ML19294B174
| ML19294B174 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/15/1980 |
| From: | Groce R Maine Yankee |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| PC-70-2, NUDOCS 8002270410 | |
| Download: ML19294B174 (10) | |
Text
-
Y H1 AIRE
- HARHEE, TURNPIKE ROAD (RT. 9)
.e ENGINEERING OFFICE WESTBORO, M ASSACHUSETTS 01581 C
617-366-9011 O
B.3.2.1 RfY 80- 29 PC-70-2 February 15, 1980 United States Nuclear Regulatory Cor: mission
%'ashington, D. C. 20555 Attection: Office of Nuclear Reactor Regulation Mr. Robert W. Reid, Chief Division of Operating Reactors
References:
(a) License No. DPR-36 (Docket No. 50-309)
(b) MYAPC letter to USNRC (DIY 79-97) dated September 18, 1979 (c) USNRC letter to MYAPC dated November 30, 1979
Subject:
Modified Spent Fue1 Pin Storage
Dear Sir:
Your letter, Reference (c), requested additional information to enable completion of your review and evaluation. Enclosed herewith as Attachment A are responses to your respective questions.
A draft of this letter wo submitted on January 21, 1980, to be followed by a formal letter after a meeting in late January. However, NRC personnel changes and circumstances beyond our control have forced delays of this meeting; therefore, we have elected to formally submit this information for the record at this time.
We trust this information is satisfsetory; however, if you have additional questions, please contact us.
Very truly yours, MAINE YANKEE ATOMIC POWER COMPANY Robert H. Groce Senior Engineer - Licersing RHG/smw Enc.
w 8002 270410
e 4
QUESTION NO. I t
Specify whether all provisions, including the January 18, 1979 addenda, to
" Review and Acceptance of Spent Fuel Storage and Handling Applications" have been considered and to what extent all provisions are met.
ANSWER The provisions of " Review and Acceptance of Spent Fuel Storage a"
!andling Applications" have been applied including the revisions described in the January 18, 1979 addenda.
QUESTION NO. 2 Provide a more detailed description and sketches of a typical fuel rack and fuel pin storage skeleton including details of the spacer and support grids and the tie rods.
Describe the modifications which are required to the present racks.
Explain in detail the load path along which all postulated forces are tranomitted to the spent fuel pool structures.
ANSWER The poison spent fuel racks are of aluminum construction with a fuel center-to-center spacing of twelve inches. There are six 10 x 7 rack modules, three 9 x 7, five 10 x 5, one 8 x 5 and one 9 x 6 to give a total fuel storage capacity of 953 assemblies. The poison material is boral sheet located between bundles overlapping the active fuel length. The top and bottom rack grida are omprised of rectangular and standard shaped aluminum members weld' d e
to form a gridwork approximately five inches deep.
The grids are positioned by four aluminum angle sections that run the full length of the modules at the Diagonal cross-bracing is provided on all four sides. Attached at corners.
the corners of the base of the bottom grid are adjustable feet which may be raised or lowered by turning a screw shaft from the top of the rack module.
The fuel bundles are contained in square tubes referred to in the analysis as cavities. These cavities are constructed eact with its own individual base which rests on the pool floor.
Each cavity vertically supports its own weight. The cavities are contained in a rack structure consisting of a top and bottom welded aluminum gridwork which holds the cavity in position with a twelve inch center-to-center spacing. This rack structure is supported by four feet which are attached to the bottom welded grid. Vertically, the rack carries only its own weight. The components of a fuel rack are shown schematically in Figure 1.
The new consolidated fuel that will be stored in the racks are 17 x 17 assemblies. The fuel pin spacing is maintained by several spacers placed at intervals along the length of the assembly. The skeleton structure of the consolidated assembly consists of perforated stainless end plates 8.136" square connected by four stainless rods. The rods screw into the base plate and are also welded. The top plate fits over the rods at the upper end and is held by a cap which screws onto the rods and acts as a lifting handle for the consolidated bundle. The spacers which are 24 GA stainless steel are welded to the stainless rods. The stiffness of the bundle is provided by the Zircaloy tubes which enclose the uranium pellets as well as the four stainless rods and additionally by the fuel pins.
Page 2 The consolidated assembly sits inside the cavity and rests on ABS plastic insulators attached to the cavity bare. The cavity, with its base, fits through the upper and lower grid of the rack structure and rests on the pool floor. Only horizontal reaction loads are carried to the grids by bearing.
No modifications to the existing racks are required, the only change would be the addition of seismic supporte which would be provided at the racks closest to the pool walls. These supports would press against the wall and act as compression members. They will not be bolted or otherwise attached to the wall or liner.
Betwes n racks at the upper and loher grid elevations, spacers may be added to better 'nable loads to be transmitted across rack grids to the supports and pool walls. The load path for all horizontal loads is through the upper and loher rack ' rids through the supports to the walls. Vertical cavity loads are transmitt ed to the pool floor through the cavity base.
QUESTION NO. 3 Describe the storage cage
- ation and structural support while it is partially filled, prior to piacing it in'the fuel rack. Verify that it can withstand all loads and load combinations for which the rack is designed, with no loss of integrity er impact to other items in the spent fuel pool.
AMSWER Figure 2 shows the working location in the spent fuel pool. The fuel consolidation is to be accomplished in the same manner as fuel has been reconstituted in the past. The empty consolidated cage is placed in the Spent Fuel Inspection Elevator. The fuel assembly to be disassembled is adjacent in a spent fuel rack cell. The fuel pins are then transferred from the rack stored assembly to the Inspection Elevator assembly.
It is anticipated that the consolidated cage will remain in the inspection elevator until it is completely filled with fuel pins, it then will be transferred to the storage racks.
It is mechanically possible to reverse the respective locations of the empty consolidated cage and the spent fuel assembly if this were to prove convenie nt.
The radiological consequences of seismic damage to a consolidated assembly in the spent fuel elevator is clearly bonded by the " Design Basis Fuel Drop Accident," (see Addendum A.2, PC-70-1, and response to question No. 10).
QUESTION NO. 4 Verify that the increased compressive loads on the ABC plastic insulator have been considered and that no degradation occurs which would decrease its insulating integrity.
ANSWER The small increased compressive loads of approximately fifty to sixty percent on the plastic insulators have been considered and are within the applicable limits with no degradation of insulation integrity. Although in all cases insulators are provided, there would be no real corrosion problem withcut them.
a Page 3 s
QUESTION NO. 5 Provide the results of the pool wall design re-evaluation. Verify that the pool wall integrity is maintained, considering the additional seismic restraints.
ANSWER The support design remains to be completed. The pool wall design re-evaluation will be finalized when the support configuration is determined.
The supports hdll be such that loads are transferred to the wall in a manner which will not ecmpromise the integrity of the structure.
All applicable sections of " Review and Acceptance of Spent Fuel Storage and Handling Applica tions" will be met.
QUESTION NO. 6 Provide the total increase in weight supported by the floor which will result from using the pin storage concert. Verify that the effect of this increase on the dynamic response characteristics of t'he spent fuel pool has been considered.
ANSWER The pin storage concept will increase the weight supported by the floor by fifty to sixty percent. The fuel pool is a rigid structure resting on bedrock and has load carrying capability sufficient to account for the relatively slight increase in weight from completed assemblies. The increased weight is considered when determining the seismic loading on the fuel pool structure.
QUESTIONS NO. 7 AND 8 (7) Verify that the racks have been analyzed for seismic and impact loads.
Provide details of the analysis.
Discuss any interaction or impacting between fuel assemblies, storage cages, and racks during maximum seismic excitation, and verify that their integrity is not compromised.
(8)
Provide a detailed description of the model, and its development, used in the dynamic analysis. Specifically describe the properties of connecting members, the gap elements, and the independent modes, and describe how they were developed.
In addition, justify the direct use of the ground spectra with the rack model.
ANSWER The racks have been analyzed for seismic and impact loads utilizing the ANSYS computer program. The analyses method is non-linear transient dynamic. The dynamic model used for the analysis is shown in Figure 3.
The fuel pool supported on bedrock is considered a rigid structure transmitting the ground motion to rack supports.
Page 4 s
The cavity (can) and consolidated assembly are modeled as beam elements.
Consolidated assembly inpact is included in the analysis by means of the gap elements shown; the gap represents the maximum possible separation between the consolidated assembly and the cavity. The water inside the rack is accounted for with the fluid coupling elements shown. Support points at nodes 1 and 9 represent the lower and upper rack grid elevations respectively.
The properties of the cavity beam elements represent the cross-section of the cavity. The gap elements include the consolidated assembly to cavity separation as well as the local cavity stiffness. The consolidated assembly properties represent the new assembly stiffness.
QUESTION NO. 9 Provide technical bases to justify the dynamic load factor used to represent the effect of f tel assembly impact.
Provide a more detailed description of your analyses and results.
ANSWER Initially, due to the non-availability of a time history acceleration record, the dynamic load factor was used in a static analysis to represent consolidated assenbly impact.
The effect of impact is considered similar to the impact of a uniform load on a beam. The factor is determined by assuming a fuel bundle constant acceleration equal to the maximum ground motion. The cavity (can) is taken as a simply supported beam.
The cavity and consolidated assem';1y are modeled using the ANSYS computer program. Gap elements are used to represent the initial separation of the bundle and can.
A constant force is applied to the bundle over a time period such that the cavity is deflected to its maximum. The dynamic factor is taken as the ratio of the maximum dynamic deflection to the maximum static deflection.
The results of the initial static analysis using this load factor show that the racks would be acceptable under the new consolidated fuel loading, however, additional support would need to be incorporated.
The dynamic analysis as explained in the answer to questions number 7 and 8 was subseqently done. This analysis proves the overconservation of the static method and the results supercede those of the static. The impact " factor" ic not incorporated in the final analysis and design.
QUESTION NO. 10 Verify that the drop of a compacted fuel assembly on a compacted or disassembled storage cage has been considered. Verify that the integrity of both the fuel assembly an3 the storage cage is not compromised, and that the storage cage will not overturn or impact other storage cages.
Page 5 ANSWER The accident suggested is clearly bounded by the " Design Basis Fuel Drop Accident," see PC 70, Addendum A.2.
It must be kept in mind that the fuel that is being disassembled has had a significant period for radiological decay shich is an exponential function of time. The accident considered in the FSAR is much more severe since fuel freshly discharged from the reactor is assumed to rupture. As an illustration, consider the relative fission gas activity of a freshly discharged assembly and an assembly that has aged only 120 days af ter core discharge. The breaking of the 176 pins of a single freshly discharged assembly is the equivalent to breaking over a million consolidated assembly pins or thousands of consolidated assemblies.
A three-year wait before commencing consolidation would only further increase this far more than adequate conservatism.
QUESTIONS NO. 11 AND 12 Provide the weight of a fully compacted fuel rack.
In addition, specify the heaviest load that will be transported over the spent fuel racks and the maximum possible drop height. Verify that the worst case drop of a compacted f uel rack on another rack will not adversely ef fect the integrity of the racks.
Discuss whether the movement of any racks will be necessary during fuel consolidation. Describe the handling procedures and the precautions which will be taken to prevent damage to (lift) the spent fuel and the racks.
ANSWER A fuel rack containing fuel is not permitted to be lifted, nor is an empty rack allowed to be moved over fueled racks. Therefore, any movement of spent fuel racks will be in accordance with the T.S. to further reduce the potential of a " radiological" accident.
Consistent with this proposed change, the movement of a heavier than standard assembly (i.e., a consolidated assembly) over a fueled rack is considered and the accidental drop is addressed in Addendum response. The maximum drop height is 18 inches over racks and the limit is established by designed mechanical limitations of the refuel crane.
The transporting of heavier objects over fueled racks than those addressed herein is outside the scope of PC 70.
Should the need for such an operation arise, it would be subjected to the standard Engineering Design Review.
QUESTION NO. 13 Demonstrate that the values of water level in the pool, the cooling water velocity and the water temperature will be identical to the values used in developing the original thermal design loads.
Discuss at least the following items:
influence of increased thermal load on stresses / strains in concrete walls and bottom of the pool and in the lines and also of the racks.
Page 6 influence of the increased weight of the spent fuel scored in the pool of the virtual mass of the water, horizontal forces on the bottom and the walls during an earthquake and the influence of the liner and racks.
Discuss the potential need for improved survaillance during normal operation of the plant and af ter seismic and/or tornadic occurrences.
ANSWER
.The limiting therm'l design condition in Section 9.8 of the FSAR is condition 92.
For this condition the design thermal heat load was 22x106 BTU /hr. The primary component cooling wa ter supplied to the shell side of the spent fuel 0
pool heat exchanger was assumed to be 85 F.
This results in a bulk spent 0
fuel pool temperature of 154 F.
As stated in WMY 79-97 (P.C. 70) Section 1.0, the bulk pool temperature will be monitored and the discharge of spent 0
?uel to the pool administratively limited so as not to exceed 154 F.
Whereas the design of the spent fuel pool equipment (pumps, piping and heat exch anger ) has not changed since the FSAR and the bulk pool temperature has not increased the cooling water velocity remains unchanged from the original values used in condition #2 of the PSAR.
The pool water level continues to.tme maintained at the original design level of 44 ft.
See response to questions 7 and 8 for answer to part two of the above question.
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