ML20104B377
| ML20104B377 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/08/1992 |
| From: | Burzynski M TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-TVA-SQN-TS-92-01, CON-TVA-SQN-TS-92-1, RTR-NUREG-0612, RTR-NUREG-612 NUDOCS 9209150294 | |
| Download: ML20104B377 (7) | |
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k Tennessee Va4ey AutWy 1101 MaAet Street.Chananmga. Tennessee 3N02 September 8, 1992 10 TVA-SQN-TS-92-01 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Gentlemen:
In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SQN) - RESPONSE TO QUESTIONS ON REQUEST FOR LICENSE AMEN 0 MENT TO TECHNICAL SPECIFICATION (TS) - SPENT FUEL POOL STORAGE CAPACITY INCREASE The enclosed pages respond to your questions concerning the subject submittal. We received the questions from you on July 23,_1992.
A followup discussion for purposes of clarification was held by members of your staff and TVA on July _28, 1992.
Please direct questions-concerning this issue to_C. R. Davis at (615) 751-7509.
Sincerely, Mark J. 'urzynski
-Manager Nuclear Licensing and Regulatory Affairs
-Enclosures-I cc: See page 2 Y
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U.S. Nuclear Regulatory Commission Page 2 4
Seatember 8, 1992
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cc (Enclosures):
j-Mr. D. E. LaBarge Project Manager j
U.S. Nuclear Regulatory Commission One White Flint North 1
11555 Rockville Pike Rockville, Maryland 20852 j.
Mr. Michael H. Mobley, Director (w/o Enclosures) j Division of Radiological Health T.E.R.R.A. Building i
150 9th Avenue, N 1
Nashville, Tennessee 37203 i-NRC Resident Inspector Sequoyah Nuclear Plant-2600.Igou Ferry Road 3
Soddy Daisy, Tenr.essee 37379 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission i
Region II l.
101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 i
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-ENCLOSURE s
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A.
Control of Heavy Loads Questivns i
1.
The licensee has comaltted to comply with the general guidance of Section 5.1.1 of NUREG-0612. " Control of-Heavy Loads at Nuclear Power _P lants,7* as documented in Table 2.4.1-of the licensing report.
However._Section 5.1.2 of NUREG-0612 is also-applicable.
I to the rerack operation.
Provide _ documentation clearly.
describing how compliance with Section 5.1.2 of NUREG-0612 will j
t'e achieved for all phases of the rerack operation. Control of i
the movement of the impact shield should be specifically addressed.
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2.
Show that postulated load drops during_the re"acking operation i
will not damage the fuel pool liner-and structure to such an extent that the stored fuel may become_ uncovered.
Responses:
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-Section 5.1.2 of-HUREG-0612 was not considered an applicable-F commitment for the rerack operation because Generic Letter 85-11
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cancelled the. Phase Il r'eluirements of_Genei!c Letter 81-07.
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However, in response to'the question. it is~noted that Option (4) of.
j Section 5.1.2.of NUREG-0612. calls for satisfaction of.the evaluation 1
criteria nf Section 5'1 and evaluation of the consequences of a postulated heavy load drop in conformance with the guidelines of i
Appendi.v A to that dociiment.
Brsed on the projectecfoutage: schedule
-for.the Sequoyah reactor units, all fuel stored in.the pool'is-
. expected to nave undergone 90 days of decay at the time the heavy 4
load movements-are begun ano is certain to have decayed for at least
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54 days, because_the Un:t 2 end-of-cycle 6 refueling _ outage duration l
is projected to De 65 days. Almost all the radiciodine and.
l short lived xenun ~and kryptun will have _deccyed to: very low-. levels.
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The-gaseous radionuclide remaining is1Kr-85.which mak.es a minor L
contribution to any offsite dose, This limits releases of-radioactive material that may result from damage to-spent = fuel Also, as indicated in Table 1.1.11of the submitted report, " Spont Fuel Pool Modification for increased Storage' Capacity," a' total of.-
F approximately 900 fuel assemblies will be stored _in the pool at the time ofireracking.
Ninety percent of these will have' decay _ times.
which significantly exceed 90 days.
This further limits the L
potential for radioactive material release. -Accordingly, based on a:
comparison of'this-information and: specific rerack calculations-performed for a Sequoyat rue _1 handling. accident with NUREG-0612-Tables 2.1-1 an" 2.1-)..ne accidental dropping of a postulated-heavy
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load will-prou;ce doses that are well within-(less.than one-fourth)-
i 10 CFR Part 100 limits.
i Section 2.2.C of-NUREG-0612 c6ncludes.that there appears to tfe no potential.for a criticality-situation'dueito a hecvy' lead drop--in_a i.
PWR spent fuet pool ~which-contains-only_ totally spent fuel.
No fresh fuel-and very little, if any.~ partially-burned fuel is expected to be
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stored in Sequoyah's pool during installation of the new racks.
The racks contain large amounts of Boral neutron poison and structural steel.
Section 2.2.6 further concludes that_there is-low potertial-for criticality if PWR spent fuel racks containing non-spent feel are-crushed.
Section 4;l of Appendix A to NUREG-0612 allows _ boron credit for-load. drop' accidents under certain conditions.__Sequoyah's technical specificaticns will' satisfy these conditions in that boron concentration must be determined by chemical analysis to be greater than or equal to 2000 parts per millien_(ppm) at_least once-per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during fuel movemcnt.
Such fuel movement or shuffling will be required on a continuing bacis as toe new racks are-installed.
Section 4.1 of Appendix A further states that the largest array of non-spent fuel a licensee should have to consider _:is that of an offload core, and Section 4.2.2(2) states an estimated maximum-reactivity insertion due to crushing is 0.05 6k.
By comparison the 4
reactivity value of 2000 ppm boron-for Sequeyah's new and existing racks is'between 0.15 and 0.20. _ Therefore, the accidental dropping of a heavy load will not result in a fuel confir ration such that k rr is larger cben 0.95.
e Appendix A of-NUREG-0612 mandates that the structural analysis'of-the load be_ predicated on se following bases / limitations.
(i)
The load drop orientation is the most' adverse which would result in the most severe consequences.
(ii)
The fuel has decayed for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
(iii) The true stress-strain relationship of the deforming structure is employed.
Postulating accidental-drop of a rack along with its lifting attachment from the maximum possible height, and further postulating the most adverse' physically admissible dropiprofile; _results in c-primary impact of the_ rack with the pool slab. Recognizing that the pool slabLis. located over a rock subgrade and buttressed with approximately 20 feet of rainforced concrete, the-postulated rack drop is incapable af actuating a gross structural-failure. This conclusion is substantiated by prior analyses performed for recent rerack licensing submittals suchias the one for Three Mile. Island Unit 1.
-The physical integrity'of the pool and its function:as a i
. container _of cooling and shielding water is, therefore, unimpaired.
While' localized danare to the liner can be hypothesized, the associater: lcakage_will be' minor'and contained within the relatively-small volume of. the leak chase system,_ and _thereforei strictly =
speaking, does notLconstitute loss of water from the spent fuel pool system. Ale? the-leak rateLwill be small. compared to'available installed akeup capacity from the Refueling Water Storage Tank; (RWST). tach reactor unit.has an RWST-with'a volume of 170,000 gallons and a boron concentrat. ion maintained-between-2500 and
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2700 ppm as required.Dy technica1 specifications. Borated water may i
be supplied from'the RWST via-a. refueling water purification pump, which has a 200 gpm design flow.
Two'such pumps are available.
Alternatively, a temporary line can be run from the boric acid blender, located in the Chemical and Volume Control System, directly into the spent fuel pool.
in summary,. the cooling and shielding of spent; fuel-in the pool would remain unaffected by a postulated heavy lsad accident during the rerack operation.
4 With respect to the movement and placement of the impact shield, we note the following items:
1.
The weight of the impact shie3d is approximately 3-1/2 tons.
This is-less than the: weight of the heaviest rack so the.t movement of the shield c"er the ' designated heavy load path-i outside of the pool area is not a bounding case, 2; The impact shield is not moved over any portion of_the main-pool, but is m;ved into'its final. position by moving oirectly'over the cask pit..The impac, shield geometry is such that during this movement, -the shield supports will-be over the cask pit _ concrete walls. He shield itself will be parallel with the horizontal-piane, and the heigh _t of travel of the shield above the top of the cask pit surrounding walls will be minimized.
Because of these_ factors, there is no credible scenario by which the impact-shield could drop into the cask pit, rather, any accident during movement would simply-bring the shield supports onto the top of the supporting walls.
3.
Spent fuel stored in the cask pit area will always havo aged at least one year.
In summary, fuel stored 4 the cask pit islnot'placed in jeoparoy by any uncontrolled verticti mvement'.of the.impcct shield during its installation or remova.
B.
Thermal _-Hydraulic Considerations:
Q_uestion 1.
The discharge scenurios of Section 5.4'of-the licensingl 2 port assume feci transfer begins.after 28T hours of: decay in the'
- reactor vessei. _ However, Section 3/4.9.3 of the Sequoyah-Technical Specifications permits movemer.t of irradiated fuel after only 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of decay.
Evaluate the impact of the l
' shorter decay time 0n spent. fuel pool design limits.
L Resb6nse:
Additional thermal-hydraulic analyses were-performed'assumirg.that the transfer to Lthe spent fuel pool: begins af ter 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of-deco.y.
The. discharge moJes--denoted as Casei la and.:1b were reanalyzed.
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Case Ib had previously yielded the maximum pool t.ulk temperature as
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shown in Table 5.5.1 of the submitted report " Spent fuel Pool.
Modification for increased Storage Capacity." The transfer of fuel-to the pool was assumed to begin after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of-decay and-be completed within 12 days of reactor shutdown. The following'results were obtained.
l Tmax (Maximum Pool Coincident Time After Case WaterTemperature)(,F)
Reactor Sh_utdown (hrs) la (revised -
2 cooling trains 139.5 291 4
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I tooling trains 177.2 293 It is noted that these'temperaturis are well within the range of..
typical norms for PWR -fuel pools. ' The corresponding temperatur es for 1
Cases la a~! lb in Table 5.5.1 are 138.0*F and 174.9'F, respecti ely. The' spent f"31 pool design limits-are not exceedea.
Question 2.
Describe the degree of redundancy in spent fuel pool cooling.
trains provided by the backup spent._ fuel pit pump with regard _to-single tctive failures.
Potential. single iailures-in support systems _such as the electrical distributi_on system and the component cooling water system should be considered.
.asponse:
'There are three spent fuel pool cooling pumps'in the spent fuel pool cooling system.
Pumps A~-A and B-B are trained. "The third pump C-S can'be aligned-to either. train and can be powered.from either train.
Each power source to the backup C-S pump-is an independent Class 1El electrical se,; ply.
'A mechant 'l interloc_k is-provided_onEthe, power supply t.ansfer panel to prevt..c the C-S-pump.from being powered from both trains.
Normally, _'one SFP pump and one heat _ exchar.ger is required to handie the existing heat load, but there are times when two pumps-and two neat exchangtrs (both trains) are used.
In the event of_ failure of'one spent _ fuel pump, the back'up pump'(C-S)'would
.be aligned'and operated.
In the_ event.of. failure of one heat exchanger,' cooling would be-done iitth the other train..The systems can be shut dc o for limited oeriods of time for maintenance or-replacement of malfunctioning _ components.
The Compunent Cooling Water: System (CCS) " designed such that.no F
single active failure will. interrupt cooling. water to both A an.d B-.
safeguard-trains.
The system consists of five CCS pumps and three-pairs of heat exchangers; serving both units'.-
The. heat exchangers are designated as Heat Exchangers 1Al/lA2, 2Al/2A2, and 0Bl/082.
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Cooler IAl/1A2 serves train A: loads in Unit-1, Cooler 2Al/2A2 serves train A loads-in Unit 2, and Cooler 081/082_ serves train B loads in-both' units. During-full _ power operation, with spent' fuel-pool maximum cooling required, two CCS pumps and-one heat-exchanger pair?
may be used for the unit carrying _the spent fuel pooi_ heat-load.
The ether unit requires only one CCS pump and one heat exchanger pair.
_A' CCS pump capable of supporting either unit is aligned to the third heat _ exchanger pair to serve-either unit's train 8 safeguard equipment.
The fifth pump may be used as-backup for either unit by transferring the spent fuel-pool. load to the unit aligned with that
-pump.
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