ML19275A383
| ML19275A383 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/01/1979 |
| From: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| TASK-09-01, TASK-9-1, TASK-RR NUDOCS 7910040224 | |
| Download: ML19275A383 (17) | |
Text
-
s 1 Consumers
/
power n
Company D
General Offices: 212 West Michigs.n Avenue. Jackson, Michigan 49201 e Area Code 517788 05S0 LJ October 1, 1979 Director, Nuclear Reactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - RESPONSE TO NRC STAFF QUESTIONS REGARDING EXPANSION OF SPENT FUEL STORAGE CAPACITY Consumers Power Company letter dated June 26, 1979 requested approval of a proposed Technical Specifications change permitting increased storage capacity in the Big Rock Point spent fuel storage pool. The Description and Safety Analysis of the proposed increase had been previously submitted by Consumers Power Company letter dated April 23, 1979. NRC letter dated August 17, 1979 transmitted a number of questions related to this request. The answers to these NRC questions are transmitted here..th.
Consumers Power Company letter dated June 26, 1979 stated, in part, that a revision to the April 23, 1979 Description and Safety Analysis would be submitted. The information intended for that revision is included in the attached. Consumers Power Company still intends to revise the Description and Safety Analysis, but has elected to postpone such revision until after all NRC staff questions in this matter have been addressed. This will permit a single revision to incorporate all new information identified as a result of responding to the staff questions.
The response to Question 4 in the attach?ent references a drawing which is considered proprietary by NUS Corporatica. This drawing will be transmitted under "parate cover as soon as the appropriately executed affidavit (10CFR2.790(b)) is received from NUS.
David A Bixel (Signed)
David A Bixel Nuclear Licensing Administrator CC JGKeppler, USNRC 1099 037 79L0040R M F
I gJ"IONAL INFORMATION RELATIVE TO INCREASE In ROCK POINT SPENT FUEL STORAGE CAPACITY u.
The ao.i' ac1 information requested by NRC letter dated August 17, 1979 is provit uelow:
Question 1 Justify your structural acceptance criterl. which allows local stresses to exceed strength limits.
In addition, provide a detailed description of and criteria for the phrase, "no loss of function of the fuel rack."
Response
The spent fuel storage racks will utilize a geometric configuration designed to maintain subcriticality through all credible environmental and abnormal loadings, such as earthquake and the impact due to drop of a pent fuel assembly during routine spea; fuel handling.
Local stresses in the racks may exceed the strength limit of 1.6S only for the case of a drcpped fuel assembly. The transfer cask winch is equipped with a load limit cutout switch which limits upward detected forces to a maximum of 2000 pounds; this limits the stresses which might be caused by potential binding of a fuel rack during withdrawal from the racks to a level below 1.6S.
In the event of : fuel assembly drop, the kinetic energy of the falling assembly is converted into local plastic strain energy in the rack at impact.
Gross stresses in the rack remain well below yield stress and consequently the rack suffers local deformation rather than global structural collapse. As discussed in Section 5.1.3 of the Description and Safety Analysis, local stresses are allowed to exceed the limits provided there is no loss of function of the rack (ie, the structural integrity of the rack is maintained sufficiently to provide a safe geometric configuration keeping the neutron multiplication factor of any array of fuel assemblies less than or equal to 0.95).
Question 2 Verify that all provisions of the NRC guidance on spent fuel pool modifications, entitled " Review and Acceptance of Spent fue. Storage and l
Handling Applications" (10:luding errata), are met and that the pool continues to conform to all FSAR analysis methodolcgy and acceptance criteria.
Justify any deviations.
Response
All provisions of the referenced guidance have been considered and are met, as applicable, in the design of the new spent fuel storage racks.
Structural, thermal kydraulic and criticality analyses have been performed using current methodology, criteria and NRC guidance. Although current analysis methodology 1099 038
2 N
may differ from earlier FHSR analysis methodology, the present design conforms to all FHSR acceptance criteria.
Question 3 Assuming the double pump failure which you postulated, verify that the maximum thermal load is with11 acceptable limits and has been considered in the analysis of the existi.Tg racks, liner, and concrete pool struc*ure.
Response
The analysis of fuel pool heat loads and temperatures discussed in Sections 6.2 and 6.3 of the Description and Safety Analysis has been revised.
The revision reflects heat loads and temperatures resulting from utilizatioa of ali 441 available assembly storage locations rather than the storage of 434 assemblies previously analyzed.
In addition, a significant overconservatism in the previous prediction of long-term spent fuel pool heat load has been corrected. Provided as Appendix A to this response are our revisions to those pages of the Description and Safety Analysis affected by this new analysis.
These revised pages will be included in the revision to the Description and Safety Analysis discussed in the forwarding letter.
The double pump failure of the spent fuel pool cooling system considered in the Description and Safety Analysis was postulated in accordance with the NRC guidance on spent fuel pool modifications discussed in Response 2 above. This failuce was not port of the USAEC design criteria used for the original design, engineering and construction of Big Rock Point and was not, therefore, explicitly considered. Nonetheless, the proposed expansion or spent fuel pool storage capacity does act alter the thermal equilibrium conditions of spent fuel storage as presently licensed.
Question 4 Provide more detailed description and sketches of a typical fuel rack and fuel ran including details of the supporting grids, fuel seating surface, leveling legs, and connections between various elements. Exulain in detail the load path along which all postulated forces are transmitted to the spent fuel pool structure. Make clear how the cans are supported vertically and laterally within the fuel racks.
Response
Details of a typical fuel rack and can, including the supporting grids, fuel seating surface, leveling legs and connections between various elements are shown on NUS Drawing 5148-M-2001.
All postulated forces will be transmitted through the can and grid system to the leveling legs and into the floor of the fuel pool. The square cans will be comprised of 1/4-inch thick stainless steel welded to 1/2-inch thick base plates that will support the deadweight of the fuel assemblies. The stainless steel base plates will in turn be welded to the lower grid system that will be 109y 039
3 comprised of a lattice of stainless steel beams 1-1/4 inches wide by 4 inches deep. The downward load from the deadweight of the fuel assembly will be transmitted from the baseplate to the grid beams to the gussets and bosses into which the leveling legs are screwed.
Lifting lugs will be weld-d to the upper grid beim system that is comprised of a lattice of stainless steel beams 1-1/2 inches wide by 2 inches deep that will be welded to the cans.
Lifting forces will be transmitted through the lifting lugs, to the upper grid beams, to the cans and into the lower grid beam system and base plates.
Seismic loads will act horizontally on the cans and vertically on the base plates. These loads will be transmitted through the cans and grid systems to the floor of the fuel pool via the leveling legs.
Impact loads from a dropped fuel assembly will be transmitted either from the lead-in guides or a baseplate (depending on im ict location) to the remainder of the structure via i
the grid beam and can system.
In the case of a stuck fuel assembly, a verti:al force would be applied to a can it. the event of an attempted withdrawal. This force would be distributed through the can to the grid beam system and baseplates to the remainder of the rack. This force would be limited by the 2,000-lb load limit cutoff switch of the fuel transfer cask winch.
Question 5 Provide descriptive information including plans and sections showing the spent fuel pool in relation to other plant structures. Specifically sketches should show the spent fuel pool location with respect to the other features represented in the model shown in Figure 5-2.
Response
Plans and section drawings detailinC the spent fuel pool and its location in relation to other structures, systems and components within the reactor building are included in the FHSR, as amended, as Drawings M-101, 103 and 104.
Question 6 Justify the statement that the 8X11 rack type has the greatest potential for tipping and will develop the greatest internal forces.
Response
Both the 8 x 11 and 8 x 13 racks will have a slightly greater potential for tipping than the 9 x 9 rack due to the shorter distance (8 feet) between leveling legs.
The 8 x 11 rack will have the greatest potential for tipping because it is supported.'n four leveling legs while the 8 x 13 rack is supported on six leveling legs. The 8 x 11 rack will develop greater bending moments in the grid beams due to seismic and deadweight loading. Other internal forces and moments will be approximately the same or somewhat greater in the 8 x 11 rack; however, thermal loading is local in nature and will be
!099 030
4 the same in all racks. The lifting lugs are sized for the largest rack, namely the 8 x 13.
Question 7 Provide a detailed description of the boundary conditions between the fuel
- r. cks and pool floor liner, and between the fuel racks and pool walls and how they were determined for each of the models used. Also describe the properties of the top grid restraint and specify how they were obtained.
Describe the location, intent, and actual physical interface of the top grid and base horizontal fixed support, shown in Figure 5-3.
Provide detailed sketches of the restraints and s pports.
Response
The spent fuel storage racks are a freestanding design, and no restraints r
supports external to the racks are provided. Therefore, the only physical contact exists between the rack leveling legs and the fuel pool floor liner.
Figure 5-3, attached, has been revised to clarify the intent of the upper gric restraint. The upper grid restraint, shown in Figure 5-3 as an equisalent rotational spring, represents the rotational restraint provided to the top of one can by its welded connections to surrounding cans via the upper grid beam system. The stiffness of this spring was determined using a detailed two-dimensional finite element model of the spent fuel rack in which a row of cans was modeled.
The lower grid restraint, shown in Figure 5-3, represents the restraint which would be experieaced by the rack if the coefficient of friction between the bottom pad of the leveling legs and the fuel pool floor liner were large, effectively restraining horizontal motion of the rack. This represents the worst case condition for the tipping analysis since no energy dissipation is allowed due to sliding of the rack.
The revised Figure 5-3 will be included in the revision to the Description and Safety Analysis report discussed in the forwarding letter.
Question 8 Provide i discussion describing the seismic input used with each of the models in Figures 5-3, 5-4, and 5-5.
Specifical:-
state how the input excitations were obtained, and provide the procedure r to obtain the equivalent static loading.
Responsa The seismic input used with the tipping and sliding nonlinear models shown in Figures 5-3 and 5-5, respectively, was a displacement time-history of the fuel pool floor. This time-history was obtained from the seismic time-history analysis of the plant using the model shown in Figure 5-2.
Seismic input to this model is discussed in Section 5.2 of the Description and Safety Analysis.
i.
ii)99041
5 The three-dimensional finite element model shown in Figure 5-4 was used to derive the stresses in the rack under deadweight, thermal and seismic loads.
In this model, N-S and E-W equivalent seismic loading was represented by uniform forces appliad along the length of the cans.
In d etermining the magnitude of this equivalent static loading, the maximum bending moment at the base of the can, determined from the time-history tipping analysis, was equated to the bending moment at the base of the cans in a model which represented a row of cans from the finite element model shown in Figure 5-4 when each can was subjected to a uniform horizoutal load. Since the racks are rigid with respect to vibration in the vertical direction and the amount of tipping is so slight such that significant impact loads on the leveling legs do not develop, the seismic analysis in the vertical direction consisted of a static analysis using the maximum vertical floor acceleration.
Q_uestion 9 Spccify the minimum distances between racks, racks and pool walls, and racks and other equipment. Also, verify that the stated 1/2" rack sliding is based on the absolute sum of displacements of two adjacent racks.
Response
The spent fuel storage pool and rack arrangement, including minimum distances between racks, racks and peol walls, and racks and other equipment, is shown on NUS Drawing 5148-M-2000 ' attached).
As discussed in Section 5.3 of the Description and Safety Analysis report, the maximum distance the racks are predicted to slide during a postulated design basis earthquake (DBE) will be less than 1/2 inch, based on the absolute sum of the displacements of two adjacent racks. The minimum distance between racks will be approximately 1-3/4 inches; between racks and pool walls approximately 3 inches; and between racks and other equipment in the spent fuel pool approximately 2 inches. Adequate spacing is therefore provided to preclude impact during a DBE of two adjacent racks, of racks and the pool walls, and of racks and other equipment in the pool.
Question 10 Discuss any interaction or impacting between fuel and cans during maximum seismic excitation and verify that their integrity is not compromised as a result of impact.
Response
The fuel is supported vertically at its base by the fuel storage can base plates. Horizontally, friction forces restrain motion of the base of the fuel while the upper portion of the fuel is restrained by the sides of the continuous square can.
During a seismic event, the fuel is free tc move from side-to-side within the can.
Impact between the fuel and cans is mitigated by the hydrodynamic coupling, scrain energy developed in the fuel assembly and rack during impact, and the fact that the total gap between the can and fuel i099 042
6 is small (approximately 1/2 inch). The nonlinear dynamic analysis performed accounts for all of these effects. Structural integrity of the racks has been assured by means of a detailed analysis in which the fuel can interaction was explicitly modeled. The fuel is stored within a smooth walled square can in a manner consistent with current practice in spent fuel storage in the United States and in conformance with the general recommendations of the fuel vendor.
Question 11 Discuss the effects of the increased borizontal loads on the liner and concrete structure (walls) of the pool. Verify that the walls and liner are capable of withstanding these loads.
Response
Since the racks are freestanding (without restraints to the walls), no increased horizontal loads on the walls of the pool will occur. The pool floor and liner have been analyzed and shown to be capable of withstanding the loads due to the increase in quantity of spent fuel stored in the racks.
Question 12 Provide calculations showing how the kinetic energy of the dropped fuel assemblies was determined, and justify that the worst postulated drop results in a kinetic energy of 195,000 in. Ibs.
For each of the fuel assembly drop cases, quantify the resultant rack stresses, reaction loads, effect on the leveling legs, and the number of equivalent fuel assemblies damaged. Verify that the integrity of both the stored spent fuel and the dropped fuel assembly is not compromised due to the effects of the worst case drop. Verify that the possibility of cans or rack overturning or separation of cans from the grid system have been considered with a worst case drop, and will not occur.
Response
Two postulated fuel assembly drops were considered to determine the response of, and effects on, the racks. Energy (transfer) methods were used in each case to determine (1) the extent of damage to the rack from resulting stresses, (2) reaction loads, and (3) effects on the leveling legs.
The worst case fuel assembly drop is postulate: :^. cur with a fuel assembly that has been lif ed to the upper limit of travel of tua auxiliary hook and subsequently falls into the fuel pool through an empty : pent fuel rack can and impacts the baseplate at the bottom of the rack. The kinetic energy of the fuel assembly at the time of impact with the fuel rack baseplate may be conservatively estimated (due to the relatively large difference in densitiec of the fuel assembly and the water) by multiplying the effective buoyant weight of a fuel assembly by the total distance traveled in the drop. The buoyant weight of a fuel assembly is 383 pounds. The bottom of the fuel assembly may be lifted approximate 1f 43 feet 3 inches (519 inches) above the bottom of the fuel pool. The fuel rack baseplate at the bottom of the storage rack is assumed to be 11 inches from the bottom of the pool floor.
Hence, the 1099 043-
7 kinetic energy of the fuel at impact with che baseplate is conservatively estimated to be approximately (519 - 11) (3R3) = 195,000 in-lb (rounded up).
This case represents the greatest drop bright, and consequently the largest kinetic energy at impact from the drop af a fuel assembly onto a rack. The maximum stress in the fuel rack baseplate is approximately 55,000 psi.
Local plastic deformation occurs and, as a worst case, it may be postulated that the baseplate, beneath the fallen fuel assembly, may be sheared off.
- HowcVer, even in the worst case, the structural integrity of the rack is maintained as gross stresses in the rack remain well below yield thereby preventing global structural collapse or the rack. The reaction load for this drop is 35,680 pounds; the maximum bending stress in the fuel rack lower grid support system is *,500 psi; and the worst case loading on a leveling leg occurs when a corner fuel rack can is involved producing an axial compressive stress of 6,000 psi.
The second case is postulated to occur with a fuel assembly that has been lifted to the upper limit of travel of the auxiliary hook and then drops on top of the funi rack lead-in guides.
In this case the lead-in guides are assumed t, be 94.6 inches from the bottom of the pool floor. Hence, the kinetic energy of the fuel at impact with the lead-in guides is conservatively estimated to be (519 - 94.6) (383) = 162,500 in-lb (rounded up).
This energy at impact would be dissipated by the crushing of the lead-in guides; however, buckling of the cans is not postulated to occur. After impacting the lead-in guidas in a vertical position, the fuel assembly is postulated to rotate and impact a row of lead-in guides in a horizontal position. The kinetic energy conservatively estimated to be dissipated dua to this rotation is approximately 15,300 in-lb (rounded up).
The fuel rack cans are welded to the grid system as discussed in response to Question 4 and form an integral unit that will maintain its overall structural integrity under the forces which could result from a dropped fuel assembly.
Overturning of a rack as a consequence of dropping a fuel assembly is not possible.
The integrity of a dropped fuel assembly cannot be assured nor can that of a fuel assembly which might suffer a direct hit from a dropped fuel assembly.
However, the health and safety of the public will not be jeopardized in such an event as discussed in Consumers Power Company letter dated June 28, 1977.
Question 13 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 (fuel assembly or cask) on the pool liner will not compromise its integrity. Also, verify that a worst care drop on the edge or corner of a rack will not adversely affect the integrity of the rack.
Response
The heaviest load planned for transport over the spent fuel racks is the 24-ton spent fuel transfer cask.
The maximum possible drop height of this cask
!099 044
8 was considered in the cask drop analysis submitted July 1, 1974, and in the staff's Safems Evaluation Report on the cask drop analysis published February 6, 1976. The primary function of the pool liner is to provide a surface within the pool to assist in maintaining water quality and cleanliness.
In addition, it does provide additional leak tightness; however, it does not provide any additional structural support. Therefore, an object drop which merely penetrates the liner, but does not affect the integrity of the fuel pool concrete structure, will not result in a loss of pool water in excess of the 200 gpm makeup capability. No heavy object, with a weight less than the 24-ton spent fuel transfer cask, is capable of creating a leak in excess of the 200 gpm limit. The drop of the 24-ton spent fuel transfer cask is precluded, as discussed in the staff's SER of February 6, 1976, by the use of safety slings. As discussed in Section 4.0 of the Description and Safety Analysis Report, the movement of other casks over or near stored fuel is restricted by administrative controls.
Question 14 Provide the fundamental frequency of the pool walls, and verify, if they are flexible (i.e., fundamental frequency less than 33 H ), that the provisions of the NRC guidance on spent fuel pool modifications (see question 2) have been met.
Response
The pool walls are not utilized to provide lateral restraint to any of the racks. Consequently, the flexibility of the walls does not influence the behavior of the racks, nor do the racks impose any loading on the wsils.
In accordance with NRC guidance entitled, " Review and Acceptance of Spent Fuel Storage and Handling Applications," pool wall flexibility need not be considered.
Question 15 Provide a detailed discussion describing the sequence of installation, and the handling procedures and requirements of the new fuel racks, and a description of the precautions to be taken to prevent damage to the stored fuel during the construction phase.
Response
The installation of the new spent fuel storage racks will involve the movement of fuel assemblies and existing fuel storage racks presently in the pool. The fuel assemblies and existing fuel storage racks will be moved to make a clear path for the installation of the new fuel storage racks using detailed written procedures.
Fuc1 rack installation procedures will meet the following requirements:
a.
Fuel racks with fuel assemblies stored in them will not be bandled or moved. Only empty racks will be moveu in the pool.
!099 045
9 b.
No rack will be handled or moved over stored fuel.
Only one operation (eg, lowering a rack or moving a rack or shuffling fuel c.
assemblies) will be allowed at any time.
Detailed written procedares will be prepared to describe the sequence of installation and path of movement of the new racks, the relocation of fuel assemblies in the new racks, the relocation of existing racks in their final positions, and the redistribution of fuel assemblies, if required.
Question 16 Describe the planned inservice inspection and surveillance o.f the new racks and supporting liner. Discuss the potential for corrosion of the liner, considering thermal and other stresses occurring during the life of the facility. Provide the water chemistry of the fuel pool water and the procedures available to monitor and maintain the quality. Verify that no corrosion of the racks, fuel cladding, or the pool liner will occur over the lifetime of the plant based on the spent fuel pool water environment.
Response
The inservice inspection of the spent fuel pool is not presently a requirement at the Big Rock Point Plant, nor, to the best of our knowledge, is it required of licensees in general. No materials, other than those presently in use in the fuel pool, will be added. The new fuel racks and existing tuel pool liner are fabricated from austenitic stainless steel and appropriate austenitic welding rods are used in welds.
Stainless steel has proven to be an acceptable material and is being used extensively in similar applications.
Storage of zircalloy clad fuel elements in spent fuel pools has been found to be acceptable and is widely used throughout the industry.(1),(2) Therefore, there does not appear to be justification for the imposition of any inservice inspection requirements on the Big Rock Point Plant spent fuel pool.
The spent fuel pool water chemistry is maintained by the spent fuel pool filter system. The pool water chemistry is routinely tested for pH.
conductivity and turbidity. Typical values are as follows:
a.
pH - 6.9.
b.
Conductivity - 0.3 micromho.
c.
Turbidity - 20 ppb.
There are no provisions for borating the fuel pool water.
(1) Johnson, A B, " Utility Spent Fuel Storage Experience," PNL-SA-6883, April 1978.
(2) Weeks, J R, " Corrosion of Materials in Spent Fuel Storage Pools,"
BNL-N'UREG 23021, July 1977.
i099 046-
Revised APPENDIX A 6.0 COOLING CONSIDERATION 6.1 General As discussed below, analyses were performed on the existing spent fuel cooling and fuel pool cleanup systems. It was concluded that the present in-stalled systems provide sufficient capacity and redundancy to acccmmedate the proposed increase in spent fuelloading. Table 6-1 defines the performance of the system under various single active failure conditions.
6.2 Cooling System Performance The adequacy of the cooling system was analyzed in view cf the expanded fuel storage capacity. The decay heat generation rates due to the spent fuel were calculated using NRC Standard Review Plan 9.2.
Two design conditions were evaluated; (1) normal refueling of 1/4 cf a core, and (2) full core offload.
The heat load for the normal refueling case was based on the following prcjected sequence cf fuel movement:
1.
86 fuel assemblies in the pec1 as of April 1979.
2.
330 additional fuel assemblies accumulated in fifteen successive refuelings.
3.
25 fuel assemblies officaded 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after reactor shutdown in the sixteenth refueling.
The above refueling description results in the total of 441 fuel assemblies in the pool.
1099 047 6-
APPENDIX A TABLE 6-1 BIG ROCK PLAIJT - SPENT FUEL POOL COOLII4G SYSTEM (SFPCS) SINGLE ACTIVE FAILURE ANALYSIS Component Failure Mode Consequence Normal Rerueling IIcat Load Full Core Offload IIeat Load 6
6 1.37 x 10 BTU /hr 3.79 x 10 BTU /hr Spent Fuel Mechanical Second pump is operational. Com-Second pump is operational. Maxi-Cooling Pump Failure lete inventory of r: pares available mum pool water temperature will be ci b
for rapid pump repairs, if needed.
132 F. with one heat exchanger in flowever, maximum pool tempera-operation.
ture will reach 93 F.
Offsite Electrical Emergency power is not available.
Emergency power is not available.
Power Failure Pool will begin to boil in 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br />.
Pool water will begin to boil in 20 Makeup water is available for 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Makeup water is available gpm evaporation rate.
for 7.8 gp... evaporation rate.
b Reactor Mechanical Same as above Same as above j
Cooling Water Failure EI (ItCW) Pump O.>
CO
Revised APFDDIX A 6
The heat load for this sequence of events is 1.37 x 10 BTU /hr. The pool temperature, assuming both spent fuel pool cooling pumps are operaticnal, 0
is predicted to be no greater than 82 F. Postulating a single active failure in the SFPC system the maximum expected temperature is 93 F.
The heat load for the core officad case was based en the following projected refueling sequence:
1.
86 fuel assemblies in the pool as of April 1979.
2.
242 additional fuel assemblies accumulated in successive refuelings.
3.
29 fuel assemblies discharged at the last normal refueling.
4.
84 fuel assemblies (full core) offloaded 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after an emergency shutdown after 60 days
- cf operation at full power following the last normal refueling.
The above refueling sequence results in a total cf 441 fuel assemblies in the 6
pool. The maximum heat load for this sequence of events is 3.79 x 10 BTU /hr.
0 The maximum pecl temperature expected is 101 F under normal conditions and 132 F with less of a single heat exchanger or pump.
6.3 Fuel Element Heat Transfer The fuel rack base is elevated above the f1cor to assure adequate ficw under the rack to each fuel assembly. Analyses have been performed which show that natural convection flow is sufficient to p reclude local boiling at the hottest storage locatien.
- This is the time of maximum heat generation in the pcci and is datermined considering the decay of stored assemblies and the buildup of fission products in the core prior to cfficad.
~1099 049 6-
Revised ADPDDIX A The analyses were based cn the following censervative assumptions:
1.
The assembly inlet temperature is the mixed het temperature cf the pool,132 F, for the single active failure condition fc11cwing a full U
core officad 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown.
2.
A hot assembly peaking facter of 1.5 is applied to a limiting batch 4
average assembly release rate conservatively estimated at 4.1 x 10 BTU /hr corresponding to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown of the 240 MWt core.
3.
The maximum Iccal peaking factor is 2.73 giving a maximum Iccal o
heat flux of 1396 BTU /hr-ft'".
4.
A film coefficient of 37.5 STU/hr-ft - F is based on pure ccnduction to a stagnant water layer at the fuel red surface.
5.
A one-dimensional fluid flow analysis is used.
6.
The downcomer region en the periphery of the pool feeds 11 assemblies in a row, each assumed to be generating the maximum heat rate defined in assumption 2.
7.
Heat generated in the intercell regicn by gamma depositicn, equivalent to a heat flux of 87.2 BTU /hr-ft at the can inner wall, is remcved frcm the intercell region by conduction through the intercell water, fuel can metal, and a stagnant water layer at the can inner surface.
With a single active failure in the spent fuel pool cooling system, the bulk temp-erature will not exceed 132 F for a full core officad. Under these ccnditions, C
c the maximum fuel rod surface temperature is less than 198 F providing a 39 7 0
margin to local boiling. The margin to bulk boiling is 77 F in the assemblies and 58 F in the intercell regicn. This represents the limiting thermal conditien in the pool.
1099 050 o-4
Revised AFPCIDIX A A natural convecticn analysis was also performed for the case when all pool ecoling systems are Icst. This analysis assumes one dimensional flow and the heat leads established by assumptions 2 and 6. As in the previous analysis, subccoled liquid enters the bottom of the fuel assembly at the mixed het temp-c erature of the pool, which in this case is 212 F. Calculations show that the saturation temperature is reached within one half inch cf the fuel assembly outlet where the maximum void fraction is.206.
Since the intercell region is opened to the water above the rack at the corners of each assembly, air or steam cannot be trapped between the assemblies.
In summary, with a single active failure as well as in normal operation, the hottest location is below the local saturation temperature and thus local boiling will not cccur. Even if all pool ecoling systems are lost, less than one-half inch cf the assembly height will be in bulk boiling. The maximum void fraction at the cutlet is.206. Design of the rack is such that spaces between assemblies will always have water in them.
1999 051 6-5
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'1099 052
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