redefined resulting in Redioa 1 OBE loads of 643 psf vertical plus 646 psf horizontal.
Region 2 OBE loads were redefined to 672 psi vertical and 908 psf horizontal. l As outlined in Standard Review Plan Section 3.8.4, the three directions (X, Y, Z) of earthquake were applied such that all permutations of the signs were considered. Table 3.2-4a and b shows four of the OBE composite loads. Four additional cases not shown in Table 3.2-4a and b were developed by multiplying those shown in the table (El through E4) by -1.0. Similarly, SSE loads were formulated by multiplying the eight OBE cases by 1.8. i ? 4-19
. _ _ _-_ -_. _ _ - =
The service and factored load combinations were formulated according to Section 3.8.4, paragraph 3.6 of the Standard Review Plan. Table 3.2-5 presents the eight service load combinations and five factored load combinations from the Standard Review Plan. Eight of the SRP composite load components were not applicable to this structure and were not considered in the evaluation. These composite load components include R o (normal operating pipe reactions), W (design wind), Wt (design tornado), Ra (pipe break reactions), Pa (accident pressure ) and Yr, Yj, Ym (impact and impulse from pipe break and impact). Excluding these loads, the final loads considered reduce to those shown in Table 3.2-6. Examination of Table 3.2-6 shows load cases i.b.1 and i.b.3 to be identical, as are i.b.4 and i.b.6. Since live load is always present, the response of the structure to i.b.7 is bounded by i.b.2. Similarly, load case i.b.1 bounds i.b.8. This results in four service load combinations considered, two of which contain OBE, which has eight sub-load cases, resulting in a total of eighteen service load combinations. The response of the structure to T o is similar to Ta, with Ta controlling. Therefore, load case ii.b was eliminated in lieu of li.e. For the same reason, load cases ii.a and ii.e are bounded by li.d. This leaves two factored cases, one containing SSE, which has eight subcases, resulting in a total of nine factored load combinatilons. Table 3.2-7 summarizes the coefficients applied to the composite loads for formulation of the service and factored loads previously described. Since the effect of the dead and live portions of a load combination are reduced during earthquake motion in the negative global direction, the factors on these composite loads are reduced by 10 percent. The final loads were formulated for all areas of the pool which were considered in this evaluation. Analysis was then performed for each particular concrete wall or floor for the two or three controlling load combinations. 4.4.1.4.7 Liner Plate Load Combination Formulation The individual and composite load cases used for evaluation of the 4-20
~
liner plate are identical to those presented in Table 3.2-2 and 3.2-4b, respectively. The service and factored loads specified by th'eStandard Review Plan for plastic design methods are shown in Table 3.2-9. The same eight components for composite loads that were not considered for the liner plate analysis: including R (pipe break reactions), Pa (accident pressure), and Yr , Yj, Y T, (impact and impulse from pipe break and missile impact). Excluding these loads, the loads considered were reduced to those shown in Table 3.2-10. From Table 3.2-10, it is evident that load cases i.b.1 and i.b.3 are identical, as are 1.b.4 and i.b.6. Application of OBE in all possible locations resulted in load combination i.b.1 being bounded by i.b.2. The number of service load combinations considered was reduced to three, two of which contained OBE, which has eight subcases, resulting in seventeen possible service load combinations., The response of structure to To was bounded by Ta, which resulted in elimination of li.b.2 in lieu of li.b.3. Similarly, load case ii.b.1 was bounded by li.b.5. Structural response due to SSE (which is OBE factored by 1.8) results in elimination of li.b.4 in lieu of li.b.5. A load case of (D + L + E') was considered separately to address the effects of earthquake without thermal loads. These factored load combinaticns re main, two containing SSE which (considering earthquake permutations) results in a total of 17 factored load combinations. The final composite load case coefficients are summarized in Table 3.2.11, for the service and factored load cases previously described. Applied displacements and strains due to cracking and curvature effects were applied for the load combinations described. Concentrated loads representing the rack pad forces were not applied directly to the liner plate model at the individual load case level. It can be shown that the coefficient of friction between the rack pads and liner plate (steel-to-steel interface) is less than that between the 4-21
liner plate and concrete slab. Consequently, the racks will slide before the load will be taken by the liner plate. If the rack pads stick (corresponding to a coefficient of friction of 1.0), the force provided by the cell's vertical reaction and the concrete liner plate friction is greater than the cell's horizontal reaction. In either case, the load is transmitted directly to the concrete slab which was qualified for the design loads. The tables for this section are presented in Appendix A to this section, pages 17 through 27. 4.4.1.5 Pool / Rack Interaction The dynamic interaction betwen the pool structure and the rack modules was accounted for by considering the mass of fully loaded rack modules in the dynamic analysis model of the auxiliary building. Motions of the spent fuel pool from a time-history analysis of the auxiliary building were then used as input for a nonlinear seismic time-history analysis of the spent fuel rack modules. The nonlinear time-history analysis of the rack modules produced seismic loads which are transmitted to the pool floor. These seismic loads consisted of horizontal shear loads and vertical loads including impacting of the rack module on the pool floor. The total horizontal loads on the pool floor are obtained by combining the loads due to the North-South and East-West earthquake directions in accordance with Regulatory Guide 1.92. The total vertical loads are obtained by combining the vertical seismic load and the tipping impact load in accordance with Regulatory Guide 1.92 and adding the deadweight load. As far as phasing of racks, the nonlinear seismic analysis of the racks assumes all the rack modules move in phase. 4.4.1.6 Summary of Results The spent fuel pool was evaluating according to the criteria in the Millstone Point Unit 2 Design Criteria NRC Standard Review Plan. The original design was performed according to ACI-313-63 code 4-22
s criteria. I or this evaluation NNECO has chosen to utilize load combinations specified in the in the NRC Standard Review Plan followed by evaluation of the reinforced concrete sections according to ACI 349-80. The pool wall and floor liner plate were evaluated according to the strain criteria specified by _the ASME Code. A plate thickness tolerance of 16% was used, along with the weld offset, for computing membrane plus bending strains. Pool floor liner plate weld stresses were compared to AISC criteria. As shown in Table 3.1-1, a stress allowable criteria is used in evaluating the anchors for nonthermal loads versus a displacement criteria for thermal load combinations. The following tables in Appendix A to this section, pages 28 through 53, identify the critical spent fuel pool and liner stresses and their comparison'to allowable values based upon the previously described criteria. As described previously, these stresses are based on fully consolidated fuelloads. By reviek of these tables, it can be shown that all stresses / strains,
'.except the liner weld stress, remain within the stated code allowables.
The stress levels in the weld, with thermal considered, is 55.6 Ksi which represents Pm + Pb + Q (secondary), a combined stress level l which exceeds the allowable stress of 53.4 Ksi by about 4 percent. The strain levels for the weld section, however, are well within code l allowables (2.14 SF). This implies that the liner / liner welds under the
~ thermal condition will relax and produce a self limiting stress level l - which will ensure a leuk' tight boundary during the thermal accident condition.
An an'alysis wa's performed which investigated the local effects on
$ the pool floor slab due 'to rack module impact loads. The analysis
, , considered two adjacent rack mounting feet impacting the slab simultaneously. The concrete being impacted was considered to be fully cracked. Therefore, only the residual reinforcing bar strength was accounted for. The controlling load combination for this analysis 4-23
._A___-_________-_-
l
- j. was 1.7 (D + L + E). It was determined that the residual shear strength for the section is 3,565 kips. The required residual shear strength capacity is 239.4 kips.
4 1 The analysis therefore shows that the structural integrity of the pool floor is maintained when subjected to the local maximum rack j module dynamic mounting foot loads. 4 J l1 1 4 f a 1 i k' t J 1 9 4 l 4-24 i-
~ ~n, ,, - . ., ,. --.,-e.,...------, ... , -_.- -, , , , . - , _ - - , -
,.,.m . . - , - - - , ,,a-, -n--- --a ,- .--,- , .
4.5 DESIGN AND ANALYSIS PROCEDURES 4
, 4.5.1 Seismic Methods Development for Consolidated Fuel 4.5.1.1 Consolidated Fuel Storage Box Structural Tests 4
Three dif ferent series of structural tests were conducted on a consolidated fuel storage box to obtain static and dynamic properties of the box and fuel rods for subsequent fuel rack seismic analyses. Static load-deflection tests were performed on a loaded box to measure the lateral stiffness and static deflection shape of the box with fuel rods. These tests were performed with two fuel rod compaction ratios. Static compressive load tests on a short section of box measured the local wall stiffness of the box. Forced vibration tests on a loaded box in water determined natural frequencies, mode shapes, critical damping ratios, excitation forces and magnification ratios. The forced vibration tests were performed with three fuel rod compaction ratios. Results of the tests were evaluated and 4 correlated with analytical models to obtain a consolidated fuel storage box or canister model for use in seismic analyses of the fuel racks. The tests showed that the fuel rods have an insignificant effect on canister static and dynamic stif fness and that the static stif fness is linear. Fuel rods can be modeled as additional lumped weights. Water effects in the forced vibration test can be predicted well using standard theoretical calculational techniques used in fuel rack analyses. The water can be represented as additional lumped weight. Fuel rods do significantly affect damping ratios, which tend to increase with decreasing compaction ratios. The canister and fuel rods in water can be represented with a lumped-mass model of about 10 nodes without loss in accuracy. l 4-25
4.5.1.2 Consolidated Fuel / Storage Rack / Pool Interface Modeling A computer code CFCGEN was developed which automatically generates a lumped-mass stick model of a fully loaded consolidated fuel storage box or canister from canister and fuel dimensions and material properties. The code is generic and can be used for dif ferent canister and fuel designs. The effect of certain spent fuel rack design parameters on the lateral seismic response of the racks was investigated. Non linear time history analyses were performed. The design parameters investigated were the following:
- 1) rack support scheme - welding of storage cell feet to base plate of rack module
- 2) length of vertical inter-cell welds in storage rack modules.
- 3) width of gap between rack and pool wall
- 4) storage cell wall thickness Using CFCGEN and the methods described in Section 4.5.2 the racks with consolidated fuel were modeled in both lateral directions and analyzed with corresponding acceleration histories for the Millstone 2 pool floor elevation.
It was found that by changing the design parameters of the rack and its support scheme, the rack frequency can be detuned from the frequency of the peak of the earthquake spectrum and seismic loads reduced. This was implemented in the Millstone Unit 2 rack design. 4.5.2 Spent Fuel Storage Rack Design Analyses The spent fuel storage racks are designed to withstand forces generated during normal operation, an Operating Basis Earthquake, or a Safe Shutdown Earthquake. Lateral and vertical seismic loads along with fluid forces are considered to be acting simultaneously on the fuel racks. The racks are designed to assure rack structural integrity while at the same time keeping the fuel in a subcritical sta te. 4-26
Linear response spectrum methods are used for the vertical direction. The lateral seismic responses of the spent fuel storage racks are determined using a non-linear time history analysis. Non-linear time history analyses are performed for the lateral directions primarily because of impacting. The effects of impacting structures significantly influence the stresses in both the storage structure and the intact assemblies or consolidated fuel storage boxes and, because they are non-linear in nature, can only be accounted for by perform ng more complex non-linear time history analyses. Rack verticalimpact loads on the floor due to tipping are calculated. The seismic input used for these analyses consists of the vertical response spectrum and the lateral acceleration time histories corresponding to the pool floor elevation at Millstone Unit 2. The analyses are performed in accordance with Reg. Guide 1.122, Revision 1, February 1978. The first step in the analytical procedure is to determine the dynamic characteristics of the fuel storage racks. This is done by developing a three-dimensional firiite element model of the structure and solving for the natural frequencies and mode shapes in air. The finite element code used in the study is SAP IV (see Section 4.5.3). The resulting dynamic characteristics are then incorporated into a non-linear representation of the entire system, which includes the rack and intact or consolidated fuel in the storage rack. The CESHOCK computer code (see Section 4.5.3), is used to determine the non-linear time history response of the system. The ef fects of impacting between the fuel and the storage rack are represented in two CESHOCK models, one for intact and one for consolidated fuel. Because of the close proximity of the structures, hydrodynamic coupling effects between the fuel, the storage rack and the pool are also included in the models. (See Reference 4-2 (attached as Appendix B) for additional information.) 4-27
This reference describes the general methodology used to develop a nonlinear seismic analysis model of a spent fuel rack module while stressing the importance of modeling fuel assemblies as discrete , structural elements and the non-linear impacting behavior between the rack module and the stored fuel. The analysis for this report differed from that presented in the referenced paper primarily in that the analysis was done using models based on the Millstone Unit No. 2 rack module designs, pool layout, and site specific acceleration time history data. The analysis included cases of standard fuel assembly storage as well as consolidated fuel storage. The racks are analyzed using a finite element model in the SAP IV code and the loads from Section 4.3. SAP IV output consists of membrane stresses and bending moments for each element. The component stress on each element resulting from the application of each directional load is combined by the square root sum of the squares method. The results are compared to stress allowables in accordance with the rules of ASME Boiler & Pressure Vessel Code Section 111, Subsection NF 3000. A schematic description of the mathematical model used for the non-linear rack module analysis is shown in Figure 4-7. The model is two-dimensional, with each mass having a translational and a rotational degree-of-freedom. Mass nodes 1 through 18 were used to represent the fuel rack module. These mass nodes were linked by massless flexible elements. Similarly, mass nodes 19 through 27 were used to represent both the standard fuel assembly and/or the consolidated 4 fuel storage box. Hydrodynamic couplings, designated by element H , are included between the rack module nodes and the pool structure nodes, and between the fuel nodes and the rack module nodes. Non-linear gap-spring elements were used to represent the possibility of impacting between the fuel and the rack module. The fuel was coupled to the base of the rack module by a " slip-stick" friction element. An element of the interface of the module base and the pool liner 4-2S
represented a " slip-stick" friction element in the sliding analysis and a non-linear torsion spring in the shear and rocking analyses. An in-phase mode of vibration was conservatively considered in assessing the hydrodynamic coupling effects between adjacent rack modules. Because of the character of the site specific Millstone 2 seismic excition, the higher rack module frequencies resulting from the in-phase node analysis were consevative because they were closer to the frequencies of the response spectra peaks. An out-of-phase mode of vibration would have resulted in the lower frequencies farther away from the response spectra peaks. The lower frequencies result from high hydrodynamic masses produced by out-of-phase motion. In the nonlinear analysis models, hydrodynamic coupling is specified between the rack module and the pool, and between the fuel and the rack module. Potential theory (incompressible inviscid theory) is employed, using simple two-dimensional models of the structures coupled by the fluid, to estimate the hydrodynamic virtual mass terms based on the model configuration. Three-dimenstional end effects were then accounted for by modifying the calculated hydrodynamic mass terms. For the rack module-to-pool hydrodynamic element, the rack modules were assumed to move in-phase and the potential theory model consisted of two bodies: the fuel rack module array within .he spent fuel pool structure. To determine the resulting hydrodynamic mass terms, a finite element analysis using a computer code based on two-dimensional potential flow, was used. The ADDM ASS computer code, C-E proprietary, was used to calculate the hydrodynamic masses of two dimensional bodies with arbitrary cross-sectional shapes with fluid finite elements between the bodies. ADDMASS is based principally on the following work: Yang, C.I., "A Finite - Element Code for Computing Added Mass Coefficients," Argonne National Laboratory Report No. ANL-LT-75-49, September 1973. 4-29
l l The horizontal non-linear CESHOCK analyses determine the tipping associated with fully and partially loaded fuel racks. Load factors are used to transfer the vertical and horizontal base loadings from the two dimensional CESHOCK non-linear analyses to the liner three-dimensional model which has been modified to represent the tipped condition. A separate load factor is calculated to represent the base movement developed during tipping of the racks. The vertical tipping displacement from the horizontal nonlinear - analysis is used as the input to a separate vertical nonlinear CESHOCK model that is used to calculate vertical impact loads. These loads are used to determine the adequacy of the foot / rack design. Non-symmetric fuel loadings were considered by analyzing partially loaded modules. The most severe -cases considered were those modules with a single row of cells along an edge of the module loaded with consolidated fuel. The worst case for tipping was a Region 117x3 module, partically loaded with consolidated fuel, excited by the East-West seismic component (7 cell direction). The worst case for shear load was a Region 117x9 module fully loaded with consolidated fuel, excited by the North-South seismic component (9 cell direction). The maximum horizontal displacement of the top of a module, including tipping, is determined from a non-linear time history analysis of an individual module. Separate analyses are made for a number of different modules with varying degrees of fuel loading, including empty, partially loaded, and fully loaded modules. In these individual module time history analyses, all the modules in the pool are assumed to move in-phase when determining the rack-to-pool hydrodynamic characteristics. 4-30
To' calculate the peak intermodular relative displacement, adjacent
= modules are assumed to move out-of-phase. The peak relative displacement is conservatively calculated by summing the absolute value of the peak displacements at the top of the module for the two modules considered. Using this approach, a peak intermodule relative displacement of 1.776 inches was determined. This value is less than the intermodular gap.
, .C-E uses a gap-spring element to model the impact between the fuel assembly and the rack cell in a nonlinear dynamic analysis. The spring represents the spacer grid one-sided impact stiffness with the appropriate gap. C-E determines fuel assembly one-sided impact stiffness using full-scale fuel assembly pluck impact tests and model-test correlations of the test data with analytical results. The value of the spacer grid impact stiffness for the Westinghouse fuel assemblies that was provided to C-E by Northeast Utilities was greater than that for a C-E fuel assembly and was conservatively used in the nonlinear dynamic analysis. Impact damping was conservatively not used in the analysis. Friction between the pool liner and the module mounting feet is addressed a two ways. In the first approach, the rack module is not permitted to slide relative to the pool. in this case, the coefficient ^ of friction is assumed to be extremely high, to model the possibility of adhesion between the rack module and the pool, which could occur over the design life of the modules due to one of several mechanisms. { This fixed-based model provides conservative shear loads to both the module and the poolliner. The second approach uses a sliding-base model in which a friction element connects the rack module base to the pool liner. The 4 friction element used in a " slip-stick" friction element with a velocity dependent coefficient of friction. Realistic values for the coefficient of friction are used in this sliding base model. A static coefficient of friction of 0.55 was used. The coefficient of friction i ' 4-31 i
decreases linearly with increasing relative velocity of the module base with respect to the pool liner until a minimum dynamic coefficient of friction of 0.28 is reached at a relative velocity of the module base with respect to the poal liner until a minimum dynamic coefficient of friction velocities above 2.5 in/sec. For relative velocities above 2.5 in/sec., the minimum dynamic coefficient of friction applies. The friction values used are based on the following sources: i) data from Combustion Engineering laboratory tests, ii) data obtained through a technical exchange agreement with Kraf twerk Union (KWU) of West Germany. Final Report of a Theoretical and Experimental Study for Further Development of Light Water Pressurized Water Reactors, " Wear Behavior of Friction Materials and Protective Layers With Regard to their Application Possibilities in Water Cooled Nuclear Reactors", written by P. Hoffman, Metallic Materials RT41, Fordervagsvorhaben BMFT-Inv. Reakt. 72/S11 Draf twert Union, August 1973., and iii) textbook Friction and Wear of Materials, Ernest Rabinowicz. Justification for the use of the stated values of friction coefficient lies in the bases of their selection being results of experimental studies. The values used in the analysis are values that have been derived from laboratory testing. Structural analyses were also performed to evaluate the results of a maximum crane uplift force, and a fuel assembly or loaded consolidated fuel storage box drop on the fuel storage rack structure. A fuel assembly or loaded consolidation box falling onto the racks can either fall into a cavity or onto the top of the racks. 4-32
In the case of a fuel assembly falling into an empty cell: The fuel assembly drop accident was evaluated to determine the effect of the dropped assembly on the functional and structural integrity of the racks. The analysis indicated that the impact of the fuel assembly on the support bars caused plastic deformation of the support bars and the fuel cell wall supporting the bars. The fuel bundle drop through the rack to the fuel support resulted in the side walls of the rack shearing however, the bundle and support bars did not impact the floor, resulting in no damage to the pool liner. (The active fuel length of the bundle will remain contained within the storage rack.) In the case of a consolidated fuel storage box falling into the cavity: It was conservatively assumed that the impact of the consolidatcJ fuel storage box caused plastic deformation of the support bars and fuel cell wall supporting the bars; resulting in an impact of the consolidated fuel storage box and support bars to the pool liner. With respect to the racks, the primary function and structural integrity of the racks to maintain separation of the fuel was not impaired. With respect to the spent fuel pool liner and concrete structural capacity: An evaluation was completed in March 1983 to determine the potential consequences of drops into the Millstone Unit No. 2 spent fuel pool. The analysis was performed as part of NUREG-0612-
" Control of Heavy Loads at Nuclear Power Plants." This analysis addressed an object which weighed 67,000 pounds being dropped from a height 41 feet above the spent fuel pool slab through water directly onto the slab at elevation (-)2'-0". The above-mentioned calculation conservatively takes no credit for any energy absorption that would 4-33
occur due to deformation of the object at impact. The 67,000 pound object dropped would be expected to perforate the spent fuel pool liner. The object would penetrate into the concrete floor slab but not perforate it and therefore not jeopardize the floor slab structural integrity. Since no perforation of the concerete slab is expected, no gross leakage from the poolis expected to occur. Assuming the consolidated storage box (weighing approx. 2500 lbs), hits the weakest part of the spent fuel pool, which would be the leakage detection system for the liner welds, the leak detection and monitoring system, as described in the Millstone Unit No. 2 Final Safety Analysis Report Sections 5.4.3 and 9.5.2, would be activated. The leak detection and monitoring system would be utilized if (a) a pool liner weld seam was to fail or (b) a local failure of the leak chase collection channel due to a dropped object directly hitting it was to occur. If the spent fuel pool liner is perforated, the 5'-0" thick reinforced concrete spent fuel pool floor slab will nct as a water retaining barrier. If either a weld seam or leak chase collection channel was to locally fait due to a dropped consolidated . fuel storage box, both overall leakage detection system integrity and pool structuralintegrity would be maintained with no significant pool water inventory loss. 2 In the case of a fuel assembly falling onto the top of the rack: The load resulting from the impact of a dropped fuel bundle was calculated. Using a finite element model, the calculated impact force was applied at various locations on the top of the rack. The analysis of a fuel bundle drop, onto the top of the storage rack, results in an impact stress that showed local deformation at the point of contact on the rack. However, the results of the analyses have shown that the stresses in the region of active fuel are well below a!!owable stress limits, therefore it is concluded that the racks primary function remains unaf fected. In the case of a consolidated fuel storage box falling onto the top of the racks: 4-34
l For the analaysis of a loaded consolidation fuel storage box dropping on to the top of the fuel rack structure a dynamic non-linear multi-spring / mass model of a loaded storage box was developed. This was used in CESHOCK analysis to determine the impact load. The maximum impact load resulted in cell wall stresses exceeding the proportional limit. An iterative elastic / plastic CE/* 'C analysis of an appropriate region of the rack was therefore . formed to evaluate the permanent deformations throughout the re,, son. The maximum permanent deformation of the cell walls in the active fuel region was found to be 1.5 mils, which is less than 1 percent of the nominal clearance betwee the cell wall and a fuel assembly or CFSB. Hence, this accident does not af fect the primary function of the rack which is to maintain separation between fuel assemblies and/or consolidated fuel storage boxes and insure the flow of coolant. In the case of a Maximum Crane Uplift Force: An analysis of a typical fuel rack indicated that the force required to deform an individual canister or to overcome the dead weight of the rack is significantly greater than the load which the spent fuel handling machine can impart. For the analysis a maximum load of 6000 lbs. was assumed. This load was applied to the fuel rack; resulting stresses were well within allowable limits. With respect to seismic loadings on the spent fuel rack modules, the following rack modules were analyzed:
- 1) Region I- 8 x 10 module
- 2) Region II - 7 x 3 module
- 3) Region II - 7 x 9 module
- 4) Region 11 - modified 7 x 9 module The pool layout was arranged so that the rack modules were placed in specific locations and orientations within the spent fuel pool.
Acceleration time histories were available for both the north-south 4-35
and east-west directions. The acceleration time histories were applied to the rack modules in a manner consistent with actual module in-pool orientations. The worst case for shear load was found to be a Region 117 x 9 module, fully loaded with consolidated fuel and excited by the north-south seismic component. The most significant factor in identifying possible worst cases is the relationship between the model natural frequencies and the acceleration response spectra for the appropriate spent fuel pool acceleration time histories. For a given response spectrum, potential worst cases may be identified by selecting cases where the model natural frequencies are near the peak of the response spectrum. There are a number of other factors, however, that have an effect on the model frequency characteristics and consequently the response loads. Among these factors are the natural frequency of the rack module in air, the type of fuel storage (standard bundle v s. consolidated), and the hydrodynamic elfects between the fuel and the rack module as well as between the rack module and the pool structure. Because a number of factors affect the identification of a " worst case", several analyses are performed, corresponding to dif ferent regions of the pool, different size modules, different earthquake directions, and type of fuel storage. The results of the non-linear time history analyses, performed in both horizontal directions, and the linear response spectrum analysis, performed for the vertical direction, provide a set of load multiplication factors to be applied to the three-dimensional SAP IV stress model. The horizontalload factor is defined as the ratio of the maximum horizontal shear load derived from the CESHOCK model non-linear time history analysit to the horizontal empty rack (modal) weight from the SAP IV model. Likewise, the vertical load factor is defined as the ratio of the maximum vertical load determined from the response spectrum analysis to the vertical empty rack (modal) weight from the SAP IV model. The load factors are applied to the component stresses obtained from the SAP IV model. These stresses 4-36
. . . _ = . - _ - .. _ - . - .
were obtained by applying a one-G response spectrum load to each of the three orthogonal directions. Maximum Base shears and load factors are tabulated below: Base Shears Region I Rack Region II Rack Maximum Horizontal: SSE 880#/ Cell 977 #/ Cell l OBE Not Applicable 603 #/ Cell l Base Shears Region I Rack Region 11 Rack Maximum Vertical: SSE 3721 #/ Cell 3423 #/ Cell , OBE SSE values for maximum vertical base shears were used. Typical Load Factors Region I Rack Region 11 Rack i Horizontal (X-direction) 10.10 12.70 Horizontal (Y-direction) 9.39 11.59 Vertical (Z-direction) 26.02 26.82 (Factors shown are based on 8 x 10 and 7 x 9 racks.) The analysis to determine the structural adequacy of the fuel storage module under tipping was conducted using the following technique: 1) Two loading conditions were applied to the SAP IV model, these are: a 1-G horizontal load placed in the direction the module tips, and a 1-G vertical downward load. 2) Using the principal of superposition, the vertical load is adjusted until the compression and tension in the
; feet which lift is reduced to zero, thereby creating a load state that approximates the module at the intstant the module lifts off.
4-37
The actual horizontal seismic load, at the point of lift off, is determined in a similar fashion as described above using a non-linear time history analysis. The 1-G horizontal and the adjusted 1-G vertical load can now be factored. This factor will be the seismic load due to the loaded module divided by the 1-G horizontalload of an empty module. 4.5.3 Computer Code Descriptions The computer codes used in these analyses are described in the following subsections. SAPIV SAP IV is a general structural analysis program for static and dynamic analysis of linear elastic complex structures. The finite element displacement method is used to solve for the displacements and compute the stresses of each element of the structure. The structure can be composed of unlimited number of three-dimensional truss, beam, plate, shell, solid, plane strain-plane stress, thick shell, spring, axisymmetric elements. The program can treat thermal and various forms of mechanical loading as well as internal element loadings. Dynamic analysis options consist of eigenvalue solutions yielding frequencies and mode shapes, response history by mode superposition, response history by direction integration, and response spectrum analysis. Earthquake type of loading as well as time varying pressure can be treated. The output consists of displacements at each nodal point as well as internal member forces for each element. The program being used at C-E is essentially equivalent to the version verified, documented, and released by the University of California (Reference 4-3). 4-38
CESHOCK The CESHOCK computer code performs transient, dynamic analyses of non-linear elastic systems. These systems can be either axial models having one degree-of-freedom per node or lateral ones having one rotational and one translational degree of freedom per node. The response of a system is determined by numerically integrating (using a Runge-Kutta-Gill technique) its equations of motion. Excitation can take the form of either initial conditions or time histories of applied accelerations, velocities displacements or forces. The non-linearities can consist of gaps, friction, hysteresis or non-linear springs. Hydrodynamic action can also be modeled, with both on-diagonal (added mass) and off-diagonal (coupling) terms being considered. The program automatically searches the response time histories and prints out the maximum and minimum values of all nodal accelera-tions, and member loads and can generate an optional output tape containing the complete response histories. CESHOCK is an extensively modified, proprietary version of the SHOCK computer code developed by V. K. Gabrielson and R. T. Reese of Sandia Laboratories (Reference 4-4). It differs from the original in the areas of damping, coefficient of restitution, friction, hydrodynamic effects, hysteresis, input of time histories, output options, allowable problem size and the manner of inputting stif fness elements. CESHOCK has been verified by demonstration that its solutions are substantially identical to those obtained by hand calculations or from accepted analytical results via an independent computer code (References 4-4 and 4-5). The CESHOCK code numerically integrates the equations of motion using a Runge-Kutta-Gill technique. The initial integration timestep, calculated by CESHOCK, is one-twentieth of the period of the highest individual mass-spring frequency in the model. The timestep is continually checked and adjusted by the code as a function of the 4-39
rate of change of the linear and angular accelerations. The timestep is held within the bounds of one-fifth times the initial timestep to two time the initial timestep. With this procedure for selecting the integration timestep, the CESHOCK numerical solution has been shown to be stable and convergent. This approach can determine the stress state of the module due to module tipping under seismic effects. This approach is only valid for lift off of a few mils. The results of the non-linear analysis indicates such a situation does exist. TYPICAL MULTIPLICATION FACTORS FOR SEISMIC EFFECT Horizontal 1-G Factor = 6.895 Vertical 1-G Factor = 20.82 (Factors shown are based on 7 x 9 rack.) CFCGEN The computer code CFCGEN develops a lumped mass stick model of a loaded consolidated fuel storage box from canister and fuel dimensions and material properties. The model can be used in the SHOCK code for non-linear time history seismic analysis. The features of the CFCGEN code include the capability to segment the fuel rod in as many as 10 segments with dif ferent densities, e.g. endcaps, plenum, pellets. The weight of each segment is automatically applied to the appropriate nodal locations. Up to twelve node locations can be used to model the canister. They can be input manually or can be generated internally to be equi-spaced. It is also possible to define the stiffness contribution of the fuel to the overall box stifiness. A fuel consolidation factor can be specified which will automatically adjust the amount of contained water based on box dimensions and fuel rods to obtain the correct weight distribution. Temperatures can also be specified which will adjust 4-40
the elastic modulus (E) of stainless steel and water density, accordingly: The code was developed by C-E and was verified by calculating the model parameters by hand for a typical loaded canister and comparing the results with CFCGEN. 4.5.4 Consolidated Fuel Storage Box Analyses Static and impact analyses were performed to verify the adequacy of the consolidated fuel storage box design. The results of the structural analyses demonstrate that the consolidation box and cover can be safely lif ted and transported using the cover as the liftpoint. The consolidation box is designed such that it will not be overstressed when subjected to a tensile load of 6000 lbs, which is assumed for analytical purposes as the maximum force that can be exerted by the spent fuel handling machine or the hoist assembly. The insert assembly, supporting the weight of the fuel rods, can withstand an impact of 5 g's. An actual free drop would probably exceed 5 g's. In the postulated worst case, dropping of a loaded consolidated fuel storage box might result in discharge of its contents and failure of all rods. This case would not result in an off site radiation exposure exceeding the limits of 10 CFR 100 or cause an ' increase in Keff above 0.95 and it therefore meets the design criteria. 4.6 MATERIALS, QUALITY CONTROL, AND SPECIAL CONSTRUCTION TECHNIQUES 4.6.1 Materials The following provides a list of materials used in the construction of the spent fuel racks.. 4-41
FUEL STOR AGE MODULE: A. Basic Module: ASTM-A240-304 Stainless Steel, max. carbon content,065%. B. Support Bars: ASTM-A276 or A-479-304 Stainless Steel, max. carbon content .065%. ADJUSTABLE FOO_T,: A. Block: ASTM-A182 max. carbon content
.065%.
B. Retaining Ring and Pad: ASTM-A240-304 Stainless Steel, max carbon content .065%. C. Screw: ASTM-A453 grade 660, high alloy steel. POISON BOX: A. Inner / Outer Shell, ASTM-A240-304 Stainless Steel, max. Funnel, Base and carbon content .065%. Braces: B. Insert: AMS 5697,in the quarter hard condition - Rockwell "C" -23, max. carbon content .065%. C. Poison: Boraflex - Bisco Products Inc. - Supplier. 4-42
CELL BLOCKING DEVICE: A. Pipe ASTM-A312 304 Stainless steel, max. carbon content .0655 B. Pin: ASTM - A479 304 Stainless steel, max. carbon content .065% C. Detent Pin: 300 Series Stainless steel. CONSOLIDATED FUEL STORAGE BOX: A. Cover Assembly: ASTM-A240 304 or A479 304 Box Assembly: Stainless Steel, Max. Floor Insert: Carbon Content .065% 4.6.2 Quality Control Northeast's and Combustion Engineering's Quality Assurance Programs ensure that all manufacturing and installation activities conform to acceptable quality requirements throughout all areas of performance. The pertinent requirements of 10CFR50, Appendix B, and Combustion Engineering Ouality Assurance Topical report CENPD-210-A, Rev. 3 and Specification 00000-WQC-5.2 will be followed, in addition, Northeast's Topical QA Report (1)(2) describes Quality Assurance requirements with which the design, procurement, and fabrication of the consolidated fuel storage boxes will comply. (1) 3. F. Opeka letter to T. E. Murley, " Proposed Revision 7 to the Northeast Utilities Quality Assurance Program Topical Report," dated June 5,1985. (2) S. D. Ebneter letter to 3. F. Opeka, "10CFR50.54(a)/50.55(f) Quality Assurance Program Description Review - Northeast Utilities Quality Assurance Program Topical Report (Revision 7)," dated August 9,1985. 4-43
4.7 POISON MATERIAL IN-SERVICE SURVEILLANCE PROGRAM A long term surveillance program will be implemented to ensure continued acceptable performance of the Boraflex neutron poison material used in the racks. This program will be based on current performance information for the Boraflex neutron poison material. Proper documentation will be obtained from the manufacturers of Boraflex and the racks to assure the quality of the neutron poison material and its proper loading in the racks. Visual inspection of the racks will be performed upon receipt of the racks to verify that the Boraflex is loaded in each of the specified locations in each rack. Uniquely identified samples, taken from material representative of that used as a neutron absorber, will be placed within surveillance capsules and placed within the Region I racks. At specified intervals the sampics will be removed and tested for boron content. Irradiation tests have been previously performed to determine the stability of Boraflex in boric acid solution. The results of these tests are documented in test reports of the Bisco Corporation (References 4-6,4-7, 4-8). From these tests, there is no evidence indicating any deterioration of the Boraflex material through a cumulative irradiation in excess of I x 1011 rads gamma affecting the suitability of Boraflex as a neutron poison material. Under the proposed surveillance program, calculations have shown that the specimens would require at least five years in the pool environment to approach this level of cumulative exposure. Direct dosimetry will be utilized, however, to establish an accurate record of cumulative exposure. Periodic testing and examination of poison specimens will take place beyond the point at which cumulative exposures exceed those of documented tests. The surveillance specimens will initially be examined after approximately five years of exposure in the pool environment. Several specimens will be checked for mechanical integrity and other general performance characteristics. This examination will include visual performance 4-44
characteristics. This examination will include visual inspection as well a other tests to verify the material stability. This initial surveillance will be used to verify that the performance of the Boraflex is consistent with current Bisco Corporation test results. Based on the results of the initial surveillance, and results from existing fuel rack surveillance programs at the Millstone Nuclear Station, additional testing will be scheduled to assure acceptable material performance throughout the life of the plant. ! 4-45
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Figure 4-7 CESHOCK Model of Millstone 2 Region II 7 X 9 Spent Fuel Rack Module
4.8 REFERENCES
4-1 NRC Guidance " Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 1978 and modified January 19, 1979. 4-2 Longo, R., and Baisley, D. F., " Seismic Analysis of Spent Fuel Racks" ANS paper TS-7308 presented at the ANS Topical Meeting on Options for Spent Fuel Storage at Savannah, Georgia, September 26-29, 1982. J 4-3 Bathe, K. J., Wilson, E. L., and Peterson, F. E., " SAP IV - A + Structural Analysis Program for Static and Dynamic Response of Linear Systems", Report No. EERC 73-11, Earthquake Engineering Research Cente r, University of California - Berkeley, June 1973. 4-4 SCL-DR-65-34, " SHOCK - A Computer Code for Solving Lumped Mass Dynamic Systems", V. K. Fabrielson, January,1966. 4-5 Topical Report on Dynamic Analysis of Reactor VesselInternals Under Loss-of-Coolant Accident Conditions with Application of . Analysis to CE 800 Mwe Class Reactors," Combustion Engineering, Inc., Report CENPD-42, August 1972 (Proprietary). 4-6 3. S. Anderson, "Boraflex Neutron Shielding Material - Product Performance Da ta," Brand Industries, Inc. Report 748-30-2 (August,1981). 4-7 3. S. Anderson, " Irradiation Study of Beraflex Neutron Shielding Materials, " Brand Industries, Inc., Report 748-10-1 (August 1981). 4-8 3. S. Anderson, "A Final Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials, " Brand Industries, Inc. Report 748-21-1 (August 1978). 4-55
Section 4 Appendix A Millstone Unit 2 Spent Fuel Pool Evaluation f t _ . . - . - . . _ . . - =
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(, ,n MARCH 25,1983 ENCLOSURE 1
MILLSTONE POINT - UNIT 2 SPENT FUEL STORAGE FACILITY FINITE ELEMENT MODEL NORTH SECTION SOUTH SECTION titutAAnt ELewtNTS 30u0 tithis11 W1W9 RANT EttMtNIS r 50uo (LEW1511 - - r g r; . .
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. _ - _ _ ..=__
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- Case Number Description l I g vertical accelerotion for dead weight of concrete 2 Hydrostatic f orces 3 1000 lb/f t2 vertical slon lood over Region i 4 2 1000 lb/f t vertical slab load over Region 2 5 Operating thermal (pool water at 150 F) 6 Accident thermal (pool water at 212 F) 7 I g ZPA north earthquoke. 2.34 g peak pool woll accelerotion plus hydrodynamic forces
(+X occeleration) 8 1 g ZPA west earthquake. 2.34 g peak pool wall acceleration plus hydrodynamic forces
, (+Y occeleration) i 9 -1000 lb/f t 2 hori~z ontal slob load over Region I in X direction (+X oc'celeration) 10 -1000 lb/f t 2horizontal. slob lood over Region 2 in X direction (+X occelerotion) ll -
-1000 lb/f t2horizontal slob load over Region 1 in Y direction (+Y oce,eleration) 12 -1000 lb/f t2horizontal slob load over Region 2 in Y direction (+Y occelerotion)
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Tobic 3.2-3 Nortin ust Utilities Service Com:my Millsfue Point Unit 2 Spent Fuel Pool Evoluotion Sumrnory of Lood Definitiori Porarreters item Description Pool Properties: Pool water Ueptn 38'-6" Pool Normal Operating Temperature 150 F Pool Accident Temperature 2120F Pool Hydrodynamic Forces TID 7024, App F Auxiliary Building Compartment Temperatures: All Comportments 55 F Thermal Stress - Free Temperature 55 F Operating Conditions: Fuel Transfer Canal Dry Coskioydown Area Dry Seismic Ground Accelerations: OBE Horizontal 0.091 OBE Vertical 0.06 g SSE Horizontal & Vertical 1.8 (OSE) na hY: W1
Toble 3.2-4 a Nortlast Utilities Service Conymy Millstone Pdint Unit 2 5 pent Fuell'ool Evoluotion Composite Lood Case Description Individual Load Case..N. umber: 1 2 3 4 5 6 7 8 9 10 li 12 Composite Load Case l D - Deod Load 1.00 1.00 .374 .607 2 L - Live Load 2.56 4.16 3 T - Operating Therr.iol 1.00 4 T - Accident Thermal I.00 5 E - OBE 0.06 0.06 0.57 0.92 0.09 0.09 1.99 3.23 1.99 3.13 6 E 2
- OBE 0.06 0.06 0.57 0.92 -0.09 0.09 -l.99 -3.23 1.99 3.73 7 E 3
- 00E 0.06 0.06 0.57 0.92 -0.09 -0.09 -1.99 -3.23 -l.?9 -3.23 8 Ef - OBE 0.06 0.06 0.57 0.92 0.09 -0.09 1.99 3.23 -l.99 -3.23 NOTES: 1) Four additional OBE cases are defined as -1.0 times E , through Eq , respectively.
- 2) SSE is taken as 1.8 times OBE.
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Tabla 3.2- 7 Northeast Utilities Service Company Millstone Poini Unit 2 Spent Fuel Pool Evoluotion Final Lood Coneination Coef ficients for Reinforced Concrete Composite Load Cases D L T Tg Eg E E E o 2 3 4 LOAD COMBINATION IDENTIFIER i.b. I l.40 1.70 I i.b.2.1 1.40 1.70 1.90 i.b.2.2 1.40 1.70 1.90 i.b.2.3 1.'i 0 1.70 1.90 i.b.2.4 1.40 1.70 1.90 i.b.2. 5 1.26 1.53 -1.90 i.b.2.6 1.26 1.53 -1.90 i.b.2.7 1.26 1.53 -l.90 i.b.2.8 1.26 1.53 -1.90 ) i.b.4 1.05 1.28 1.28 i.b.5.1 1.05 1.28 1.28 1.43 i.b.5.2 1.05 ! .28 1.28 1.43 i.b.5.3 1.05 1.28 1.28 1.43 i.b.5.4 1.05 1.28 1.28 1.43 i.b.5.5 0.95 1.15 1.28 -1.43 i.b.5.6 0.95 1.15 1.28 -1.43 i.b.5.7 0.95 1.15 1.28 -1.43 i.b.5.8 0.95 1.15 1.28 -1.43 ii.c l.00 1.00 1.00 ii.d. l 1.00 1.00 1.00 2.25 ii.d.2 1.00 1.00 1.00 2.25 ii.d.3 1.00 1.00 1.00 2.25 ii.d.4 1.00 1.00 1.00 2.25 ii.d.5 0.90 0.90 1.00 -2.25 ii.d.6 0.90 0.90 1.00 -2.25 ii.d.7 0.90 0.90 I.00 -2.25 ii.d.8 0.90 0.90 1.00 -2.25
T at>ie 3.2-9 Noriteost Utilities Service Compmy Millstore Point Unit 2 Spent Fuel Pool Evoluot;on
. Liner Plate Stor.dord Review Plan Lood Combination Summary Load Comoinction Number Description SERVICE LOAD COMBINATIONS - Ll!ER PLATE i.b. i 1.7D + 1.7L i.e.2 1.7D + 1.7L + 1.7E i.b..s 1.7D + 1.7L + 1.7W i.b.4 1.3 (D + L + To+R) o i.o.5 1.3 (D + L + E + T o + Ro) i.b.6 1.3 (D + L + W + T o + Ro)
FACTORED LOhD COlABl$ATIONS - LINER PLATE 11.$1 D + L + To + Ro+ E' ii.b.2 0 + L + To + R, + W, ii.b.3 D + L + To + Ro+ 1.5 P o li.b.4 D + L + To + R o+ 1.25 P o+ 1.0 (Y 4r. Y +j Ym) + 1.25 E , ii.b.5 0 + L + Ty + R o + 1.0 P o + 1.0 (Y
, r + Y; + mY ) ?E'
~.
O A
- wm
.! Y
Tottle 3.?-10 14ortiseust Utilitic. Service Cornpany Millstor.e Point Unit 2 Lpent Fuell'ool Evoluotion Applicable Lirer Plate Liardard i<eview Plan Lood Cornbirotions Lood Combination 14 umber Description SERVICE LOAD COMBINATIONS - LilER PLATE i.e.1 l.70 + 1.7L (Sounded by i.b.2) i.e.2 1.70 + 1.7 L + l.7E i.e.3 1.70 + 1.7L (Identical to i.b.1) 1.o.4 1.3 (D. + L + T,) i.b.5 1.310 + L + E + To) 1.b.6 1.3 (D + L + T;) (identical to i.b.4) FA ,TORED LOAD COMBINATIONS - LINER PLATE s . ii.b.I D + L + To + E' (Bounded by ii.b.5) ii.b.2 D+L+T o (Bounded by li.b.3) ii.b.3 D+L+T o , ii.b.4 0 + L + To+ 1.25E (Bouhded by ii.b.5) ii.b.5 D + L + To + E' O m N: M
Table 3.2-1 i Northeast Utilities Service Conomy Millstone Point Ulit 2 Spent Fuel Pool Evoluation . Final Lood Continction Coef ficients Service Composite Load Cases - Liner Plate O L T, T E, E E E o 2 3 4 LOAD COMBINATION IDENTIFIER I.b.2.1 1.70 1.70 1.70 i.b.2.2 1.70 1.70 i.b.2.3 1.70 1.70 1.70 1.b.2.4 1.70 1.70 1.70 i.b.2.5 1.70 1.53 1.53 -l.70 i.b.2.6 ' l.53 1.53 -l.70 1.b.2.7 1.53 1.53 1.b.2.8 1.53 .1,53
-1.70 i.b.4 ,
-1.70 1.30 1.30 1.30 i.b.5.1 1.30 1.30 1.30 1.30 i.b.5.2 '
1.30 1.30 1.30 f.b.5.3 1.30 1.30 1.30 1.30 i.b.5.4 1.30 1.30 1.30 1.30 1.b.5.5 - 1.30 1.17 1.17 f.30 -1.30 1.b.5.6 1.17 1.17 1.30 1.b.5.7 -1.30 1.17 1.17 1.30 i.b.5.8 -1.30 1.17 1.17 1.30
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Tabla 3.1-1 Northeast Utilities Service Cornpmy Millstone Point Ubil 2 Spent Fuel Pool Evoluotion Lhwr Plate Criterio Summory J Liner Plate AllowablesOI Liner Anchor Allowobles I2) Membrone Strains Lood Combinolions Without Thermal sc = .005 in/in st = .003 in/in , Non-Factored Load Combinations Fo - 0.5 F " Factored Load Combinations Fo = 0.85 F 9 Membrane Plus Bending Stroins t sc = 0.0111in/in st = 0.010 in/,in
, Lood Combinations with Thermal o = 0.5 u Fu md u are based on on ultimate displacement of 0.2 inches.
bblt I otes These allowables are consistent with those specified by ASME Section 11, Subsection CC for 1) f 17 containment liner plate when ultimale strength is lhe basis, i.e., factored load combinations. s 2) These allowables are consistent with AISC, Specification for Slect Structures, Pori 2; ASME Section ill Subsection CC for containment liner onchors md formulos from 11eferences 13 mdII.I ! \
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Table 4.1-1 Northeast Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultmt Moments Controlling Section See: ion (2) Section(3) 3,,,;,n Load Axiol Resultant Allowable Code Location Case Force Moment Moment Retio Pool North Wall
~
Horizontof Section (D+L T, I.25E3') 6.686 76.97 388.2 0.20 Lower Portion of Wall- East End Elements 444-445-446-447 . (MFPSTA1Al-058) Vertical Section (D'+L'+To -l.25E3') -22. !! 710.9 1325.0 0.54 Low er Portion of Wall Mid-Span Element 437 (MFPSTAIAl-05) Horizontal Sect:on (D+L+To+1.25E3') 1.794 44.35 545.8 0.08 Upper Portion of Wall-East End Elements 477-478-479-480 . (MFPSTAIAl-058) Vertical Section (D+L+To+1.25E3') 10.42 2 72.5 5 98.6 0.46 Upper Portion, Mid-Spm Elements 482-493-504-515 (MFPSTAIAl-05A) Pool South Wall Horizontcl Section (D'+L'+To -l.25E4') -30.32 810.I' 1367.0 0.59 Lower Portion, West End of Pool Element 685 (MFPSTAIAl-06) Units: Forces are in kips /in. Moments are in kip in/in. Notes: 1) Positive moment causes tension on outside surface of walls and lower surface of floor slob.
- 2) T moment:, are relieved, maintaining equilibrium and curvoture of section.
- 3) Ahowable moment is bcsed on strength design method per ACI 349/80.
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Table 4.1-1 Nor1heast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultant Moments (Continued) Controlling Section Section(2) SectionI3) Section
. Load Axio! Resultant Allowable Code Location Ccse Force Mcrnent Moment Ratio Pool South Wall (Continued) .
Vertical Section (D'+L'+ To - 1.25E4') -33.12 813.1 1516.0 0.54 Lower Portion, Mid-Spm , Element 668 (MFP5TAlAl-06) Horizontc! Section (D'+ L'. T g .25E4') l -23.27 685.6 l 142.0 0.60 Upper Portion, West End of Pool Element 707 (MFPSTAIAl-06) Vertical Section (D+L+To+1.25E4') I1.99 177.3 54 5.7 0.32 Upper Portion, Mid-Spon Elements 712-723-734-745 (MFPSTAIAl-06A) Pon! East Woil Horizontal Section (D'+L'+To -1.25E2') 7.807 109.3 339.1 0.32 Bottom of Wall Elements 577-578-579-580-581-582-583-5PA (MFPSTAIAl-078) , Vertical Section (D'+ L'+ T g .25E3') l -18.52 669.0 1332.0 0.50 Lower Portion of Wall - South End Element 578 l (MFPSTAIAl-07) Units: Forces are in kips /in. Moments are in kip in/in. Notes: 1) Positive moment causes tension on outside surface of walls and lower surface of floor slob.
- 2) Tomoments are relieved, maintcining equilibrium and curvature of section.
- 3) Allowable moment'is based on strength design method per ACI 349/80.
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Table 4.1-1 Northemt Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultcnt Moments (Continued) Controlling Section Section(2) 3,c,;,n(3) 3,c,;on load Axial Result ant Al'owable Code Location 'Ccse Force Moment Moment , Ratio Pool East Wall (Continued) - Horizonto! Section (D'+L'+Ta -l.25E3') -0.821 133.0 612.8 0.22 Upper Portion of Wc11 . Elements 609-610-6II-612-613-6I4-615-616 (MFPSTAIAl-078) Vertical Section (D'+L'+To -l.25E3') 7.527 19.77 65.6 0.03 Top of Wall - South End Elements 609-617-625-633 (MFPSTAIAl-07A) Fuel Trcnsfer Ccnol Separation Woll ' South (4 f t.) Portion of Wall (MFPSTAIAl-08) Horizontal Section (D'+L'+To -l.25E3') 15.30 58.27 60.56 0.96 Mid.Spcn (Element 844) Vertical Section (D'+L'+Ta -l.25E4') -15.82 366.4 74 9.0 0.49 South End of Well . Lower Portion (Element 829) Horizontal Section (D'+L'Ta -l.25E4') -18.12 345.6 640.0 0.5L South End of Woil ! Lower Portion (Element 829) . Units: Forces are in kips /in. Moments ore in kip in/in. Notes: 1) Positive moment causes tension on outside surface of walls and lower surface of floor slob.
- 2) Tomoments are relieved, maintaining equilibrium cnd curvoture of section.
- 3) Allowable moment is based on strength design method per ACI 349/80.
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Table 4.1-1 Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultont Moments (Continued) Controlling Section Section(2) Section(3) Section Load Axial Resultant Allowable Code Location Case Force Moment _ Moment Rctio Fuel Trmsfer Cmo! - Separation Wall (Continued) Vertical Section (D'+ L's 3-l.25E4') -8.915 268.1 6PA.5 0.39 Mid-Span (Element 843) North (3 f t.) Portion of Wcli (M.FPSTAIAl-08) Vertical Section Below (D'+L'+To-1.25E4') -23.64 363.6 581.5 0.63 Elevation of Bottom of Gate Opening (Elernent 823) Horizonto! Section Below (D'+L'+To -l.25E3') -14.47 304.6 5 91.9 0.51 Elevation of Bottom of Gate Oi>ening (Element 823) Vertical Section (D+L+Tc+ 1.25E4') -l1.11 196.1 473.9 0.41 Above Elevation of Bottom of Cote Opening (Element 839) Horizontal Section (D'+L'+To -l.25E4') -7.476 192.5 332.1 0.58 Above Elevation of Bottom of Cote Opening (Element 839) Units: Forces are in kips /in. Momen'ts ore in kip in/in. Not es: 1) Positive moment causes tension on outside surface of walls and lower surface of floor slob.
- 2) Tnmoments are relieved, maintaining equilibrium md curvature of section.
. 3) ATlowable moment is based on strength design method per ACI 349/80.
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Toble 4.1-1 Northeast Utilities Service Cornpmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultmt Moments (Continued) Col.trodloo ing S 3,;,;on Location kxiontionlecti@c(nt2) tesu Allowable Section(3) Code Case Force Moment Moment Ratio Cosk Loydown Areo West Separation Well , (MFPSTAIAl-10) Vertical Section (D"+L'+To -l.25E2') - 10.26 -134.7 -232.3 Below Elevation of 0.53 Bottom of Cote (Element 874) Horizontal Section (D'+L'+To -l.25El') -7.759 at Bottorn of Wall -91.34 -10A.4 0.50 (Element 872) Vertical Section (O'+L'+To -l.25E2') -5.537 -84.16 -351.2 , 0.24 Above Elevation of Bottom of Cote Opening (Element 880) Horizontal Section (D'+L'+To -l.25E2') -10.92 -91.31 -203.6 0.45 Above Elevation of Bottom of Cote Opening (Element 880) l i Units: Forces are in kips /in. l Moments are in kip in/in. Notes: 1) Positive surface moment of floor slob. causes tension on outside surface of walls and lower 2) l 3) Tomoments are relieved, maintaining equilibrium and curveture of section. ADowable moment is based on strength design method per ACI 349/80. i l i 1
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Table 4.1-1 Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultant Moments (Continuad) E Controlling (4) Section Section(2) Section(3) Section
~ Load Axial Resultant Allowable Code Location Case Force Moment Moment Ratio Cask Loydown Area South Separation Wall (MFPSTAI Al-10)
Vertical Section Below (D'+L'+To-l .25E2') -5.087 -104.0 -203.6 0.51 Elevation o' Bottom of Cote Ope.iing (Element 906, Horizontal Sectho (D'+L'+To-l .25E2') -7.573 -88.94 -183.2 0.49 of Bottom of Wall (Element 903) Vertical Section (D'+L'+Ta -l .25E2') 1,031 -118.2 -355.4 0.33 Above Elevation of Bottom of Gate Opening (Element 910) { Horizontal Section 'O'+L'+To -l .25E l') -9.703 -85.72 -196.6 0.44 Above Elevation of Bottom of Gate Opening (Element 910) Pool Floor Slob (MFPSTA5Al-09) North-South Section (D+L+Ta + l .04') -31.06 784.1 1303.0 0.60 Neor Center of Pool (Element 347) Units: Forces are in kips /in. Moments are in kip in/in. Notes: 1) Positive moment causes tension on outside surface of walls and lower surf oce of floor slob.
- 2) T moments are relieved, maintaining equilibrium and curvature of section.
- 3) AEowable moment is based on strength design method per ACI 349/80.
- 4) D' and L' indicate tho fcr earthquake' motion in the negative global direction (-E), the f actors on these composite loads are reduced 10 percent. E and E' indicate OBE and SSE, respectively.
m__. - . . - - Table 4.1-1 Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tobulction of Controlling Section Resultant Moments (Continued) ControllingI ") Section Section(2) Section(3) Section s Load Axict Resultant Allowable Code Location Case Force Moment Moment Ratio Pool Floor Slob (Continued) East-West Section (D+L+Tc+ 1.25E4') 22.28 108.4 130.6 0.83 at Cosk Loydown Area (Element 309) North-South Section (D+L+T a+ 1.25E l') 217.4 -31.16 -136.5 0.23 in Cask Loydown Area Elements 302-303-304 (MFPSTA5Al-098) Ecst-West Section (D+L+To+ 1.25E4') -1.726 225.2 793.9 0.28
'in Cask Loydown Area Elements 302-310-318-326 (MFPSTAS A l-09 A)
Foundation West Wall Beam Horizontal Section at (D+L+To+ 1.25E3') -1.283 -39.59 -237.6 0.17 South End of Beam Element 99 (MFPST AI Al-17) Units: Forces are in kips /in. Moments orein kip in/in. Notes: 1) Positive moment causes tension onoutside surface of walls and lower surf ace of floor stab.
- 2) T moments are relieved, maintaining equilibrium and curvoture of section.
a
- 3) Anowable moment is based on strength design method per ACI 349/80.
- 4) D' and L' indicate that for earthquake motion in the negative global direction (-E), the f actors on these composite foods are reduced 10 percent. E and E' indicate OBE and SSE, respectively.
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Table 4.1-1 Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Controlling Section Resultant Moments (Continued) Controlling Section Section(2) Section(3) Section lood Axial Resultent Alloweble Code Location Ccse Force Moment Mommt P.ctio Foundation West Wall Column - Horizont ol Section (D+L+To+ 1.25E4') -28.12 277.0 865.3 0.32 of Top of Column Element 102 (MFPSTAIAl-18) South Foundation Woil . Vertico! Section (D'+L'+To-1.25E2') -3.954 -102.5 -312.6 0.33 East Portion East End of Wall of Bottom Elements 1-2-3-4-5 (MFPSTAIAl-1IB-l) - Vertical Section (D'+L'+To -l.25E4') 10.22 54.0 54.92 0.98 West Portion West Erid of Well of Bottom Elements 10-11-12-13-14-15-16 (MFPSTAIAl-LIB) Inner West Foundation Wall Verticol Section (D'+L'+To- 1.25E2') -0.994 58.02 289.7 0.20 of Bctiorn Elemmts 165-166-167-168-169-170-171 (MFP.iTAIAl-ISB) Units: Forces ore in kips /in. Moments ore in kip in/in. Notes: 1) Positive moment causes tension on outside surfoce of walls and lower surface of floor slob.
- 2) T moments are relieved, maintaining equilibrium cnd curvoture of section.
- 3) Ailowoble moment is based on strength design method per ACI 349/80.
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4 Table 4.1-1 Northeast Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoulation Tabulation of Controlling Section Resultmt Moments (Continued) Controlling Section Section(2) Sec,;on(3) 3, ,;on Load Axiol Resultant Allowable Code Location Cese Force ' Mom ent Moment Rotio Inner South Foundation Woil Vertical Section (D'+ L'+ aT - I.25E4') 1.553 -89.81 -128.8 0.70 of Bottom Elements 193-194-195-196 - (MFPSTA!Al-138) 1 s Units: Forces ore in kips /in. Moments are in kip in/in. . Notes: 1) Positive moment causes tension on outside surface of walls and lower surface of floor slob.
- 2) T moments are relleved, maintoining equilibrium md curvature of section.
- 3) Ailowable moment is based on strength design method per ACI 349/80.
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Table 4.1-2 Northea:t Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Resultmt Trmsverse Shear Forces Controlling Allowcble(3) Code Load Section(2) Section Shear Location Case Shear Shear Ratio Pool Nor1h Wo!! Vertical Section (D+L+T o+ 1.25E3') 3.062 6.'377 0.48 of West End of Wall Elements 443-454-465-476-487-498-509 - Vertical Section (D+L+To+1.25E3') 8.881 27.77 0.32 of West.End of Wall at Top Element 520 Vertical Section (D'+L'+To -l.25E4') 14.50 28.93 0.50 of Intersection with Cask Loydown Areo West Wall at Top , Element 512 Vertical Section (D'+ L'+ To- 1.25E4') Il.27 31.21 0.36 at Intersection with Cask Loydown Area West Wall Elements 435-446-457-468-479-501 Horizontal Section (D+ L+ T,+ 1. 25E3') 1.805 6.167 0.29 of Bottom of Wall Elements 433-434-435-436-437-438-439440-44 l-442-443 Units: Kips / inch Notes: 1) Data from MFPSTA1Al-04
- 2) Shear forces are linearly interpolated to the distance from the face of the effective support equal to the distance from the section compressive face to the centroid of the tensile steel where oppliccble.
- 3) Allowable sheer is based on strength design per ACI 349/80.
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Table 4.1-2 Northeast Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Resultmt Transverse Shez Forces (Continued) Controlling Allowable (3) Code Load Section(2) Section Shecr Location Cese Sheer Sh-or Ratio Pool South Woil Vertical Section at (D+L+T o+ 1.25E4') 10.18 25.89 0.39 West End at Top of Wall Element 740 - S'ertical Section at (D+L+T o+ 1.25E4') 1.087 6.234 0.17 West End of Wall Elements 663-674-685-696-707-718-729 Horizontof Section at (D'+L'+To -l.25E4') 5.397 7.827 0.69 Top of Wall Elements 740-741-742-743-744-745-746-747- - 748-749-750 Pool East Woil Vertical Section at (D+L+T o+ 1.25E3') 3.876 25.88 0.15 South End of Wall of Top Element 633 Vertical Section at (D+L+To+ 1.25E3') - 3.018 6.362 0.47 South End of Woll Elements 577-585-593-601-609-617-625 l Units: Kips / inch Not es: 1) Data from MFPSTAIAl-04
- 2) Shear forces are linectly interpolated to the distance from the foce of the effective support equal to the distance from the section compressive face to the centroid of the tensile steel where opplicable.
- 3) Allowable shear is bcsed on strength design per ACI 349/80.
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Table 4.1-2 Northeast Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion ToWlation of Resultmt Trmsverse Shear Forces (Continued) Controlling Allowable (3) Lood Code Section(2) Section Location - Case Shear Shear Shaar Ratio Pool East Woil (Continued) Vertical Section (D+ L+ To + l.25E2') 8.720 26.26 0.33 of Intersection with Cask Loydcun Area South Wol.~ or Top Element 637 Vertical Section (D+L+T a+ 1.25E2') 14.55 of Intersection with 31.18 0.47 Cask Loydown Area South Wall Elements SM-589-597-605-613-621-629 Horizontal Section (D'+L'+T o+ 1.25E2') 5.573 at Top of Wall 5.922 0.94 Elements 625-626-627-628-629-630-631-632 Fuel Trmsfer Canal Seporation Well Vertical Section at (D+L+T o+ 1.25E3') I l.73 South End of Wall 19.05 0.62 (4 f t. portion) at Top Element 870 Units: Kips / inch Notes: 1) Doto from MFP5TAlAl-04 2) Sheer forces are linearly interpolated to the distance from the foce of the effective support equal to the distance from the section compressive face to the centroid of the tensile steel where applicable. 3) Allowable shear is based on strength design per ACI 349/80. f \~ .uus
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Table 4.1-2 t4ortheast Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Resultmt Trcnsverse Shear Forces (Continued) Controlling Allowable (3) Code Location Load Section(2) Case Section Sheer Shaar Shear Ratio Fuel Transfer Cmol Seporation Wall (Continued)
~
Vertical Section at South End of Woil (D+L+T a+ 1.25E3') 1.84 9 4.837 0.38 (4 f t. portion) Elements 814-822-830-838-846-854-862 Horizontal Section of MidHeight of (D+L+To+1.25E3') 4.130 4.346 0.95 South (4 f t.) Portion Elements 833-834-835-836-837-838 Vertico! Section . Below Gate Opening (D+L+To+1.25E3') 0.718 3.307 0.22 North (3 f t.) Portion Elements 808-816-824 Horizontal Section of Bottom of Wall (D+L+T +1.25E3') 2.910 4.041 0.72 Elements 807-808-809-810-811-812-813-814 Units: Kips / inch Notes: 1) Data from MFP5TAlAl-04 2) Shear forces are linearly interpolated to the distence from the face of t elfective support equal to the distance from the section compressive fece
- 3) to the centroid of the tensile steel where opplicable.
Alloweble shear is based on strength design per ACI 349/80. 4
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Toble 4.1-2 Northeast Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tobulation of Resultmt Trmsverse Shear Forces (Continued) Controlling Allowable (3) Code Load Section(2) Section Location Sheer Ccse Shear Shear Rotio Cosk Loydown Area South Seporofion Woil - Vertical Section C+L+T + o 1.25E4') 2.533 3.325 0.76 of Intersection with
- Pool East Wall Elements 903-905-907-909-91 l-913-91 S-917
- Horizontal Section O'+ L'+ To -l.25E3') 1.594(4) 2.084 0.76 at Bottorn of Wall Elements 903-904 Cask Loydown Areo West Seporotion Wall ,
Vertical Section O'+ L'+ To- l.25E2') 1.887 4.079 of Intersection with 0.46 Cask Ldydown Area South Wall Elements 873-876-879-882-885-888-891-894 Horizontal Section O'+L'+ To- I.25E l') 1.691 1.943 0.87 of Bottorn of Well Elements 871-872-873 Units: Kips / inch Notes: 1) Data from MFPSTAIAl.04
- 2) Sheer forces are linearly interpolated to the distance from the face of the ef fective support equal to the distance from the section compressive face to the centroid of the tensile steel where applicable.
- 3) Allowable shear is based on strength design per ACI 349/80.
- 4) Transverse sheer odjusted based upon crocked section equilibrium moment gradient.
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Toble 4.1-2 Northeast Utilities Service Compony Millstone Point Unit 2 Spent Fuel Pool Evoluotion . Tobulation of Resultant Transverse Shear Forces (Continued) Controlling (4) Allowable (3) Code Load Section(2) Section Shear __ Location Case Shear Shear Rotio Pool Floor Slob (II Ecst-West Section (D'+L'+To-l .25E l') 3.851 6.813 0.57 at Mid-Span Elements 301-309-317-325-333-341-349 357-365-373-381 North-South Section (D+L+T o+ 1.25E I') 12.27 14.15 0.87 Beneath Cosk Loydown Area West Separation Wall Elements 313-314-315-316-317-318-319-320 (D+L+T a+ 1.25E l') 1.698 6.454 0.26 North-South Section of Mid-Spon Elements 321-322-323-324-325-326-327-328 Foundation South Walt (D'+L'+To-l .25E l') 2.141 7.581 0.28 West Portion Horizontal Section at Top Elements 58-59-60-61-62-63-64 (D+L+T o+ 1.25E l') 2.446 7.064 0.35 East Portion Horizontal Section at 1op Elements 49-50-51-52 54-55-56-57 Units: Kips / inch Notes: 1) Data from MFPSTA5Al-04
- 2) Shear forces are linearly interpolated to the distance from the face of the eff ective support equal to the distance from the section compressive f ace to the centroid of the tensile steel where opplicable.
- 3) Allowoble shear is based on strength design per ACI 349/80.
- 4) D' and L' indicate that for earthquake motion in the negative global direction (-E), the f actors on these composite loads are reduced 10 per cent. E and E' indicate OBE and SSE, respectively.
Table 4.1-2 Northecst Utilities Service Compmy Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Resultont Trmsverse Shecr Forces (Continued) Controlling Allowoble(3) Lood Code Location Section(2) Section Shear Ccse Sheer SNor Rotio Foundation East Woil Horizontal Section of Top (D+L+ Ta+ 1.25El') 2.976 6.94 9 0.43 Elements 238-239-240-241-242-243-244 - Founchtion Imer South Wall Horizontal Section (D'+ L'+ To- 1.25E4') 1.848 of Bottorn 3.316 0.56 Elernents 193-194-195-196-197-198 Foundation Inner West Wall Horizontal Section (D'+L'+T o- 1.25E3') of Bottorn 1.848 2.920 0.83 Eternents 165-166-167-168-169-170,171 Foundation North Woil Horizontal Section (D+L+T a+ 1.25E2') 5.803 of Bottom 10.46 0.55 Elements 109-il0-Ill-112-113-114 Foundation West Wall tJorth Portion (D+L+T o+ 1.25E4') 3.001 11.79 0.25 Horizonto! Section at Bottorn Eternents 77-78 - Units: Kips / inch Notes: 1) Data from MFPSTAIAl-04 2) Shear forces are linearly interpolated to the distance from the fece of the effective support equal to the distonce frorn the section compressive face to the centroid of the tensile steel where opplicable. 3) Allowable sheer is bcsed on strength design per ACI 349/80. f \~ wueva L-N omo
;/ remmat
Table 4.1-2 Ncrthemt Utilities Service Cornpany MillstonePoint Unit 2 Spent Fuel Pool Evaluation Tabulation of Resultcnt Transverse Shecr Forces (Continued) Controlling Alloweble(3) Load Code Location Section(2) Section Shear Ccse Shear Sh-o- Ratio Foundation West Woll South Portion (D+L+ To+ l.25E3') 6.140 12.91 Horizontal Section 0.48 at Bottom . Elements 83-84-85 Notes: 1) Data from MFPSTAIAl-04 2) Sheer forces are linectly interpolated to the distance from the foce of the elfective support equal to the distance from the section compressive face to the centroid of the tensile steel where opplicable. 3) Allowable sheer is based on strength design per ACI 349/80. suus J\" . . or==
.i /. m
Table 4.1-3 Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tabulation of Resultmt fr@lme Shear Forces Controlling Allowable Code Lood Section(II Section Shear Location Ccse Shear Shear Ratio Pool North Woil Horizontal Section (D'+ L'+ To- 1.25E3') 0.774 25.4 0.03 of Top of Wall Elements 510-511-512-513-514-515-516-517-518-519-520 . Pool South Wall Horizontal Section (D+L.To. l.25E3') 3.032 25.4 0.12 of Bottom of Wall Elements 663-664-665-666-667-668-669-670-671-672-673 Pool East Wall ' Horizonto! Section (D+L+Ta+1.25E2') 9.206 26.58 0.35 of Bottorn of Wall Elements 577-578-579-580-581-582-583-584 Fuel Transfer Cono! Seporation Wall South (4 f t.) Portion (D+L+Ta+1.25E3') 8.670 24.79 0.35 Horizontof Section of Bottom of Wall Elements 817-818-819-820-821-822 Units: Kips / inch Notes: 1) Allowable shear is based on strength design per ACI 349/80.
/ (~ katutt.sm
'4 W
)/ rewmst
Table 4.1-3 Northeast Utilities Service Cornpony Millstone Point Unit 2 Spent Fuel Pool Evoluotion Tobulation of Resultant Ir>-Plane Shear Forces (Continued) Controlling (2) Allowable Code Load Section(II Section Shear o Location Case Shear Shear Ratio Fuel Transfer Canal Separation Wall (Continued) Q Horizontal Section at (D+L+T a+ 1.25E3') 14.29 23.90 0.60 Bottom of North (3 f t.) Portion Elements 807-808 Cask Loydown Area South Separation Wall Horizontal Section in (D+L+T a+ 1.25E2') 5.566 30.35 0.18 Upper Portion of Wall Elements 913-914 Cask Loydown Area West Sepwation Wall Horizontal Section at (D+L+To -l.25E3') 6.770 12.80 0.53 Bottom of Wall Elements 871-872-873 Pool Floor Slob North-South Section (D+L+T a+ l .25E l') 14.43 15.87 0.9i Near East End of Pool Elements 313-314-315-316-317-318-319-320 l Units: Kips / inch Notes: 1) Allowable shear is based on strength design per ACI 349/80.
- 2) D' and L' indicate that for earthquake motion in the negative global direction (-E), the factors on these composite loads are reduced 10 l percent. E and E' indicate OBE and SSE, respectively.
Table 4.1-S Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Pool Floor Liner Plate Analysis Summary Controlling Non-Thermal Lood Combination 1.7 (D + L + E2) i.b.2.2 Allowable S train Strain (in/in)x10-3 (in/in)x 10-3 Ratio Element cs to cs /to Membrane Strains Tensile 90 0.200 3.0 0.07 Compressive 140 -0.056 -5.0 0.01 Membrane plus Bending Strains Tensile 84 0.437 10.0 0.04 Weld Allowable Stress Stress (ksi) Ratio Cs os /3a Node (s) Co Weld Stress 105 2.67 17.8 0.15 Data from MFPSTA4Al-12 Controlling Thermal Lood Combination (D + L + To + El') ii.b.S.! Allowable Strain Strain i (in/in)xl0-3 (in/in)x10-3 Ratio Element cs to c/co s Membrane Stroins Compressive 31 -l.59 -5.0 0.32 Membrane plus Bending Strains Compressive 3 -6.54 -14.0 0.47 Weld Stress Allowable (ksi) Stress (ksi) Ratio Node (s) Cs Co Cs/Co Weld Stress 1-6 by I 55.6 53.4 1.04 Data from MFPSTA4BI-12B L
Table 4.1-5 Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Pool Floor Liner Plate Analysis Summary (Continued) Controlling Non-Thermal Lood Combination 1.7 (D + L + E2) i.b.2.2 Actual Allowable O/O u Node Load Ratio Load Ratio (Anchor Location) O/O u O/0 u0 O/Q uo 204 0.88 0.85 1.0 Data from MFPSTA4Al-09 Controlling Thermal Lood Combination (D + L + To + El') i.b.5.1 Allowable Node Displacement Displacement 6 /6 u (Anchor Location) 6 (inches) 6u (inches) (Ratio) 39 0.062 0.10 0.63 i Seom Embedded Angle Shear Allowable F3/F3a Stress-F S tress - F* Node-DOF (ksi) s (ksi) (Ratio) 53-64 15.6 16.5 0.95 Data f rom MFPSTA4BI-128
Table 4.1-5 ' Northcost Utilities Service Cosmmy Millstone Point Wit 2 Spent Fuel l'ool Evoluution Wall Liner Plate Strains Mem6rane Tensile Strains Nominoi hilowable Location - Description ' S train Strain (Analysis identifier) Load (in/in (in/in flotio Combination x 10'] x 10'] E3/lio I4n th & South Wolls Element 510 - X Section (MFl'STA I A2-1 l) (D'4 L'+ T -1.25E4') 1.1IU 3.0 0.37 North Wall at Top ll.lf.5.8 East Wall Element 601 - X Section 1.7(D'+L' E2) 'O.t:30 (MFI'STA I A2-12) Mid-liight of Wall 3.0 0.15 1.B.2.2
/
Fuel Tronsfer Canal Wall Element 863 - X Section 3 Fool Por tion 1.7(D+ L4 E4) 0.820 3.0 ut Top of Wall I.O.2.In 0.27 (MFI'STA l A2-13) Fuel Trmster Cunal Wull Element utsti - Y Section 4 Foot Portion Mitl-I-leight of Wall (D'+ L'+ T -1.2SE4') 0.69's 3.0 0.23 (MFI'S I A I A2-13) ll.LU.S.8 Cosk Loyttown Area South Wull Element 871 - Y Section 1.7(D e L eE2) 0.197 West Separolion Wall 3.0 0.07 (MFI'STA I A 2-IIs) I.B.2.2 atBottom r 7, . (9W
Toble 4.1-5 (Continued) ' Norlheast Utililies Service Conymy . Millslonc l'oint L.Asil 2 Spent Fuel Pool Evoluution Wall Liner Plate Strains Membrine Compressive Strains Nominul Allowable Strain Strain Location - Description Load ' (Analysis identifier) (in/in (in/i Combination x10-] x 10g j s a North & South Wolls Element 668 - X Section (D'+ L'+T -1.25E4') -0.623 -5.0 0.12 (MFI'S T A l A2-1 l) South Woll at Dollom II.li.5.0 East Wall Element 612 - Y Section (D'+L'+T -l.25E3') -0.597 -5.0 0.12 (MFl'ST A I A2-12) Mid-Span of Wall l1.1f.5.7 Fuel Trmster Conal Wall Element 823- X Section (D'+ L'+ Tg -l .25E4') -0.949 3 Foot liiick Por tion -5.0 0.19 Mid-l leight of Wall li.u.5.0 (MFI'S I A I A 2-13) Fuel Trmsfer Cunal Wall Element U22 - X Section 1.25E4') -0.501 4 Foot Thick Portion -5.0 0.12 South End at Doliom (D'+ II.dL'+ T '.5.8 (MFPSI A I A2-13) Cusk Loydown Arco Walls Element 078 - X Section (D'+ L'+ T '- 1.25E 2') -0.911 -5.0 0.18 (MFI'S I A I A2-14) West Separation Wall ll.d.5.6 Uelow Gore
- k. s v 2, .
;? sG
_ -- - . - - - . .. . _ - _ - - - . . - _ . - . -. =- .-~. _ . ~ . . Table 4.1-6 (Continued) Northeast Utilities Service Company Millstone Point Unit 2 Spent Fuel Pool Evoluotion Wall Liner Plate 5 trains Membrane + Dending Tensile Stroins Membrane Nominal + Bending Allowable Strain Strain Strain Location - Description Load c (in/in (3(in/iq c 3(in/irg Ratio j (Analysis identifier) Combination x10'] x 10' x 10- (34 o l t North and South Wolls Element S12 - X Section (D'+ L' To-l.2SE4') 1.11I 4.444 10.0 0.44 (MFPST AI A2-1 l) North Wall at Top ii.b.S.8 East Wall Element 601 - X Section (D'+ L'+ To -l .2SE3') 0.438 1.751 10.0 0.18 (MFPST AI A2-12) North Wall Adjacent ii.b.S.7 CLA South Wall Fuel Transfer Canal Wall Element 863 - X Section 1.7(D + L + E4) 0.820 3.280 10.0 0.33 3 Foot Thick Portion Top of Wall i.b.2.4 (MFPSTAI A2-13) Fuel Transfer Canal Wall Element 870 - X Section (D'+ L'+ T - 1.2SE4') 0.571 2.284 10.0 0.23 4 Foot Thick Portion Top of South End of Wall ii.b.S.8 (MFPST AIA2-13) Cask Loydown Arco Wolls Element 871 - Y Section 1.7(D + L + E2) 0.197 0.788 10.0 0.08 (MFPST AI A2-14) West Seporation Wall i.b.2.2 of Bottom t
Toble 4.1-S (Continued) - NorIheosi Utilities Service Compmy - Millstone Point Unit 2 Spen! Fuel Pool Evoluution Wall Liner Plate Strains I-tembrane + Bending Compressive Strains Membrone Nominal Dending Allowable Strain Strain Stroin Location - Description Load (in/in (in/in (in/in ', I (Analysis identifier) Combination x10'] x 10'] x 10'] s "o Marth and South Walls North Wall - Element 443 (D'+ L' + T - 1.2 5E4') -0.544 -2.176 -14.0 0.16 (MFPST AI A2-1 I) Y Section, Dot som at ll.E5.0 . West End of Wall Eust Wall Element 500 - Y Section (D'+ L' . Tg -l .25E3') -0.561 -2.245 -14.0 0.16 (MFPSI Al A2-13) Oot som of Wall at II.u.5.7 Mid-Spun Fuel Trmster Canal Wall Element 023 - X Section (D'+ L'+T -1.25E4') -0.949 -3.796 -14.0 0.27 3 Fooi Thick Portion Mid-Heiglit of Wall ll.ES.O (MFPST A l A2-13) Fuel Trmster Canal Wall Elernent 822 - X Section (D'+ L' + T - 1.2 SE 4') -0.587 -2.348 -14.0 0.17 4 Foot Thick Portion South End at Dot tom II.E5.0 (MFPSI Al A2-13) Cask Loydown Arco Wolls Element 877 - X Section (D'4 L'+Tg -l.2Sli2') -0.762 . -3.050 -14.0 0.72 (MFPS I A I A2-14) west Sepunction Wall 11.13. 5 . 6 Oclow Gute E. G s
r Section 4 Appendix B Reference 4-2 i l I I i i l l I i i
i L.__.-.....-- l l SEISMIC ANALYSIS OF SPENT FUEL RACKS I i R. LONGO !. D. F. BAISLEY l !
' l i
i I, Nuclear Power Systems f Combustion Engineering, Inc. 1 Windsor. Connecticut , f. f
\
\
1
- l. ;
Presented at ! AMERICAN NUCLEAR SOCIET ' f l l TOPICAL MEETING ON t l j OPTIONS FOR SPENT FUEL STORAGE ' ' September 26 29,1982 - l Savannah, Georgia f POWER j SYSTEMS ! j
'- _ - - ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l ABSTRACT The paper describes the nonlinear time hostory sessmic analy sss method used by C Efor the design and hcenung of spent fuel racks. The method is apphed to spentfuel racks that store both standard and conschdatedfuel assembhes. The analysis us based upon a Jarect numerical sntegratwn of the coupled equatwns of morson for the fuel and the rack. The equatwns ofmotion accountfor the gaps. hyJrod;namte couphng and smpactsng between the structures of the fuel andfuel rock system. A n nmary of representatuse results from nonhnear tome hssto,y analyses coverung a w Je range of designs and sessmic e.scitations is presented. A compari. son of these results with those obtained through the use of the response spectrum analysis method is presented to dem. onstrate that the response spectrum method-which is un. able to accountfor interaction effects-may lead to incorrect results. The importance of modeling the fuel as a separate structural element is established. Esamples of how thefuel responds to seismic e.scitation as its own naturalfrequen. cies-not at that of the rock structure-are presented. The applicabihty of the seismic analysis method to a consoli. datedfuel andfuel rack design is descussed. 1 i Additional copies of this technical pacer may be obtained
~
by wrotong Communications. Drot. 1021 1904, mnasor. Please refer to the number (TIS 7308) that appears on the lower right corner of the front cover.
SEISMIC ANALYSIS OF SPENT FUEL RACKS INTRODUCTION riteracting submerged structures are in close proximity C E led the industry in performing nonhnear time history (small gaps). seismic analyses of spent fuel racks in 1975. Since then. The nonlinear time history method was deseloped by C E has applied the methodology to nine spent fuel rack C E for use in spent fuel rack analyses because the linear applications covering a wide range of designs and reactor response spectrum method does not prope:ly characterize sites. This experience is supplemented with many parameter the fuel-to fuel rack to. pool interaction and, as demon-studies using the nonlinear time history method, strated later in this paper, it may yield incorrect results. The nonlinear time. history analysis method employed by C Eis based upon a direct numericalintegration of the equa- THEORY tions of motion for the fuel and the rack. It utilues multi. To aid in understanding the analysis method requirements degree of freedom spring and lumped mass models of the corresponding to the physical problem, consider the follow-fuel and the rack, and accounts for the effects of gaps and ing simplified analog of the spent fuel rack problem (see submergence m water directly in the equations of motion Fi ure 2). The three concentric cylinders represent the pool defined by the model. It uses the seismic excitation time- (P), the rack (R), and the fuel (F). There is water between history corresponding to the spent fuel pool elevation in the the fuel and the rack, and between the rack and the pool, auxiliary building. Figure I provides an example of a typical The connection (spring Kc) between the fuel and the rack
" represents the gap between these structures as well as the 103-" Impact stiffness with which the fuel spacer grids interact
$8 02 l
' I I
P j 02
# 03-, '
- R 0 ' ' ' ' b ' ' ' 10 15 20 25 ,
Time in Seconds Figure I: E2 ample of Sessmut Excitation Tome litstory 1940 El Centro Earthquale
- ph seismic excitation used for nonhnear time hisinry analysis-the acceleration time history for the 1940 El Centro carth- h quake. The response of the fuel and rack, together with the seismic loads, is obtained directly from the analysis. The analysis is performed by means of the computer program ,
CESHOCK. )' KR To allow insertion and withdrawal of fuel, each spent fuel rack cell has a gap between the cell walls and the fuel. Dur- _ ing seismic excitation, the fuel moves freely through the p available gap and impacts the cell walls. The fuel responds ; - to excitation at its own natural frequencies-not at that of g the rack structure-since it is a separate structure and not attached to the rack. As the fuel moves within the rack and 6 p. 6R as the rack moves relative to the pool, the water between these structures is moved by them. The acceleration of the water introduces hydrauhc loads on the structures which re-sults in a lowtring of natural frequencies of fuel and rack. Figure 2: Somplofied Analog ofSpent Fuel Rack Phyical These hydrodynamic effects are accentuated when the Problem I
I with the f.sck when in contact. The connection esprmg K.)
+
between the r.ack and the pool represents the manner in s a which the rack is supported by the pool. Nomenclature is as follow s:
,,4 ]4 ]2 2 a 4
l , 4
, =,* ,
t , . , e R, = sentnic escitation sacceleration time history) , '1
'r ll t .
at spent f uel pool elevation i
'f ' Cod 34pon 5 = acceleration of rack trelative to pooll 'l k=hl . 7- l p' #)/
l1
.i . ,,
X, = acceleration of fuel trelative to pool) I i ,
,, g 6, = dnplaecment of rack (relatise to pool) l l.i .
6, = displacement of fuel trelative to pool) j l :! ! . l i i Fuel Storage l hql ; ! , ; Tubes M. = mass of rack l l j, 6
- l ; . :
- M., = mass of water displaced by rack ,
{ f l f ! My = mass of water contamed within rack j j l
# V Channel M, = mass of fuel ,, Base Suppon M,, = mass of water displaced by fuel F.. = nuid force on inner boundary of rack Figure 3: C E Hl. CAP Spent Fuel Rack Module F.,o = nuid force on outer boundary of rack through the use of CESHOCK. In contrast to the above, the F,,o = nuid force on outer boundary of fuel response spectrum method can accommodate only a single K. , K,a = as defined above uncoupled equation for the response of a one degree-of free-a,.ai,0.y = factors describing the effect of geometric p m system. Modifying the response spectrum method to include an approximation of the effect of water on frequency.
proximity of hydrodynamics the analogous equation of motion for the system of Figure 3 that corresponds to the response spectrum method of With reference to the above nomenclature a .. . igure 2,
""*I I SIS and neglecting damping terms for purposes of simplifying
. discussion, the following equations of motion can be ,
(M + Mc + Mo)E + K & = -1M + Me) R, developed: Here the representation of the system is clearly incom. M.t R, + E.) = - K (6.) + Kc(6, - 8.) + F o
- Pl ete, with all sorts of approsimations (of unknown effect)
F., required to select the single salun of mass, stiffness Ilinear M,t R, + F,) = - KJ 6, - Sc) + F,.o - d), etc.. all wed. Comparison with the two equations above demonstrates the point that the response spectrum The Guid forces are given by: method does not model the real, physical situation, for u. F.,o = M.. (X, - a. E.) ample. it does not account for the gap between the fuel and the rack, wnich causes the system to have different natural F.* = M., ( - X, + 2 3, - a:I.) frequencies (and to respond to different frequencies of ex. F,,,, = Mr. (R, + 2 3. - 0 %,)3 citation) and allows fuel to rack impacting to occur. Also, it does not account for the hydrodynamic coupling between Substitution of these expressions for Huid f6rces into the the fuel and rack, with the introduction of micractive fluid two equations of motion and simplification of terms yields forces. the required coupled equations corresponding to the physical problem: RESULTS (M.
- a.M., + a M., sE. - (20M.,13, + (K. + A number of spent fuel rack seismic analyses have been Kc,)&. - Kc6, = - ( M. + M., - M.,, IR, performed by C E covering a wide range of rack designs and seismic excitations. The two basie types of spent fuel
-( 27M ,,)%. + I M, + a:M,,,13, - Kua . + K. 8, = - .
_ , y, _ y y' racks of fered by C E are shown in Figures 3 and t. The High Capacity (HI CAPIdesign in Figure 3 is composed of square The equations account for the gap between the luel and storage casilies fabricated from stamless steel plate with the rack, the hydrodynamic couplmg between the sub. each cavity capable of accepting one fuel anembly. The merged structures and impacting between structures. The storage cavities are structurally connected to form modules complete equations of motion (includmg damping) corre. from the use of channels, plates and chevron beams whwh sponding to the physical situation are modeled and sched provide the load. carrying frame and mamtain spacing be.
10 0 ~ a
/- j g g#
. Site tv N
10-
't, h cio ,,n9 /
,, ',;;;/ ~
!f 8!,' , ,,
4 S*' 8 / / ' Site til g*4 g 8
/ f30k k , Q.
,l,fj f
]f 'S'te i Site vi l /, g\
l;' e -
' ~ Site
\ -i Site vil v
-- . ~._ s,te 11 i '
g 01" 10 100 1 l , l l 1 Frequency (CPS) l Figure 3: Spen FuelPools Seismic Response Spectra l 8 l , ;, nonhnear springs Ko through Kc.; the frictional restraint y 6 l-%9 between the fuel and the rack and that between the rack and
', j I r ' the pool are represented by the friction couplings F, and
,f - F. ,. respectively. The corresponding parameters for Model I
,e '
Flow B are shown in Figure 7.
"""*2
Figure 8 is a brief segment of typical displacement re. Figure 4: C E Super Hi CAP Spent Fuel Storage Module 'P***** ( M*d'I ^) '* 'h' "I'*I* * ** 5'I** **"P**di"E H-tween storage casities. The C E standard Super Hi CAP A' KG6 ' 14 7 spent fuel storage rack shown in Figure 4 is a stainless steel 7M -H-- / monolithic honeycomb structure with square fuel storage Kas, H Km6 / Spacer Gnd locations. The fuel assembly storage cells are welded to- 3
/
gether to permit the assembled modules to be load beating structures as we!! as the storage ce!! enclosures. Each indi. Fuel KFs " / Mocet# H / . Pool 1 vidual cell is a structural member and serves as a guide and retainer for a Neutron Poison Insert or a Consohdated Fuel Shgg; - >2-H- / , Box. Fo!!owing is a summary of representatise results fium / #
, Kr' , g , ' Kna nonlinear time history analyses (utihzing CESilOCK). Beam ' KG3 1 /
i compared with corresponding response spectrum method 4( -H--
/
analysis results. "3 l Kr3 ~ ,d=-Hycrodynamic Figure 5 shows several different seismic escitations used -
- p. 10
/ Coupling in obtaining the results. The response spectra are show n only -H-- Elements to illustrate the differences in the escitations corresponding Rigid g,, ,H g n, 7
to seven sites; time histories for these sites were used in the Mass , / CESilOCK analyses. 2( }9--H-- / Figures 6 and 7 represent awo Iypical CESitOCK models. g,, Kni / Model A corresponds to a freestanding Ill CAP design and ,8 / Model B represents a freestanding Super lil CAP design. 1(.g/
'p ,,, [Fnetion in si,e,ngElement For Model A. the fuelis modeled by rnasses I through 7 and 7 Analysts. Non.
springs K., through K,. the rack is modeled by mawes 8 Fnction Element unear Torsion through 14 and springs K., through K..; the hydrodynamic couphng between the rack and the fuel and the rack and pool T "" ' wp 77j7y r is represented by the couphngs - H. the fuel to rack paps and fuel to riek impact characteristics are modeled by the Agure 6: HI CAP fuelRad Non/meur CESHOCK Alodel 3
e , s# H'~Qr*H's a / S,mulated
,H- gg Kf 97 'f Sr.acer Gnd Fuel Modet [
KF4 Kas g g, Rack Model
,, Hs10 , H s16 /
a( [~ 'N KF3 K83 Kn3 / Pool Massiess eeam' ,,H ,9 ,Hst5 / r
- Hydrodynamic k , g g ,,, g _g- H %ff_ eumsees R9d Mass xer s /
K r,t ,,,,,, - Ka1 Kni / Poison Don Model- Kai 13 /
/
/
y Fnction Element Fs n ~'p/ ,,,,
-7 in Sliding Analysis.
Fnction Element / " $ """9 '" Fne Rocking Anaiysis
/ ,
T /
////////////////////
Figure 7: Super HI CAP Fuel Rack Nonimear CESHOCK Model to a Hl CAP design for site !!!. Figure 9 provides a similar sponse spectrum analyses (refer to Table 1) shows that the response for a Super HI CAP design (Model B) for site Vll- response spectrum method may give incorrect results. The t'ote the low amplitude. high frequency iesponse of the results demonstrate the importance of the interaction be-rock portion of the modelin contrast to the high amphtude. tween fuel and racks. The interaction is caused by the rela-low frequency tesponse of the fuel. Typical fuelimpact load tive motion between the fuel and rack, through the water. pulses and their effect on peak base shear are seen by com- filled gaps, and impacting of the fuel and rack, paring the response quantities shown also on Figure 9. The ,, , peak base shears occurs just after the time of peak fuel im. %oe ne a tin D put loads. ,,.n,,,,,n,nn.,,,o Table i presents a tabulation of senmic loads developed ,,, a, , ,, , , j 'a withm the rack and trarnmitted to the pool for a number of tius enstoa, essai,w designs and the sites of Figure 5. The lo.id values have been ,[,"'E*,, g , *' ] "
, s,, 5" ]
normalized. The first column identifies the site and the rack design. Four variations of a Hi CAP design (A - D) and 3 variations of a Super HI CAP design (E -G) are presemed
, 'og$,'g"lh',]
ne smN e im.cari
')
4 ::
,l 'd 4 os o E!
i os Four variations of Hl CAP design D are shown; the original si otsicN A 1 99 I 79 1 li sersion, a second version in which dynamic analysis param- n 3m am 3 00 eters were changed by 10% (e.g., fuel stiffness), a third ni otsicN D HiCAPi 27) : 56 t ot l' l' O' '" version with one fourth the original fuel.to rack gap, and a 08 'G 3# '# !" fourth version with an impact sprirg stiffness ten times that of the original. Four variations of Super Hl CAP deugn F n otsicN D , j ', are presented whnh include variainm m p.ips, impact stiff- in,q 4 :t em 4 :t neu and hulrodsnatme mau representatum l)es'rn G v og w,N t iuti a tu c ari t n4 : ?! *n shows remlh los i=Hh a siill and a mit t.n L wppori struc- o,oo , ;. , , . . ;, ture 'l he scion,liolumn persenn the senum loads ohi.nned r,4r nm m ai n n si sim 'u from it.e Cl SilOCK analyses. The third solumn peesents {'[",',' ""} A the cortespondmg sennue loads ot tained, for mmparaine vi n w,N , iu ri a
,.,. p .og . .c g , ,3 ,,
"' C^ h oier avo=o purpmes, by means of response spectrum method analyses.
The last column gives the ration ofloads obtained by the two a rut CAP 8"*' methods, Comparison of results from nonlinear time history anal. ytes (fuel to rack interaction analyses) with those from re-vu DeSica o E','$5"g isorts wicari l$ 'f 3N 4
Fuel Rack Canty u,d. Fuel /-Fuel Pin Storage Boa (Poisoned as Required)
,, 0 2 - I i
e
- o3 . ~ ,
h k Y 5". [ 0~-
--' tee $ 'tI&
d
.g g., )
{, 3, Mid-Rack
. 3 D s$ k o2-j 1Sec.--
.o 3 Fogure 8: CESHOCK Desplacement Response for HI CAP h FuelRock {k f FUEL CONSOLIDATION s g$
Nonlinear time history analysis is also used by C E to analyre consolidated fuel rack designs. The consohdated fuel racks consist of the Super }il cap design with consol-h} s S S idated' fuel rods in each cell. A typical consohdated fuel ;s, / k arrangement is show n in Figure 10. A consolidated fuel can- 3 / g ister with a closely compacted array of fuel rods contained N m ' pio n e(m o:w:e,,sm d - within it exhibits nonlinear characteristics similar to stan-L ompacted C Fuel Pins o 29 - u,o.pg - figure 10: ConsolidatedfuelPin Arrangement
, o 19 E y,o.m.c. dard fuel assemblies. Separate models must be developed to
} one
\ represent different degrees of compaction and, for cases of less than complete compaction, fuel rod impacting must be y
- accounted for. The hydrodynamic effects on fuel canister
{ 0 01 natural frequency and damping are also incorporated into o ott the model. Basic modeling information concerning the dy. [ namic interaction between the consolidated fuel and the can is provided only by testing. Because the interaction between 0 21 consolidated fuel and the can is similar to standard fuel, the
,g 3, nonlinear time history method is used to analyze consoli-d 15 -
dated fuel rack designs. The use of the response spectrum
$ga3 method for consolidated fuel rack designs may lead to in-il I k b correct results.
F_ F_ With consolidation factors of 2 or greater under consid. l .o s -
'l* ,, , , , , , , , , , eration by many uti.ities, it is the job of the analyst to min-imize storage pool design loads due to earthquakes. Because 35 most pools were not designed for consolidation, they cannot 23 , readily accept higher loads. To mintmize modifications to g strengthen pools or to show that modifications are unnec.
e 'S - essary, there are a number of steps the analyst can take. Some of the methods offered by C E to obtain margin for 85 5*,3, d q 'l l l 9
% consolidation designs are listed below:
$ h f' I. Pe analyze the Auxiliary Building with Soil Strue-15 - ture interaction.
- 2. Perform Finite Element Analysis of the Pool.
ao a: ae ae e eo or ,e ee oe too 3. Couple the Fuel Rack Modelto the Auxiliary Build-W 'S ing Model. Forure 9: CESHOCK Response Parameters For Super' 4. Detune the Consolidated Fuel Racks from the Hl. CAP fuel Rack Earthquake. 5
5.0 NEED ASSESSMENT AND ENVIRONMENTAL IMPACT 5.1 NEED ASSESSMENT 5.1.1 .Need For increased Storage Capacity A. NNECO currently has no contractual arrangements with any fuel reprocessing facilities. B. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemblies would be stored on site. Spent fuel could be sent off site for final disposition under existing legislation, but the government facility is not expected to be available before 1993. As matters now stand and until alternate storage facilities are available, spent fuel assemblies on site will remain there. C. It is estimated that the spent fuel pool will be filled, with the proposed increase in storage capacity, in 2009. 5.2 ENVIRONMENTAL EFFECTS 5.2.1 Heat Dissipation Effects This section evaluates the changes in thermal effects due to the proposed increased spent fuel storage capacity. Because spent fuel assemblies will be added periodically, and it is not anticipated that any will be shipped of f-site in the foreseeable future, the annual average heat loads in the spent fuel pool will increase as more assemblies are added to the spent fuel pool. As indicated in Section 3.2, the calculated increase in heat load for a normal refueling sequence is 3.9 x 106 BTU /hr. This would result in an increase in the spent fuel pool water temperature of Il0F, from 1200F to 1310F. This added heat load would cause a minimal temperature rise in the once-through cooling system discharge water temperature. The new 3 pent fuel racks were designed from the outset to accommodate both spent fuel assembhes and consolidated spent fuel. The storage capacity increase used to calculate the heat load increase includes the capacity for consolidated fuel. The proposed storage of consolidated fuel will not, therefore, result in an increase in the spent fuel water temperature beyond 1310F or in the temperature of the once-through cooling system discharge water. 5.2.2 Radiological Considerations 5.2.2.1 Normal Operations Miscellaneous Sources of Exposure The spent fuel pool cleanup filters are required to be changed when clogging of the filter element results in a high dif ferential pressure across the filter. 111storically, this high differential pressure occurs 5-1
rarely during normal operations and three (3) to four (4) times during periods of increased activity in the spent fuel pool (e.g. a refuel outage). The proposed storage of consolidated spent fuel will result in an increased inventory of spent fuel in the storage pool. However, as a general rule, the water borne contamination levels have been found to be a function of work activities in the pool and not a function of fuel inventory. Therefore, the modification of the spent fuel pool is not expected to result in a significant change to the spent fuel pool cleanup filter changeout periodicity. Dose Rates from Fuel Assemblies, Control Rods and Burnable Poison Rods The current inventory of spent fuel, control rods, burnable poison rods and waterborne fission and corrosion product contamination results in spent fuel pool dose rates of: o 15 millirem per hour on pool surface at the center of the pool o 12 millirem per hour three (3) feet above pool surface at the center of the pool. o 7 millirem per hour on the fuel handling bridge at the center of the pool The predominant cont ibutor to these dose rates is waterborne radioactivity. With the new spent fuel rack design implemented in the fuel pool, the minimum water coverage over irradiated fuel in movement will be greater than 10 feet above the irradiated f uel stack. The new spent fuel racks are free standing and situated at a lower elevation in the fuel pool. By changing the fuel handling tool / sling arrangement, more water coverage can be provided for fuel in transit. Af ter the current rerack is complete, as built measurements will be taken to assess the total benefit of such a tool / sling arrangement change. Dose Rate Changes at the Sides of the Pool Walls l The proposed spent fuel rack design does provide for the storage of spent fuel assemblies approximately fourteen (14) inches closer to the walls of the spent fuel pool. However, the minimum thickness of the high density concrete wa!!s and floor of the spent fuel pool is six (6) feet. The fractional change in the the total shield transmission factor when going from a six (6) foot concrete and eighteen (IS) inch water composite shield to a six (6) foot concrete and three (3) and five sixteenths (5/16) loch water shield is insignificantly small. 5-2
Dose Rates at the surface of the SFP water resulting from radioactivity in the water Samples of spent fuel pool water exhibit a peaking effect in isotopic cor'centrations immediately af ter opening the fuel transfer canal gates and allowing the free flow of reactor cavity water into the pool area. The peak tritium concentration in the spent fuel pool water during the 1985 refueling outage was measured as 3.2 E-2 uCi per ml. A measured dose rate of 15 millirem per hours on the pool surface at the center of the the pool is predominantly a result of waterborne radioactivity. The radiological consequences of increasing the capacity of the Millstone Unit No. 2 spent fuel pool from 301 fuel assemblies to 667 assemblies were previously analyzed in Reference 5-1 (Sec. 7.0). The predicted dose rates from 667 assemblies were: Location Dose Rate (mrem /hr) Refueling Platform 2.54 Poolside 1.54 Extrapolating the calculated increase in Reference 5-1 yields the new dose rates for 1,112 assemblies: Location Dose Rate (mrem /hr) Refueling Platform 2.76 Poolside 1.69 The results of this analysis have been extrapolated to predict the radiological impact of storing consolidated fuel in the racks thereby increasing the capacity of the pool from 1112 to 1965 assemblies. Extrapolation yields new dose rate estimates for 1965 total assemblies of: Location Dose Rate mrem /hr) Refueling platform 3.13 Poolside 1.93 Extrapolation in this fashion will yield conservatively high dose rates because the original model used to predict dase rates was based primarily on leaky fuel. Leaving additional old spent fuel assemblies in the pool will not increase the source term linearly as the above ratio implies because decay will result in some of the short lived activity decreasing and, according to Reference 5-1, the cobalt source activity (approximately 75% of the dose rate)is independent of fuel loading. 5-3
The incremental increase in cumulative dose based on the Reference 5-1 occupancy times is: Refueling platform (3.18 mR/hr - 2.76 mR/hr) (400 Man-hrs) = 0.168 Man Rem Poolside (1.93 mR/hr - 1.69 mR/hr) (200 Man-hrs) = 0.048 Man Rem TOTAL 0.22 Man Rem Thus, storage of consolidated fuel will result in an increase in cumulative dose that is conservatively estimated to be below 0.25 Man Rem. The " Radiological Considerations" used in this section for predicting the radiological impact of increasing the spent fuel capacity to 1965 assemb!!cs draws its basis from previous results of dose rate predictions contained in the 1976 Millstone Unit No. 2 Spent Fuel Pool Rerack Submittal, Section 7.0. The basis, models, input data and assumptions for originally predicting the radiological impact of increasing spent fuel pool storage capacity from 301 to 667 assemblies were as follows: The additional spent fuel assemblies in the pool would result in an increase in dose rates in the spent fuel pool area due to a buildup of radionuclides in the pool water. To determine the amount of increase, a calculational model was devised which considered the leakage of the isotopes from the fuel to the pool water, the decontamination factor and flow rate of the spent fuel pool purification system, the isotopic half-lives, and the decay time of the stored spent fuel. Using the resulting model, the activity in the spent fuel pool was predicted for the original 301 assembly pool capacity and for the increase 667 assembly capacity. Activity due to isotopes of cobalt was assumed to be from radioactive crud in the pool water and therfore not affected by the number of assemblies stored in the pool. From this model, the dose rate at the pool surface was found to increase from 1.55 to 1.66 millirem per hour when the pool capacity was increased from 301 assemb!!cs to 667 assemblies. On the refueling platform, five feet above the center of the pool, the dose rate increased from 2.36 to 2.54 millirem per hour. At poolside, one foot from the pool wall and five feet above the surface, the dose rate increased from 1.43 to 1.54 millirem per hour. These resulting small increases in dose rate were found to have a negligible ef fect on personnel exposure. For example, assuming an occupancy thne of 400 man-hours per year at the refueling operations, i the total incremental dose resulting from the original expansion of pool i 5-4
capacity from 301 to 667 assemblies was 0.100 man-rem per year. Sources of the Spent Fuel Pool Water The typically measured, post-shutdown levels of fission and corrosion products in the primary coolant during refueling outages are as follows: Radionuclide Concentration (uCi/ml) Co-58 1.17 E-3 Co-60 2.033 E-4 Cr-51 1.965 E-4 CS-134 3.815 E-3 Cs-137 5.784 E-5 l-131 6.016 E-5 Mn-54 9.77 E-6 Nb-97 1.713 E-4 Sr-92 1.337 E-5 The movement of fuel from the reactor vessel to the spent fuel pool has typically been shown to have a negligible effect on the levels listed above. Spent fuel pool samples do exhibit a peaking ef fect in isotopic concentrations immediately af ter opening the fuel cransfer canal gates and allowing the free flow of reactor cavity water into the pool area. However, the radionuclide concentrations listed above are representative . of the bounding values. Airborne Radioactive Sources & Dose Rates The principle sources for the evaporation mechanism to generate airborn radiological hazards are the fission gases entrained in the spent fuel pool water. In this category, the isotopes Kr85 and H-3 are of the most concern but only H-3 is evident in appreciable quantities during normal operations and refuelings. The peak H-3 concentration in the spent fuel pool water during the 1985 refueling outage was measured as 3.2 E-2 micro-curies per milliliter. The design maximum spent fuel pool water temperature under normal operation, with the increased amount of stored spent fuel, increases from 1220F to 1310F. This small increase in the water temperature will result in an increase in the evaporation rate from the pool. While this increase in the etaporation rate will result in an increase in tritium exposures to plant personnel working in the area of the spent fuel pool, the present levels of dose commitments from tritium exposure in the spent fuel building are below the levels required for tracking per 10 CFR
- 20. The calculated increase in tritium exposures, due to the increased fuel storage capacity, is approximately 28% and is not deemed as significantly increasing personnel dose commitments.
5.2.2.2 Accident Conditions Similar to Reference 5-1, the proposed modification does not affect the basis for the safety analysis of the fuel handling accident discussed in Section 14.19 of the Millstone Unit No. 2 FSAR. As such, the results of that analysis are still applicable. 5-5
In Reference 5-2 the requirement for a specified decay time of 120 days for fuel stored within a distance L from the center of the spent fuel pool cask set-down area was identified. The distance L equals the major dimension of the shielded cask. While the proposed modification does not affect the basis for the safety analysis of the cask drop accident, review of the consequences of a cask drop accident with respect to the increased fuel storage density, in the form of consolidated f uel, determined that the site boundary two hour whole body dose would actually be 30 millirem lower than the dose previously calculated in Reference 5-3 because of the increased decay time of the fuel stored in Region !!. 5.2. 3 Chemical Discharges The water from the Millstone Unit No. 2 spent fuel pool is purified by passing it through the spent fuel pool cleanup system and returning it to the pool. This cleanup system consists of filters and a demineralizer to remove radioactive nuclides and chemical impurities in the water. The wastes generated by the system consist of a small volume of filter cartridges and resins which are packaged and shipped offsite as solids to an approved burial si te. There is no change expected in the environmental elfects from chemical discharges over that previously evaluated.
5.3 REFERENCES
5-1 Letter, D. C. Switzer (NNECO) to G. Lear (NRC) under Docket No. 50-336 dated November 22,1976, " Modifications to Spent Fuel Storage Pool." 5-2 Letter, G. Lear (NRC) to D. C. Switzer (NNECO) dated June 30,1977. 5-3 Letter,3. F. Opeka (NNECO) to E. J. Butcher (NRC) under Docket No. 50-336 dated July 24,1985, " Modifications to Spent Fuel Storage Pool." 5-6
Docket No. 50-336 B12000 Attachment 3 Millstone Unit No. 2 No Significant Hazards Consideration Determination May,1986 t
- 1. Introduction 10 CFR 50.91 requires that requests for an amendment to an operating license be accompanied by an analysis "about the issue of no significant hazards consideration." Such analysis is to focus on the three standards set forth in 10 CFR 50.92(c), as quoted below:
"The Commission may make a final determination pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22 or for a testing facility involves no significant hazards considerations, if operation of the facility in accordarice with the proposed amendment would not (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety."
Northeast Nuclear Energy Company (NNECO) submits that the activities associated with the storage of consolidated spent fuelin the Millstone Unit No. 2 spent fuel pool are outside the standards set forth in 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration determination is warranted with respect to the proposed license amendment.
II. Background Millstone Unit No. 2 was designed and constructed with a spent fuel pool having a capacity of 301 spent fuel assemblies. The Millstone Unit No. 2 Final Safety Analysis Report addresses the safety and environmental implications of the pool, including the relevant acceptance criteria with respect to criticality, structural integrity, and cooling. The NRC found the environmental and safety impacts of spent fuel storage to be acceptable. By November 1976 it became necessary for NNECO to provide greater onsite capacity for storage of spent fuel. Backend fuel cycle services, particularly spent fuel reprocessing, had not materialized as originally anticipated. A spent fuel pool rerack amendment was approved and issued by the NRC on June 30, 1977. The reracking amendment allowed modifications to increase pool storage capacity to 667 fuel assemblics. Following the Cycle 6 refueling outage in 1985, it again became necessary to provide greater onsite storage capacity for spent fuel. The circumstances in the backend of the nuclear fuel cycle services made it necessary that fuel owners establish and implement a plan for " life-of-reactor-storage" of nuclear spent fuel. Thus, a second spent fuel pool rerack amendment was requested by NNECO. On January 15,1936 an amendment authorizing a storage capacity increase was approved and issued by the NRC which allowed modifications to increase pool storage capacity to 1112 fuel assemblies. The rerack utilized a region strategy,
which included regions of both poisoned and non-poisoned spent fuel racks, and in one region taking credit for reactivity depletion in spent fuel to provide substantially closer center-to-center spacing of fuel assemblies for increased capacities. A further reduction in center-to-center spacing of fuel assemblies is not anticipated because of the physical limitations of pool size and structural require men ts. However the most recent rerack modification will accommodate the spent fuel generated by normal operation of the plant only until approximately 1993, assuming reloads of a third of a core and a!!owing space for a full core offload. NNECO is now applying for the license amendments necessary to support the storage of consolidated fuel, together with spent fuel assemblics, in Regions I and 11 of the Millstone 2 spent fuel pool as reracked in 1986. NNECO, in cooperation with the Electric Power Research Institute (EPRI), Baltimore Gas & Electric Co. and Combustion Engineering is completing the development of a process which packs the fuel rods from two spent fuel assemblies in a storage box which can be handled like a fuel assembly and fits a cell of the new Millstone 2 fuel racks. Storage of consolidated fuel would effectively double the net capacity of the new fuel racks after reserving space for the full core of f load.
te. No structural changes to the storage rack modules are required as they were initially designed to withstand the static and dynamic loads resulting from the storage of both intact fuel assemblies and consolidated fuel in Regions I and 11. Similarly, both regions of the spent fuel pool were initially designed to store both intact fuel assemblies and consolidated fuel in a safe, coolable, subcritical configuration with K gg e less than .95. The proposed storage of consolidated spent fuel will increase the capacity of the spent fuel pool to 1965 assemblies and is described more fully in the accompanying " Millstone Nuclear Power Station, Unit No. 2 -Consolidated Fuel Storage Safety Analysis Report" (Attachment 2). In sum, NNECO will utilize consolidated fuel storage boxes designed to contain spent fuel rods from two assemblies in a close-packed triangular array to be stored in the spent fuel pool.
~5-III. Application of Standards A. First Standard The proposed modification does not involve a significant increase in the probability or consequence of an accident previously evaluated.
The NNECO safety analysis of the proposed method of fuel storage has been accomplished using current NRC Staff accepted Codes and Standards as specified in Section 4.2 of Attachment 2. The results of the safety analysis demonstrate that the proposal meets the specified acceptance criteria set forth in these standards. As a result of the analysis and reviews, NNECO has identified the fo!!owing potential accident scenarios: 1) spent fuel cask drop;
- 2) loss of spent fuel pool forced cooling; 3) seismic event; 4) spent fuel assembly drop; 5) criticality accident; and 6) load handling accident. The probability of these events is not af fected by the amount or type of fuel stored in the pool.
All potential events which could involve accidental criticality have been examined in Section 3.0 of Attachment 2. It was concluded that the bounding accident was dropping an unirradiated fuel assembly into a blocked fourth location in Region II. The probability of
6-dropping an unirradiated fuel assembly during fuel movement operations is not af fected by the capacity of the fuel storage racks or the form of fuel stored. The preposal to store consolidated spent fuel in the Millstone Unit No. 2 spent fuel pool will not involve an increase in probability of any previously evaluated load handling accident as accepted standards and procedures will be utilized as described in Section 3.3 of , . The consequences of the spent fuel cask drop accident have been evaluated as described in Sections 5.4 and 9.8 of the Millstone Unit No. 2 Final Safety Analysis Report (FSAR). By controlling the decay time for fuel stored within a specified distance from the cask set down area to not less than 120 days prior to cask movement together with an administrative control specifying a minimum required boron concentration in the water of the spent fuel pool, the consequences of this type accident will not be increased from previously evaluated even ts. The consequences of the loss of spent fuel pool forced cooling accident have been evaluated and are described in Sections 3.2 of . The racks were evaluated against the appropriate NRC standards as described in Section 4.2 of Attachment 2. The results of tne seismic and structural analysis show that the proposed racks meet all of the NRC structural acceptance criteria and are
._ - . _ ~. _ _ _ _ _ . _ . _ _ _ _ . . .
consistent with results found acceptable by the NRC Staff in previous i
- poison rerack SERs. Thus, the consec;uences of seismic events will i not significantly increase from previously evaluated seismic events.
I 1 The consequences of a spent fuel assembly drop accident are described in Section 14.19 of the Millstone Unit No. 2 FSAR. A complete list of assumptions is provided in FSAR Table 14.19-1. Results of the analysis are well below the limits of 10CFR100 and are i presented in Section 14.19.3. The consequences of this type accident will not be significantly increased from previously evaluated accidents. i i j The consequences of a criticality accident have been evaluated in Section 3.1.2 of Attachment 2. All potential events which could involve accidental criticalMy have been examined. The bounding criticality accident was found to be the dropping of a fresh fuel assembly into a blocked fourth location in Region 11. Administrative controls in the form of a Technical Specification of minimum boron concentration "fd the yater of the spent fuel pool (Attachment 1)
~
wi11 preclude the bounding criticality accident; there fore, the consequences,of this type ac'cident will not be significantly increased from prshious accident evaluati6nsI 5
, ,f '
. .e n
'h s , e:-, . . . , , ._ _ - ._.. _ _ . . _ - . . - - - - - - . - -- _ __
1 B. Second Standard i The proposed modification does not create the possibility of a new or 4 different kind of accident from any accident previously evaluated. NNECO has evaluated the proposed storage of consolidated fuel in accordance with the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plan sections, and appropriate industry Codes and Standards as described in Section 4.2 of Attachment 2. In addition, NNECO has reviewed the NRC Safety Evaluation Report for the R. E. Ginna Nuclear Power Plant application. NNECO does not consider a fuel cannister accident to be materially dif ferent from a fuel assembly accident since both assume the failure l of a fuel handling tool or system. The incorporation of a two-region i spent fuel pool storage system does create the requirement to perform additional evaluations to ensure the criticality requirement is maintained. These include the evaluation of the limiting condition (dropping a fresh fuel assembly into a blocked fourth location in Region II). This evaluation shows that when the boron concentration i required by the existing Technical Specifications is me t, the criticality criterion is satisfied. Hence, the proposed amendment i does not create the possibility of a new or different kind of accident l I
- from any accident previously evaluated.
l l l l t
- . , _ - . , , ~ , . . - _ - , --
C. Third Standard The proposed modification does not involve a significant reduction in a margin of safety. The issue of " margin of safety" when applied to the storage of consolidated spent fuelincludes the following considerations:
- a. Nuclear criticality considerations
- b. Thermal hydraulic considerations
- c. Mechanical, and structural considerations The margin of safety that has been established for nuclear criticality is that the neutron multiplication factor (Keff) in the spent fuel pool is to be less-than-or-equal-to 0.95, including all uncertainties, under all conditions. For the proposed modification, the criticality analysis is decibed in Section 3.1 of Attachment 2. The methods utilized in
, the analysis conform with ANSI N210-1976, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations"; ANSI j N 16.9-1975, " Validation of Calculation Methods for Nuclear Criticality Safety"; the NRC guidance,"NRC Position for Review and i Acceptance of Spent Fuel Storage and Handling Applications" (April 1978), as modified (January 1979); and Regulatory Guide 1.13, " Spent
} Fuel Facility Design Basis," proposed Revision 2. The computer programs and data libraries relating to the spent fuel storage rack design have been used in numerous spent fuel rack replacement applications by other NRC licensees and have been reviewed and approved by the NRC. i Additional laboratory tests were performed to establish an achievable metal / water ratio for use in criticality calculations. Evaluations of existing computer codes using previous criticality experiments involving fuel rods in triangular and square arrays resulted in a homogenized fuel module representation which would not adversely ] affect the accuracy of the multiplication factor predictions. The l results of NNECO's analysis indicate that Keff is less-than-or-equal-to 0.95 under all postulated conditions, including uncertainties, at a 95/95 probability / confidence level. Thus, the proposed storage of consolidated spent fuel, in meeting the acceptance criteria for criticality, does not involve a significant reduction in the margin of i safety for nuclear criticality. For thermal hydraulics, the relevant considerations for determining if there is a significant reduction in margin of safety are: , (1) maximum fuel temperature, and (2) the increase in temperature of the water in the pool. The thermal hydraulic evaluation is described in Section 3.2 of Attachment 2. Results of this analysis show that bulk boiling of the pool will not exist under normal a i
-. . _ _ _ _ _ _ , _ . . . _ _ _ _ _ , . , . _ . , . . . _ . _ _ - , . - _ , _ _ . _ _ _ _ . _ _ . _ _ . , _ _ - _ , _ . _ _ _ _ _ . _ . _ _ . , _ ~ _ . _ , _ .
operation, fuel cladding temperatures under abnormal conditions are sufficiently low to preclude structural failure, and that boiling does not occur in the water channels between the fuel assemblies nor within the storage boxes. As shown in Section 3.2 of Attachment 2, the maximum temperature will not exceed the current margin of safety (1500F). For the maximum normal heat load case (full-core discharge at 150 hr af ter shutdown, which fills the spent fuel pool to its capacity), the pool temperature will not exceed 1500F. Thus, there is no significant reduction in the margin of safety from a thermal hydraulic standpoint or from a spent fuel pool cooling standpoint. The mechanical, material, and structural considerations of the proposed storage method are analyzed in Section 4.0 of . As described in Section 4.2, the racks are designed in accordance with the applicable NRC Regulatory Guides, Standard Review Plan sections, and position papers, as well as the appropriate industry Codes and Standards. The racks are designed to Seismic Category I requirements. The materials utilized are described in Section 4.6.1 and are compatible with the spent fuel pool and the spent fuel assemblies. The structural considerations are also described in Section 4.0 of Attachment 2. The conclusion of the analysis is that the margin of safety is not significantly reduced by the proposed storage of consolidated fuel.
IV. Conclusion A recent final rule promulgated (I) by the Commission has specifically addressed the issue of standards for no significant hazards considerations. Contained in the Supplementary Information accompanying the final rule were examples of amendments that are and are not _likely to involve significant hazards considerations. Included in the list of examples (51FR7451) considered not likely to involve significant hazards considerations is an expansion of the storage capacity of a spent fuel pool. Further, four criteria were specified which were to be satisfied for this example to apply. (1) "The storage expansion method consists of either replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits." As shown in Sections 2, 3 and 4 of Attachment 2 to this suomittal, the spent fuel storage racks recently authorized for ust: in the Millstone Unit No. 2 spent fuel pool are indeed designed to allow the safe storage of either a standard spent fuel assembly or a consolidated fuel storage box containing spent fuel consolidated to a ratio of 2:1. Thus, this criterion is satisfied. (1) The rule change entitled " Final Procedures and Standards on No Significant Hazards Considerations" was published in 51FR7744, ef fective May 5,1986.
(2) "The storage expansion method does not involved rod consolidation or double tiering." This submittal does not deal with double tiering. It does involve one aspect of rod consolidation, namely, the storage of consolidated fuel storage boxes which will provide for the storage of the fuel pins from two spent fuel assemblies (assuming a 2:1 consolidation ratio) in the same storage volume as that used by a single spent fuel assembly. The submittal does not seek authorization for actual rod consolidation processing on site, such as ( dismantling 6r cutting apart fuel assemblies. I
, Thus, this submittal does not seek approval for the entire cycle of rod i
consolidation. The analytical tools necessary to evaluate and justify the safe storage of consolidated spent fuel - the only activity for which , approval is sought - are readily available from standard techniques. In this sense, the submittal does not involve new technology, the presence of which might prevent the NRC from concluding that the amendment involves no significant hazards consideration. NNECO submits that the ) intent and underlying safety objective of this criterion are met for the'. proposed amendment. '. i f i If the study being undertaken by NNECO, in conjunction with the Electric Power Research Institute, and others, results in demonstrated, viable methods for a commercial process far consolidation of spent fuel, we would l anticipate utilizing those methods at Millstone Unit No. 2. However, these processes are not germane to this amendment request. i
The above evaluation regarding criterion (2) cited above is supported by the information contained in SECY-83-337, Study on Significant Hazards, dated August 15, 1983. While addressing the fuel storage technique of rod consolidation, the Staf f stated (at page 5) that:
" Rod consolidation, however, involves new technology and increased handling of highly radioactive components of fuel assemblies."
(Emphasis added). The thrust of the concern appears to be focused on the process of rod consolidation, rather than storage. The following excerpt from the Enclosure to SECY 83-337 (at page 26) reinforces the fact that the focus of the concern relates to the process:
"The storage expansion method utilizing rod consolidation is a unique situation. One concern is that due to the mechanical handling of the bundles there will be additional failed fuel. While this additional failed fuel is not associated with an accident, the release of gap activity, primarily Kr-85, if rod consolidation is performed only on aged spent fuel, could be increased considerably. (Emphasis added).
NNECO acknowledges that SECY 83-337 also discusses the increase in handling that will accompany the storage for which authorization is sought. However, we believe that this handling process can be evaluated using standard techniques and does not involve new technology. l l t
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The background information contained in SECY-83-337 supports NNECO's position that the intent and underlying safety objective of criterion (2) are met. (3) "The Keff of the pool is maintained less than or equal to 0.95." As shown in Section 3 of Attachment 2 to this submittal, this requirement is fulfilled for all cases of fuel storage. Hence, this criterion is satisfied. (4) "No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to justify the expansion." The racks authorized for installation and use in the Millstone Unit No. 2 spent fuel pool were designed and constructed based on proven design and analysis techniques. As shown in Attachment 2 to this submittal, the analytical techniques used to justify this expansion request are those used for many spent fuel pool expansions. Hence, this criterion is satisfied. In light of the above considerations, the intent and underlying safety objectives of all four criteria previously listed as an example of an amendment not likely to involve significant hazards considerations are fulfilled. The proposed Millstone Unit No. 2 storage of consolidated spent fuel does not:
- a. Involve a significant increase in the probability or consequences of an accident previously evaluated; or t
- b. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- c. Involve a significant reduction in a margin of safety.
As such, NNECO has determined and submits that the proposed storage of consolidated spent fuel in the Millstone Unit No. 2 spent fuel pool involves no significant hazards considerations. f g}}