ML20235G491
| ML20235G491 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/22/1987 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20235G494 | List: |
| References | |
| NUDOCS 8709300070 | |
| Download: ML20235G491 (38) | |
Text
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Northern States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 Telephone (612) 330 5500 September 22, 1987 Director Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Attn:
Document Control Desk Washington DC 20555 PRAIRIE ISIAND NUCLEAR GENERATING PiANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Fuel Rod Consolidation Attached for your information is a copy of our project description and safety evaluation for the Spent Fuel Consolidation Demonstration Project.
The demonstration project has been found to not involve an unreviewed safety question or require an amendment to the Prairie Island Operating License. This safety evaluation is being submitted as requested by the NRC Staff at the NSP/NRC informational meeting held on August 14, 1987.
Several documents referenced in the Safety Evaluation are attached.
Preparation work for fuel consolidation began on September 16, 1987.
Our original schedule called for fuel rod consolidation to begin on October 1, 1987.
Preparation work is currently ahead of schedule and consolidation could begin as early as September 28, 1987.
Please contact us if you have any questions related to the information we have provided.
QJ kw S David Musolf Manager - Nuclear Support Services c: Regional Administrator III, NRC Sr NRR Project Manager, NRC Sr Resident Inspector, NRC G Charnoff State of Minnesota Attn: William Clausen, Brad Moore, Dr J W Ferman Attachments:
j!N Project Description - Spent Fuel Consolidation Demonstration Safety Evaluation - Spent Fuel Consolidation Demonstration (Revision 1)
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- h Topical Report - Fuel Consolidation Demonstration at Prairie Island f lI'$[
Nuclear Generating Station Internal Westinghouse Letter dated 9/3/87 Letter dated 9/18/87 to L McCarten (NSP) from J Smith (Nutech) 8709300070 070922 k
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l Page 1 of 18 Modification 861944 i
PROJECT DESCRIPTION SPENT FUEL CONSOLIDATION DEMONSTRATION I.
Introduction The current spent fuel storage capacity at Prairie Island will be exhausted in the 1994-1995 time frame.
The spent fuel consolidation demonstration is intended to determine if rod consolidation is a feasible method for increasing Prairie Island's capacity to store spent fuel.
II.
Scope This project will entail the consolidation of up to 50 fuel assemblies. The rods from two assemblies will be placed in a storage canister which will then be placed i
in a storage rack cell. The top and bottom nozzles will i
be stacked into bundles and will occupy two storage rack i
cells.
The cages leftover will be stored in rack cells for processing at a later date. The demonstration program will not involve removing any fuel assembly parts from the pool.
At a later date it is planned to process the cages and package any pieces that can be disposed of and ship them I
to a commercial burial site for disposal. The processing of the empty cages is not within the scope of this modification.
1 III. Description Rod consolidation involves removing the fuel rods from two assemblies and placing them into a canister that will fit in an existing spent fuel storage rack. In the process of transferring the fuel rods, they go from q
a loosely spaced square array in the fuel assembly (designed to optimize reactivity) to a tightly packed triangular array in the storage canister (Figure 1). In order to achieve consolidation ratios greater than about 1.5:1, it is necessary to transfer all the rods simultaneously into the rod storage canister. Otherwise,
Paga 2 of 18 Modification 86L944 bowed, twisted, and/or bent rods will prevent the last several rods from entering the canister. The key to the operation is the transition canister. This consists of a series of tubes, plates, and rollers which rearrange the fuel from it's original square configuration to the tight packed triangular array.
Equipment Description The equipment used for rod consolidation is listed below, followed by a brief description of each component.
Fuel Assembly Dismantling Station (FADS)
Storage Can Loading Statich Nozzle Stacking Station Debris removal, water filtration, and underwater TV l
systems l
Storage can and nozzle stack lifting tools A) Fuel' Assembly Dismantling Station (Figure 2)
At this station, the top and bottom nozzles are removed from the assembly, the fuel assembly is rotated to the upside down position, and the fuel rods are pushed out of the assembly and into the transition canister.
The following sub-acuemblies/
components make up this station.
- 1) Elevator / Rotator - This supports the fuel assembly during dismantling. It allows the fuel assembly to be raised and lowered for access to the rods. It also rotates the assembly for access to the bottom i
l nozzle.
- 2) Fuel Transition Canister - The rods are pushed from the assembly to the transition canister in preparation for loading into the storage canister.
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The transition canister consists of 179 tubes supported by a framework and support plates. At the top, the tubes are in the same array.and pitch as l
the fuel rods in the fuel assembly. At the bottom, l
the tubes are in the tightly packed array needed to l
load the rods into the storage canister.
1
- 3) Transition Canister Support Stand - This stand I
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7
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di' Page 3~of 18 3
Modification 86L944
[.: J pq supports the transition canister under.the fuel assembly,and also positions it where 'it can be lifted by an overhead crane.
- 4) Rod Push Tool - The rod push. tool is used to push.
~
the rods out of the fuel assembly andsinto'the transition canister. rods'are pushed'two at a~ time.
A guide plate. locates the tool over the fuel rods.
5). Nozzle Removal Tools - The top nozzle is removed by drilling out the weld between the nozzle and the t
guide tv.be sleeve. The bottom nozzle'is removed by drilling out the threads of the thimble screw that holds the bottom nozzle to the gui6g tube.
- a g
i B) Storage can loading station (Figures 3j 4)
~
.1 The fuel rods are transferred from the transition can to the consolidated rod storage canister (CRSC) at this station. The following is a brief description of the components at this station.
- 1) Winch - The winch is mounted on the deck, across the fuel transfer canal, and is used raise and lower the CRSC loading frame and.to move the frame laterally..
I
- 2) CRSC Loading Frame - The CRSC loading frame supports the fuel transition canister and the CRSC.
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- 3) Rod Position Indicator - The rod position indicator i
consists of 179 bars, where each of the bars m'
engages the top'of a fuel rod in the transition canister.
The position indicator served two purposes; it provides a downward force on the fuel l
rods to ensure that they enter the storage canister j.
as the storage canister loading frame is raised, K
and they provide a means of identifying any rods
/
that is not entering the storage canisuer.
- 4) Consolidated Rods Storage Canister (Figure 5,6) y The consolidated rod storage canister (CRSC) is designed to hold the fuel rods from two assemblies
.(358 rods) and yet fit into a fuel rack cell in the spent fuel pool. It is a.050 inch stainless steel structure.
A partition dividee the canister in half.
This allows the rods from one assembly to j
Page'4 of 18' modification 86L944 placed in at a time. The partition extends above the top of canister to serve as a lifting lug.
. Snap in covers are used on each side of.the lifting.
lug to close off the top end of the storage canister.- The covers each have'the assembly ID of the assembly they.are covering engraved on them.
the bottom of the storage canister includes a letdown pan which'the fuel rods rest on. The pan is also used to support the rods,during the earlier stages of_the process. It is initially installed in the bottom of the transition canister to support the rods as they are pushed from the fuel assembly.
It is transferred-from the transition canister to the storage canister with the rods, where it remains.
- 5) Letdown Pan' Support Columns - When the transition r
canister is in the storage canister loading frame, the letdown pan rests on'the columns. The columns are two lengths of 2 1/2" piping which extend through two holes in the bottom of the canister.
There are two holes on each side of the canister
-partition.
C) Nozzle Stacking Station The nozzles are stacked on top of each other as they are removed from the fuel assemblies (Figure 7). The top nozzles are stacked separately from the bottom nozzles, in-groups of ten per stack. Tie rods align the nozzles and hold the completed stacks together.
The completed stacks will fit into a rack cell one stack upon another.
l 1
l D) Debris Removal, Water Filtration and Underwater TV Systems An underwater vacuum system will be provided by l
Westinghouse to remove debris that will be created l
1..
during the process, It is expected that the most
)
debris will be as a result of the drilling operations
]
to remove the top and bottom nozzles.
]
Water filtration will be provided by the Spent Fuel E
Pool Portable Filtration System (SFPPFS) already in use in the spent fuel pit. This system will be i
0-
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q Page'5 of 18 Modification 86L944 positioned so as to maintain water clarity without inhibiting any operation involved (i.e.~use of the bridge crane).
TV systems will be used to monitor various aspects of the process, mainly at the fuel assembly dismantling station, and also at the storage can loading station.
E) Storage Can and Nozzle Stack Lifting Tools Special tools designed to lift the storage canisters and nozzle stacks have been built. These tools are the property of NSP and will remain on site permanently after the completion of the demonstration.
Process Description The consolidation equipment will be set up in the Unit 2 end of the fuel transfer canal. A layout of the equipment locations can be seen in Figure 8.
Once an assembly has been selected for consolidation, the basic process steps are as follows.
- 1) The assembly is loaded and secured into the Fuel Assembly Dismantling Station (FADS)
- 2) The top and bottom nozzles are detached from the assembly.
The top nozzle is detached by using a drill to cut the welded area between the nozzle and the guide tube sleeves. The bottom nozzle is separated from the assembly by inserting a long drill bit down through the guide tubes and drilling out the threads of the thimble screw.
- 3) The top nozzle is removed from the assembly and transferred to the nozzle stacking station.
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- 4) A catch grid is placc1 over the top of the assembly to prevent the rods from f alling out of the assembly prematurely.
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- 5) The fuel assembly is rotated to the upside down
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position and the bottom nozzle is removed to the nozzle stacking station.
)
- 6) A guide plate is placed on the assembly where the
Page 6 of 18 Modification 86L944 bottom nozzle used to be.
- 7) The fuel transition canister is rolled under the assembly.
- 8) The catch grid is released and the fuel rods are pushed into the transition canister two at a time.
- 9) The transition canister is placed on support columns that pass through through the bottom of and extend to the top of a consolidated rod storage canister (CRSC) in the consolidated rod storage canister loading frame.
- 10) The rod position indicator is placed over transition canister.
- 11) The CRSC loading frame is raised. This lifts the i
Transition canister off the rods and pulls the CRSC over the rods. The rod position indicator above and the support columns below prevent the rods from moving up or down.
- 12) The CRSC is lowered off of the support columns and a cover is placed over the fuel rods.
This process is then repeated with another assembly whose rods are placed in the other side of the CRSC.
IV. Organizations Involved Westinghouse has been contracted to provide the equipment and personnel to consolidate the fuel. The contract calls for a demonstration program consisting of up to 50 fuel assemblies, with options for a total of 1000 assemblies if NSP decides to pursue consolidation further.
l Westinghouse has also been contracted to build a thimble grip handling tool to move the Region D, E, and F j
assemblies which have suspect bulge joint connections. In 1
addition, Westinghouse was selected to perform the criticality, thermal hydraulic, and radiological safety j
analysis for this project.
I NSP's Special Nuclear Programs department has the responsibility for overall project management of the consolidation demonstration. Their duties include vendor I
t
_b
Page 7 of 18 Modification 86L944 selection, contract administration, design review, and serving as the NSP contact with outside organizations.
NSP's Fuel Supply department administers the Interim Spent Fuel Storage Fund that is paying the expenses associated with the demonstration. All purchase orders are issued by the fuel supply department.
NSP's Nuclear Support Services department acts as the company liaison with the Nuclear Regulatory Commission and provided review of all safety analysis and safety related determinations.
NUTECH was contracted to perform structural analysis of the spent fuel pool and spent fuel racks.
NSP's Nuclear Analysis Department performed design reviews of the criticality and thermal hydraulic safety analysis.
Fluor Engineers, Inc. performed a design review of the spent fuel pool and spent fuel rack structural analysis.
Utility Associates International performed a design review of the thermal hydraulic and radiological safety analysis NSP's Power Supply Quality Assurance will be performing an audit of the consolidation demonstration throughout the process.
V.
, Engineering Considerations A list of engineering requirements was generated by NSP's Special Nuclear Programs department prior to soliciting competitive bids from various vendors. This document 1
specified the requirements NSP placed on the design and operation of the consolidation equipment.
VI. Materials, Codes and Standards All work performed by Westinghouse is in accordance with the current draft of ANS-57.10, " Design Criteria for Consolidation of LWR Spent Fuel". The design and operation of the equipment complies with the Occupational Safety and Health Act (OSHA), ANSI N14.6, and NUREG-0612.
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1
Page 8 of 18 Modification 86L944 The lifting equipment has been designed in.accordance with ANSI N14.6-1978, American National Standard for Special Lifting Devices for Shipping Containers 10,000 pounds (4500 Kg) or more for Nuclear Materials" Westinghouse has generated a list of acceptable materials and fluids which has been reviewed by the Radiation Protection group at Prairie Island. These materials were selected in accordance with ANS-57.10 which, in turn, specifies materials per the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF, or AISC-5326 1978, " Specification for Design, Fabrication and Erection of Structural Steel for Buildings" Structural design is based upon the codes and standards referenced in ANS 57.10. Allowable stresses and strengths of section members or connections are given in ANSI /ANSC N690, " Specifications for Design, Fabrication, and Erection of Safety Related Structural Steel in Nuclear Service" The criticality, thermal hydraulic, and radiological accident analyses were performed in accordance with several standards. These standards are as follows; ANSI N18.2-1973
" Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants",
Section 5.7, Fuel Handling Systems ANSI 57.2-1983
" Design Objectives for Spent Fuel Storage Facilities at Nuclear Power Stations", Section 6.4.2 ANSI N16.9-1975
" Validation of Calculational Methods for Nuclear Criticality Safety" USNRC Standard Review Plan, Section 9.1.2,
" Spent Fuel Storage" ANSI 8.17-1984
" Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors" USNRC, Letter to All Power Reactor Licensees, from B.K.
Grimes, April 14, 1978, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications"
Page 9 of 18 Modification 86L944 NUREG-0800, Branch Technical Position ASB-9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev.
2, July 1981 ANS-57.10, " Design Criteria for Consolidation of LWR Spent Fuel", Section 6.11.1, October, 1986 USNRC Regulatory Guide 1.25 USNRC Regulatory Guide 1.109 VII. Construction / Plant Interfaces The major impact the rod consolidation will have on plant operations is the demand for plant personnel to support the project. Consolidation activities will be on a.two-ten hour shift, six day a week schedule. An operator and shift supervisor will be needed at all times for movement of fuel.
Health physics personnel will be needed for radiation protection during the project.
A nuclear engineer will be on site at all times that consolidation is in progress.
The impact on plant operations has been minimized to the extent possibic by planning the demonstration during a time frame when no other operations are planned in the spent fuel pit. The demonstration will be completed and Westinghouse off site before the next shipment of new fuel arrives on sits, currently scheduled for early December.
VIII. References
- 1. " Objectives for a Fuel Rod Consolidation Demonstration at northern States Power (NSP) i Prairie Island (PI) Plant", Northern States Power -
Special Nuclear Programs Department
- 2. Topical Report, " Fuel Consolidation Demonstration at Prairie Island Nuclear Generating Station", Northern States Power Company
- 3. NSP Fuel Consolidation Equipment, Preliminary Design Review Package
- 4. NSP Fuel Consolidation Equipment, Intermediate
O Page 10 of'.18' Modification.86L944 Design Review Package
~5. ANS-57.10,." Design Criteria for Consolidation of LWR Spent Fuel", Final Draft, October 1986 6.. Northern States Power Company Purchase Order E28226FB
, submitted By: (
A 7-I b7 Reviewed By: _M fA L -Et h P)]
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b Page 1 of 19 Modification 86L944 Revision 1 SAFETY EVALUATION SPENT FUEL CONSOLIDATION DEMONSTRATION
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The following safety concerns and their resolutions will be the basis for determining that the spent fuel consolidation demonstration can be performed safely.
I.
Criticality Analysis of Consolidation Process II.
Criticality Analysis of Consolidated Fuel Storage in Racks III.
Criticality Analysis of Postulated Accidents IV.
Thermal Hydraulic Analysis of Consolidated Fuel Storage Canister V.
Spent Fuel Pool Cooling System Evaluation VI.
Radiological Analysis of Handling Accidents VII.
Mechanical, Material, and structural Considerations VIII. Structural Evaluation of Spent Fuel 2 col IX.
Structural Evaluation of Spent Fuel Storage Racks X.
ALARA 4
I.
Criticality Analysis of Consolidation Process This analysis (Reference 1) was performed to ensure that criticality will not occur at any time during the consolidation process. During the consolidation process, the fuel configuration will change significantly from that of a fuel assembly. To bound these fuel configurations from a criticality basis, the following fully or partially loaded configurations were considered under normal operation conditions:
a) Fully and partially loaded Fuel Transition Canister (FTC) i l
1 i
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l Page 2 of 19 l
Modification 86L944 Revision 1 b) Fully and partially loaded Consolidated Rod Storage Canister (CRSC) c) Fully and partially loaded fuel assembly j
l The design hasis for preventing criticality outside the
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reactor is that, including uncertainties, there ic h 95 j
percent probability at a 95 percent confidence level that the 1
effectivo multiplication factor (Keff) cf the fuel array will
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be less than 0.95, as recommended in ANSI 57.2-1983.
j Fuel Transition Canister The following assumptions were used in the criticality analysis of the fuel transition canister (Reference 1) :
1
- 1) The fuel esntains the highest enrichment authorized l
(4.0 w/o), is at its most reactive peint in life.,
J and no credit is taken for any burnable poison in the fuel rods.
Calculations have shown that the Westinghouse (W) 14x14 STD fuel pins yield an equal or larger Keff than does the W 14x14 OFA or the EXXON HI-PAR, LOW-FAR, or TOPEOD fuel when when all the pinc have the same enrichment.
Thus, only the W i
14x14 STD fuel pins kere analyzed.
i I
- 2) The moderator is pure water at a temperature of 68 F l
A conservative value of 1.0 gm/cc is used for the j
water density.
- 3) No credit is taken for any spacer grids or spacer sleeves.
- 4) The arrays are infinite in the axial extent which I
precludes any neutron leakage from the e.nds of the array.
i The results of these analyses show that for the fuel i
trarisition canister, the maximum Keff, including i
uncertainties at a 95/95 probability /confidance level, is 0.8940.
This meets the design limit of Keff < 0.95 without placing any restrictions on the process.
1 CRSC _and Fuel Assembly f
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Page 3 of 19 i
Modification 86L944 l
Revision 1 f
i As a result of the similar sizes of the fuel assembly and J
CRSC fuel envelopes, the fully and partially loaded fuel
)
assembly and CRSC cases can be bounded by the same case, j
This analysis (Re%6rence 1) used the same assumptions as those used for the Fuel Transition Canister, with the
(
exception that the initial enrichment was taken as 3.0 w/o.
1 In order to meet the design limit for fuel assemblies with
]
. initial enrichments greater than 3.0 w/o, credit was taken for the reactivity decrease caused by fuel depletion. This method is known as reactivity equivalencing, and is described in detail in section 3.4.3.2.1 of Reference 1.
The result.of this analysis is that the maximum Keff, including uncertainties at a 95/95 probability / confidence level is 0.9112. Therefore, the acceptance criteria for criticality are met for consolidation of spent fuel in the CRSC for fresh fuel at 3.0 w/o initial enrichment, or fuel with greater than 3.0 w/o initial enrichment as long as it has a minimum amount of burnup, as defined by the graph shown on Figure 1.
j
(
II. Criticality Analysis of Consolidated Fuel Storage in Racks This criticality analysis (Reference 1) was performed to show
)
that fully or partially loaded consolidated rod storage i
canisters can be stored in the existing fuel stcrage racks.
j Che following assumptions were used in this analysis:
l
- 1) The fuel contains the highest enrichment authorized (4.0 w/o), is et its rest reactive point in life, and no credit is taken for any burnable poison in the fuel rods.
Calculations have shown that the Westinghouse (W) 14x14 STD fuel pins yield an equal er larger Keff than does the W 14x14 OFA or the EXXOM HI-PAR, LOW-PAR, or TOPROD fuel when when all i
the pins have the same enrichment.
Thus, only the W 14x14 STD fuel pins were analyzed.
- 2) All fuel rods contain uranium dioxide at an enrichment of 4.0 w/o U-235 over the infinite length of each rod
L l
Page 4 of 19 Modification 86L944 Revision 1
- 3) No credit la taken for any U-234 or U-236 in the fuel, o
nor is any credit taken for the buildup of fission product poison material.
- 4) The moderator is pure water at a temperature of 68 F.
A conser:Jative value of 1.0 gm/cc is used for the j
density of water.
i
- 5) The arre.y it infinite in the axial and radial extent which precludes any neutron 16akago from the array.
- 6) The minimum poiGon material loading built into the racks (i.e.
0.04 grams B-10 square centimeter) is -
used throughout the array.
l This calculation resulted in h Keff of 0.8419 with a 95/95 percent probability / confidence level uncertainty of 1 0.0037.
A second case was considered which considered mechanical and material thickness tolerances resulting from the manufacturing process in addition to asymmet.ric positioning of the CRSC within,the storage cells. This calculation resulted in a maximum Xeff of 0.9177 at a.95/95 percent probability / confidence. level. This value is within the 0.95 acceptance criteria for criticality.
A sensitivity analysis was performed to determine the minimum number of fuel rods that can be placed in the CRSC at a uniform pitch and meet the spent fuel rack Keff limit of O.95. Results of this study show that down to 113 fuel rods can be placed in each half of the CRSC for a total minimum of 226 fuel rods in a canister. Calculations show that the most reactive configuration is with fuel rods in both sides of the canister.
i III. Criticality Analyais of Postulated Accidents Accidents can be postulated which would increase reactivity through the unco.ntrolled spreading of the fuel rods. This uncontrolled spreading could occur as a result of spillage from a damaged or mishandled canister, or by misloading of the transition canister or the CRSC. These accident
Page 5 of 19 Modification 86L944 Revision 1 conditions are bounded by the uncontrolled release of two assemblies worth (358) of fuel rods. At no time are more rods than this being handled during the consolidation process. The maximum Keff that can result from the uncontrolled release of 358 fuel rods at 4.0 w/o enrichment with no burnup in cold unborated water is 1.1508 with an uncertainty of + 0.0025 (Reference 1).
For these accident conditions however, the double contingency principle of ANSI 8.17-1984 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool can be assumed as a realistic initial condition since not assuming it's presence 1
would be a second unlikely event.
The presence of 1000 ppm boron in the pool water will decrease reactivity by approximately 30 percent delta K to 0.850.
Thus, for postulated accidents, should there be a reactivity increase, Keff would be less than or equal to 0.95 if the boron concentration in the pool water is greater than or equal to 1000 ppm.
Accidents can also be postulated which may damage fuel rods and result in the release of fuel pellets during the consolidation operation. Generic studies by Westinghouse (Reference 8) have shown that more than 100 pounds of Uranium Dioxide pellets at 3.0 w/o are needed to form a critical mass 1
in unborated water. As a result, small releases of fuel l
debris from broken fuel rods will not Keff to exceed 0.95.
Large accumulations of fuel debris (>100 lbs) will most likely not cause Keff to exceed 0.95 with the consideration of the soluable boron in the water, but would have to be l
evaluated on a case by case basis.
IV. Thermal Hydraulic Analysis of, Consolidated Fuel Storage Canister An analysis was performed to evaluate the thermel hydraulic performance of the consolidated rod storage canister (Reference 1).
This analysis determined the minimum amount of cooling time required for a fuel assembly to meet the i
1
1 Page 6 of 19 j
Modification 86L944 i
Revision 1 design criteria.
This criteria is defined in section 6.11 o#
Reference 7 as "Under normal conditions the water temperature j
shall not result in boiling at anytime during the process".
This criteria also requires that an assumption be made as to the maximum bulk water temperature, which then must be made a requirement during normal conditions.
The mechanism for cooling the fuel rods in the canister is natural circulation. The decay heat of the fuel rods heats the fluid,. creating buoyancy forces which produce flow through the canister. The higher the decay heat level, the greater the flow. However, the canister fluid exit l
temperature increases with increasing decay heat level faster I
than the flow rate. Consequently, the decay heat level must be-at or below a certain limit before boiling can be prevented.
The analysis performed assumed a bulk water temperature of 150 F and a boiling temperature corresponding to boiling point of water at an elevation corresponding to the top of the CRSC at rest in its storage position in a rack cell.
This elevation is approximately 26 ft below the surface of the spent fuel pool, and the corresponding boiling point is 241.2 F. This calculation determined the minimum cooling time required for a fuel assembly as a function of the amount of time the assembly spent in the core. This requirement is shown graphically in Figure 2. These results also apply.to the case during the CRSC transfer process, where the CRSC actually have as little as 10 feet of water cover. For this condition, the analysis places a more restrictive limit on the pool bulk temperature to be less than 120 F.
For accident conditions (total loss of pool cooling), the water temperature would rise until boiling at the pool surface began. For this condition, a conservative check on canister thermal hydraulic conditions determined that the maximum fuel cladding temperature would not exceed 263 F.
This is well below the normal fuel operating temperatures.
The canister thermal hydraulic analysis was also used to determine the amount of flow area that was required at the bottom of the canister. This result was then transmitted to the canister designer, who then ensured that the design met
l l
Page 7 of 19 Modification 86L944 i
Revision 1 j
I these requirements.
V.
Spent Fuel Pool Cooling System Evaluation The number of assemblies in the spent fuel pit will not change as a result of the rod consolidation demonstration.
The maximum number of assemblies allowed to be stored
)
in the spent fuel pool is defined as 1386 assemblies, not j
including those assemblies which can be returned to the reactor (T.S. 5.6.D). The current spent fuel pool cooling systen evaluation (USAR 10.2.2) assumes that the pool contains the maximum number of fuel assemblies allowed under the operating license.
Since the consolidation demonstration will not alter the number of. assemblies stored in the pool, j
nor will it alter the heat removal capacity of the cooling system, the current analysis bounds all conditions that will occur as a result of the demonstration and no additional analysis is required.
VI. Radiological Analysis of Handling Accidents Radiological analyses were performed to determine the thyroid and whole body doses at the exclusion area boundary resulting l
from postulated drop accidents involving a consolidated rod storage canister (CRSC) and comparing the results with previously analyzed accident conditions (Reference 1). The following two accidents were analyzed:
- 1) Canister drop on floor
- 2) Canister drop onto storage racks The postulated accidents were analyzed only for the fuel handling building (not containment) since that is where fuel consolidation will take place, and the consolidated canisters will never enter containment.
l Canister Drop on Floor The assumptions used in this analysis are as follows i
l L________
i Page 8 of 19 Modification 86L944 Revision 1
- 1. The canister contains 358 fuel rods (from two assemblies of 179 rods each).and all fuel rods suffer clad damage in the drop.
2.
For the postulated accident there would be a sudden i
release of the gaseous fission products held in the void space between the pellets and the cladding.
- 3. The activity in the void space of the fuel rods in the canister is based cn1 the core activities given in Table
{
A.4 of Section 14A of the Prairie Island Updated Safety Analysis Report. In addition, the assumptions in USNRC Regulatory Guide 1.25 (i.e., 30% Kr-85, 10% other noble gases and halogens in the gap) are used. A radial peaking factor of 1.70 is used, which is conservative compared to the value of 1.65 allowed by Reg Guide 1.25.
- 4. The accident x/Q value 6.5 E-4 sec/sq m is used. This is the same x/Q value which is used in the fuel handling accident described in the USAR.
- 5. A decay period of 2 years is used.
- 6. Dose Conversion factors from Reg Guide 1.109 are used (Table B-1 for noble gases).
- 7. Retention of noble gases in tue fuel pool is negligible (i.e., DF-1).
Under these assumptions, an activity release of 3.308 E+3 curies of Krypton-85 occurs (all other noble gases and halogens decay to negligibic levels in 2 years). This activity release results in the following doses at the exclusion area boundary:
Thyroid Dose = 0 Beta Skin Dose = 91.4 millirem Gamma Body Dose = 1.1 millirem Table 1 compares the radiological effects of a canister drop with a freshly discharged fuel assembly drop, with the
l l
i Page 9 of 19 Modification 86L944 a
Revision 1 i
10CFR100 guidelines, and with the NUREG 0800 dose limits. As the table shows the whole body dose due to the canister drop is well below the guidelines of both 10CFR100 and NUREG-0800.
There is no thyroid dose from the canister drop because there
(
is no Iodine present after two years decay.
Canister Drop onto Storage Racks
{
A For a postulated accident involving a canister dropping onto a freshly discharged assembly, assuming that all the pins in the canister and the assembly are damaged, the radiological effects would still be well below both the 10CFR100 and the NUREG-0800 dose limits. This is shown by combining the results of line 1 and 2.
VI.
Mechanical, Material, and Structural Considerations All work performed by Westinghouse is in accordance with the current draft of ANS 57.10," Design Criteria for Consolidation of LWR Spent Fuel". The design and operation of the equipment complies with the occupational Safety and Health Act (OSHA),
ANSI N14.6, and NUREG-0612.
The lifting equipment has been designed in accordance with ANSI N14.6-1978, "American National Standard for Special Lifting Devices for Shipping Containers 10,000 pounds (4500 Kg) or more for Nuclear Materials".
Westinghouse has generated a list of acceptable materials and fluids. These materials were selected in accordance with ANS 57.10 which, in turn, specifies materials per the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF, or AISC-S326 1978, " Specification for Design, Fabrication and Erection of Structural Steel for Building".
Structural design is based upon the codes and standards referenced in ANS-57.10. Allowable stresses and strengths of section members or connections are given in ANSI / ANSC N690,
" Specifications for Design, Fabrication, and Erection of Safety Related Structural Steel in Nuclear Service".
VIII. Structural Evaluation of Spent Fuel Pool
I Page 10 of 19 Modification 86L944
. Revision 1 An evaluation is being performed on the structural design of the spent fuel storage racks that assumes that all existing rack locations contain consolidated storage canisters.
However, only a maximum of 25 consolidated storage canisters I
will be stored in the pool as part of the demonstration project. Since only 25 consolidated canisters will be stored f
in one rack with the remaining cells empty, the current licensing basis, as described in Reference 2, is still applicable and bounding because the total dead weight and the mass of the rack are no more than the corresponding values assumed in the current analysis. The floor and supporting wall stresses that were documented in the report continue to be maximum bounding values.
IX.
Structural Evaluation of the Scent Fuel Storace Racks Structural analyses were performed on the spent fuel storage racks (Reference 9) to ensure that the demonstration project satisfies the design criteria stated in USNRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications (Reference 10). These analyses assume that the storage canisters will be loaded into the racks in such a manner as to maintain the center of mass of the spent fuel as close to the center of rigidity of the rack as possible. An example of how this can be done is shown in Figure 3, In addition, the total mass added to the rack used to store the canisters will not exceed the mass of 56 fuel assemblies.
The current Licensing basis structural evaluation of the spent fuel storage racks is documented in QUAD-1-79-509,
" Licensing Report for Prairie Island Nuclear Generating Plant Units 1 and 2 Spent Fuel Storage Modification" (Reference 11). This report assumes a maximum of 56 spent fuel assemblies stored in a 7x8 rack. The sliding, dead load, seismic, overturning and rack to rack impact analyses documented in this report remain the bounding, licensing basis analyses, since the demonstration project will be controlled so that the maximum mass added to the rack will not exceed maximum mass added in these analyses. The purpose of minimizing the eccentricity between the center of mass and the center of rigidity of the rack is to ensure that the current analyses remain bounding for the demonstration.
Page 11 of 19 Modification 86L944 Revision 1 The current analysis does not address the local impact forces and associated stresses that would occur between the consolidated storage canister and rack storage tube. Due to a maximum possible gap of.393 inches between the canister and storage tube, additional loading caused by impact of the canister against the storage tube may be generated during a postulated Safe Shutdown Earthquake (SSE). The referenced OT position states that "the additional loads... may be determined by estimating the kinetic energy of the fuel assembly. The maximum velocity of the fuel assembly may be estimated to be the spectral velocity associated with the natural frequency of the submerged fuel assembly. Loads thus generated should be considered for local as well as overall effects on the walls of the rack and the supporting framework. It should be demonstrated that the consequent loads on the fuel assembly do not lead to a damage of the fuel".
The impact loading between the canister and storage tube is calculated by equating the kinetic energy of the storage canister to the strain energy of the storage tube, assuming the storage tube to be a simply supported beam with the impact loading being a uniformly applied load along it's length. The impact load was calculated to.be 5.67 lbs/ inch.
Based in this impact loading, a maximum through wall bending stress of 13,500 psi was calculated for the consolidated canister. Since the yield stress for A240 TP304 stainless steel is 24,700 psi at a maximum pool temperature of 212 F, the maximum canister stress due to the impcet is 55% of yield. Also, the canister impact will cause additional stresses in the storage tube. The maximum bending stress in the storage tube due to a SSE, including the effects of canister impact, is 2507 psi, or 10% of the yield stress of 24,700 psi. Because the stresses due to impact in the canister and storage tube wall are substantially below the yield stress, they will continue to perform their function of safely containing the spent fuel pins.
X. ALARA Considerations Various features of the consolidation process were j
incorporated in order to minimize the radiation exposure to l
l L---_---- -
l i
Page 12 of 19 Modification 86L944 Revision 1 those involved. Materials used were chosen minimize their effect on pool water chemistry and to minimize the amount of decontamination work required at the end of the demonstration.
i All work will be performed with the fuel assemblies at least i
5 feet underwater. This value is based on Prairie Island's experience with fuel reconstitution. This 5 feet of shielding j
will be ensured by a combination of equipment design and administrative controls on the pool water level.
The consolidation equipment includes an underwater vacuum system that will be used to collect debris that will form as a result of the process. The vacuum collects the debris in a filter made of 25 mil screening that will be stored in the spent fuel pit for future disposal. Also, the Prairie Island Spent Fuel Pool Portable Filtration System (SFPPFS) will be configured to maintain the water clarity around the work area.
l Limiting Conditions for Rod Consolidation i
The safety analyses performed for the rod consolidation used several key assumptions which place limiting conditions on the process that must be maintained at all times to ensure the safety of the process and the validity of this safety evaluation. Some of these conditions place restrictions on f
which assemblies may be chosen for consolidation, while others place restrictions on the actual consolidation process. The following is a summary of each of these restrictions, their basis, and the method that will be used to ensure that they are met.
1.
Fuel with an initial enrichment of greater than 3.0 w/o l
U-235 shall have a minimum amount of burnup, as defined by i
Figure 1. This is to prevent criticality in both the fuel assembly and storage canister. This condition will be verified before any fuel assembly is chosen for consolidation.
- 2. The consolidated rod storage canister shall have a minimum
]
of 113 fuel rods loaded into each half. This condition l
1 I
Page 13 of 19 Modification 86L944 Revision 1 ensures that the canisters will not achieve a critical l
condition at any time in the fuel storage racks. This will i
be verified for each canister as it is loaded.
- 3. The fuel shall have been out of the reactor for a length of time proportional to the time it spent in the reactor i
at power, as defined by Figure 2.
This is to ensure.
I that boiling will not occur in the canister during normal pool conditions. This condition will be verified before any fuel assembly is chosen for consolidation.
- 4. The spent fuel pool shall be maintained at a boron concentration of at least 1000 ppm boron during consolidation or when moving a loaded consolidated canister. This is'to prevent criticality during an y
accident.that could spill the fuel rods out of the CRSC.
i For the demonstration, NSP has made a commitment to the NRC that the boron concentration will be maintained at a minimum of 1800 ppm. This will be verified by water sample chemistry on a weekly basis and recorded.
- 5. The SFP temperature shall be maintained at or below 120 F j
during consolidation. This will ensure that there will be J
no boiling in the storage canister during normal pool conditions. This condition will be verified and recorded at the beginning of each shift during consolidation.
- 6. The fuel being consolidated shall have cooled out of the t
reactor for at least 2 years. This is an assumption used in the radiological analysis. This condition will be verified before any fuel assembly is chosen for j
consolidation.
j
- 7. A fuel storage rack s'all be only half full of fully loaded consolidated stsrage canisters, with every other rack cell empty so as to evenly distribute the weight.
{
This will ensure that the demonstration will stay within i
the basis of the current storage rack structural analysis.
This will be ensured by administrative controls.
- 8. Five feet of water shielding shall be maintained above the i
fuel assemblies at all times. This is a minimum shielding requirement based on Prairie Island's previous experience l
1 l
i
1 l
l l
l Page 14 of 19 Modification 86L944 Revision 1 with fuel reconstitution. This will be ensured by a cr=bination of equipment design and. administrative 4
controls.
Conclusions 1
In conclusion, this modification:
l
- 1. Does not create the possibility for an accident or malfunction of a different type than evaluated previously in the USAR or subsequent commitments.
2.
Does not increase the probability of occurrence of an accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments.
3.
Does not increase the consequences of any accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments.
4.
Does not reduce the margin of safety defined in the bases for any Technical Specification.
References
- 1. Topical Report, " Fuel Consolidation Demonstration at Prairie Island Nuclear Generating Station", Northern States power Company, August 1987 2.
" Fuel Pool Building Structural Evaluation for Prairie Island Nuclear Generating Plant", Report No. QUAD-1 558, Nuclear Services Corporation, 11/2/79 3.
NUTECH letter JKS-87-053, J.K. Smith (NUTECH)
L. McCarten j
(NSP), " Prairie Island Fuel Pool Evaluation for Rod j
Consolidation Project", September 3, 1987
- 4. Westinghouse letter MED-RPV-1524, R.E. Schwirian (Westinghouse) to L.R.
Benson (Westinghouse), " Fuel Rod l
Clad Surface Temperature in NSP Fuel Canisters", September j
l
?
)
I
Page 15 of 19 Modification 86L944 Revision 1 3,
1987 5.
Prairie Island Nuclear Generating Plant Technical Specifications 6.
Prairie Island Nuclear Generating Plant Updated Safety Analysis Report 7.
ANS-57.10, " Design Criteria for Consolidation of LWR Spent Fuel", Final Draft, October 1986 i
- 8. WCAP-2999," Criticality Control Parameters and Support Calculations for Uranium Bearing Nuclear Fuel at Low Enrichment"
- 9. Letter from J.
Smith (NUTECH) to L. McCarten (NSP),
" Evaluation of Prairie Island Spent Fuel Racks", September 14, 1987, NUTECH Letter JKS-87-057
- 10. "OT Position for Review and Application of Spent Fuel Storage and Handling Applications", USNRC, April 14, 1978
)
- 11. " Licensing Report for Prairie Island Nuclear Generating i
Plant Units 1 and 2 Spent Fuel Storage Modification",
f QUAD-1-79-509, December 1979 i
i Submitted By:
//h
/ [
[i 87 Reviewed By: 'C W [A [dd, 7!/7/S7 1
1 u___
_ __ A
PIgo 16~of.19.
L Modification 86L944 l 1.
Revision 1.
~
l' TABLE
'l l.
FUEL HANDLING ACCIDENTS RADIOLOGICAL EFFECTS Whole Body Thyroid' Line Analysis Dose (REM)
Dose (REM)
I 1-358 rods'(2-yr. decay) 0.093 0
2 1 assembly (USAR updated 2.6 13.8 evaluation..high-burnup fuel 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay) 3 10 CFR 100 dose guidelines 25 300 4
NUREG 0800 dose limits 6.25 75 (25% of 10 CFR 100) 4
' Paga 17-!' of.19 -
Modification.86L944 Revision l'-
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Page 19 of 19 Modification 86L944 Revision 1 Example Location of Consolidated Fuel Canisters in the Large Spent Fuel Pool 1
O 0 0 O O l
0 0 0 0 0 0 0 0 0 0 I
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Spacer Bar O
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Fuel Storage Pool wall Physically UnusaDie Storage Locations Storage Locations with Canisters.
O Storage Location with Standard Assemblies Empty Locations (Four -acdttlonel standard assemblies can be stored in these locations)
Figure.3
__.