ML19259D288
| ML19259D288 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 07/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19259D279 | List: |
| References | |
| NUDOCS 7910170574 | |
| Download: ML19259D288 (19) | |
Text
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f JUCLE AR llEGULA10RY C O '/>'.;l S'ii O f J n/amo c l ou, p.
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$eq SAFETY EVALM.'Ji10" DY TMS OFFICE OF ' UCI E.' R EE/,CTOP EEGUL/.T10'.
b jJd SUPPORTI :G FACILITY !CDIFICA'lIO::S TO I?.CRE.SE THE CAPACITY OF 'll:E SPINT I UEl, S'ICRAGE POOL A' D TO PRC'/IDE PR.OTECTIO': FRP?! A POS'IUIAT O CASP DROP ACCIDE"T, AND AWiNDME'.T TO I,ICENSE NO. DPR-45 (CilANGE TO TliE TECHNICAL SPECIFICATIONS)
DAIRYIMD POI.TR COOPERATIVE LA CROSSE P.0II.ING 1lATER REAC'ICJ g
g { N{ lL DOCKET NO. 50-409 Unif31d ll Introduction By Jetter of Dccenber 12, 1974, Dairyland Poc;cr Cooperative (DPC) informed thc-N0 clear R:p,ulatory Co.aission (the Conmission) of its intention to increase t h; capacity of the La Crosse Boiling Water kcactor (LAChim) spent fuel storage pool by installing cdditional storage raths.
As part of our r-eview of the proposed nodification of the spent fuel storage paal, we considered accidents which potentially could be affected by the modification.
A spent fuel cast drop is an accident which potentially could be affected by the spent fuel storage pool expansion.
!!andling of a spent fuel cash at t he LACBNR has been under m view, but has not been previously approved.
To determine the effcet of the spent fuel storage basin expansion on the fuci cask drop accidcut, it tras necessary to complete our evaluation of the spent fuel shipping cask,
handling system as part of this safety evaluation.
Therefore, this evaluation has been divided into two parts; the first considers the spent fuel storage basin expansion except for the drop of a spent fuel shipping cask, and the second considers the spent fuel shipping cask system including the effect of a cask drop with the modified spent fuel storage racks.
A.
Spent Fuel Storage flasin Expansion The present configuration provides storage cells to acecnmodate S4 irradiated f uel assemblies.
The proposed nodification would provide 134 storage cells, a capacity increase of approxicately 6 0 *..
'lhe proposed :redification would ultirc.ttely provide a means 1162 334 gnort067f'
s.
cf rer.oving the complete core frer the reactor vessel.
!!aving the capability to remove all the fuel fro. the reactor vessel is desirabic to facilitate taintenance of reactor internals if required.
Additional inforration was requested on !hrch 7, Ibrch 24, June 12, and July 24, 1975.
Rcsponses from DPC were received.'! arch 31,
!!ay 5, June 14, and July 25, 1975 respectively.
Discussion The proposed new spent fuel storage racks will be fabricated from individual cells and welded together in assemblies.
A total of four asser.blics will be fabricated, two will contain 11 cells and two will contain 14 cells.
The 11-ec11 asscmbly will be welded into a single row.
Two rows of seven single-welded assenblics will be bolted together to form the 14-cell assembly.
The 11-cell assemblics will be captured between the existing fuel storage racks on the east and :: cst walls.
'Ihe 14-cell assemalies will be located on the north and south walls in front of an existing rach.
The neu racks will be restrained fron movement by attaching them with angle braces to the existing racks.
Boral sheets will be installed as neutron absorbers to ensure that a self-sustaining neutron reaction cannot result from the proposed spent fuel pool storage configuration.
The Boral sheet's will be permanently fixed to the sides of the storage rack:: by a welded structural angle frar..e and thereby will becoue an integral part of the storage structure.
The 11-cel1-asser..blies will have a boral sheet attached to both sides.
The 14-cell assemblies will have boral sheets attached only to the side facing the existing rack.
Evaluation The ILUMER code was employed by.DPC to calculate the neutron spectrum and the resulting broad group cross sections for_the appropriate combinations of materials found in the storage pool.
. The neutron diffusion code EXTEIUiINATOR used the broad group cross sections to calculate the Ke - eigenvalue for an appropriate geometric model of the fuel assemblics in the storage pool lattice.
The ILUMER/ EXTERMINATOR calculational model was verified by a comparison of the. calculated eigenvalue of the initial 10-assemb1y LACBWR 1cading with the experimental value.
The reported difference of 0.11791 is well within experimental accuracy and we find this acceptabic.
11A)
(7c f i U...
4.)
3_
Thesc cor.pu c r c odes M ct orically h. /c Leen aced by the nucicar industry to perform this type o' e n a l." > i r.
ile enree 1:ith th" tre of these codm.
The value of K rg calcul:.ted 'cith the fuel asu.blics in their e
nominal location? in the three-rot rN h and ttro-rou rnch can 0.563 and 0.822 rc;.pectively.
rlc indeparcntly perform d a criticality analysis to demnstrate t hat the proposed spent fuel pool storage racks were properly spaced and the I; oral Absorber r..aterial adequately utilized.
The analysis yielded a E rp of 0.56 denoastrating that c
a sclf-sustained reaction could not occur.
The above values are t. ell belo.. the existing Tec'.iical Specificatica upper limit of 0.90 and are considered acceptable.
io assess the effei't of fuel asser:bly location in other than noninal positions, ana1) scs were perfor::ca t o dt t e:..j ne !lc for geometries resultity; frca a m"n deviations of allo..alle tolerances and clearances.
These effects changed Ee by 1ess than 0.35.
Thus, ne conclude that the uargin to criticality, in execss of 101 Ah, is core than adcquate to allo. for any calcula ticnal uncert aintic s.
The proposed :
c,.s include provisions for storage of shroud cans The fuel assc:.;hly shrouds in the reactor enclose the fuel assc:sblies laterally and provide guide channel: for the cont rol rocis.
The shrouds also provide lateral su,: port for-the fuel assemblics and a seating surface for the fuel assembly no::les.
A rotating lock provides positive holddown of the shroud in the core.
lihen a shroud is placed in a storage cell it is locked in place by the shroud locLing ring in the sr.ne nanner that shrouds are locked into position in the reactor core.
Spent fuel can then be lowered into the shroud can.
1lc conclude that use of the lecting ring in the spent fuel racks nahe it physically inpossible to renove a shroud can when removing the spent fuel element and'is therefore acceptable.
The Loral shect described above is similar to standard alclad sheet.
Corrosion data valid for alclad is also applicable to Eoral sheet.
The corrosion resistance of alteainum is prinarily due to a passive layer of aluminu:a oxide.
Alclad sheet has excellent corrosion resistance to water.
The controlled pil band (5.5 to 7.5) vill inhibit corrosion.
If, however, corrosion of the Doral sheet cladding did occur the consequences would be ninimal 1ecause the cladding acts as a sacrificial cover and the Poral sheet is not used as a structural menber of the rack.
The loss of B,;C uould be i:1significant and suberiticality of the pool will be assured.
1162 336
The proposed fuel s t or.. - rack design ras reviewed fo;; naterial select ion, material stabili'y, corrosion resist ance in the proposed environment and tenperature, and fabricatica procedures including welding.
tle conclude thnt the proposed rack design is satisfactory in all the above nspects.
'The Decedber 12 1974 submittal stated 1 hat the pr<, posed additional storage racks were to be free standing and captured between existing racks.
h'e have reviewed the design criteria and the design methods presented by DPC for the nodifications.
tlc conclude that the additional racks provide structural safety margins at 1 cast ar large as the existing racks and are therefore acceptable.
By application dated September 26, 1975, DPC reqtasted approval of a proposed nodification to he rade to the additional spent fuel storage racks.
The nodification proposc: to attach the north and south call racks to the existiT,g : ck using angle Icaces, thereby, providing a greater denree of stability and additional safety ;argin for seismic events.
This chmge to the December 12, 1974, application was requested by DPC with the understanding and condition that in doing so the Society Against Nuclear linergy_.(SKlE) and Mr. David S.
Sinpson would withdraw in total their petitions for leave to intervene, llc have. reviewed the design criteria and the design nethods presented by DPC for the nodi fications.
h'e conclude that the captured racks and the additional braces will increase thu safety margins and that the design provides reasonable assurance that the racks will withstans the postulated loads without impairment of structural integrity and will perform their functions as ntended and are therefore acceptable.
DPC perferned an evaluation of the existing spent fuel pool cooling system to deternine if it was adequate for the increased heat load that could result from the storage of additional irradiated fuel.
The cyaltiation showed that the existing cooling systen has sufficient capacity such that the temperature of the spent fuel pool will not exceed 1200F for the nost adverse loading condition.
The nost adverse condition postulated is unloading of the complete core, asstraing infinite irradiation of the fuel to be unloaded and infinit e decay time of the spent fuel alrendy in the pool at the time the core is unloaded.
The present technical specifications limit the tenperature of the spent fuel storage pool water to 1500F.
tie have reviewed the evaluation anJ conclude that the existing design _
of the spent fuel cooling system is acceptabic.
e p
D
An am. lysis of the spent fuel pool heat up rate in the event the spent fuel poo! cooling syrten fails was perfor ed by DPC ar.4 evaluated by us.
The "jnin,tira to reach boiling froa an initial uater terperature of 120 F was 17.2 hcurs under the most mdverse conditions.
0 Thus, ue concludt there is sufficient tine for the operator to take corrective ac. tion such as repairing the system or establishing a tenporary cooling source.
The stainless steel liner in the spent fuel pool, leak detection devices, and assurance against accidental and deli!-' rate drainage were provided and were addressed in the original 1 AR.
These items were evaluated and accepted at that time.
Since ac proposed nodification to the spent fuel pool does not eff(_t these itens, they arc not addressed in this evaluation.
We have considered the radioactive effluent effects of the proposed enlargenent of the spent fuel pool storege capacity during routine operation at I.AC BhR. 'ihere will be no change in ihe physical si c of the fuel pool or of the spent fuel pool c1' nnup syste:r.
In our e
evaluation de have considered only the effect of the releases of radioactive reiterials that may result from the presence of 50 additional irradiated fuel assenblics in the pool.
b'e have considered that releases of radioactive caterials occur only frora fuel.that already contained cladding defects at the time of discharge from the reactor and that no new cladding defects will occur while fuel is being stored in the pool.
At the present time, the fue1 LACBhR contains S1 irradiated fuel assemblics, 6{ storage pool at of which were discharged in August of 1972, 26 in March of 1973, 24 in November of 1973, and 25 during the 1975 spring outage.
The increased storage capacity will allow DPC to store the oldest 50 elements in the pool for an additional ti;o years. ~We have considered the potential effect on the release of radioactive materials due to inercasing the storage time for
~
50 elements from one year to three years.
The only significant radioactive nobic gas isotope remaining in the spent fuel that had been stored f'or 2 year or longer would be krypton-85, since short-lived nobic gases have decayed to negligibic amounts.
We have calculated the inventory of krypton-S5 in an average fuel + rod that achieved the design equilibrium burnup of 15,000 'Md/MTU to be 0.9 Ci/ rod.
In our c caination wc considered that l '6 of the fuel has defective cladding.
Since each fuel assembly contains 100 rods, we conservatively esticate the anount of krypton-S5 that could potentially 116.2 3.58
^
300R ORIGlu be released from the additional 50 assemblics over a two-year period to be 45 curies, which is considered negligible.
The expected release of krypton-85 from fuel stored in the fuel storage pcol will be a sm'.11 fraction of '3 curies a year, since expected fuel defect IcVels will be less th n IS and since the fraction of the brypton-85 that escaped fro::. t he fuel while still in the reactor vessel vould be released througL the main condenser offgas systca.
The resulting offsite doses from this incremental release will be negligible and are therefore acceptable.
Iodine-131 releases from the plant will not be ine' cased by increasing the spent fuel pool storage capacity since the spe t fuel previously stored in the pool has decayed for up to ce'.cral ears and the iodine-131 inventory in the fuel has decayed to negligible levels.
Non-volatile fission preacts and corrosion products that enter the water in the fuel storage pool vill be removed by the fuel pool cicanup systeu.
The cleanup systen consist s of a filter and a 30 gpa nixed-bed demineraliner, which is in continuous service.
The fuel storage pool contains 18,500 gallons of water, so that the cleanup systen is capable of processing the storage well contents in approx 5nately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The dcminerali: or resin is normally replaced once every six uonths.
The principal effect of the ine,reased spent fue.1 storage capacity due to leakage of non-volatile radioactive matcrials into the pool uater will be an increase in the quantity of radioactive naterials accumulated on the storage well filter and demineralizer, which can be disposed of as solid waste..In.our evaluation we. considered that the quantity of long-lived radioactive naterials removed by the cicanup system increased in proportion to the increased storage capacity, there_ fore, the quantity and curie content of the solid i:astes frc the storage well cleanup system would increase by 59%.
Ilouever, these wastes are a sna11 fraction of the total quantity of solid wastes shipped froa the site, so that the
'overall iupact on solid waste shipnents would be negligibic.
Based on our evaluation we conclude that the proposed nodification will have'a negligible effect on the radioactive materials released from the site, and is therefore - acceptable. -
bWe have evaluated the radiological consequences of a fuel handling accident at the LACDNR facility and have made independent calculations of tlic oE5ite doses.
It was assumed that a fuel handling accident damaging all of the fuel rods in a single assembly occurred 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown, the car]1est possible tine in which the vessel head could be removed.
This is a conservative assumption because the renoval of the vessel head curing S refueling shutdouns, the first conacncing on December 2',,
196S, and the last on May 9, th,n M hourr.
1975, has ne r: r eccurred in len e
I i 6.2 M 9
, ?00R BRIGINAL In the event of an accident at LACF..T<, a high radiat i on si gnal will autceatically isolate the contain:: ut within 30 seconds by closing isolation d:...pers in the intale anc cxL m t du;ts cf thc contai m nt ventilntion systen.
'Ihe point at which the exhaust air a s ceasured for high radioactivity levels is approximately 15 feet upstrenn from the location of the ; solation dr pers, tsed on n axi: r exhaust air flo., rate of 6, *00 cfn in the 20-inch duct, the radioactivity from an accident could be transported past the isolation danpers before the high-radiation signal results in danper closure.
Therefore we have assuaed in our calculations that all of the activity released into the containuent building fron a fuel handling ;ccident is exhausted to the atnosphere.
Our dose calculation nodel was based on tne assunptions given in Regulatory Guide 1.25.
A ground level release was assumed in the calculations, and no crcdit was given for iodine rcr. oval by the filters in the contajnnent exhaust centilation sys m.
The doses calculated -from a postulated accident damsing all of the fuel rods in h single fuel assembly at the r.inimun exclusion area boundary
~ distance (11.09 feet) were within the linits of 3 0 CFR Part 100 exposure guidelines.
Increasing the capacity of the spent fuel storage pool does not affect the consecuences of the postulated fue) handling accident involving a single fuel assembly at the earliest possible reactor head removal tine.
lie conclude this to be accfptable.
Dased on our revieu of the above, we conclude that the proposed nodification to the LACBriR spent fuel storage pool is acceptable.
B. -Spent Fuel Shipping Cask llandling Systen
-DPC 5 Septeuber !S, 19-74 cask drop subnitta] proposed installing an energy abscrbing structure in the fuel storage pool thich would nitigate the possibility of a shipping task breaching the storage
. pool integrity during a postulated cask drop accident. _Also proposed were administrative and ncchanical controls which would restrict cask movement to a predeternined path of travel.
DPC also subnitted
~
a radiological consequences analysis for our review and approval dated June 13, 1975, entitled " Radiological Consequences of the Spent Fuel Shipping Cask Drop /sceident." By letter dated June 26, 1975, DPC submitted for NRC review the test procedures developed Ifor te,sti.ng the crash pad.
By letter dated September 5, 1975 as supplemented Septcnber 24, 1975, DPC submitted the results of the
-noduie test and applied for an anendment to Provisicnal Operating License No. PPR-45.
The application requested a change to the Technica) Specifications for the LACB?iR to isolate the containment building if a spent fuel chipping cask is to be used while irradiated fuel with less than 30 days decay time is present in the storage
- pool.
The intent of this change is to reduce the potential radio-logical consequences of a cash drop accident i16.? 340
-S-300J OR8 Na A cask drop accident analysis was perfomed by Di C and as a result DPC determined that a postulated cask drop nccident could cause dananc to stored fuel, to the fuel ston.ge pool floor and, if the cask tcre dropped in the access hatch area, to the control rod drive lower f.:cchanisn.
The protective measures proposed by DPC were:
1.
Installation of a steel bean barrier on the biological shield pedestals to protect the control rod drives.
2.
Installation of an energy absorbing pad in t b bottom of the spent fuel pool.
3.
Administrative procedures and hoist position interlock provided to ensure that the cask cannot be raised more than six inches above the 701 foot level floor.
4.
Verification that magnetic particle or liquid penetrant examination was performed on cask lifting equipment uithin the previous six-month period.
5.
Verification that the LACBhR 50-ton crane check li'st was satisfactorily completed during the previous 30-day period, and the LACghR load test was perfarned during the prev.ious 12 months.
6.
Prior to lifting the spent fuel cask to the operating floor, the crane should satisfactorily pass a dynamic load test consisting of a visual inspectica, operational ~ test, and dynamic braking test.
7.
Isolation of the containment building-if the cask-is to be handicd while irradiated fuel with less than 30 days decay time is present in the spent fuel storage pool.
The proposed steel beam barrier will consist of four 12"x12"x1/2" box ber.Jas with a span of 10 feet 6 inches in the opening of the biological shield.
In the event the cask is dropped in the hatch area and falls to the sub-basement, the steel barrier will prevent it fren reaching the lower rod drives thus precluding any damage that would result.
The tensile energy absorber for cask drop protection consists of a specific number of tensile modules which are mounted between the base plate (itcm A of Figur.e 1) and the load-distribution plate (it e:- B of Fic tre 1).
Enc!' tmile.odule consirt r of an outer cyli:.a (C)..' a i;
.t' it mechanical method; an inner cyJinuer (b) which is attached to the 1162 341
9-load-dintribution plate (P,) by suitable i.echanical r ethod (teciding is illustrated); nnd nn intercediate cylir. der (E) which has its top
_ flange flange r.et ting on the outer cylinder (C) mnd its battt:
supp: rtino the inncr cylinder (D).
Energ) absorption as ach2eved by stretcEian the interc.cdiate cylinder as follous.
The dro;; ped P ate 'hlCh object (spent fuel cast;) strikes u.e 10:>d dir t ribut 3 GP l
has sufficicn't stiffness to distribute the iupr.ct load to the tens 21e modules.
The rcdule load is applied to the intercediate cylinder (E) by the inner cylinder (D) which acts in conpression.
The intermediate or tensile cylinder is fabricated from a ductile naterial which will accor.edate substnntial strain before d ; ultinate stress level is reached (for example, stsinless steel).
.s the impact load is applied, the inter *.ediate cylinder st retches +
absorb the impact energy.
Th? load requircid to stretch the cylinder is transnitted by the top flanpc of the internediate cylinder (E) to the outer cylindct (C) which, in turn, tTann its the load to the base plate and ultit1tely to th spent fuel pool fic ar.
The cylindrical module configurat ion was choren because it resists
' lateral loads equn11y in all dircctions.
Additional lateral support between nodules can be provided by 'ceb plates (G) or by adjacent cuter cylinders.
'otion of the internediate and inner cylinders will displace water within the nodule.
Icater relief. hole (F) is provided to.pernit the out flow of water in a nanner which does not affect energy absorption detrimentally.
It is possible to control the location and size of the water relief hole to provide added energy absorption.
Calculations presented in the DPC evaluation show that the 701 level floor will sustain a cask drop fron up to 7.5 inches without structural
~
faifure.
To preclude the occurrence of this event, DPC proposes to install linit switches on the crane which will linit the traversing height for the cask to less than 6 inches above the 701 level.
The reactor building is provided with a 50 ton main and a 5 ton auxiliary hoist overhead crane in a circular track.
The main hoist notor is a 50 horsepower wound rotor induction motor rated for 60 ninntes service.__The nain hoist speed is variable by button pressure to a taxinun of 12 feet per ninute.
The
- auxiliary hoist notor is a 10 horsepower wound rotor induction notor and its hoisting speed is variabic from 11 feet / minute minitun to 25 feet /ninute naxitum.
The overhcad crane supportfhg structure including rails and bridge is designed with a safety fgetor of 5, and all wire ropes are designed with a snfetv factor of at least c.
The wire rope on the nain hoist E:
0.
u breaking strength of 22.0 tons pe: li:a.
in. auxillory hviz.t
- 1...
P00R M NAL
. 3001ORBN1 0.4375 inch din.et er wire rope of 2 pr.rt doubic cupp rting the load, with a breaking strengt h of 7.62 tons per line.
ihe main and auxiliary rope.afcty factora arv 3.4.ad 6.? rcal.i.mtively.
The nain hoist brakes consist of tuo bre';c s, a holdinn I rake and a lead control ~brate.
The holding hMc M amito.nu, electric release, spring set, shoe-type 1 rake t:i t h a 150 pt r cent torque rating, insta]Ied on the motor shaft (13 inch rhuit brake).
The load control brake is an automatic nechanical-loaf, di.sc-type inter-posed in the gear mechanism bet teen the pc..er scu:ce and the rope d rum.
The Overhead and Gantry Crane Standard (US/ ' B 50.2.0-1967) for standby crancs requires an inspection at ler ser.:iannually in accordance with the procedure for frequent insm-tions (section 2-2.1.2) and an inspection of all rcpe ser.iannually in accordance with section 2-2.1.1-6.
DPC perforas a r:mthly crane cht ek This monthly procedure in in compli..nce with the U::'3 Starulard for Overhead.md Gantry Craner.
A load t est is only regn i r e: by the USAS Standard for Overhead and Cantry Cr;.nes en iniu al use of new, ext'ensively repaired, and altered cranes.
DPC perforas the periodic
' load t est at. Icast once cach year.
Furtht:. ore, DPC perform a dynamic load test prior to each-cask lift nequence.
'l h e spent
.- fuel shipping cash lif t sequence begim :t the tine t he empty casi; is brought into containment and ends at the tir.c the loaded cast 1 caves contaj iraent.
In evaluating potential paths for moving the cask from the contair.acnt hatch to the fuel storage area DPC considered three paths at the.
701 level.
1.
Clocheise froa the hatch area, along cast side of'biologica,1 shield to the fuel storage pool; 2.
Directly over the reactor cavity (shield plugs installed) from the hatch area to the fuel storage pool; 3.
Counterclockwise from the hatch area along the west side of the biological shield to the fuel storage pool.
The first and second paths were clininated as possibic cask travel paths because of structural inadequacies.
The-third is the preferred path and is shown in Figure 3.
This path provides the greatest degree of protection to critical systems or equip;:.ent.
Adjacent to tle path, the 701 foot floor is heavily reinforced with 21-foot wide by 36-inch deep bea;ns as shown in Figure 3, since the area is used to store the reactor cavity shield plug and vessel head.
Furthermore, the 10-inch thich slab was an interral peur vith t he bem so that full lo,J capacity can be d.
i 1 6.<
M 3
s
. To further reduce the potential radiation dese c ascqucnces resulting frca a postulated cask drop accident D?C requested the authority to change the Technical Specifications.
The requcited chnnge r-: quires DPC to i solate cont ain:.ent whenever spent fuel with less than 30 days decay time is p esent in the spent fuel storage pool and a spent fuel shipping cask is t>eing noved by the crane on the 701 level, or located within one cask Icngth of the top of the spent fuel storage cell, or is uithin the spent fuel storage well.
Evaluation DPC has determined that a postulated cask drop accident in the vicinity of the lifting hatch could breach the grade level flecr and fall to the ficor of the sub-basement.
In-the event the cask were to ricochet off the inclined containment building wall in the sub-basement, the cash could roll or tip thrcurb the opening providcd by the biological chield peat ctals and d:mge the control rod lower drive techanisms.
To preclude this damage fron' occurring DPC has proposed codifying the facility by installing a steel beam barrier across the _
biological shield pedestals. The criteria used in the analysis and design of this structure to account for anticipated loadings and postulated conditions that nay be imposed upon these structures during their service lifetine are in confor. nance with the established criteria, codes, standards and specifications acceptable to the NRC staff.
h'c therefore conclude this nodification to be acceptable.
' DPC ~has proposed to install a crash pad on the floor of the open area of the spent fuel storage pool.
The intended function of the proposed crash pad is to mitigate the consequences of a postulated cask drop'in the spent fuel storage pool so that such an event will not breach the integrity of the pool.
DPC has provided information describing the structura) effects of a 50 ton cask drep on the spent fuel pool floor, the details of.the_ crash pad, and the test progran of a model of the crash pad elements.
In addition to this information DPC has provided the test results on a rodel and planned provisions for inservice surveillance.
L'c have reviewed and evaluated the relevant naterial provided by DPC and find that the design criteria and the design acthods used by DPC are in accordance with the applicabic codes and standards acceptable to the.NRC.
The use of these criteria as defined by the applicabic codes, standards,
~
and cepecifications; the loads and -loading cor.binations; the desig:
and analysis procedures; and the testing progra:. provide reasonable assurance that, in the event of various postulated cask drop accidents occurring at the structures, the crash pad will withstand the specified design conditions without impairment of structural-integrity or the performance of required safety functions.
I 16.! M4
. 300R ORDINa In accordance with our request DPC described the planned additional analysis needed to validate the final design.
Also procided ras a discussion of the effects of a "froren" ten u.mc w dule ca the crash phenomenon.
DPC presented, in acco d ':e with our reauest, a discussion of the need fsr an incer/ ice inspectien.
Th: crash pad is a stainless welded structure which. ill be sul arsed in demineri.lind water for approximately 30 years.
Eased on the fact that the stresses in the pad during.. rnal operation are very lo'..,
that the fuel storage water is of high degree of parity and at In. temperature, DPC concludes that there is no need for inservice inspection.
h'c concur with this conclusion.
DPC presented a detailed outline of a planned design repert.
This report vill cover in detail the inform tion aircady subnitted, and cover any additional infornation necessary to c pletely validate the design, including test results and all udditie: il analysis required by chany,es generated during the final dc sign.
All pertinent infornation submitted by DPC up to this dite has been revicued and evaluated by the NRC staff.
We conch:de that the inforr.ation
' furnished to date is acceptable.
Ilowever, we rec.ucst that DPC submit for our revicu the final Design Report before installation of the crash pad in the pool and use of a spent fuel shippint,, cash.
-It is recognized that DPC is proceeding with their cask drop protection on the basis that the containment polar crane handling equipment
-could fail during any lifting event in the fuel cask travel.
DPC.
has proposed tests and inspections on the crane with v;lich we concur.
It is recogni cd that no credit is taken for these tests, because protection from a cask drop rather than prevention ef-a cask drop. is proposed.
- We have reviewed the description of the crane with an uppe'r 1init switch added to prevent a two-blocking situation from occurling.
' Administrative procedures combined with a hoist position interlock installed in the hoisting control circuits are proposed to assure that the cask will not be lifted more than six inches from the operating floor.
Administrative procedures are proposed to control the path of cask movement by the bridge and trolley across the operating floor.
We
~ ~
reqtiire that trolley stops be placed on the bridge once the cask is on the operating floor, prior to novement of the cask over the concentric route between storage pool and hatch area.
These stops would limit the trolley movenent in any direction, thus assuring that the cask will remain over the operating floor while the polar crane bridge rotates.
l16.2
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With the addition of this safety devi.ce, we conclude that the areas
- revjewed, i.e., crane capacity, tests and inspections, hoist position interlock, adninistrative procedures, and prescribcd cask path of travel are acceptabic.
'Ihc spent fuel storage pool is located within the LACDNR cor.tainment building.
During the refueling operation, a batch of 24 fuel assenblics (1/3 of the core) uill be removed frca the reac.or and transferred to the spent fuel storage pool.
The first fuel assembly will be placed in the spent fuel pool appro: ' ately 5 days fc11cwing reactor shutdoun and the complete batch of 24 luel assemblies will be transferred into the pool in a minintm time interval of 10 days.
The other fuel asscrblics in the spent fuel pool uill consist of assemblics which have been stored for 3 nonths or narc.
For the postulated cask drop accident it was assum d that a shipping cash dropped into the spent fuel pool, damaging al] of the fuel rods in the 24 frechly discharged fuel assemblics 10 day; af ter reactor shutdown.
Our analysis of this accident indicatcd that calculated doses in excess of the 10 CFR Part 100 exposure guidelines could result based on the assumption that all of the released activity was exhausted directly to the ainosphere.
In order to minimize the offsite radiological cor. sequences of the cask drop accident and to insure that the calculated doses are within the 10 CFR Part 300 exposure guidelir,es, DPC has requested a change to their Techniciii Specifications.
This change would require that the containment building be isolated whenever the spent fuel storage well contains irradiated fuel which has decayed less than 30 days after exposure in a critical reactor and a shipping cask is being moved by the crane on the 701 leve-1, or located within one cask length of the top of the spent fuel. storage well, or is within the spent fuel storage well.
We have revicued and evaluated this request and find that if a postulated cask drop accident were to occur during the period that the containment building were isolated, the offsite impact would be ninor as the containment pressure would be essentially at atnospheric pressure,_
and the contain'nent Icakage under this condition t;ould be very low.
The resulting doses from a postulated accident with containnent isol'ated would be less than ]2 of the 10 CFR Part 100 exposure guidelines.
We conclude this to be acceptabic.
P00R ORGINAl.
,,3, 343
14 -
?)0RBREWL 1lc have revie:ed and evaluated the coco icherein the shipning cask 1:ere to be used.:f t.cr the ir v" ated feel haJ dce: > :d for 30 days.
Uninn the chore o'-tier,r"" nostulatin" e ensk drop accident the calculated doses stuuld be 1:ithin the 13 CFR "c rt 100 e:.posure acc;ptable.
f.uidelines.
Vic conclude this r:n:mer of cpcratica Based on our review and analysis of the cask drop accident, the proposed :r.odifications (i.e., crash pad, structural be:u;. barrier, and liuit steitches), and the requested Technical Specification change, tre conclude that (1) fuel cask hr.ndling at I.iCLhR is acceptable contingent upon our evaluation of the final design and asserbly report, (2) the existing safety r.argins or ;otential consequences of a cask drop accident are not affected by incrcasing the capacity of the spent fuel storage pool.
Conclusion ile have concluded, based on the considerations ciscussed aboce, that:
(1) there is reasonable assurance that the health and safety of the public till not be endangered by operation in the propored nanner, and (2) such activities trill be conducted in co pliance ' ith the Cor.nission's regulations, and the proposed licensing action trill not be inimical to the co='.on defense and security or to the health and safety of the public.
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UNITED STATES OF APRICA
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/
MUCLEAR REGULATORY CD:::ilSSION Jd f
BEFORE THE ATOtIC SAFETY /dD LICEMSIMG CCARD In the flatter of
)
DAIRYLAf1D POWER COOPERATIVE
)
Docket t'o. 50-409 (La Crosse Boiling Water Reac, tor)
)
(SFP License Ar.endment)
CERTIFICA_TE OF SER'!!CE I hereby certi fy tha t copies of %.C a 6u r '5 nELFu...
Ti i. ICE.lSI M BOARD MEMORA? GUM ASD ORDEP, OF SEPTEG ER 7, 1979", "/JFIFAVIT OF 00Mii R. UEEKb C' BOARD QlJESTION A.1, A.2, AND A.3",
" AFFIDAVIT OF JAMES SMEA ON DOARD QUESTIO:m A.1 <, A.3; A.4; A.5; A.6; B.3d; E; G; AND H",
and " AFFIDAVIT OF JACK it. 00:.GHE'<.' CONCERWING BOARD QUESTI0ils B.1, B.2, B.3.a, b, c; C; D; F.1; F.2; F.3" in the above-captioned proceeding have been served on the following by deposit in the United states mail, first class, or, as indicated by an asterisk, through deposit in the fluclear Regulatory Corrmission's internal mail systen, this 18th day of September,1979:
Charles Cechhoefer,
,q.,
Chairman
- Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. George C. Anderson *
- 0. S. Hiestand, Esq.
Department of Oceanography George L. Edgar, Esq.
University of Washington fiorgan, Lewis & Bockius Seattle, Washington 98195
.1800 il Street, N.W.
Washington, D. C.
20036 Mr. Ralph S. Decker
- Route 4, Box 190D
~ itz Schubert, Esq.
Cambridge, fiaryland 21613 Staff Attorney Dairyland Power Cooperative 2615 East Ave., South George R. Nygaard La Crosse, Wisconsin 54601 liark Burmaster
' Anne K. Morse Atomic Safety and Licensing Coulee Region Energy Coalition Appeal Board
- P.O. Box 1583 U.S. Nuclear Regulatory Commission La Crosse, Wisconsin 54601 Washington, D. C.
20555 Frank Linder Atomic Safety and Licensing General fianager Board Panel
- Dairyland Power Cooperative U.S. Nuclear Regulatory Commission 2615 East Ave., South Washington, D. C.
20555 L a C ro m, '.., s c <
i r.
E'M
~
1162 551
, Docketing and Service Section*
U.S. Nuclear Regulatory Co::aission Washington, D. C.
20555 l
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y Colleen P. Woodhead
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