ML19269C680
| ML19269C680 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 02/07/1979 |
| From: | Vandenburgh D YANKEE ATOMIC ELECTRIC CO. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| WYR-79-10, NUDOCS 7902120071 | |
| Download: ML19269C680 (57) | |
Text
7 Proposed Change 158 k
Supplement No. 3 THIS DXUh!E'.T CO'lTAitlS FOOR QUAUTY PAGES M *P one 617 366-901 h
TwX 710-390-0739 YANKEE ATOMIC ELECTRIC COMPANY
'c 158-8 B.3.2.1
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.Yaux!, hlh Ih 20 Turnpike Road Westborough, Massachusetts 01581 es
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February 7, 1979 WYR 79-10 United States Nuclear Regulatory Commission Washir.; ton, D.C.
20555 Attention:
Office of Nuclear Reactor Regulation Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors
References:
(a) License No. DPR-3 (Docket No. 50-29).
(b) Yankee Atomic Electric Company Letter WYR-78-61 to USNRC, dated July 13, 1978, Proposed Change No. 158.
(c) Yankee Atomic Electric Company Letter WYR-78-61 to USNRC dated July 13, 19'78, Proposed Change No.158, Supplement 1.
(d) Yankee Atomic Electric Company Letter WYR-78-62 to USNRC dated July 13, 1978, Proposed Change No. 158, Supplement 2.
(e) Yankee Atomic Electric Company Letter WYR-78-80 to USNRC dated September 15, 1978, Additional Information, Supplement 2.
(f) Yankee Atomic Electric Company Letter WYR-78-82 to USNRC dated September 25, 1978, Additional Information, Supplement 2.
(g) USNRC Letter to Yankee Atomic Electric Company, dated October 6,19 78.
(h) Yatiee Atomic Electric Company Letter WYR-78-90 to USNRC dated October 18, 1978, Additional Information, Supplement 2
Dear Sir:
Subj ect: Crane Travel - Spent Fuel Pit Pursuant to Section 50.59 of the Commission's Rules and Regulations, Yankee Atomic Electric Company hereby proposes the following modification to Appendix A of the Operating License. This letter is Supplement No. 3 to Reference (b).
r/90212OO7 i
4 United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page J PROPOSED CHANGE: Revise Section 3.9.7, exceptions e and f, of the Technical Specifications to read:
e.
Temporary gate, and f.
Shielding panels.
REASON FOR CHANGE:
As outlined in Reference (b), Yankee Atomic Electric Company wishes to modify and upgrade the spent fuel pit by installing a stainless steel liner, constructing a division wall separating the fuel storage area from the fuel handling area, adding a redundant pump to the cooling system, installing structural supports for possible future spent fuel storage increases and modifying the cooling piping arrangement and enclosure building euperstructure. These changes must be carried out while continuing to store the spent fuel now in the pit.
To accomplish this, a temporary gate will be used to divide the pool roughly in half, permitting dewatering and construction on one side with storage on the other.
In conjunction with the gate, shielding panels are required to reduce the radiation in the work zone.
This proposed temporary change to the Technical Specifications is required to permit installation and removal of the temporary gate and the shielding panels. The previous temporary change to the Technical Specifications allowing transport of the gate support bracket and shielded work platform, Reference (d), may be rescinded since that work has been completed.
Stainless Steel Liner The location of the liner and connection details are shown on Figures PL-1 to PL-4.
The defined surfaces of the spent fuel pit will be lined with 1/4 inch thick type 304 stainless steel plates. Widths and lengths have been selected consistent with handling and installation restrictions and to minimize in pool construction time.
Af ter dewatering, unanchored backing strips will be set in place on the floor even with the bottom of the temporary gate support bracket.
The floor will then be built up with cementitious grout or cut down to the level of the top of the backing strips. The liner floor plates will be laid down, overlapping the backing strips, and will be joined with full penetration welds.
An overlay strip will cover the joint and be fillet welded to each liner plate, providing a double weld barrier to leakage.
The wall liner plates will be joined in a similar manner; however, the backing strips will be bolted in place with countersunk flat head anchors.
Again, high or low spots on the walls will be removed by cutting or grouting as necessary, to provide an adequate surface.
Mating liner plates will be installed with a 1/2 inch gap, fillet welded to the backing strip and butt welded together. An overlay strip will then cover the joint and be fillet welded to adjacent plates to develop the double weld barrier.
i United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page ?
Corner joints will be made by fillet welding an angle to mating liner plates and covering it with a flat strip fillet welded to the plates to achieve the double weld barrier.
Assembly to the temporary gate support bracket, which will become a part of the liner, will be achieved by prebending the attachment edge of the liner plate. The bent edge will be velded to the side of the support bracket.
An angle, not shown in the figures, is welded onto the bracket to cover the bracket anchor bolts.
A formed plate covering these joints will be fillet welded to the waterside surface of the support bracket and liner plate to achieve the double weld barrier.
All liner welds will be liquid penetrant tested.
The liner has been designed to provide a leaktight membrane in the fuel storage area. Design conforms to ASME Section III, Division 2, " Code for Concrete Reactor Vessels and Containments," which has been used for guidance.
In addition to deadweight and hydrostatic loads, a thermal load has been imposed corresponding to a 1500F maximum bulk pool temperature.
No loads are imposed on the liner by equipment or structures in the pool, as all loads are transmitted directly to the concrete by special anchors.
Division Wall The location and geometry of the division wall is shown on Figures DW-1 to DW-4.
It will be crected across the full width of the pool in the north end and, as a permanent structure, will isolate and allow dewatering of the fuel transfer area without affecting the fuel storage area. This will allow access for repair of the fuel transfer equipment and fuel transfer chute and will provide capability for cask handling, as discussed in Reference (b). The overall dimensions of the division wall will be approximately 16-1/2 feet wide, 37 feet high and 1-1/2 feet thick.
A3 foot wide slot in the center will extend from the top of the wall down to 21 feet above the pool floor. This will be used for transporting fuel from the fuel transfer side to the fuel storage side.
A fuel transfer gate is provided to close off this opening when required. When the gate is in place, four struts will be installed for additional support off the north wall as shown on Figure DW-1.
The wall will consist of a core of A36 steel wide flange members, spanning horizontally and vertically, and will be faced on both sides with stainless steel plates welded to the core member flanges. The wall will be filled with concrete and anchored on three sides. The design is in accordance, as applicable, with AISC Manual of Steel Construction, 7th Edition (1970), ACI 318-77, " Building Code Requirements for Reinforced Concrete" (1977) and ASME Section III, Division 2, " Code for Concrete Reactor Vessels and Containments" (1977).
Load combinations are based on criteria outlined in NRC Standard Review Plan 3.8.4, "Other Seismic Category I Structures". The loads
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 4 considered are deadweight, hydrostatic and thermal.
Seismic loads will be addressed under the Systematic Evaluation Program; however, additional lateral load capacity to 2000 psf has been provided based on conservative estimates of seismic requirements. The stainless steel faceplates on the storage side of the d vision wall will be welded to the wall and floor liner plates to provide a.aaktight membrane.
Fuel Transfer Gate The fuel transfer gate is a stainless steel structure that is placed in the central opening in the division wall. Dimensions of the gate are approximately 3 feet wide, 16 feet deep, and 6 inches thick. The basic design consists of a 3/8 inch thick face plate, on the storage side, welded to horizontal stiffeners that frame into perimeter end plates. A continuous inflatable seal is mounted on the gate to create a barrier allowing dewatering of the spent fuel upender area.
The gate is mounted with hinges and retainer pins on the fuel transfer side of the wall for support in both the closed and open positions. Details of the fuel transfer gate are shown on Figures FTG-1 and FTG-2.
The gate design conforms with the AISC Manual of Steel Construction, Seventh Edition, 1970. Loadings considered are the same as for the division wall.
Spent Fuel Rack Supports During erection of the liner and division wall, spent fuel rack support fixtures will be installed. The function of these fixtures is to transfer loads from the spent fuel racks through the liner and into the concrete structure.
Part of the fixtures are designed to support the present spent fuel racks. The rest of the fixtures are to provide capability for future expansion of spent fuel storage capacity.
The support fixtures to be installed for the present racks are cups anchored to the floor as shown on Figures RS-1 and RS-2.
The cup is stainless steel and machined to an inner diameter 1/4 inch larger than the foot on the spent fuel rack. As shown in Figure RS-1, the load is transferred from the cup to the concrete anchor through the connecting welds.
The welds also provide the double weld barrier against leakage. When a rack is set into position in the pool, each of its four feet will rest in a support fixture.
The supports for the possible future racks consist of structural framework as shown on Figures RS-3 to RS-8.
These supports are completely separate from the present spent fuel racks. However, on completion of the installation, grating will be placed above the existing spent fuel racks to provide additional protection against a fuel handling accident. The grating and supports have been designed to resist the impact of a fuel bundle
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 5 dropped from 11 feet above the existing spent fuel racks.
Connection to the pool walls is made by penetrating the liner as shown in Figures RS-7 and RS-8.
The double weld seal is provided using backing and cover plates at the penetration.
The design of the support fixtures conforms to the AISC Manual of Steel Construction, Seventh Edition,1970.
Analysis for dropped fuel bundle permits plastic defo mation but limits distortion to prevent contact with r
the spent fuel rack. Lateral strength to 0.5g static loading has been provided based on conservative estimates of seismic requirements.
Seismic loadings will be addressed in conjunction with the Systematic Evaluation Program and Supplement 4.
The concept for providing future storage capacity is described in Reference (b) and will be addressed in detail in Supplement 4.
Building Superstructure Reference (c) described modifications made to the spent fuel pool enclosure building to provide a central roof hatch.
An additional modification is being carried out to provide more floor area to accommodate changes in the cooling system layout.
As shown on Figure EB-1, the building will be expanded eastward over the roof of the new fuel vault.
The addition consists of a steel frame structure with metal siding and roof deck similar to that of the existing structure. The steel framing is supported on new columns bearing on the walls of the new fuel vault and spent fuel pit.
The addition will be erected outside of the existic; building, the new walls and roof connected to the old, and then the old east wall removed.
In this way building isolation will not be interrupted.
The primary function of the building addition is to provide space for cooling system equipment, mainly the new pump and the old pump which will be moved from its present location.
Secondarily, it will provide inside work space during the in pool construction so as to minimize removal of the roof hatches.
Temporary Cate The temporary gate consists of two pieces which when joined, form a wall 34 feet high dividing the pool into two parts. The manner of installation will be similar to that of the support bracket described in Reference (d).
The bottom section will be lifted with the yard crane and moved to the enclosure building central hatch. This will be accomplished without passing over stored spent fuel, either by moving it sideways along the east side of the crane runway or by bringing it from the end opposite that to which the fuel racks have been moved. Once in position, the piece will be lowered onto support beams which rest on the pit walls. The shear key on the gate will be fitted into the keyway in the support bracket.
t U,ited States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 6 The support beams will be the same ones used to support the bracket as described in Reference (d).
With the first section supported on the beams, the second piece is lifted, transported to the center hatch without passing over the fuel, and lowered into position. This is shown in Pictorial No. 1, " Gate Installation". With the crane still attached, the pieces are bolted together. Then the full gate is raised off the beaus and lowered to its final position. The final arrangement of the gate, shielding and fuel is shown in Pictorial No. 2, " Liner Installation".
Removal is the reverse of these steps.
Each section of the gate weighs approximately 14.1 tons.
The pieces vill be lifted using the 15 ton auxiliary hook, with redundant lifting cables to protect against failure of a sling. When the pieces are joined, they will be lifteo with the 75 ton main hook, again using redundant cables.
The larger capacity hook is not used for carrying the individual pieces for two reasons.
First, there is more clearance between the enclosure building roof and the smaller hook, providing enough room to move the sections without tipping them.
Second, the auxiliary hook can be moved farther to the east, making it possible to move the pieces outside the perimeter of the spent fuel pool.
In the initial design phase, an evaluation was made of using more gate sections, with each piece weighing less than those discussed above.
Iloweve r, it was felt tnat the greater number of lifts and joining operations resulted in a higher chance of a handling accident.
Because of the larger number of horizontal gate splices, there would be a greater chance of laakage during use.
Finally, calculations performed to evaluate the effect of dropping the load into the pool showed that it would not cause gross damage to the concrete structure. These calculations are described below under
" Safety Evaluation".
The temporary gate geomete; is sho"n on Figures TG-1 to TG-5.
It consists of two sections which wnen jotued will ce approximately 34 feet high,16 feet-4 inches wide, and 14 inches thick.
The sections are of equal size and will be joined mechanically at the center while coincidentally compressing a neoprene gasket seal.
Shear keys are welded to the end and bottom plates on three sides of the gate. When installed, the shear keys set into a slot in the support bracket to provide the mechanical support for hydrostatic loading. Scaling around the perimeter of the installed gate will be by redundant continuous inflatable seals mounted on the support bracket. Each seal has its own pressure regulator, check valve, and alarm so that loss of plant compressed air or one of the seals will not cause overall gate seal failure.
Gate sandwich construction consists of a core of wide flange members, full length horizontal and spliced vertical, with face plates on each water
United States Nuclear Regulatory Commission February 7,1979 Attention: Office of Nuclear Reactor Regulation Page 7 side welded to the member flanges and end plates to form a watertight conpartment. The compartment in the lower section will be flooded with clean water to provide shielding. The upper section will not be flooded.
The shear keys are 2 inches thick and are attached by full penetration welds to 2 inch thick end plates forming the sandwich closure.
The horizontal joint between the gate sections is fitted with a compressible seal for watertightness. The bottom gate section has fittings provided for pumping of f leakage collected in the space between the peripheral pneumatic seals.
Fittings are also provided for filling and draining the botto:a core.
Before shipment to the site, the gate sections will be pressurized to verify watertightness. The gate is designed to resist deadweight and hydrostatic loads in accordance with AISC Manual of Steel Construction, Seventh Edition, 1970.
Loads on the gate are transmitted to the concrete by the gate rupport bracket, shown on Figure SB-1, which was described in detail in References (d) and (e). The bracket is supported by anchor bolts set in the concrete.
Design was based on an allowabic bolt shear load of 20% of the average ultimate strength in the manuf acturer's tests. This is more conservative than the allowable load recommended by the manufacturer. Also, in the manufacturer's tests, the bolts failed before the concrete failed. The anchor embedment in the pit is greater than that of the tests and passes beyond the wall rebar.
Installation of the bolts was carried out in accordance with the manuf acturer's instructions and the bolts were grouted in with the epoxy grout described in References (d), (e), (f), and (h).
This grout closed off any spaces between the bolts and brackets to assure a slip free connection.
In the NRC evaluation of the temporary gate support bracket installation, Reference (g), approval of the epoxy grout was conditioned upon the results of Yankee's investigation into the presence of any heavy metals, halogens, or sulphur compounds which might be released as a result of long term irradiation. The results of this investigation have been submitted in Reference (h).
As stated therein, the only substances of concern are chlorides, which are present (at less than 0.3%) only as trace impurities in the epoxy resin. Knowing the amount of grout actually used, the proportion made up by the resin, and assuming instantaneous release of all chlorides into the pool water, it is found that the chloride content of the water will rise less than 1 ppm.
This does not represent a hazard to the spent fuel pit or the spent fuel assemblies and is conservative since any release of chlorides will be gradual and concentrations will be reduced as the pool water is continuously passed through a demineralizer resin.
As an additional check on the anchor bolts, a test was carried out in the pool prior to placing the bracket.
A test plate, shown on Figure SB-2, of the same thickness, width and material as the bracket was mounted on the pool wall.
Four anchor bolts were used, of identical size and
t United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 8 material as on the bracket, placed in the same bolt pattern. A space of one inch from the wall was imposed by using plywood shim plates. The bolts were installed underwater by divers using the same drilling method as for the bracket. Once the test plate was set, it was lifted incrementally to twice the allowable design loading and the bolts still maintained anchorage.
The plywood spacers were used to siuulate the worst case grout space, although the structural benefit of grouting in the bolts was not included.
There are two independent barriers to leakage between the bracket plate and the concrete.
Fi rs t is a rubber seal on the north side of the bracket. This seal is clamped onto the concrete and onto the bracket.
Second is the grout behind the plate.
The rubber seat is equipped with a hose fitting so that any bypass leakage can be pumped out.
The seal is clamped onto the plate by a channel running along the sides and bottom secured with threaded studs set in the bracket. This channel will also be used to stabilize the shielding panels described below. No seal is required on the south side of the bracket as the liner will serve this function when the north end is dewatered.
Shielding Because the spent fuel will be stored close to the temporary gate, it is necessary to employ special measures to reduce the radiation reaching the dry side.
First, the spent fuel must be positioned as far from the gate as possible and graded so the oldest fuel is closest to the gate.
Second, as much shielding should be placed between the fuel racks and the gate as is possible.
Finally, the construction must proceed quickly so that the number of spent fuel bundles in the pool is minimized.
Rack arrangements for the construction phases are shown on Figures RA-1 to RA-4.
Since Phase 1, support bracket installation, has already been completed, Figure RA-1 shows the current rack arrangement. The worst situation for radiation occurs when the fuel is all in the north end.
With racks deployed as originally described in Reference (b), fuel would be stored just 8 inches from the gate.
However, by using a 10 element rack next to the north wall, the first row of cavities can be emptied, providing an additional 11 inches distance.
This 10 element rack is shown in Figure R A-5.
It is made up of two existing 5 element racks and is braced from one of the 40 element racks. The rack will only be used during the construction work. Gaining the additional distance and water shielding will provide a significant reduction in radiation. When the fuel is stored in the south end and an additional 36 fuel bundles have been discharged in April, 1980, the space between the fuel and the gate will be approximately 21 inches.
If the north end work is completed prior to this discharge, the space will be about 43 inches.
During installation of the gate support brackets, measurements of radiation were taken close to the spent fuel in the pit.
This fuel had
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 9 been graded in a manner similar to that contemplated for use with the gate.
Using this data and the fuel arrangement for the north end described above, the radiation level on the dry side of the temporary gate is conservatively estimated to be 10 to 20 R/hr.
It is therefore necessary to provide three tenth-value layers of additional shielding between the spent fsel and gate.
The shielding used will be steel cased lead panels providing 4.75 inches of lead.
The sh elding panels are shown on Figures SB-1, RA-6, and RA-7.
Each panel weighs approximately 9.5 tons.
As is done with the temporary gate, each panel will be lifted using the 15 ton auxiliary hook with redundant cables. They will be brought to position without travelling over stored fuel.
The panels are made up of steel boxes filled with lead. The steel is 1/4 inch A36 plate. The lead is 4.75 inches thick and is isolated from the pool water by the steel box.
The joints between panels are V-shaped to reduce streaming and are so spaced as to lie adjacent to the flanges of the horizontal beams in the gate.
These flanges provide 2 inches of additional steel shielding across the joints. Lifting fixtures are provided and the steel box supports the lead during transport. The panels will bear on the gate support bracket and will be stabilized by the bracket seal channel on the sides.
Removing fuel from the first row, using the shield panels, and flooding the core of the lower gate section will provide enough shielding to lower the radiation to less than 20 mR/hr.
As an additional measure, concrete shield plugs, shown on Figure RA-8, will be placed in the empty first row cavities. Since these will be directly in front of the closest fuel bundles, they will be quite effective in further reducing radiation, but the actual amount cannot be accurately estimated. The shield plugs, which weigh less than a spent fuel assembly, will be padded to prevent scratching the aluminum cavities.
Before the water is pumped down, a thorough survey will be taken and additional shielding placed on the dry side as required to eliminate any zones higher than 20 mR/hr.
The final effort to reduce radiation lies in the construction schedule. Taking advantage of the progress prior to November, 1978, the schedule has been shortened so as to complete all liner installation before the next refueling outage in April, 1980. The schedule, as outlined before, is ambitious but will greatly reduce radiation levels because there will be 36 less fuel elements stored when the north end is dewatered. This will result in an extra 22 inches between the stored fuel and the gate.
Eeploying the same radiation shielding measures as above will result in insignificant levels on the dry side of the gate.
This construction schedule calls for shielding and gate installation in the pool to begin March 1, 1979.
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 10 Cooling System As described in Refer;uce (b), the cooling system will be changed to support the various construction phases and at job completion will be permanently altered. The major change will be the addition of a second pump of slightly larger capacity (600 gpm versus 500 gpm) which will be installed on the roof of the new fuel vault in the space provided by the addition to the spent fuel pit enclosure building. The present pump will be moved to the :.ew fuel vault roof also, and the suction and discharge of the two p_sps will be crcss-connected tc enable either one to take suction on either end of the pit.
The old and new pumps, designated P21-A and P21-B respectively, and the associated piping and valves are shown on Figure CS-1.
Five of the valves will be re-used from the existing system and the rest will be new.
The new arrangement will also have local level and temperature indication for both ends of the pool and local pressure indication on the discharge header of the pumps. High and low level alarms for both ends of the pool will be provided in the control room along with a low discharge header pressure alarm in the waste disposal building. Each pump will be powered from a dif ferent electrical bus, which is capable of being backfed from different power sources.
The new pump is designed and fabricated to the requirements of ASME Section Ill, Subsection ND.
The valves will meet YA-GEN-3, " Specification for Valves for Nuclear Use."
The new piping will be designed in accordance with ANSI B31.1, 1977, to meet the current system design criteria.
Seismic loadings for the cooling system will be addressed under the Systematic Evaluation Program. All piping and supports used in the temporary arrangements during construction which will not later be a part of the permanent system will also meet the requirements of ANSI B31.1 to ensure that the temporary cooling system will always provide a level of safety and reliability equal to that of the present system.
All the temporary arrangements during the construction phases will provide a Safety Class 3 cooling system with a Safety Class 3 pump. The system will remain connected into the shutdown cooling system for emergency backup. The cooling system must be shut down for brief periods when changeovers are made, but most changes will be made using flanged connections without the need for cutting and welding. Decay heat calculatione, described below under " Safety Evaluation", show that sufficient time will be available before the pool temperature exceeds 150 F.
Construction Sequence The four phases of construction are described in Reference (b).
The first phase, installation of the temporary gate support bracket has already been completed. The final three phases are detailed below.
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 11 Phase 2 (south end) 2/79-7/79.
1.
Install new pump P21-B and associated valves and piping up to point of connection to existing system.
Install north end suction piping using spool piece and put blind flange on south end suction. Move existing level indicator to new well in north end.
Install pressure indicator on P21-B discharge.
See Figure CS-2.
2.
Secure cooling on P21-A.
Make connection on P21-A discharge by rotating flanged elbow 90. Restore cooling on P21-B.
3.
Place shield panels into position on gate support bracket. Set concrete shield plugs into racks next to shield panels.
4.
Move spent fuel to north end.
Install pool cover.
5.
Install temporary gate.
Flood core of bottom section, inflate pneumatic seals and start pump on bracket edge seal.
6.
Dewater south end using submersible pump discharging over gate into north end.
Hold level in north end constant by bleeding off equal flow to gravity drain tank and processing through waste disposal system.
7.
Decontaminate south end.
Remove P21-A and associated valves, piping and instrumentation. Relocate skimmer pump and hydraulic drive for fuel transfer equignent.
8.
Erect liner and rack supports in south end.
Install new level and temperature instrumentation in south end.
If time allows, install pump P21-A in new location next to P21-B.
9.
Reflood south end. Water can be taken from either the demineralized water storage tank or the primary water storage tank.
10.
Remove gate and pool cover.
Phase 3 (north end) 7/79 - 1/80.
1.
Remove shield panels and reinstall on sou:h side of gate support bracket. Move concrete shield plugs.
2.
Move fuel to south end.
Install pool cover.
3.
Secure cooling in pump P21-B.
Close valve in north end suction, disconnect piping and put on blind flange. Connect south end suction piping.
Cut spool piece out of discharge piping to shorten for south end return, install flanges and bolt up.
Restart cooling on P21-B.
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 12 See Figure CS-3.
4.
Install temporary gate.
Flood core of bottom section and inflate pneumatic seals.
5.
Dewater north end in same way as Phase 2, step 6.
Decontaminate concrete surfaces.
6.
Install liner, rack supports and division wall in north end.
Install pump P21-A in new locar. ion next to P21-B, if not already done.
Install cru instrumentation in north end.
7.
Reflood north end.
8.
Remove temporary gate.
Phase 4 (final) 1/80 - 3/80.
1.
Move fuel away from center area.
2.
Remove shield panels.
3.
Using diver, connect up rack support beams across area previously occupied by temporary gate.
Remove pneumatic seals and hardware from temporary gate support brackets.
4.
Install pump priming lines and new discharge pressure indicator.
install alarms in control room.
5.
Secure cooling on P21-B.
Disconnect pool return, install flanged spool piece and reconnect.
Connect north end suction piping. Restart cooling on either P21-A or P21-B.
6.
Move fuel from north end to fuel storage area.
SAFETY EVALUATION:
The heaviest loads presently allowed over the spent fuel pit are the spent fuel racks which weigh up to 6.3 tons.
The shielding panels and gate sections will weigh approximately 9.5 and 14.1 tons respectively. Submerged, they will weigh 8.6 and 4.5 tons.
As mentioned before, these objects have been sized not only to minimize weight but also to minimize the amount of handling required. Movemcat directly over stored spent fuel is prohibited, and all handling will be in accordance with approved written procedures. To provide further assurance against a handling accident, redundant slings and lifting eyes are to be used on the gate sections and shielding panels.
Calculations have been made to assess the ability of the pool to retain water in the unlikely event of a drop from the crane. The same method
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 13 of analysis was used as that described in Reference (e).
The objects were assumed to drop f rom a position above the roof, through the hatch and onto the floor slab. The energy absorbing characteristics of the pool water and deformation and buoyancy of the objects were neglected. The objects were assumed to impact on a corner to simulate the worst orientation. Under these very conservative conditions, the limiting calculated concrete penetration is 15 inches. This indicates that the 36 inch thick floor slab will not be perforated, and the ability of the pool to perform its water retaining function will not be impaired.
A calculation has also been made to evaluate the off-site consequences of the massive rupture of spent fuel in the unlikely event of an accidental drop which results in impingement upon a loaded spent fuel rack. The calculation is based on the following:
1.
All 189 spent fuel assemblies stored in the pool are ruptured and release their gap activity.
2.
All fuel has decayed for 120 days since reactor shutdown.
3.
The fuel has an irradiation equivalent to 3 equilibrium core cycles at 600 Mwt.
4.
All fuel assemblies have a radial peaking factor of 1.65.
5.
No solid fission products are released.
6.
30% of the Kr85 inventory and 10% of the remaining noble gas inventory are within the fuel gaps and are released to the spent fuel pit water and subsequently to the fuel building atmosphere.
7.
10% of the halogen inventory is within the fuel gaps and is released to the spent fuel pit water.
8.
99.75% of the Iodine is inorganic while 0.25% is in the organic f o rm.
9.
The spent fuel pool decontamination factors for the inorganic and organic Iodine forms are 133 and 1, respectively.
- 10. The atmospheric dispersion factor for the exclusion area radius is 3.29x10-4 sec/m,
3
- 11. The air filtration system is not operating.
Based on a breathing rate of 3.47x10-4 3
m /sec, the resultant thyroid dose is 0.34 rem.
Using a semi-infinite source, the resultant whole body gamma dose is 0.05 rem.
These values are much less than the 10CFR100 limits of
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 14 300 rem ani 25 rem, respectively.
During the construction work, a pool cover consisting of steel decking over aluminum box beams will be placed over the storage area of the pool to prevent accidental entry of tools or debris into the water.
All pieces making up the pool cover weigh under 900 pounds and will be hand placed following approved written procedures. This cover will be the same one used during installation of the center roof hatch.
During the erection of the liner, rack supports and division wall, construction materials weighing more than 900 pounds but less than 3000 pounds will be lifted into the pool. However, these loads will be lifted only over the dewatered portion of the pool and lowered through the end hatch, and an accidental drop will not affect the water volume on the other side of the temporary gate.
These lifts will be made following approved written procedures and only when the temporary gate is in place. The materials are beams and plates and drop analyses onto the concrete are bounded by those of Reference (e).
The temporary gate serves to retain cooling water around the spent fuel during construction. The only active components are the pneumatic seals which are completely redundant, with separate air regulators and ala rms. An analysis was made tt ietermine the consequences of gross leakage past the gate. The volume of water stored on either side of the gate was calculated, assuming a water height of 33 feet which is the minimum permitted by Technical Specifications.
It was found that if the smaller volume, in the north end, suddenly equalized over the full pool, the resultant water level would be 16 feet, leaving 7 feet over the active fuel. This is sufficient to shield competely against radiation, and since no fuel will be handled when the gate is in the pool, the consequences of a fuel handling accident are not increased.
During construction, flanged spool pieces are to be installed in the cooling system suction piping to lower the inlet to elevation 1022'-
0", 14 feet above the pool floor.
In the event of water equalization, the inlet will be 2 feet below the water surface, but the pump will not have sufficient net positive suction head (NPSH) for operation.
Therefore, the shutdown cooling pump will be started, throttled to 500 gpm.
At the same time, refilling of the pool will be started.
Both the Demineralized Water Storage Tank and the Primary Water Storage Tank are available as makeup water sources, with the fire protection system as a third backup.
A calculation was made to assure the ability of the shutdown cooling pump to draw frcm the pool.
It was assumed that the barometr ic pressure was 14 psi and the initial water temperature 900F (it is currently 85-90 ).
The required NPSH at 500 gpm is 11 feet of water. The available NPSH is calculated to be 21 feet. Losses in the suction piping were evaluated using loss coefficients from Grane Technical Paper No. 410, Flow of Fluids Through
United States Nuclear Regulatory Commission February 7, 1979 Attention: Office of Nuclear Reactor Regulation Page 15 Valves, Fittings and Pipe.
If the shutdown cooling pump is not available, the pool will be filled until the spent fuel pit cooling pump can be used. Assmuing the initial pool water temperature is 900F, 650F fill water is supplied at 100 6
gpm and decay heat generation is 1.8 x 10 BTU / hour, it was found that the water temperature will rise to only 92 F before there is adequate NPSH.
Each of the Low Pressure Surge Tank Make-up pumps, which would be used to fill the pool, are sized to deliver 100 gpm at 236 feet of head. There are 37 feet of piping head losses from the Demineralized Water Storage Tank to the spent fuel pit.
The decay heat rate is from calculations, described below, made to evaluate how long the cooling system may be secured.
In the unlikely event of a break in the cooling system leading to syphoning of water from the pool, the water level cannot drop below elevation 1022'-0" which leaves 5 feet of water cover over the active fuel. At job completion, with the temporary spool pieces removed, the piping inlet will be at elevation 1032'-0", 24 feet above the floor. At all times, the cooling system return line outlet will be higher than elevation 1032'-0".
During the construction work, the cooling system will be shut down to connect or disconnect piping.
All the switchovers will use flanged connections to assure rapid completion of the work. During one shutdown, a spool piece will be cut from the discharge pipe and a flanged connection installed. This will be the longest operation, although c timated to take less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. There is adequate time available to accomplish this work, but in the event something happens which prevents completion, the piping will still discharge into the pool, enabling cooling to be restored.
The time available before restarting the cooling system was calculated to be at least 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, based on the following:
1.
The water temperature must not exceed 150 F, which is in accordance with ANSI N-210.
Calculations for a bulk pool temperature of 150 F have shown that water within the racks will remain under the boiling point.
2.
Water level at shutdown is 33 feet.
3.
Water temperature at shutdown is 90 F.
Temperature data collected immediately af ter the latest core discharge showed temperatures between 85 and 900 4.
Cores IX to XIII (189 assemblies) are in the spent fuel pit with Core XIII having approximately 130 days decay.
5.
The overall heat transfer coefficient of the spent fuel pit cooler 2
is 225 BTU /hr-f t - F.
This value has been verified from data collected
United States Nuclear Regulatory Commission February 7, 1979 Attention: Of fice of Nuclear Reactor Regulation Page 16 in 1977 on spent fuel pit temperature.
6.
Decay heat generation rates are from a computer program based on Branch Technical Position ASB 9-2.
The program calculates the decay heat contribution from each discharge stored in the pool and sums these to arrive at a total heat generation rate. The program has successfully duplicated decay heat generation rates presented in ASB 9-2 for a reactor operating time of 16,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
At one point during installation of the support bracket in October, 1978, the spent fuel pit cooling pump was shut down. Pool temperature was monitored during the shutdown, which lasted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and no perceptible change was observed.
This safety evaluation addresses the ef fect of the Proposed Change and the installation work on the safety related function of the spent fuel pit and spent fuel pit cooling system. The conclusion is that there is no increase in the probability of an accident (or equipment malfunction) or of an accident of a different type which has not been analyzed, and the margins of safety which have been defined in the bases of the Technical Specifications have
,t been reduced. The final presence of the new structures in the pool will represent a significant upgrade and increase in the margins of safety. This Proposed Change has been reviewed by the Nuclear Safety Audit and Review Committee.
FEE DETERMINATION: This Proposed Change is a part of the overall project described in Reference (b) which was determined to be a Class IV amendment and for which a payment of $12,300 has been submitted.
Yankee Atomic Electric Company believes that the Commission's revised fee schedule is illegal and has petitioned for review thereof to the U.S.
Court of Appeals for the Fifth Circuit. Accordingly, the license fee has been submitted under protes t and without waiver of the Company's right to contend that it should be refunded in whole or in part.
SCHEDULE OF CHANGE: To ensure this project remains on schedule, the work detailed herein will begin on April 1,1979 and terminate on April 1,1980.
We, therefore, solicit NRC approval of this proposed change before April 1, 1979.
United States Nuclear Regulatory Commission February 7, 1979 Attention: Of fice of Nuclear Reactor Regulation 9 age 17 We trust this information is acceptable to you; however, should you have any questions, please feel free to contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY MN~
D. E. Vandenburgh Senior Vice President COMMONWEALTH OF MASSACHUSETTS)
)ss.
COUNTY OF WORCESTER
)
Then personally appeared before me. D. E. Vandenburgh, who, being duly sworn did state that he is Senior Vice President of Yankee Atomic Electric Company, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Yankee Atomic Electric Company, and that the statements therin are true to the best of his knowledge and belief.
T m 2-
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REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL PIT LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 900 pounds shall be prohibited from travel over the spent fuel pit except for the:
a.
Spent fuel pit building roof hatches, b.
Spent fuel inspection stand, c.
Fuel handling equipment, d.
Spent fuel racks, e.
Temporary gate, a.d f.
Shielding pane's.
APPLICABILITY: With fuel assemblies in the spent fuel pit.
ACTION:
With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.7 Loads in excess of 900 pounds shall be prevented from traveling over the spent fuel pit by administrative control except that the:
a.
Spent fuel pit buildir roof hatches, the spent fuel inspection stand, the fuel handling equipment, spent fuel racks, the temporary gate and the shielding panels may travel over the spent fuel pit in accordance with approved written procedures.
b.
Spent fuel storage racks, the spent fuel inspection stand, the temporary gate and the shielding panels shall be prevented from traveling over fuel assemblies in the spent fuel pit by administrative control, and c.
Fuel handling equipment when moved for maintenance sr.all be prevented from traveling over fuel assemblies in the spent fuel pit by administrative control.
YANKEE-ROWE 3/4 9-8 Amendment No.
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