ML19309A299
| ML19309A299 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/05/1976 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8003270720 | |
| Download: ML19309A299 (18) | |
Text
u.s. NUCLEAR REGULATORY COMMISS12N DOCKET NUM!E3 NRC roRu 195
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('3 50-312 c
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NRC DISTRIBUTION FOn PART 50 DOCKET MATERIAL
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TO:
FROM: SMD DATE OF DOCUMENT NRC.
3-5-76 L
Sacramento, Calif. 95813 J.J. Matti:noet DATE RECEIVEo 3-8-76 CLETTER DNoroRizEo PROP INPUT FORM NUMBER OF COPtES RECEWEO i
ConsciNAL xluwCLASSir:Eo OCOPY g
j DESCRIPTION Ltr notarized 3-5-76 trans the E NCLOSU RE Additional Ihformation Rancho following:
Seco Nuclear Gen.-Station Unit 1 in regard to spent fuel so,trage...~.
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(40 cys encl rec'd)
THIS DOCUMENT CONTAINS PCOR QUALITY PAGES
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PLANT NAME:
Rancho Seco u
,e SAFETY FOR ACTION /INFORMATION ENVIRO DHL 3-9-76 9
ASSIGNED AD :
ASSIGNED AD :
I g DRANC11 CHIEF :
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Re id BRANCH CHIEF :
PROJECT MANAGER:
b ep g,y M u g gggg PROJECT MANAGER :.
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INTERNAL DISTRIBUTION C REG _ FILE 3 SYSTEMS SAFETY PLANT RVRTEMR RmrTon TPr77 If NKU run RETNEMAN TrnFRcn ERNST L d I&E /d RcnnnFnnR BENAROYA BAL_ LARD I
d OELD LAINAS SPANGLER
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GOSSICK & STAFF ENGINEERING IPPOLITO L
MIPC MACCARy SITE TECH CASE KNIGHT OPERATING REACTORS GAMMILL HANAUER SIINEIL STELLO STEPP HARLESS PAULICKI HUIJ1AN OPERATING TECH PROJECT MANAGEMENT REACTOR SAFETY
/
EISENHUT SITE ANALYSIS BOYD ROSS M
SHAO VOLLMER P. COLLINS NOVAK
( BAER BUNCH HOUSTON ROSZTOCZY
( SCIMENCER af J. COLLINS l
PETERSON CHECK
/ GRIMES KREGER MELTZ
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llELTEMES AT & I SITE SAFETY & ENVIRC SKOVHOLT SALTZMAN ANALYSIS RUTBERG DENTON & MULLER EXTERNAL DISTRIBUTION CONTROL NUMEER
/JPDR sacramenEo, ca m. NATL LAB BROOKHAVEN NATL LAB V
/ REG. V-IE ULRIKSON(ORNL)
CONSULTANTS
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Pegulatory. Docket $th mg SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street. Box 15830. Sacramento. California 95813: (916) 452 3211 March 5, 1976 Director of Nuclear Reactor Regulation Attention:
Mr. Robert W. Reid, Chief
_ __ _N Operating Reactors, Branch No. 4 U. S. ftuclear Regulatory Commission Washington, D. C.
20555
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.v Docket No. S0-312 h.,, 'y'.
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Rcncho Seco Nuclear Generating
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Station, Unit No.1 Dea" Mr. Reid:
This letter is in response to your letter dated February 9,1976 which requested acditional infomation to support the District's submittal of December 19, 1975 in regard to increased spent fuel storage at Rancho Seco Nuclear Generating Station, Unit No.1.
The infomation requested is provided in the attachment to this
'l etter.
In accordance with your request, I am also enclosing two additional signed and notarized originals and 37 copies of this letter and attachment.
Sincerely yours,
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b' J.'J. Mattimoe Assistant General Manager N.
and Chief Engineer j
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G{eneral Counself for D v10 S'. KAPLAYf
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Sacramento Municipa'i Utility
/.Y District 4TFQ V Subscribed and sworn to before me this
\\ R Nor 5th day of March,1976 tT* \\$
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M '. the County of S(acramento, State of etty Mattier, otary Public iir and for California My Comission Expires January 12, 1980.
2328 AN ELECTRIC SYSTEM S E R VIN G MORE THAN 600,000 IN THE HEART OF C ALI F O R N I A
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Ifegulaton] UOcket file ADDITIONAL INFORMATION RANCHO SECO NUCLEAR GENERATING STATION, UNIT NO.1 DOCKET N0. 50-312
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e 1.
Question:
When is the next refueling date and what is the proposed schedule for subsequent refueling?
Answer:
Rancho Seco Unit No.1 has had an extended outage this year due to turbine and generator problems.
The present schedule for plant operation indicates that the first refueling will occur during July 1977.
Subsequent refuelings will be on an annual basis in the spring of each year.
2.
Question:
How many fuel assemblies will be replaced during each refueling?.
Answer:
The Rancho Seco Unit No.1 core contains 177 fuel assemblies. One-third of these assemblies will be replaced during each refueling.
Core symmetry considerations for proper fuel management dictate that 56, 60, or 61 assemblies may be discharged during any given year. However, the average number of assemblies replaced each year will be 59.
3.
Question:
What is the total construction cost associated with the proposed modifica-tion of the spent fuel pool (SFP)?
Answer:
The contract price for the design and fabrication of the replacement racks is $799,000 including estimated freight charges. The current estimate for the removal of the old racks and installation of the new racks is $200,000.
This gives a total modification cost of $999,000.
4.
Question:
What are the alternatives to increasing the storage capacify of the SFp?
The alternatives considered should include, but not necessarily be limited to, the following options:
A.
Shipment to a fuel reprocessing facility.
Provide status, if any, of any contractual agreements.
B.
Shipment to another reactor site.
C.
Tennination of operations of the reactor.
.N
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'AdditiCnal Information Page 2 Docket No. 50-312 These options should include a cost comparison in terms of dollars per kilogram of uranium stored and the cost for providing replacement power within or outside of the licensees' generating system.
-Answer:
The Sacramento Municipal Utility District had a fuel reprocessing contract with General Electric Conpany. When General Electric was unable to operate their facility at Morris, Illinois, this contract was canceled. At that time, General Electric approached the District.with a proposal to provide fuel storage at the Morris, Illinois facility.
The most recent estimates by General Electric for the average cost of this storage equates to approx-imately '$10 per kilogram of uranium per year.
At the present time, General Electric has not fully committed to this fuel storage proposal. The District feels it would not be prudent to wait for such a program to develop and lose the opportunity of modifying the Rancho Seco Unit No.1 fuel storage racks before the first refueling at which time the spent fuel storage facility would become contaminated, increasing the costs and radiation exposures associated with the modification.
Like-wise, at the present time, none of the other fuel reprocessing facilities have any definite plans to provide spent fuel storage.
Since the Rancho Sece ' nit No.1 fuel storage facility is empty at the J
present time, it is not logical to attempt to store fuel from this unit at'another unit licensed to store spent fuel. Any other facility would have no more storage space available than Rancto Seco Unit No.1 has itself.
The first discharge of fuel from Rancho Seco Unit No.1 will be in the sunmer of 1977. ' With annual discharges, the existing spent fuel pool would be filled with the discharge occurring in the summer of 1980.
This implies that the District would be unable to discharge fuel in the summer of 1981 and the unit operation would have to be terminated. The District does not have sufficient generating capacity to provide the replacement power which would be required if Rancho Seco Unit No. I were to' terminate operation. Therefore, the District would have to purchase power elsewhere or construct alternate generating facilities.
This power would cost 2.5 to 3.0 cents per kilowatt hour in 1981 for a total annual cost of $150 to
$100 million, assuming a 75% plant capacity factor.
5.
Question:
Provide data on the quantity of stainless steel used in the new racks.
Answer:
The quantity of stainless steel to be utilized in the new spent fuel racks at Rancho Seco is approximately 135,000 pounds. The following list provides
Additiinal Infor 'i n C
. Docket No.50-31F Paga 3 a component breakdown of the amount of steel uiil ~ze' :
d 5 - 6x6 racks 9 8,320 lbs.
6 - 6x7 racks 9 9,550 lbs.
41,600 lbs.
=
1 - 6x7-4 rack 9
8,830 lbs.
57,300 lbs.
=
2 - 7x7 racks 9 11,130 lbs.
8,830 lbs.
=
1 - 2x6
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rack 9
2,450 lbs.
22,260 lbs.
=
_ 2,450 lbs.
=
'k 132,440 lbs.
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Fixtures Shims, Cask Catcher Attachments - Add
- 1-1/2%
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_ 2,000 lbs.
6.
Question:
134,440 lbs.
Provide the following information related to the w t system:
a er purification A.
What is the volume of water in the SFp?
B.
How many demineralizers are used and what length required to clean up the total volume of water in the p of time is C.
demineralizers and filters resulting from th ool?
~
wastes from xpansion?
Answer:
The spent fuel pool cooling system utilizes o s 606,000 gallons.
flow rate turnover. of 160 gpm.
Sixty-three (63) hours are required for tneeminera The filter and demineralizer depletion is mo Both the filter and demineralizer are expected t e
differential pressure increase rather than loss o be changed base rom the stored fuel.
radioactive contaminants.Therefore, no si of capacity to wash volume is expected from the expansion.gnificant increase (< remo
,2%) in 7.
Duestion:
Provide a discussion of the models and calculatio to personnel from radionuclide concentrations in thns used to e ing the following:
e spent fuel pool includ-A.
Expected maximum radionuc?ide concen water source terms including Csl34, trgtjon (58,/cc of {v) uci g
Cs'd, C0 and Co i
p, O~
t
. Additional Informatio'n Page 4 Docket No. 50-312 B.
The dose rate above the spent fuel pool resulting from these source terms.
C.
The total dose rate above the pool from (B) plus the contribution from the stored spent fuel pool elements in the expanded pool.
D.
The annual man-rem dose equivalent based on all operations performed by personnel in the pool area.
Answer:
The expected radionuclide concentration is:
Csl34 5x10-4 uc/cc Csl37 lx10-3 uc/cc CoS8
<lx10-4 uc/cc Co60 1xio-4 uc/cc The contaminate listed above originate from material deposited on the assemblies during their in-core life.
The cobalts are activation products of corrosion products in the primary coolant.
The cesiums are the most soluble of the fission products and originate from in-core leakers. When the assemblies are removed from the reactor to the fuel pool, the deposited material slowly backs or sluffs off the assemblies.
The pool clean up system (filter and demineralizer) removes the material as it becomes available thus resulting in the equilibrium concentration shown.
The dose rate expected from the source terms above is <lmRem/hr. This dose rate is based on experience in similar fuel storage' pools with similar concentrations of radionuclides.
The total dose above the pool is expected to be <2.5 mrem /hr made up of
<l.5 mrem /hr. from a single fuel assembly lif ted in the fuel handling machine within 10 feet of the surface and <1 mrem /hr. from the pool water contaminants. As shown in the attached Figure 7, no significant contri-bution to dose is made by the fuel covered to a depth of 23 feet with water.
Fuel being transferred is the controlling contributor to fuel basic dose rates, not the stored fuel.
As noted above, there should be no increase in radiation levels above the spent fuel pool as a result of the rack modification.
The total man-rem dose for one year in the pool area is listed below:
Activity Dose 80 man-hours clean up and inspection @ <1 mr/hr.
<.080 man-rem
^
Additional InformatNi O
Page 5
- Docket No. 50-312 Activity Dose 810 man-hours new fuel receipt and inspection @ <1 mr/hr.
<.810 man-rem 120 man-hours equipment check-out prior to refueling @
<1 mr/hr.
<.120 man-rem 60 man-hours fuel handling during refueling @ <2.5 mr/hr.
<.150 man-rem Total Annual Doss
<1.160 man-rem 8.
Question:
Provide a discussion of the models and calculations used to estimate releases of radioactive materials to the environment from the modified 1
spent fuel pool.
Answer:
Radioactive gases may be released from the spent fuel storage directly into the atmosphere of the fuel building. This air is exhausted through particulate filters and charcoal filters. The major radioactive gas that may be released during fuel storage is Kr85 with a half-life of 10.7 years.
As shown in Appendix 14D of the Rancho Seco Unit No.1 FSAR, the initial estimate of KRd5 release from the auxiliary building was 32 curies / year which is predominately from the fuel storage area.
Increasing the fuel storage from 244 assemblies to 579 assemblies (a factor of 2.37) will not necessarily increase the Kr85 release rate. The fuel discharge frou the reactor will continue on a 1/3 core / year rate and the release of Kr85 is most likely to occur during the initial handling and the first year of storage.
Nevertheless, a conservative approach is to assume that the Kr85 yearly release will increase by the 2.37 factor.-
Therefore, the maximum Kr85 release from the fuel storage pool is 76 curies /
year, an increase of 44 curies / year.
The total plant release initially pro-jected was 981 curies / year thus the maximum percentage increase due to fuel storage expansion would be less than 4.5%.
9.
Question:
Discuss the potential for fuel handling and fuel cask accidents, such as movement of transfer cranes over the storage area, that would be affected by the expansion.
Answer:
.Since the spent fuel pool does not contain any spent fuel elements at the present time, the expansion program itself has no potential for fuel
,N h
v Additional Infonnation Page 6 Docket No. 50-312 handling accidents. Since the pool geometry itself is not being changed by the expansion, the expansion program does not introduce any new or different considerations for fuel handling and fuel cask accidents. The movement ~of the fuel handling bridge and turbine gantry crane will remain as described in the Rancho Seco Unit No.1 Final Safety Analysis Report.
- 10. Question:
Discuss the storage or disposal of the original fuel racks.
Ansver:
Since spent fuel has never been stored in the Rancho Seco Unit No.1 spent fuel pool, the existing fuel storage racks are essentially uncon-taminated.
It is anticipated that very little effort will be required to bring any existing contamination levels down to those acceptable to allow disposing of the original fuel racks as ordinary scrap metal.
11.
Question:
i
.Please provide details as to the location of the failed fuel storage f
locations in the pool.
If they are not part of the array, discuss the reactivity effect of their presence.
If they are part of this array, provide assurance that they contain at least as much neutron absorber as the regular storage location.
Answer:
Figure 11 provides details as to the location of the failed fuel storage locations in the pool.
The neutron absorption effect of the failed fuel storage racks is such as to make reactivity less than for the normal spent fuel rack array. This is due to the additional 2 inches (minimum) of water between the failed fuel storage locations as opposed to the normal spent fuel racks.
The additional 2 inches of water is worth more as a neutron absorber than the.077 inches of stainless steel utilized in the spent fuel guide tubes.
- 12. Question:
Provide sufficient detail as to location and arrangement of temporary storage modules in the transfer canal to support the assertion that they are safe as regards criticality.
In particular, provide assurance that the transfer path is far enough from these locations to provide negligible neutron coupling between transferred and stored assemblies.
Answer:
The proposed fuel storage rack modification does not involve any modifi-cation to the temporary storage locations in the transfer canal inside the reactor containment. The NRC staff safety evaluation of this portion of the fuel handling system is therefore unaffected.
3 9
w!
e Additional Infcrmation Page 7 Docket No. 50-312 13.
Question:
Assuming the loss of all cooling systems resulting in a bulk temperature of 212 F at the surface of the pool:
A.
calculate the outlet conditions of the coolant from the hottest subchannel of the hottest bundle. This should include coolant temperature and pressure and, if applicable, steam quality and void fraction; B.
show that the cladding will not swell or rupture due to the cladding temperature and the internal pressure from fission gas present in the fuel rod at end of life; and C.
show that the void fraction of the we.ter is zero inside and between the fuel bundles over their entire length, or else that keff is at a safe level when boiling occars inside and between the stainless steel storage tubes.
Answer:
Loss of coolant is not a credible accident as two Class I systems are available as backup to supply cooling water.
For the sake of analysis, it will be assumed that coolant flow is lost.
Then, the means of dissipating the heat is to allow the pool to come to a boil and the heat be removed by vaporization.
Figure 13-a illustrates -
the system for this heat removal.
The fuel is stored in the bottom
- 14 feet of the pool.
There are
- 23
.oet of water above the stored fuel.
Within the fuel assemblies the water is heated and rises by natural convection due to the reduction in density. As the warmed water leaves the top of the fuel assemblies, it mixes with the large reservoir of water above the racks and it in turn heats this reservoir and causes convective currents which mix this water.
There are three temperatures of interest in evaluating the coolant temperature as it leaves the fuel assemblies.
A.
The temperature of the surface of the pool, which at equilibrium is 212 F (boiling).
B.
The temperature at which boiling occurs at a depth of 23 feet (top of the fuel assemblies) is E38 F.
C.
The temperature of the coolant which is available to flow through the fu91 assemblies which is the same as that shown in (B) of 238 F.
D Q
Additienal Infomatien Page 8 Docket No. 50-312 To detemine the temperature of the cooiant which is available to flow through the fuel assemblies, it is necessary to analyze the free con-vection heat dissipation loop in the top 23 feet of the pool.
This free convection loop was analyzed using the method presented in " Heat Trans-mission" by W. H. McAdams, 3rd Edition, 1954, Page 180. Analyses show that the temperature of the inlet coolant will be somewhere between 212*F and 238'F, depending on location of the specific storage cell, location of other bundles, etc. For conservatism, an inlet temperature of T = 238*F will be assumed. At this temperature, and using a maximum energy fuel bundle, the coolant outlet conditions are T = 238*F.
Steam quality = 0.171%. Void o
fraction = 63%.
An existing B&W FSAR calculation perfomed for Rancho Seco end-of-life clad conditions (Table 3.2-19, SMUD FSAR) demonstrates that the max clad circumferential stress is 50,000 psi with an ultimate stress of 62,500 psi.
These stresses are based on a clad temperature of 425 F and hot or operating internal pressure conditions.
Using the conservative assumption that the internal gas-pressure within the stored fuel in the racks is the same as the design basis (3300 psi) the calculated mean circumferential stress is:
0 sa 96 23,500 psia
[pg.3.2-63,66]
=
Without performing a more rigorous calculetion and not taking credit for the maximum clad temperature lying below 270 F for the conditions specified during loss of cooling systems, there still remains a safety factor of 2 between the accepted E.0.L. design basis stress of 50,000 psia and the i
23,500 psia conservatively estimated for the stress of stored fuel assemblies.
The void fraction within the irradiated assemblies is not zero and as shown in 13A could be as high as 63% under the ultra-conservative conditions assumed. Within the fuel assembly an increase in the void fraction reduces the neutron moderation which in turn decreases the keff. The other factor which favors criticality control is that the higher burn-uo fuels (less fissionable material) are the ones within which boiling takas place.
Therefore, they become even less reactive with boiling.
Bulk boiling between the stored fuel canisters is not credible because,
the high flow rate of water induced within the assemblies carries the heat to the large water reservoir above the racks where the heat is dissipated by evaporation at the surface.
The boiling point of the water returning to the bottom of the assemblies is graduall; increasing as the static head above increases thereby suppressing steam formation.
- 14. Question:
Reevaluate the consequences of dropping of the fuel cask, taking into account the closer spacing for the proposed spent fuel locations..This evaluation should include the possibility of the fuel cask. tipping or 1
Additional InfomatiC b
Pag 2 9 Docket No. 50-312 rolling into the spent fuel. Also provide diagrams showing the location of the spent fuel racks in the pool and area of impact in the event the cask tips or rolls into the pool.
Answer:
The area of the spent fuel pool occupied by the modified fuel storage racks is the same area occupied by the original acks.
The fuel storage i
rack area is separated from the cask set down area.by a large steel cask catcher.
This cask catcher prevents a fuel cask from tipping over when it is in its loading area.
The cask catcher will also prevent the cask from falling into the fuel storage racks if it were to tip over while setting on the intermediate step where its head is removed and installed.
The consequences of the fuel cask dropping into the cask area is as described in the Rancho Seco Unit No.1 Final Safety Analysis Report, Answer to Question 5.A.43, Page SA-38. ' The closer spacing for the proposed spent fuel racks has no effect on the evaluation of the conse-quences of dropping a spent fuel cask.
15 Question:
Provide a list of all seismic and non-seismic systems which can be used as makeup in the event the spent fuel pool cooling systems fail and it cannot be repaired within the time limits speci.fied in your proposal of 1
December 19, 1975.
Answer:
i The design of the spent fuel pool cooling system incorporates connections with the decay heat removal system to provide the necessary backup for spent fuel pool cooling should the spent fuel pool cooling system fail.
There are two Class I decay heat removal systems which can be used for spent fuel pool cooling.
16.
Question:
Diagrams or sketches of the new spent fuel storage racks have not been provided.
Provide such diagrams which indicate the general arrangements of the lateral bracing and the locations and details of vertical and horizontal supports.
Answer:
See Figure 16 attached.
17.
Question:
Provide a diagram which schematically represents the dynamic model used in the seismic analysis.
Indicate the support points, gaps, locations of translational and rotational springs, if utilized, and the method employed to account for the dynamic effects of the pool water.
I
3-
)
Additional Information Paga 10 Docket No. 50-312
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Answer:
See Figure 17 attached.
The support points are nodes 23 and 32 in the model.
No gaps were used in the computer model.
Side adjusting screws were modeled as translational springs.
There are three of these springs, two located at node 33 in the model and one at node 15. No rotational springs are used.
The dynamic effects of water surrounding the spent fuel guide tube was considered by lumping additional mass to fuel guide elements as explained in the EDS report, Section 3.2.2.2.
"The hydrodynamic resistance of the pool water to motion of the racks was included as a guide tube added mass.
The hydrodynamic resistance to motion of all other rack structures (with the exception of the fuel assembly) was assumed to be negligible. Since the guide tubes move together during seismic loading and are spaced with large clearances, the hydrodynamic added mass is closely approximated as the mass of the water enveloped by the boundaries of the guide tubes (i.e., the mass of a volume of water equal to the length multiplied by the width and depth of the guide tube).
The added mass calculated in this ~ manner was then combined with the self-mass of the fuel guide."
18.
Question:
On Page 7 a discussion of the lateral clearance of 1/8" for thermal expansion is presented. State the pool temperature at which contact with the' wall is anticipated and the contact pressure during normal operation.
Specify whether or not these connections are relied upon to transmit shear.
Provide a description of the frictional resistance of such connections and the effect of this resistance on the seismic analysis.
Answer:
Contact between racks and the pool walls will not occur at any temperature condition up to and including a worst case thermal transient.
The worst case transient is based upon reduction of spent fuel pool cooling from two cooling systems (spent fuel pool cooling and decay heat removal) to one cooling system (spent fuel poo? cooling).
The transient considered results in a fuel rack temperature of 180*F coincident with a concrete mean temperature of 136*F.
Differential growth between the racks and concrete is calculated to be 0.264 inches, which would be compensated for by leaving a clearance of 0.132 inches between the racks and the pool walls on all sides. The actual clearance at installation will be set at 3/16 inch to avoid interference between the racks and the pool walls under all temperature conditions.
The wall jacks are not relied upon to transmit shear.
Computations show that for the expected value of friction coefficient between the levelling screws and the pool floor liner (u = 0.5), the racks will not
q y
Additional Information "
P39311 Docket No. 50-312 move on the pool floor during the OBE or the SSE.
In the event the friction coefficient is low enough to allow the racks to move on the pool floor in the SSE, the jacks oriented in the direction of motion are strong enough in compression to absorb all forces imparted by the racks.
- 19. Question:
Regulatory Guide 1.61 is referenced for determining the damping values of the welded steel storage racks.
However, this guide does not discuss structures innersed in a fluid.
If damping values are incorrec'if assessed, a shift in the response frequency may occur, which could lead to an unconservative evaluation of the system response.
Discuss this possibility and demonstrate that such a shift in system response would not adversely affect the fuel storage racks.
Answer:
The damping values used for the response spectrum analyses of the spent fuel racks were the values specified for welded steel structures in Table 1 of USNRC Regulatory Guide 1.61. Additional dampening wnuld result from the hydrodynamic effects and also from the presence of stored fuel.
This additional dampening was neglected in the analysis.
The shift in natural frequency of the system which would result from this additional dampening is negligible.
For example, for a total damping to ten percent of critical, the natural frequency is shifted by only 0.5 percent. While the effect on natural frequency is negligible, the higher damping would significantly reduce the response amplitudes and consequently the forces.
Based on the above considerations, it was concluded that neglecting the additional dampening is conservative.
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8 10 12 DISTANCE FROM TOP OF ACTIVE FUEL, FT Figure 7
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':iMENS10NS SHOWN ARE MINIMUM DISTANCES f
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SMUD FAILED FUEL RACK - LOCATION OF FAILED FUEL FIGURE 11-
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