ML19261D799

Forwards Addl Info Re Spent Fuel Storage Expansion in Response to NRC 790315 & 21 Requests

ML19261D799
Person / Time
Site:	FitzPatrick
Issue date:	06/22/1979
From:	Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Ippolito T Office of Nuclear Reactor Regulation
References
JPN-79-35, NUDOCS 7906260423
Download: ML19261D799 (19)

Text

i3 POWER AUTHORITY OF THE STATE OF NEW YORK to CoLUMaus CIRCLE NEW YORK. N. Y.1Co19 (212) 397 6200 GEORGE T BERRY thECUTivE DintCrom FREDERICK R CLARK C""'""'N LEWIS R B EN N ETT GEORGE L. ING ALLS S S 5 7 4 'e f E C I

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Di4ECTOM vlCE CM ac RM Ame 5

E ER RICH ARD M FLYNN 7

y7 g ROBERT i wiao*2' JOHN W BOSTON June 22, 1979 l.,",',,',",,o,'1,o,,,

wiLuau r. Luoor JPN-79-35 r,.,0 N.., m R,N N. ;R.

CO N T ACLL ER Director of Nuclear Reactor Regulation U.

S.

Nuclear Regulatory Commission Washington, D.

C.

20555 Attention:

Mr. Thomas A.

Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Spent Fuel Storage Expansion

Dear Sir:

In response to your letters dated March 15, 1979 and March 21, 1979 requesting additional information relative to our submittal of July 26, 1978, " Increase Spent Fuel Storage Modification" for the James A.

Fit:: Patrick Nuclear Power Plant, we are submitting the following attached information.

Very truly yours, Paul J.

Early Assistant Chief Engineer-Projects 2311 283 gi

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s ?,C n,d s a-. 0 J, a-u OUESTION 1 You stated in your July 26, 1978, submittal that the d se ratee above the s

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r field will be as 1:w as reasonably achievable (ALARA) during the preposed poci modification.

Relevant experien:e at nuclear power plants shcw typical

values, in the vicinity of spent fuel peels of 1 to 2 mre:/ hour. Your res; nse should censider increased purificatica system Operation, increased filter-de:Inerali:er change out frequency, or relocating certain tocis and conpenents stored in the pec1.

Resp 0NSE

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could be received abcVe the spent fuel pcci would be 10 mrer/hr.

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around the fuel pcci has significantly reduced the dose rate in this area.

No in:reased purifi:atien syster Operation is necessary as water chemistry

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Q.1-1

OUESTION 2 Identify the principal radionuclides and their typical concentrations in the spent fuel pool water found by gar..ca isotepic analysis prier to and following refueling.

Provide the dose rate above the spent fuel real and at the pcci edge f cm these concentrations, including crud build-up along the sides of the pe:1, as compared to the dese rate contribution to the "10 mr/hr...

frem tocis and cmponents stored around the

cel."

Estimate the ccmpatible occupational exposu:e, in annual man-rem, due to all operations assceiated with fuel handling in the spent fuel pool area, frc the aforementioned sources.

Demonstrate that storage of the tools and =0mponents in the spent fuel pool, with : mmensurate increase in ba:kground radiation levels in the STP operations area, is compatible with achieving ALARA exposures in cperating personnel, r E F F O'; S E The results of a recent analysis of the fuel pool vater is shown belev. The

encentrations for these radi0nuclides are less than the maximum permissible 0ncentrations alleved in ICCTR:0, Appendix S, Table II, Column 2 for radica:tive material in an unrestricted area. Tuel handling is performed from a bridge unich is approximately 5 feet above the water.

Dese rate readings taken frem the bridge are less than mrem /hr in all cases and less than 1 mrem /hr in mest cases. Assuming that a refueling outage takes place ence per year, and assuming that 30 days are spent ever the fuel pc01, a cumulative total annual exposure Of 1.44 man-rem would result. This veuld be split up among approximately 20 men, giving a total exposure annually of 72 mrem per man due to work ever the spent fuel pool. The dose rate free crud and to ls s cred in the spent fuvl pcc., is small ecmpared to the potential dose rate from tools ste:ed arcs j :Me fuel pool which have nov been relocated to other areas.

Radionuclide A?.alysis Ouantitative Analysis

!setere Ouantity Std. Deviation Cs-137

. 9491:I-07 uti/mi +-

9. 6351E-03 uti/ml Mn-5a 5.6605:E-06 ue /mi +-

2.36111E-07 u i/mi Cc-60 1.64016E-05 uti/mi +-

u.69730E-07 u:i/mi Total 2.25E-5 uci/ml 2311 285 0.2-1

m.re...m.:

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Provide the estimated man-rem exposure and discuss the cecupational exposure expected during this proposed STP modificati.cn.

In this evaluatica, address the expected dose rates frem spent fuel poci vater (including items stored in the pool), spent fuel and the equipment to'be dispesed of, number of verkers (including divers, if any) and cecupancy times for each phase of the operation; the removal and disposal of the present spent fuel racks and installatien of the new higher density racks; and the disposition of mi-scellaneous equipment presently stored in the pool.

FESPONSE The Installation Fr0 gram will be a: cmplished in three phases with the dispesal of the cid racks and bracing accomplished separately.

Phase :

censists of the preparations necessary te install the new racks such as: develeping precedures, arrange for diving and dispesal services, and rencving unnecessary equipment and material from the fuel peel.

Phase II censists of removing the seismic bracing and existing racks frem the southern half of the fuel paci and installing ten high density racks in that area.

Thase III repeats Phase !! for the ncrthern half of the peel. The disposal of the old racks and bracing vill be at:ceplished as a separate effort during and after the Installation Program.

The Installation Fregram vill be a::cmplished by an 6-memcer Installation Team cens sting of 5 maintenance personnel, I te:hnical adviser, 1 rad protection technician, and 1 Q.C.

inspector.

n addition, divers will be used during Ph'ase I: to cut a 10 inch seismic brace and to cut five sving bolts.

The diving team vill consist of a supervisor, a diver, a backup diver, and a diver tender.

The preparation and disposal of the rencved fuel racks and bracing vill be accomplished by a 4-man team of hu sith physic technicians.

The estirated man-rem exposure for ecmpleting the installation of the high density spent fuel stcrage rack and the disposal of the ex; sting racks and bracing is 5.935 man-rem.

This exposure is broken devn as fc11cvs:

p.ack and bracing rencval 1.S40 man-rem New rack installati n 1.065 man-rem Rack and bracing disposal

.550 man-rem Diver

.295 man-rem

!.ei transfer

.195 man-rem The estimated man-rem exposure is based en a recent survey of the fuel peel areas which shows that the average dose rate 3 ft frca the fuel peci as being 1.9 mr/h-and the dose rate en the refueling bridge as less than mr/hr.

Since ecs of the operatzen vill be perfereed from these areas, 2 mr/hr was used as the average dose rate.

The spent fuel stered in the fuel peci vill be stored at the opposite end of the peel from the end where the installatien verk is being performed.

This results in a negligible effect on the dose rate to the installation team.

2311 286 Q.3-1

All riscellaneous equi; tent and m.aterials presently stored in the fuel pool vill either be removed and disposed of or placed in the Oppesite end of the fuel pool from the installation verk and vill have a negligible impact en the dose rate.

2311 287 Q.3-:

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Frevide the additional cc:upational exposure (in man-rem) frc= normal cperatica in the spent f.el pool area, including refueling, after the STp modification prcpesed in your July 26, 1978 submittal.

Include the expected exposure from more frequent changing of STP filters and demineralizers, from spent fuel pool water, from any equipment stored in this water, and from spent

fuel, r, r e r. n v. e r

_~

The fuel has approximately 266 inches of water ever it.

The resulting dose rate frcm s cred fuel is a negligible contributor to the cumulative occupational ex;csure. The effect of the extra fuel added to the fuel peci on radiatica exposure received while verking around the peci is still negligible.

The effect of more fuel in the fuel pool en spent fuel pool filters is also negligible. The same amount of fuel would have passed through the spent fuel pcci over the same time period. The effect en filters is experienced over the first few days during and after the time fuel bundles are moved and has little er no con : bution after it has been in place for an extended period. The f requency Of filter back-flushing is centrolled by high dif f erential pressure, and more frequent back-flushing is not anticipated.

No additional equipment vill be stored in the paci as a result of this modification.

2311 288 c.u_1

OUESTION 5 Describe the method the vill be used to dispose of the present racks (i.e.,

crating intac: racks or : 2tting and packaging).

If the racks are to be cut and packaged, show that the expcsure received by this disposal method, as ccepared to crating the intact racks for disposal vill provide as lov as is reasonably achievable (ALARA) exposure to personnel, pESp0NSE The existing fuel racks will be hydrolazed during removal from the fuel pool to remove the loose surface contamination. The existing racks will then be placed in the reactor internal storage pit to dry.

Localized decontamination vill be performed if necessary, prior to packaging.

The existing racks vill be packaged in waterproof wooden boxes and shipped intact to a disposal site as low specific activi*.y (LSA) radioactive vastes.

2311 28L9 Q.5-1

CUESTION 6 Discuss in sete detail the impact of the proposed STP modification en radicactive liquid effluents from the plant, including leakage of water from the pool.

Discuss the spent fuel poci leak cc11ection system, including the disposition of leakage if it should occur.

p E SFmi? E The modification has no impact on radicactive liquid effluents from the plant.

The same activity and approximately the same volume of water is in the fuel

cci as prior to the modification.

Any leakage frca the peci vill be collected in the reactor building fico drain sumps through the spent fuel pecl leakage detection system and processed in the ficer drain radvaste system. The spent fuel peci leakage detection system is described in Section 9.3.4.1 of the TSAR.

2311 2?90 Q.6-1

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v Discuss the capability cf the spent fuel pool eccling system to keep the ex;ected, not destgn, spent fuel p001 bulk water temperature at er belev the p s,\\ R design of 1:5'r during norma' refuelings until the mee. ted pool is filled.

!f the bulk water temperature is expected to be above TSAR design value, discuss when this will occur and for what period of time.

F.E S P O V S E The fuel pool ecoling system is designed to maintain the bulk fuel pcci temperature at er below 125'T.

Analysis of the expected decay heat leads and ecoling system indt:ates that the bulk spent fuel pool temperature vill exceed 1 5'T when the decay heat load is greater than 7.5 x 106 Stu/hr.

Based on the

enservative assumptions that the refuelings vill be accomplished na 12 tenth schedule and ene-quarter of the fuel bundles vill be simultaneously discharged to the spent fuel peci 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after plant shutdevn, the temperature vill first reach i:5'T during the third refueling for less than one day and exceed 1:5'T each subsequent refueling. During the sixteenth refueling, the temperature vill be above 125'T for approximately 14 days.

2311 291 Q.7-1

O ??E S TIC S' 8 F cvide the present frequency of replacing the filter-deminerali:er resin beds for the spent fuel pool.

R E S F C N'S E Tit:Fatrick has stone septum preccat filters.

The filters are changed app cximately weekly during refueling and approximately monthly during normal operations.

2311 292 v.wO.

CUESTIC' 9 Frevide the estimated volume of contaminated matertal (e.g.,

spent fuel racks, seismic restraints) expected to be removed from the spent fuel pecis during the modification and shipped from the plant to a licensed burial-site.

?ESPONSE The estimated volume of contaminated material to be shipped to a licensed burial site is as follows:

1.

Existing spent fuel racks 171.6 ft* per rack x 39 racks

6,692.4 ft*

2.

Control red racks per rack x 3 racks 331.65 ft8 110.55 ft8 3.

Safety curtains 1.6 ft3 per curtain x 32 curtains :

51.2 ft' 110.0 ft5 4

Upper and lever rack restraints 7,185.25 fts Total Volume Q.9-1

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Trovide the weight and circasions of each icad.

Discuss the lead transfer path, including whether the lead must be carried over the pool, the maximum height at which it could be carried, and the expected height during transfer.

revide a descriptien of any written procedures instructing crane operators about loads to be carried near the pool.

Provide the number of

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re typical load into the pool.

pI57 m:5E During nereal plant crerations, after completica of the high density spent fuel stcrage rack modification, the loads which will be moved over the stored

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.r 3 3.- a.e... ; s Nuclear instruments Centrol rods Tuel channels Fuel handling tecis During the installatica of the hi;h density fuel racks, the existing racks and

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fuel p001.

No leads will be transported ever fuel racks which contain spent

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Existing spent fuel racks:

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Safety curtains:

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Centrol red s:crage racks:

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7,250 lb, 14.99' x

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5 x 10 new spent fuel rack:

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The ecvement of these leads will be controlled by approved p rocedures and marked areas on the refueling ficer to prevent carrying over any spent fuel presently in the pool.

Q.10-1

)

Se:ause of the height of the new racks and the icv level of contamination cf the old racks, this equipment will be transported ever the ficer,and fuel peci area apprcximately 6 inches above tb4 fl:Or elevation of 369'-6".

The maximum height that the loads could be transported are as follows:

7'-8" a' eve 369'-6" 1.

Old spent fuel ra:ks:

b 2.

New spent fuel racks:

12'-0" above 369'-6" 3.

Failed fuel racks:

.10'-C" above 269'-6" 4

Safety curta ns:

3'-0" above 269'-6" 5.

3 racing (cid) 15'-0" above 369'-6" By using the previously centioned procedures and restricting the e.ovement of the loads as described above, no spent fuel can be damaged.

2311 295

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er.re.-vy i t.

s Otscuss the instrumentation to indicate the spent f uel poci water temperature.

nclude the capability of the instrumentatien to alarm and the location of the alarms.

,e r s. ?. C N e. r.

A thermo Ouple, 19-!E-71, is mounted in the spent fuel pc0L and is logged on re: Order IC-!R-131.

The recorder is located on the Nuclear Steam Tempera:ure Recorder Panel.

An alarm switch en the re: Order ts set to annuncia:e at 130'F.

The high temperature alarm is alarmed on the "Tuel Pool C0 cling and Clean Up System Trouble" annunciator drop in the control r00m and the plant C0mputer-Thermoccupies located en the Outlets of the fuel peci :: cling system heat exchangers, 19-E-1A and 13, are also legged on recorder IC-TR-131 and provide alarm signals to the plant computer and the " Trouble" annunctator identified above.

2311 296 Q.11-1

OUESTION 12 Prepose a techni:al specification which prohibits carry:ng loads greater than the weight of a fuel assembly ever spent fuel in the storage pool; er ;astify why this specification is not needed to limit the potential consequences of accidents involving dropping heavy leads, other than casks, onto spent fuel to those of the design basis fuel handling a:cident.

RESPONSE

A technical specification is not required because the reactor building crane is equipped with interlocks which prevent the operation of the main and 20-ton auxiliary heists ever the fuel pool with the exception that the main hoist is alleved to operate within a small area over the cask area.

The 20-ton auxiliary hoist vill be required to remove the existing fuel racks.

A Temporary Maintenance Frecedure has been written and will be reviewed and approved by the Plant Operations p.eview Committee which will detail the tnstallatien of centrol jumpers to allev the operatica of the 20-ten heist ever the fuel pcci during STP rack replacement.

This prceedure also identifies load transfer paths into and out of the fuel pool so that no leads are transported ever fuel racks which contain spent fuel.

The fuel rack insta11atton team vill be instructed on all aspects of the installation program.

2311 297 Q.12-1

O';ESTION 13 Discuss the analysis and results ef the fuel pool concrete walls to verify their integrity under the preposed increase in the mechanical and thermal loadings.

PESTONSE The fuel pool valls were included in the finite element m0 del.

Figure 4-1 in the submittal is a sket:5 of the area used in the finite element medel which was analyzed.

Both mechanical and thermal stresses were less than design allowables.

2311 298 0.13-1

0"ESTICN 14 provide the folleving information for D3E leading case, considering both an empty rack and a rack with full fuel load.

a.

The computed sliding and re: king displacements for critical loading conditions.

b.

The corresponding factor of safety against impacting the wall and other objects in the pool.

PESPONSE a.

The maximum horizontal rocking displacement at the top of the rack is 0.38 inch for the D3E under the critical recking loading conditions as defined en page 3-6 of the submittal which is 0.8 coefficient of friction for two full racks.

The maximum sliding displacements cecur for empty racks at 0.2 coefficient of friction. These displacements were calculated at 1.47 inches for DBE and 0.737 inch for 03E.

No lift-up or rocking was noted under these conditions.

b.

The fellcving table summari:es the fa: tor of safety against impacting into walls or other objects in the peal for 03E and DBE conditions.

Descrirtien 0_L; 2]Li apent iue.,..a.3 l s 3.tz

. l.,

a Existing ilcor Swing Selts 7.46 3.74 Cask Dr p Frctection System 5.42 0.5 Sparger Pipe Brackets

.07 2.03 Existing Rack NE Corner 7.46 3.74 2311 299 Q.14-l

QUEST!CN 15 Discuss the extent of rack-to-rack impact during a DBE and/or SSE.

Discuss also the design effort taken to minimi:e the effect of such impact en the integrity of the rack and fuel elements contained therein.

FEsp0NSE The maximum rack-to-rack impact forces have been calculated to be 81,000 P0unds for the D3E and 64,000 pounds for the CBE.

The forces occur at the top grid only and are included in the stress analysis of this member.

Since the upper fitting of the fuel assemblies is not attached to the top

grid, these impact leads are not directly transmitted to the fuel assemblies.

Q.15-1

e n, r e. i a v..-

1. 6 s.

-m Clarify that three ccmpenents of ea'rthquake were used in the seismic analyses of ANSYS and SAP IV models.

If not, ;ustify the conservatism of using less than three cceponents.

PES?cNSE All three components of earthquake have been conservat;/ely censidered in the rack design.

As explained in the submittal, the ANSYS time histery was done fer enly two compenents of earthquake which were the maximum herirental (X-direction and vertical Y-direction).

Mcwever, the forces computed frem this planar time-l.istory medel vere applied en the detail (3-D) medel simultaneously in beth the X-Y and Z-Y planes.

These resultant leads were then conbined by an SRSS to obtain the overall leads.

Note that this method, in effectively considering all three components of earthquake, doubles up on the vertical (Y-direction leading).

2311 301 0.16-1

OUESTION 17 In the dropped fuel bundle analysis of 18-inch drop, verify that the rack model stays elastic during impact and hence the elastic energy balance method is adequate in predicting a static impact lead.

Resp 0NSE The maximum ecebined stress calculated for the 18-inch d:c; conditions as defined on page 3-6 of the submittal was 10,540 psi.

This maximum stress c: curred on an inner rib of the tcp grid and censisted primarily of major axis bending stress.

Since the minimum yield strength of the top grid is 16,000 psi, the grid remains elastic except for a small localized area at the impact interface. Hence thc elasti: energy balance method used is adequate in predicting the impact leads.

Tull si:ed tests on an actual grid casting indicate plastic behavier is small producing en the order of a 1/10-inch indentation.

2311 302 Q.17-1

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'le rif y that the leads, lead cenbinations, and acceptance criteria used for the rack design are consisten: with Sections 3.8.4.11.3 and 3.8.4.11.5 cf the Standard Review Flan for Steel Structures.

pESFnNSE In addition to the three load conbinations and acceptance criteria as s p e cified in p aragrap h 3.3.1.1 of the subnittal, the following CBE lead tenditica was analy:ed.

1.05 D+L+T,+E These four lead ecebinations and acceptance criteria enservatively ecver all applicable leading for spent fuel racks as specified in Se:tions 3.8.4.11.3 and 3.8.4.11.5 of the Standard Review Flan for Steel Structures.

2311 303 Q.18-1

.