ML19319D957
| ML19319D957 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/19/1975 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19319D956 | List: |
| References | |
| NUDOCS 8003270633 | |
| Download: ML19319D957 (13) | |
Text
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Biulatory Docket Fle f
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ATTACHMENT I 1
RANCHO SECO UflIT 1 PROPOSED AMEfiDMENT 40 TO OPERATIflG LICENSE NO. DPR-54 i
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33 8003270
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS j
Design Features 5.4 NEW AND SPENT FUEL STORAGE FACILITIES Specification 5.4.1 New Fuel inspection and Temporary Storage Rack A.
New fuel shall be removed from the shipping containers, inspected and temporarily stored in the new fuel inspection rack.
This rack is located on the operating floor and consists of two parallel modules containing ten spaces each on 21-1/8 inch centers. This spacing is sufficient to maintain Keff less than 0 9 when flooded with unborated water, based on a fuel enrichment of 3.5 weight 2
percent U 35 After fuel assemblies have been inspected, they may be moved to the new fuel elevator and lowered to the floor of the spent fuel storage pool, one at a time. (l) 8.
New fuel may also be stored in their shipping containers.
5.4.2 New and Spent Fuel Storage Racks and Failed Fuel Storage Container Rack New fuel while awaiting transfer to the Reactor Building and irradiated or failed fuel prior to off-site shipment will be stored in the stain-less steel lined pool.
The spent fuel pool is sized to accommodate 579 fuel assemblies including 3 assemblies in failed fuel containers.
During refueling, the borated fuel pool water will have a minimum concentration of 1800 ppm.
The pool has the capability of storing new and spent fuel assemblies in fourteen stainless steel rack modules and three failed fuel assemblies in a special rack module. All assemblies are on minimum 14.5 inch centers in both directions.
This spacing is sufficient to maintain Keff less than 0 95 when flooded with unborated water, based on a fuel enrichment of 3 5 weight percent.
5.4.3 New and Spent Fuel Temporary Storage The Reactor Building has one single row stainless steel storage rack in the deep portion of the refueling canal.
This rack is designed to hold six assemblies and one failed fuel detection can, all on 21-1/8 inch centers.
5.4.4 Spent Fuel Pool and Storage Rack Design The spent fuel pool and all storage racks are designed for the design base earthquake.
REFERENCE (1)
FSAR subsection 9.8 l
ftegularery Dogket f!is tx-\\'te?S ATTACHMENT II RANCHO SECO UNIT 1 FINAL SAFETY ANALYSIS REPORT PAGE REVISIONS
o roel Storage Building 5.4 FUEL STORAGE BUILDING 5.4.1 GENERAL DESCRIPTION The Fuel Storage Building is a Class I reinforced concrete rectangular tank with a stainless steel liner and a super structure with concrete walls and steel roof system.
The spent fuel pool is sized to hold three times the core capacity and is designed to withstand all external influences, particularly seismic conditions.
l The design of the walls takes into account the hydrodynamic effects of the water including both impulsive and convective forces generated by the inertial characteristics of the mass of water during earth movement.
The building is structurally separated from the Auxiliary Building and bellows joints are provided in the fuel transfer system to allow for differential movement between the Reactor Building and the fuel pool due to thermal or earthquake loads.
5.4.2 DESIGN BASES 5.4.2.1 Design Loads The Fuel Storage Building is designed for all credible conditions of loading including normal loads and loads from a design base earthquake accident.
The following loads are considered:
A.
Structure dead loads B.
Live loads C.
Earthquake loads 0.
Wind loads.
The critical loading conditions are those caused by the hydrostatic effects of the borated water, the fuel cask, and an earthquake.
5.4.2.1.1 Dead Loads Dead loads consist of the weight of the cor. crete, structural steel, and equip-ment.
5.4.2.1.2 Live Loads The following live loads are considered in the design:
A.
Roof loads of 20 pounds per square foot B.
Hydrostatic loads from the pool filled with borated water 5.4-1
Fuel Storage Building t
C.
Loads.from fuel up to 579 fuel elements l
D.
Loads from the fuel transfer cask 5.4.2.1.3 Earthquake Loads Earthquake loading is predicated upon an operating base earthquake a t the site having a horizontal ground acceleration of 0.13g and a vertical accaler-ation of 0.099 In addition, a design earthquake having a ground acceleration of 0.259 and a vertical acceleration of 0.17g is used to check the design to ensure no loss of function.
Seismic response spectrum curves are given in Appendix 5B for both horizontal ground motion and vertical ground motion. A dynamic analysis which includes the hydrodynamic effect of the water is used to arrive at equivalent static loads for design.
Seismic loads are combined as outlined in Appendix 58.
5.4.2.1.4 Wind Loads Wind loading is based on Figure 1 (b) of ASCE Paper 3269. Wind Forces on Structures using the fastest wind speed for a 100-year recurrence period or the recormiendation of the UBC whichever is greater.
5.4.2.2 Design Criteria The main considerations in establishing the structural design criteria for the Fuel Storage Building are to provide a leak tight pool to contain spent fuel under all conditions of loading including earthquake.
Except as noted in these criteria ACI 318-63 and AISC, Sixth Edition design methods and allowable stresses are for the design of reinforced concrete and steel respectively.
The strength of the structure at working stress and over-all yielding is com-pared to various loading combinations to ensure safety.
The structure is designed to meet the performance and strength requirements under the following conditions:
A.
At design loads B.
At factored loads 5.4.2.2.1 At Design Loads This loading is the basic " working stress" design.
The structure is designed for the following loading cases:
A.
A = (1 +.05) D + 1.0L B.
A = (1 +.05) D + 1.0L + 1.0E 5.4-2
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1 Appendix 5A OVESTION 5A.39 With respect to the spent fuel pool, discuss whether (AEC NO. 5.31) cracking from thermal stresses was considered in the design of the pool walls and the provisions made to control cracking in this structure.
To substantiate whether or not cracking is a potential problem, state the predicted maximum thermal stresses which can be developed in the pool walls under the most severe anticipated thermal conditions, and considering the combined stresses due to the worst anticipated loading conditions.
ANSWER:
Reinforcement for crack control is in accordance with ACI 318-63, as a minimum.
Abnormal thermal stresses were not considered in the design of the fuel pool since three cooling systems must fail for this condition to occur (two of the cooling systems are Class I).
Sufficient time is available, in case of the functional loss of these systems, to re-establish cooling.
Based upon the functional loss of all three available cooling systems, the most severe thermal condition that could exist in the pool is 212 F. (refer to paragraph 9.6.2.4).
If only the spent fuel heat removal system is used the temperature of 180 F would exist in the spent fuel pool with three cores stored.
If the spent fuel cooler and pump was inoperative both of the decay heat removal systems would be available and the fuel pool temperature would be less than 180 F.
The spent fuel pool has been checked and can withstand up to 212 F upon loss of all cooling.
QUESTION SA.40 Where yielding is permitted in structures, define the (AEC N0. 5.32) allowable limits and list the structures, parts of structures, and elements where yielding may occur and under what load combinations.
Provide information to show how those limits meet acceptable design strain limits.
ANSWER:
Yielding is predicted only in the containment plate and minor local concen-trations in the concrete of the reactor interior structure under loading conditions of pipe rupture and missiles.
However the average stresses of the Reactor Building interior structure is below yield.
SA-34
a Spent Fuel Cooling System
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9.6 SPENT FUEL COOLING SYSTEM l
9.6.1 DESIGN BASES The spent fuel cooling system is designed to maintain the spent fuel storage pool at 120 F with a heat load based on removing the decay heat generation from one 1/3 core, which has been irradiated for 989 effective full power days, (EFPD) and cooled for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.
In addition to the above design bases, the system has the capability by it-self to maintain the spent fuel storage pool at 120 F while removing decay heat from the following combination of stored fuel assemblies.
A.
1/3 core irradiated for 3 years and cooled for 8 years B.
7 C.
6 D.
5 E.
4 F.
3 G.
2 H.
1 I.
14 days As a final design basis, the system supplemented by a decay heat removal pump and cooler, has the capability to maintain the spent fuel storage pool at 120 F while removing the decay heat from the following combination of stored fuel assemblies.
A.
1/3 core irradiated for 3 years and cooled for 5-1/4 years B.
4-1/4 C.
3-1/4 D.
2-1/4 E.
l-1/4 F.
1/4 G. 1/3 core irra;iated for 2-1/4 years and cooled for 5 days H.
l-1/4 5
I.
1/4 5
Nb3 L
Spent Fuel Cooling System 9.6.1.1 Performance Reouirements The first design basis of the system predicates an operating schedule in which the nuclear unit is on an equilibrium refueling priod (284 days per cycle) with approximately 1/3 of a core being removed from the unit at the end of each period.
The removed fuel assemblies will have been in the reactor for three cycles, i.e., two 284 day cycles and an initial cycle of 421 days for a total of 989 days.
The second design basis for the system considers the possible use of the pool to store its maximum amount of fuel. A conservative full 3 years irradiation was assumed for each 1/3 core and shows the capability of the spent fuel pool cooling system to handle the entire cooling load after the ninth batch has cooled for 14 days.
The final design basis considers that it is possible during the life of the plant to be necessary to unload the reactor vessel totally for maintenance or inspection when six batches (2 full cores) are already residing in the spent fuel storage pool.
9.6.2 SYSTEM DESCRIPTION The schematic flow diagram for the spent fuel cooling system is shown in figure 9.6.1.
The system consists of (a) cooling loop, including the spent fuel storage pool, pump, cooler, piping and valves, (b) purification loop, including the spent fuel coolant demineralizer pump, filter, ion exchanger, piping and valves and (c) skimmer loop, which is valved into the normal purification loop.
The removal of the decay heat generated in the spent fuel storage pool is accomplished by continuous recirculation of the coolant through the spent fuel cooler.
The purification loop is tied into the cooling loop downstream of the cooler. A fraction of the total flow is diverted through the purifi-cation loop and, after passing through the filter and ion exchanger, returned back to the primary flow.
In addition, a skimmer is installed to keep the pool surface I
Spent Fuel Cooling System 9.6.2.3.3 Heat Exchanger The spent fuel cooler is designed and manufactured to the ASME Boiler and Pressure Vessel Code Section III-C remainder. General requirements conform to TEMA-R Standards.
Nondestructive testing including ultrasonic and eddy-current testing of tubes conforms to the specifications of the above code.
Both the shell and the tube side of the cooler are tested hydrostatically at 1.5 times the design pressure.
9.6.2.3.4 Filters and Ion Exchanger The spent fuel coolant filter and ion exchanger are made of stainless steel and designed, manufactured and tested in accordance with the ASME Boiler and Pressure Vessel Code.
9.6.2.4 Mode of Operation During normal conditions up to nine batches (3 cores) will be stored in the pool. At this time the spent fuel pump and the cooler will handle the load and maintain 120 F.
For the case where 3 cores are stored due to complete unloading of the reactor vessel, one of the two redundant decay heat removal pumps and coolers will be used to supplement the pump and cooler to maintain the spent fuel storage pool temperature at 120 F.
If only the spent fuel pump and cooler are used when this storage condition exists, water tempera-ture will eventually rise to 180 F five days after shutdown; although 9.8 l
hours will be required to heat the large spent fuel storage pool to this temperature.
If no cooling is provided, the time required for the spent fuel storage pool to reach 212 F for each of the foregoing quantities of stored fuel is:
A.
3 cores 36.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> B.
3 cores (one a complete unload) 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 9.6.2.5 Reliability Considerations During the time when up to 3 core. are stored in the pool, the installed l
equipment will be utilized to maintain the pool at 120 F.
9.6.2.6 Leakaae Considerations Whenever a leaking fuel assembly is transferred from the fuel transfer canal to the spent fuel storage pool, a small quantity of fission products may enter the spent fuel cooling water. A purification loop is provided for removing these fission products and other contaminants from the water.
The fuel handling and storage area housing the spent fuel storage pool will be ventilated on a controlled basis, exhausting circulated air to the out-side through the plant vent.
9.6-4
W Fuel Handling System 9.8 FUEL HANDLING SYSTEM 9.8.1 DESIGN BASES 9.8.1.1 General System Function The fuel handling system (Figure 9.8-1) is designed to safely and effectively transport and handle fuel from the time it reaches the plant in an unirradiated condition until it leaves the plant after postirradiation cooling.
The system is designed to minimize the possibility of mishandling or maloperations that can cause fuel assembly damage and/or potential fission product release.
The reactor is refueled with equipment designed to handle the spent fuel assemblies under water from the time they leave the reactor vessel until they are placed in a cask for shipment from the site.
Underwater transfer of spent fuel assemblies provides an effective, economic, transparent radiation shield, as well as a reliable cooling medium for removal of decay heat.
Borated water insures subcritical conditions during refueling.
9.8.1.2 New Fuel Storage Area The new fuel storage area is a separate area for the temporary dry storage of new fuel assemblies, and it will accommodate twenty (20) new fuel assemblies.
New fuel assemblies are stored in a rack having two parallel rows approxi-mately 4 feet 3 inches apart with the fuel assemblies in each row having a center-to-center distance of 21-1/8 inches.
This spacing will maintain a keff of less than 0.9 when wet.
9.8.1.3 Fuel Storage Pool The fuel storage pool is a reinforced concrete pool, lined with stainless steel, in the Fuel Storage Building. The pool is sized to accommodate 576 j
spent fuel assemblies, which allows for a full core of irradiated fuel assem-blies in addition to the concurrent storage of six batches of new or spent fuel assemblies.
The fuel assemblies are stored in racks in parallel rows with a center-to-center distance of 15 inches in both directions.
Control rod assemblies that are removed from the reactor are stored in the spent fuel assemblies. Additional spaces are provided for the storage of three (3) failed fuel containers in the fuel storage pool.
9.8.1.4 Fuel Transfer Tube Two horizontal tubes convey fuel between the Reactor Building and the fuel storage pool.
These tubes contain tracks for the fuel transfer carriages, j
gate valves on the spent fuel storage pool side, and a flanged closure on the Reactor Building side.
The fuel transfer tubes penetrate into the fuel trans-fer canal at the lower depth, where space is provided to rotate fuel transfer carriage baskets containing a fuel assembly or a failed fuel container.
9.8-1
Fuel Handling System 9.8.2.3 Safety Provisions Safety provisions are designed into the fuel handling system to prevent the development of hazardous conditions in the event of component malfunctions, accidental damage, or operational and administrative failures during refueling or transfer operatiu s.
The fuel storage racks are designed so that it is impossible to insert fuel l
assemblies in other than the prescribed locations, thereby insuring the necessary spacing between assemblies.
Under these conditions, a criticality accident during refueling or storage is not considered credible.
Fuel handling equipment is designed to minimize the possibility of mechanical damage to the fuel assemblies during transfer operations.
If fuel damage should occur, the amount of radioactivity reaching the environment will pre-sent no hazard.
The fuel handling accident is analyzed in Section 14.
All spent fuel assembly transfer operations are conducted underwater.
The water level in the fuel transfer canal provides a minimum of 9 feet of water over the active fuel line of the spent fuel assemblies during movement from the core into storage to limit radiation at the surface of the water to less than 10 mrem /h.
The fuel storage racks provide a minimum of 23 feet of water shielding over stored assemblies.
The minimum depth of the water over the fuel assemblies and the thickness of the concrete walls of the storage pool are sufficient to limit the maximum continuous radiation levels in the work-ing area to 2.5 mrem /h.
Water in the reactor vessel is cooled during shutdown and refueling by the decay heat removal system, as described in 9.5.
Adequate redundant electrical power supply assures continuity of heat removal.
Fuel storage pool water is cooled by the spent fuel cooling system described in 9.4.
A power failure during the refueling cycle will create no immediately hazardous condition because of the large water volume in both the transfer canal and fuel pool.
With a maximum quantity of spent fuel assemblies in the storage pool and no l
cooling available, the water temperature in the spent fuel pool will increce very slowly.
During reactor operations, bolted and gasketed closure plates on the Reactor Building flanges of the fuel transfer tubes guarantee that spent fuel pool water will not leak into the transfer canal if there is a leak through the transfer tube valves.
Both the storage pool and the fuel transfer canals are completely lined with stainless steel plate for leak tightness and for ease of decontamination.
The fuel transfer tubes are attached to these liners to maintain leak integrity.
The fuel pool cannot be inadvertently drained by gravity since water must be pumped out.
9.8-5
Fuel Handling System During the refuleing period, the water level in the fuel transfer canal and the spent fuel pool is the same, since the fuel transfer tube valves are open.
This eliminates the necessity for interlocks between the fuel transfer car-riages and transfer tube valve operations, except to verify full-open valve position.
The fuel transfer canal water will have a boron concentration of 1,800 ppm.
Although this concentration is sufficient to maintain core shutdown if all of the control rod assemblies were removed from the core, not more than two control rods will be removed at any one time during the fuel shuffling and replacement.
Although not required for safe storage of spent fuel assemblies, the fuel storage pool water will also be maintained at 1,800 ppm boron during refueling.
l The fuel transfer mechanisms permit initiation of the fuel basket rotation only from the building in which the fuel basket is being loaded or unloaded.
Carriage travel and fuel basket rotation are interlocked to prevent inadvertent carriage movement when the fuel baskets are in the vertical position.
Rota-tion of the fuel baskets is possible only when the carriages are in the rotat-ing frame at the end of travel.
The fuel handling and control rod handling mechanisms are so designed that the fuel and rod assemblies are withdrawn into the mast tube for protection prior to transfer.
Interlocks prevent operating the bridges or trolleys until the assemblies have been hoisted to the upper limit in the mast tube. Mandatory slow zones are provided for the hoisting mechanisms during insertion of fuel and rod assemblies.
The slow zones are ir. effect during entry into the reactor core, fuel transfer basket, or fuel rack, and just before and during bottoming of the fuel and rod assemblies.
The controls are appropriately interlocked to prevent simultaneous movement of the bridge, trolley, or hoists.
The grapple mechanisms are interlocked with the hoists to prevent vertical movement unless the grapples are either fully opened or fully closed.
The grapple cannot be opened as a result of operator error, or electrical or hydraulic failure.
Hard stops prevent raising an assembly above minimum shielding depth in the event of an up-limit failure.
All fuel transfer system operating mechanisms are in the fuel handling and storage area for ease of maintenance and accessibility for inspection prior to start of refueling operations. All electrical equipment, with the excep-tion of some limit switches, is located above water level for greater integrity and ease of maintenance.
The hydraulic systems that actuate the fuel basket rotating frames use demineralized water for operation.
l l
9.8-6
Standby Safeguards Analysis combined water-ir artition factor of 103 is conservatively assumed in the two condensers. 1,2 An activated charcoal bed efficiency of 90 per:ent is assumed for iodine removal.
Consequently, a total iodine decontamination factor of 104 is assumed for this accident.
Individual isotopic activities released to the secondary system in this acci-dent are listed in Appendix }4D, table 14D-17.
Based on these values and a decontamination factor of 10 for iodine, total releases are given in table 14.2-6 along with a summary of results arising from this accident.
The environmental effects of the suam generator tube rupture accident with-out credit for condenser relief are presented in paragraph 14.3.4 and table 14.3-1.
14.2.2.3 Fuel Handling Accident 14.2.2.3.1 Identification of Accident Spent fuel assemblies are handled entirely under water.
Before refueling, the reactor coolant and the fuel transfer canal water above the reactor are increased in boron concentration so that, with all control rods removed, the kef3 of a core is no greater than 0.99.
In the spent fuel storage pool, the fuel assemblies are stored under water in storage racks having a keff of less than 0.95 assuming no boron present.
Under these conditions, a criticality accident during refueling is not considered credible.
Mechanical damage to the fuel assemblies during transfer operations is possible but improbable.
The mechanical damage type of accident is considered the maximum potential source of activity release during refueling operations.
14.2.2.3.2 Analysis and Results The assumptions made for this analysis are shown in table 14.2-7.
The reactor is assumed to have been shut down for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is the minimum time for reactor coolant system cooldown, reactor closure head removal, and removal of the first fuel assembly.
It is further assumed that all 208 fuel pins in the hottest fuel assembly fail releasing all gap activity.
The gases released from the fuel assembly pass upward through the spent fuel storage pool water prior to reaching the Fuel Handling Building atmosphere.
As a minimum, the gases pass through at least 10 feet of water. Al though there is experimental evidence that a portion of the noble gases will remain in the water, no retention of noble gases is assumed.
In experiments whereby air-steam mixtures were bubbled through a water pond, Diffey, et.al.,(3) demonstrated decontamination factors of about 1,000 for igdjne.
Similar results f r Eggleton. 5) iodine were demonstrated by Barthoux, et.al.,141 and predicted by Based conservatively on these references, 99 percent of the iodine released from the fuel assembly is assumed to remain in the water.
The iodine and noble gas activities released to the pool water are given in Appendix 14D, table 14D.10-1.
The Fuel Handling Building is ventilated and discharges through 90 percent efficient charcoal filters to the unit vent.
fhe environmental effects of this accident are discussed in paragraph 14.3.5.
14.2-14
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