ML20197G622

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Non-proprietary Rev 7 to HI-971661, Licensing Rept for Reracking of Ja FitzPatrick Sfp
ML20197G622
Person / Time
Site: FitzPatrick 
Issue date: 11/06/1998
From: Pellet S, Rosenbaum E
HOLTEC INTERNATIONAL
To:
Shared Package
ML20138L977 List:
References
HI-971661(NP), HI-971661(NP)-R07, HI-971661(NP)-R7, NUDOCS 9812090124
Download: ML20197G622 (43)
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to JAFP-98 0385 i

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REVISED PAGES TO NON-PROPRIETARY LICENSING REPORT FOR THE ADDITION i

OF STORAGE RACKS, " LICENSING REPORT FOR RERACKING OF J. A.

FITZPATRICK SPENT FUEL POOL," REVISION 7 l

PROPOSED TECHNICAL SPECIFICATION

. CHANGE REGARDING DESIGN FEATURES l

l 1

l New York Power Authority I

JAMES A. FITZPATRICK NUCLEAR POWER PLANT I

Docket No. 50-333 i

DPR-59 9812090124 9812039 DR ADOCK 050 3

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LL Holloc Center,550 Lincoln Drive West. Mortion, NJ 08053 HOLTEC

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INTERNATIONAL i

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goveming codes Note : Signatures and pnnted names required in the revow block.

A roymon of this document will be ordered by the Project Manager and carried out if any of its i

contents is matena#y affected during evolution of this project The determination as to the need for reymon will be made by the Project Manager with input from others, as deemed necessary by him.

I Must be Project Manager or his designee

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Desgnated Manufacture F:

Florida Office j

      • Report category on the cover page indicates the contractual status of this document as
  • j A = to be submitted to client for approval I = for client's informabon N = not submitted extemally i

THE REVISION CONTROL OF THIS DOCUMENT IS BY A "

SUMMARY

OF REVISIONS LOG" PLACED BEFORE 1

THE TEXT OF THE REPORT.

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Holtec Conter,555 Lincoln Drtve West. M&rvc t MJ 06053 HOLTEC

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  • Fax (609) 797 - 0909 INTERNATIONAL NEWJORK POWER AUTHORITY REVIEW AND CERTIF GAUUNNpm npsw anme DOCUMENT NAME :

Licensing Report for Speu tgg cua=W Expanson 1

ACCEPTED 2 'O ACCEPTED A8 NOTED HOLTEC DOCUMENT l.D. NUMBER :

97166t REsuBWTTE NoT REOuW4ED

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Note : Signatures and printed names required in the review block.

A revmon of this document will be nrdered by the Project Manager and carried out if any of its contents is materially affected during evolution of this project. The determination as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.

1 Must be Project Manager or his designee.

x Distribution : C:

Client M:

Designated Manufacturer F:

Florida Offee

"* Report category on the cover page indicates the contractual status of this document as "*

A = to be submitted to client for approval I = for client's information N = not submitted extemally THE REVISION CONTROL OF THIS DOCUMENT IS BY A '

SUMMARY

OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT.

4.

4.7 References 4-10 Chapter 4 Appealix A: Benchmark Calculations L

j 5.0

. THFRM AL-HYDP AULIC CONSIDER ATIONS...................... 5-1 5.1 Introduction 5-1 5.2

- Syct am Daaerint ian..................................... 5-2 5.3 Dacay Hamt inad C=1culeiana 5-5 5.4 Mathamatical Ida=11*=tian of the Sv= tam 5-5 5.5 Mathematical Model and Resnits...........................

5-6 5.61 Time-to-boil and Rail-off rate 5-8 5.7 i ncal Panl Water Tamnaramre............................. 5-9

_ 5.7.1 Basis.............................................

5 -9 5.7.2-Madal D= L ;ian 5-10 5.8 Claddine Tamnernmee.................................. 5-11 j_

5.9 Riackad' Call A nalvsis.................................. 5-13 5.10 References........................,................ 5-13 1

1 6.0 RACK STRUCTURAL CONSIDERATIONS........................ 6-1 1

6.1.

Analysis Outline (for New Pronaead Rack Madulec) 6-1 l

L 6.1.1 Madelling Paramatars............................. 6-2 i

6.1.2 Time Hiuarv Gener*ian 6-3 6.2 EncLReck - Dynamic Madal............................... 6-5 6.2.1 Outline of Madel for Camnnter Codc DYNARACK.........

6-6 6.2.2 Madal D= yan 6-9 6.2.3 Finid Cannlin e................................... 6-9 6.2.4 Dampine.'..................................... 6-11 I

~

6.2. 5 Impact....................................... 6-11 6.3 A =eamhly of the Dynamic Madal........................... 6-12 l

6.4 Time intafratian of the Fanatiana of Marian 6-14 i

6.4.1 Tima-Hintarv Analysis U=ing Multi-Deeree of Fraadam Rack Madel 6-14 6.4.2 Evaluatian of Patential for inter-Rack imnact 6-15 6.5 -

Struemral Accaatane* Criteria 6-16 6.6 Matarial PU---Uien 6-17 6.7 Stress i imita for Variana Canditions 6-17 6.7.1 Normal and Gaear Canditiana (12 vel A or Level B).......... 6-18 6.7.2 12 vel D Service I imita 6-20 6-20 6.8 C*alag of Dvnamic Simulariana 6.9 R*= nite for Sinele Rack Madal and 3.D Seismic Marian............ 6-26 l

6.10 imnact A nalysE.........._...........................

6-27 l-6.10.1 Imaart i nading Retween Fuel A==ambly and Cell Wall 6-27 I

- Holtec Report HI-971661 ii l

- -... ~.. -

l a

6.10.2 P=ck Dvnnmic Imance=

6-28 l

j 6.11 Weld Stresses 6-28 j-6.11.1 Racantata to Rack Weldc and Cell-to-Cell Welds............ 6-28 6.11.2 Heat.ingaf an Icolatad cell.......................... 6-30 l

6.12 Definitian of Terms Uced in Ractinn 6.0...................... 6-30 l

6.13 References......................................... 6-31 3

Appendix A: Dynarack Solver Output Summary

. Appendix B: 3-D Single Rack Analysis of Fuel Racks 7.0 ACCIDENT ANALYSIS and THERMAL (SECONDARY) STRESSES......

7-1 7.1 Introduction 7-1 7.2 Results of Accident Re-evnluation..........................

7-1 7.2.1 Fuel Pool.....................................

7-1 7.2.2 Fuel Rearmee nnildine 7-1 7.2.3 Refnaline Ami' =

7-2 i

~

'7.2.3.1 Dranaad Fuel A==amhly..................

7-2 7-3 1

7.2.3.2 Dranaad Gata..........................

7.2.4 Rack Dron 7-4 1

4 7.3 I nent Ruckline of Fuel Cell Wntla........................... 7-5 7.4 A nalysis of W'elded Jninte in R ack........................... 7-6 7.5 References..........................................

7.7 l

8.0 POOL STRUCTURAL ANALYRIS 8-1 4

8.1 References.......................................... 8-2 1

9.0 IN-SFRVICE SURVFIT T ANCE PROGRAM........................ 9-1 9.1 Purnace 91 9.2 Canaan Kurveillanca..................................... 9-1 9.2.1 Da=criarian of Test Canaan <

9-1 9.2.2 Benchmark Data.................................

9-2 9.2.3 i nne Term Rurvai11=nca............................. 9-2 Holtec Report HI-971661 iii

l l.:

i 1.0 =

INTRODUCTION i

1,1 Intr a n ninu l

The James A. FitzPatrick (JAF) Nuclear Power Plant is a boiling water reactor (BWR) installation located on the southeast shore of Lake Ontario, approximately 6 miles northeast of the city of Oswego, New York. The plant is rated at 2536 Mwt and has been in commercial operation since July,1975.

j.

The spent fuel pool of the FitzPatrick plant was initially reracked under License Amendment 55 l

issued June 18,1981. " Poisoned" high density racks made of aluminum alloy were installed to increase the storage capacity to 2244 locations. The second rerack campaign was performed under License Amendment 175 issued December 31, 1991. Five modules made of stainless steel containing a total of 553 storage locations were added to the pool, increasing the total installed capacity to 2797 locations. As indicated by Table 1.1, the current installed increased capacity in the JAF pool will lead to loss-of-full core offload capability by 1998. The projected loss of full core discharge capability in 1998 prompted the New York Power Authority to undertake steps to increase the spent fuel storage capacity in the fuel pool. Fortunately, there is additional floor 4

space available in the JAF spent fuel pool wherein supplemental modules can be installed.

l Under the proposed storage expansion, seven modules containing a total of 442 storage locations l

will be added to the pool, increasing the total installed capacity to 3239 locations. Three of the l

l modules (identified as N1, N2, and N3 in Figure 2.1) are to be installed in the third rerack campaign. A future fourth rerack campaign at a future date will involve the installation of the L

remaining racks (identified as F1, F2, F3, and F4 on Figure 2.1).

i All of the new racks shown in Figure 2.1 are self-supporting. The principal construction materials j

for the new racks are ASME SA240-Type 304L stainless steel sheet and plate stock, and A564-

. Type 630 (precipitation hardened stainless steel) for the adjustable support spindles. The only l

Holtec Report H1-971661 Page1 1

1 I

Chapters 4, 5, 6, and 7, respectively, deal with the criticality, thermal-hydraulic, seismic, and i

mechanical accident considerations. The adequacy of the pool structure is addressed in Clapter 8.

f In-service inW commitments for Boral are set forth in Chapter 9, followed by radiological and environmental assessments in Chapters 10 and 11, respectively.

1 i

l l

The New York Power Authority (NYPA) has enlisted the services of Holtec International of f

Marlton, New Jersey, to perform the m=y design, analysis, and safety evaluation activities. All i

analyses reported in this submittal, except radiological and shielding evaluations, were carried out i

by Holtec International.

i

?

The manufacturing of the new racks will be performed by Holtec's contractor UST&D of Pittsburgh, Pennsylvania, which has fabricated practically every fuel rack for the U.S. plants in the 90s.

i ne installation of the racks in the JAF pool will utilize the same procedures and methods which j

have been used by Holtec Intemational in all ofits turnkey rerack projects (over two dozen).

A summary of the defense-in-depth approach utilized by Holtec in the site construction effort is l

presented in Chapter 2.

l Inasmuch as the design of the racks parallels the most recent rerack submittals, the analyses l

presented in this report parallel those presented in the 1990 0.L. amendment application. Herefore, this submittal does not contain recent vintage analyses such as the Whole Pool Multi-Rack (WPMR) simulation for seismic analysis or computational Fluid Dynamics (CFD) modelling for local fuel cladding temperature evaluations. In other words, the methods, models, and analyses are kept consistent with the most recent rerack.

l Rack N3 was initially intended to be sized with 8x13 cells. However, the clear space intended for l this rack proved to be too small allowing for a length of only 12 cells. Thus, the revised rack N3 is l 8x12.

l f

Hohec Report HL971661 Page 1-4

t i

I Table 1.1 FUEL DISCHARGE DATA OPERATING CYCLE DISCHARGED FUEL i

Number of Open Storage l

Locations Total No. of Cycle No.

Shutdown Assemblies Assemblies After Date Discharged Stored in the Present Campaigns poog III and IV 1

1 6/1977 132 132 2

9/1978 136 268 3

5/1980 160 428 4

11/1981 188 616 l

5 6/1983 200 816 l

6 2/1985 188 1004 7

1/1987 196 1200 8

8/1988 184 1384 9

3/1990 148 1532 I

10 11/1991 152 1684 11 11/1994 204 1888 12 10/1996 192 2080 717 l

t i

13 10/1998 192 2272 525' 789 l

14 10/2000 200 2472 325 767 l

15 10/2002 200 2672 125 567 l

l i

Indicates time when loss of full core offload capability will occur if new racks l

are not added.

Holtec Report HI-971661 Page1 5 I

i 2.0 MODULF LAYOUT FOR INCREASED STORAGE 2.1 New Pmposed Racks i

The James A. FitzPatrick high density spent fuel storage racks consist of individual cells with 6.16" (nom.) inside square dimension, each of which accommodates a single Boiling Water Reactor (BWR) fuel assembly. The fuel assembly can be stored in the storage locations in channelled or unchannelled configuration. Table 2.1 gives the essential storage cell design data.

As stated previously, the J.A. FitzPatrick pool has undergone two rerack campaigns in the past.

The racks installed in the first rerack campaign were made from aluminum and Boral. The racks i

for the second campaign in which 553 storage locations were added, were designed by Holtec International with stainless steel as the structural material and Boral as the neutron absorber. The new racks scheduled for installation are similar to the Campaign II racks. Table 2.1 provides a

]

summary of the key design variables for Campaign II and the new rack modules. It is seen that the new racks have been designed to realize even larger criticality and structural margins than the existing Campaign II racks.

l The J. A. FitzPatrick pool does not have any Boraflex, tetrabor or borated steel in its racks.

There are 442 added storage locations in the fuel pool. Fuel racks designated as N1, N2, and N3 l

(see Figure 2.1) will be installed in the upcoming reracking campaign. Fuel racks designated as F1, F2, F3, and F4 (see Figure 2.1) will be installed in a future campaign. Table 2.2 provides data on each of the modules.

The existing and new modules for the FitzPatrick fuel pool are qualified as freestanding racks, i.e., each module is freestanding and is shown to undergo minimal kinematic displacements during the postulated seismic events. Thus, rack-to-rack or rack-to-wall impacts in the active fuel region are precluded. Figure 2.2 shows a typical new rack module for the FitzPatrick fuel pool.

Holtec Repon HI-971661 Page 2-1

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i Table 2.2 t

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MODULE DATA FOR CAMPAIGNS II, III, and IV (STAINLESS STEEL RACKS) g4 e

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Table 2 PROPERTIES OF BORAL Irradiation Tests (No changes in physical properties or neutron attenuation observed) 2 Total Gamma Dose 1.0 x 10 Rad Thermal Neutron Dose 5.7 x 10 " n/cm2 2

Fast neutron Dose 5.7 x 10" n/cm2 Specific Heat of the Aluminum 0.919 w-s/gm-K @ 100*F of the Al-B C Core 0.936 w-s/gm-K @ 100*F 4

Thermal Conductivity ofthe Aluminum 1.621 w/cm-K @ 100*F of the Al-B.C Core 0.859 w/cm-K @ 100*F Coemcient of Thermal Expansion 1.97 x 10-s in/in *C Modules of Elasticity (ASTM E-8) 9,000 ksi l

Tensile Strength (ASTM E-8, E-21) 10 ksi Ductility (ASTM E-8), Elongation in 2" coupon 0.1%

Average Core Compaction 93 %

Holtec Report HL971661 3A-7

I.

i l

perform this evaluation, versions 1.4 and 1.9 of Holtec's proprietary TBOIL program are used.

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The TBOIL program calculates the minimum time-to-boil and corresponding boil-off-rate based on the SFP thermal capacity and water volume, and the discharge conditions discussed in Section -

5.4.- The transient pool water level and boil-off rate are also detcrmined. The makeup water temperature and the time after loss of forced cooling when makeup becomes available are assumed as 95'F and 10 hrs after loss of forced cooling. An iterative solution is performed to determine the mmimum required makeup water flow rate to prevent the water level from dropping to within 10 feet of the top of the racks.

Results for all cases are presented in Table 5.6.1. It is seen that sufficient time to introduce manual cooling measures exists and the available time is consistent with other BWR reactor installations.

5.7 LOCAL POOL WATER TEMPERATURE In this section, a summary of the methodology, calculations and results for local pool water l

temperature is presented.

5.7.1 Basis In order to determine an upper bound on the maximum fuel cladding temperature, a series of l

conservative assumptions are made. The most important assumptions are listed below:

The fuel pool will contain spent fuel with varying time-after-shutdown (t,). Since the heat emission falls off rapidly with increasing r,, it is conservative to assume l

that all fuel assemblies are from the latest batch discharged simultaneously in the shortest possible time and they all have had the maximum postulated years of operating time in the reactor. The heat emission rate of each fuel assembly is assumed to be equal and maximum.

As shown in the pool layout drawings, the modules occupy an irregular floor space L

Holtec Repon HI-971661 Page 5-9

Table 5.3.1 FUEL DISCHARGE DATA OPERATING CYCLE DISCHARGED FUEL Total No. of Cycle No.

Shutdown Assemblies Assemblies Date Discharged Stored in the Pool 1

6/1977 132 132 2

9/1978 136 268 3

5/1980 160 428 4

11/1981 188 616 5

6/1983 200 816 6

2/1985 188 1004 7

1/1987 1%

1200 8

8/1988 184 1384 9

3/1990 148 1532 10 11/1991 152 1684 I

11 11/1994 204 1888 12 10/1996 192 2080 13 10/1998 192 2272 l

14 10/2000 200 2472 l

15 10/2002 200 2672 l

16 10/200>

560' 3228 t

Full core offload.

Holtec Report HI-971661 Page 5-17 I

1 Table 5.6.1 Case Numbers Minimum Time-to-Boil Required Makeup (Hours)

Water Flow Rate (gpm) 1 10.52 17.2 2

5.04 47.3 Table 5.7.1 Factor Value Radial 1.759 Axial times Radial 2.552 Total 2.814 Holtec Report HI-971661 Page 5-22

consolidated fuel canisters'. 3-D single rack analyses were performed considering both in-phase l

and opposed-phase motion of adjacent racks. All simulated conditions were performed for both coefficients of friction (0.2 and 0.8) discussed above.

l 6.1.2 Time History Generation The rack structure seismic analyse ' were performed utilizing the time-history method. Pool slab acceleration data in three orthogonal directions was developed and verified to be statistically independent. The objective of the seismic analysis of single racks is to determine the structural response (stresses, deformation, rigid body motion, etc.) due to simultaneous application of the three statistically independent, orthogonal seismic excitations. Thus, recourse to approximate statistical sununation techniques such as the " Square-Root-of-the-Sum-of-the-Squares" method

[6.6] is avoided. For nonlinear analysis, the only practical method is simultaneous application of the seismic loading to a nonlinear model of the structure.

Pool slab acceleration data are developed from specified response spectra from two earthquakes:

OBE and DBE. Since the OBE peak accelerations exceed the DBE peak accelerations, only one set of time histories was prepared to envelope both target earthquakes. The results of the dynamic simulations using the bounding time histories will conservatively be compared against the lower allowables appropriate for OBE loading. Using the provided response spectra as input, the appropriate three components of the earthquake, in the form of a time history fcr each direction, are developed using the Holtec QA validated code GENEQ [6.14). Synthetic acceleration time histories are generated for a 20 second event duration from 'he plant response spectra at level 326.8' based on 1% damping.

Figures 6.1 through 6.3 show the comparison between the design basis spectra for the spent fuel

' This license application, however, is limited to storage of intact fuel assemblies.

Holtec Repon HI-971661 Page 6-3

_m.

. _ _ ~. _ _. _ - _ _. _,

e' Combined bending and compression on a net section satisfies:

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f+Ca fu + % fn < 1 F,

D, F D, F4 u.

where:

f,

=:

Direct compressive stress in the section

~ Maximum bending stress along x-axis fu

=

Maximum bending stress along y-axis f,

=

_0.85 C,,,,

=

0.85 C,,,,

=

1 - (f,/F',) '

D,

=

1 - (f,/F',,)

D, i

=

F',,,,,

=

(x E)/(2.15 (ki/r)2 )

2 and sukripts x.y reflect the particular bending plane, f.

Combined flexure and compression (or tension) on a net section:

+ b +!*E-<1.0 0.65, F

F,,.

u The above requirements are to be met for both direct tension or compression.

g.

Welds Allowable maximum shear stress on the net section of a weld is given by:

F,, = 0.3 S, where S, is the material ultimate strength at temperature. For the area in contact with the base metal, the shear stress on the gross section is limited to 0.4S,.

j 6.7.2 Level D hrvice I imita h

As stated above in Section 6.1.2, the time history set was prepared to envelope both OBE and DBE eartM=M. To be conservative, the developed rack stresses will be compared against OBE allowables. Therefore, Level B limits will be used for all simulations.

l Page 6-19 Holtec Repon HI-971661 l

l 6.8 carnine of Dynamic simut einn.

Initial simulations were performed for racks N3, F1, F3, and F4 with adjustable suppon pedestals for all combinations of the following conditions:

l

  • Unchannelleil fuel
  • - In-phase and opposed-phase l
  • Nearly empty, fully loaded, and half loaded along the rack diagonal e 0.2 and 0.8 Coefficients of Friction (COF)

An additional set of simulations was performed for rack F4 considering all of its pedestals as fixed to account for the actual condition of the north-west suppon pedestal. All of the conditions described above were evaluated for racks N3, F1, F3, and F4 (for both suppon pedestal types) by performing simulations for every combination considering unchannelled fuel storage. These l

four racks were chosen from the total of seven proposed racks with the intention of producing j

bounding simulations. Because racks F1 and F2 are geometrically identical, analysis of only one l

of the two is necessary. Racks N1, N2, and N3 are also very similar. All three racks are 8 cells l

wide by 11,10, and 12 cells long, respectively. Rack N3 was initially intended to be 13 cells long j

and is analyzed as such. However, the clear space intended for this rack proved to be too small l

allowing for a length of only 12 cells. The only difference between these three racks is the weight l

and the aspect (width to length) ratio. Past experience with single rack analysis indicates that for l

a given canhquake set and with all other rack design parameters remaining constant, the greatest l

mass produces the greatest inertia during the dynamic event, and the greatest aspect ratio usually l

produces the greatest twisting motions, displacements, and propensity for overturning. Selection l

of fictitious rack N3 (modeled as an 8x13 cell rack) satisfies both of these parameters by being l

more massive and longer than the three actual rack modules. Therefore, rack N3 was chosen with l

the intent to represent racks N1, N2, and N3. By observation of the large design margins in all l

categories, the methodology of selecting a bounding model to represent the actual configurations l

is acceptable. Subsequent to the realization that the 8x13 sized rack N3 would not fit into the l

alloted pool space, and to provide funher argument for the methodology chosen, an additional 20 l

simulations were performed. These additional simulations replicated all 20 of the simulations for l

the 8x13 cell rack N3, but the model in the additional simulations was corrected to represent the l

8x12 condition. The results are discussed below.

l Holtec Repon HI-971661 Page 6-20

J' The following listing provides a' tabulation of the conditions used for each simulation:

Rack N3 Unchannelled Fuel In-Phase Empty.

0.2 i

Rack N3 Unchannelled Fuel In-Phase Em)ty 0.8 Rack N3 Unchannelled Fuel In-Phase Ful 0.2 Rack N3 Unchannelled Fuel. In-Phase Full 0.8 Rack N3 Unchannelled Fuel In-Phase Half 0.2 Rack N3 Unchannelled Fuel In-Phase Half 0.8 Rack N3 Unchannelled Fuel Out-of-hase Empty 0.2 Rack N3 Unchannelled Fuel Out-of-hase Em>ty 0.8 Rack N3 Unchmanelled Fuel Out-of-Ful 0.2 Rack N3 Unchannelled Fuel Out-of-Fall 0.8 Rack N3 Unchannelled Fuel Out-o Half 0.2 Rack N3 Unchannelled Fuel Out-o Half 0.8 Rack F1 Unchannelled Fuel In-Phase Empty 0.2 Rack F1 Unchannelled Fuel In-Phase Em>ty 0.8 Rack F1 Unchannelled Fuel In-Phase Ful 0.2 1

Rack F1 Unchannelled Fuel In-Phase Full 0.8 Rack F1 Unchannelled Fuel In-Phase Half 0.2 Rack F1 Unchannelled Fuel In-Phase Half 0.8 Rack F1 Unchannelled Fuel Out-of-Empty 0.2 Rack F1 Unchannelled Fuel Out-o hase Em>ty 0.8 Rack F1 Unchannelled Fuel Out-of-hase Ful 0.2 Rack F1 Unchannelled Fuel Out-of-hase Full 0.8 Rack F1 Unchannelled Fuel Out-of-base Half 0.2 Rack F1 Unchannelled Fuel Out-of-hase Half 0.8 i

Rack F3 Unchannelled Fuel In-Phase Empty 0.2

~ Rack F3 Unchannelled Fuel in-Phase

'Em>ty 0.8 j

Rack F3 Unchannelled Fuel In-Phase Ful 0.2 Rack F3 Unchannelled Fuel In-Phase Full 0.8 Rack F3 Unchannelled Fuel In-Phase Half 0.2

- Rack F3 Unchannelled Fuel In-Phase Half 0.8

~

. Rack F3 Unchannelled Fuel Out-o Empty 0.2 Rack F3 Unchannelled Fuel Out-of-Em >ty -

0.8 Rack F3 Unchannelled Fuel Out-o Ful 0.2 Rack F3 Unchannelled Fuel Out-o Full 0.8 Rack F3 Unchannelled Fuel Out-o Half 0.2

' Rack F3 Unchannelled Fuel Out-of-Half 0.8 l

Rack F4 - A ustable Pedestals Unchannelled Fuel In-Phase Em>ty 0.2 Rack F4' A ustable Pedestals Unchannelled Fuel In-Phase Ful 0.8 Rack F4 A ustable Pedestals Unchannelled Fuel In-Phase Full 0.2 Rack F4 A ustable Pedestals Unchannelled Fuel In-Phase Empty 0.8 Rack F4 A ustable Pedestals Unchannelled Fuel In-Phase Half 0.2 Rack F4 A ustable Pedestals Unchannelled Fuel In-Phase Half 0.8 Rack F4 A ustable Pedestals Unchannelled Fuel Out-of-Empty 0.2 Rack F4 A ustable Pedestals Unchannelled Fuel Out-of-hase Em>ty 0.8 Rack F4 A ustable Pedestals Unchannelled Fuel Out-of-Ful 0.2 Rack F4 A ustable Pedestals Unchannelled Fuel Out-of-hase Full 0.8

- Rack F4 ' A ustable Pedestals Unchannelled Fuel Out-of-hase Half 0.2 Hohec Report HI-971661 Page 6-21

~

Rack F4 Adjustable Pedestals Unchannelled Fuel Out-of-phase Half 0.8 Rack F4 Fixed s ' Unchannelled Fuel In-Phase Empty 0.2 Rack F4 Fixed tais Unchannelled Fuel In-Phase Em)ty 0.8 Rack F4 Fixed tais Unchannelled Fuel In-Phase Ful 0.2 Rack F4 Fixed tais Unchannelled Fuel In-Phase Full 0.8 Rack F4 Fixed tais Unchannelled Fuel In-Phase Half 0.2 Rack F4 Fixed tais Unchannelled Fuel In-Phase Half 0.8 Rack F4 Fixed Unchannelled Fuel Out-of-hase Empty 0.2 Rack F4 Fixed tals-Unchannelled Fuel Out-of-base Em >ty 0.8 Rack F4 Fixed s

Unchannelled Fuel Out-of-Ful 0.2 i

Rack F4 Fixed s

Unchannelled Fuel Out-of-hase Full 0.8 Rack F4 Fixed Unchannelled Fuel Out-of-hase Half 0.2 Rack F4 Fixed tais Unchannelled Fuel Out-of-hase Half 0.8 j

Appendix A provides solver output from all 60 simulations and Table 6.5 provides a summation.

of the results. The following three character nomenclature was used to identify the simulations l

t performed:

First Charnerer.

Reennd Charneter-Third Charneter:

i = In-phase e = empty 2 = 0.2 COF i

o = Opposed-phase f = full 8 = 0.8 COF h = half full For example, "ie2" corresponds to an in-phase simulation performed on an empty storage rack I

considering a coefficient of friction of 0.2, i

These 60 simulations were reviewed, as follows, to determine which of the remainmg 120 cases l

(for channelled and consolidated fuel types) must be simulated to ensure bounding results.

Bounding results may be defined as the greatest values for the three primary evaluation categories:

Loads, Displacements, and Stress Factors.

l A review of the initial 60 simulations provides the following observations:

1. The fuel to cell wall impact loads are negligible in comparison to manufacturers data on fuel assembly side load capacities. By observation of the similar weight and gap parameters, the channelled fuel simulations would result in simdar loads. Therefore, the impact loading to fuel l

Page 6-22 Holtec Report Hl.971661

assemblies is not of concern. However, additional runs were performed for channelled fuel to assess the variation in fuel impact loading. Rack F3 was evaluated for all conditions with channelled fuel, as described in item 3 below, and rack N3,was evaluated for its four bounding conditions, as described below in item 2. The additional simulations cover the rack conditions for.the most likely to overturn and heaviest loading.

2. For rack N3 the controlling results (i.e., loads, disp!raments, and stress factors) are obtained from the fully loaded conditions (as expected), except for the baseplate corner Y displacement.

The baseplate displacements are oflittle concern, because they are negligible (i.e., less than 0.01"). Both COF conditions must be considered for the fully loaded condition to envelope the worst cases for rack N3. Therefore, additional runs for the remaining fuel types were performed for the four conditions represented by both COF values and both phase conditions;

3. Rack F3 results do not control over any of the other racks. In fact all values, including displacements are low in comparison. However, since this rack is extremely narrow and tall, all of the remaining conditions were simulated for this rack configuration.
4. Rack F4 with the fixed pedestal has the largest pedestal stress factor (0.320). This is because the pedestal material has a lower yield stress and resulting allowables. Therefore, to determme the stress factor values for the fixed pedestals all of the remaining conditions were simulated for this rack configuration.
5. Rack F1 does not control over other racks in any load, displacement or stress factor categories.

Therefore, additional runs are not warranted for this rack. This argument is further strengthened by the fact that rack 3 should control over rack 1 by virtue of similar fluid gaps, dimensions, and lower stability against overturning.

6. Rack N3 results exceed all other rack results for unchannelled fuel in the category of adjustable pedestal stress factors. Therefore, the results from the additional fuel type simulations, discussed in item 2 above will envelope adjustable pedestal stress factor results from all racks.

Based on the above observations of the initial runs, an additional 44 simulations were performed from the possible remaining total of 120, bringing the total number of discrete simulations performed to 104.

Because the consolidated fuel canister would weigh considerably more than intact fuel and its dimensions are approximately the same as the other two fuel types, the results from these simulations are expected to exceed the results from the other two fuel types in all categories:

)

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Hohec Repon H1-971661 Page 6 23 4

i

l Loads, Displacements, and Stress Factors. Table 6.5 confirms that this is the case.

Additional dynamic simulations were performed to evaluate the two other styles of fuel by j

selecting the worst results (i.e., highest stresses and displacements) from the unchannelled runs j

and re-running using the parameters for the other fuel types. The following listing provides a short description of the conditions used for each of these additional simulation:

Rack N3 Channelled Fuel In-Phase Full 0.2 f

Rack N3 Channelled Fuel In-Phase Full 0.8 L

Rack N3 Channelled Fuel Out-of-phase Full 0.2 L

Rack N3 Channelled Fuel Out-of-phase Full 0.8 l

l Rack N3 Consolidated Fuel in-Phase Full 0.2 t

l Rack N3 Consolidated Fuel In-Phase Full 0.8 L

Rack N3 Consolidated Fuel Out-of-phase Full 0.2 p

Rack N3 Consolidated Fuel Out-of-phase Full 0.8 Rack F3 Channelled Fuel in-Phase Empty 0.2 Rack F3 Channelled Fuel In-Phase Empty 0.8 Rack F3 Channelled Fuel In-Phase Full 0.2 Rack F3 Channelled Fuel In-Phase Full 0.8 Rack F3 Channelled Fuel In-Phase Half 0.2 l

Rack F3 Channelled Fuel In-Phase Half 0.8 l

Rack F3 Channelled Fuel Out-of-phase Empty 0.2

-l L

Rack F3 Channelled Fuel Out-of-phase Empty 0.8 l

Rack F3 Channelled Fuel Out-of-phase Full 0.2 l

Rack F3 Channelled Fuel Out-of-phase Full 0.8 Rack F3 Channelled Fuel Out-of-phase Half 0.2 l

Rack F3 Channelled Fuel Out-of-phase Half 0.8

. Rack F3 Consolidated Fuel In-Phase Empty 0.2 Rack F3 Consolidated Fuel In-Phase Empty 0.8 Rack F3 Consolidated Fuel In-Phase Full 0.2 Rack F3 Consolidated Fuel In-Phase Full 0.8 Rack F3 Consolidated Fuel In-Phase Half 0.2 Rack F3 Consolidated Fuel In-Phase Half 0.8 Rack F3 Consolidated Fuel Out-of-phase Empty 0.2 Rack F3 Consolidated Fuel Out-of-phase Empty 0.8

)

Page 6-24 i

Holtec Repon HI-971661 l

l l

Rack F3 Consolidated Fuel Out-of-phase Full 0.2 Rack F3 Consolidated Fuel Out-of-phase Full 0.8 Rack F3 Consolidated Fuel Out-of-phase Half 0.2 Rack F3 Consolidated Fuel Out-of-phase Half 0.8 Rack F4 Fixed pedestals Consolidated Fuel In-Phase Empty 0.2 Rack F4 Fixed pedestals Consolidated Fuel In-Phase Empty 0.8 Rack F4 Fixed pedestals Consolidated Fuel.In-Phase Full 0.2 Rack F4 Fixed pedestals Consolidated Fuel In-Phase Full 0.8 Rack F4 Fixed pedestals Consolidated Fuel In-Phase Half 0.2

. Rack F4 Fixed pedestals Consolidated Fuel In-Phase Half 0.8 Rack F4. Fixed pedestals Consolidated Fuel Out-of-phase Empty 0.2 Rack F4 Fixed pedestals Consolidated Fuel Out-of-phase Empty 0.8 Rack F4 Fixed pedestals Consolidated Fuel Out-of-phase Full-0.2 Rack F4 Fixed pedestals Consolidated Fuel Out-of-phase Full 0.8 l

Rack F4 Fixed pedestals Consolidated Fuel Out-of-phase Half 0.2 Rack F4 Fixed pedestals Consolidated Fuel Out-of-phase Half 0.8 I

~ As discussed at the end of Section 6.9, an additional simulation was performed for rack F4 to l

conservatively evaluate the possibility of overturning and include a proper safety factor.

l An additional 20 simulations were performed to replicate all 20 of the simulations for rack N3 l

with the model corrected to represent the 8x12 condition. As expected, the results are comparable j

to the results from the previous rack N3 runs and sufficient design margin remains. None of these l

new N3 rack runs produced results which control over the full gamut of simulations.

l l

The total number of simulations performed was 125. Appendix A provides summation file outputs I

from the 105 simulations, which do not include the 8x12 rack N3 simulation results. The results l

from the 8x12 rack runs are compared with the results from the other runs in the last set of-l tabulations of Table 6.5.

I L

i t

~

Holtec Report HI-971661 Page 6-25

'8 I

6.9 p nia for single na una,1 and 3-D Reimmic Marian A complete synopsis of the analysis of the modules subject to the postulated earthquake motions, is yia;r.cd in summary Table 6.5 which gives the bounding values of stress factors R, (i = 1, 2, 3,4,5,6). The st ess factors are defined as follows:

l Ratio of direct tensile or compressive stress on a net section to its allowable value R

=

(note support feet only support compression)

Ratio of gross shear on a net section in the x-direction to its allowable value R

=

2 Ratio of maximum bending stress due to bending about the x-axis to its allowable R

=

3 value for the section R=

Ratio of maximum bending stress due to bending about the y-axis to its allowable value R=

Combined flexure and compressive factor (as defined in 6.7.le above) 3 R. =

Combined flexure and tension (or compression) factor (as defined in 6.7.lf above)

R,. =

Ratio of gross shear on a net section in the y-direction to its allowable value.

The allowable value of R (i =1,2,3,4,5,6) is 1.0. The dynamic analysis gives the maximax i

(maximum in time and in space) values of the stress factors at critical locations in the rack module.

Values are also obtained for maximum rack displacements and for critical impact loads. Table 6.5 presents critical results for the stress factors, and rack to fuel impact load.

Table 6.5 also presents maximum results for horizontal disp!ecaments at the top and bottom of the rack in the x and y direction. "x" is always the short direction of the rack. In Table 6.5, for each run, both the maximum value of the sum of all support foot loadings (4 supports) as well as eacn individual maximum is reported. The table also gives values for the maximum vertical load and the corresponding net shear force at the liner at essentially the same time instant, and for the Holtec Repon HI-971661 Page 6-26 I

L maximum net shear load and the corresponding vertical force at a support foot at essentially the same time instant.

l l

The results presented in Table 6.5 are representative of the totality of runs carried out. The critical case for structural integrity calculations is included. Appendix A to this Section 6 contains l

output summaries of all DYNARACK simulations, except for the additional runs performed for l

the 8x12 N3 rack condition. Appendix B to this Section 6 contama a general discussion regarding l

single rack simulation analysis methodology.

l l

l Analyses show that significant margins of safety exist against local deformation of the fuel storage cell due to rattling impact of fuel assemblies.

l The largest displacement occurred at the top of rack F4 when diagonally loaded with consolidated fuel and moving in-phase with adjacent racks under 0.8 coefficiem of friction conditions.

Therefore, these conditions were simulated again using an increase factor of 1.5 applied to the j

earthquake time-history. The resulting displacements show that the rack center of gravity remains within the boundary formed by the pedestals. Therefore, no overturning will occur.

6.10 Imnact Analyces 6.10.1 Imanet I nadine bewaen Fuel A=wmhly and Fell Wall i

l l

The local stress in a cell wall is conservatively estimated from the peak impact loads obtained from i

i the dynamic simulations. Plastfc analysis is used to obtain the limiting impact load. The bnit load L

is calculated as 4585 lbs. per cell which is much greater than the loads obtained from any of the i

simulations.

l-

)

l Holtec Report HI-971661 Page 6-27 l.

L.

l

~

6.10.2 Dek Dvaamic Imancte i

Dynamic analyses were performed for both in-phase and opposed-phase motion of the adjacent racks. Thus, the highest potential for inter-rack impact is enveloped. The displacements obtained from the dynamic analyses show that no impacts occur between racks or between racks and walls.

)

i I

It is also noted that the new fuel racks do not breach the theoretical plane between the new racks and the contiguous existing racks, indicating that impact with existing rack modules will not occur.

This is a plausible conclusion in view of the fact that the racks installed during campaign I and new racks have markedly different structural characteristics and their displacement time histories j

will be randomly phased with respect to each other.

Therefore, it is concluded that no impacts between racks or between racks and walls occur during I

a seismic event.

6.I1 Weld Straccec Critical weld locations under seismic loading are at the bottom of the rack at the baseplate connection and at the welds on the support legs. Results from the dynamic analysis using the simulation codes are surveyed and the maximum loading is used to qualify the welds on these locations.

6.11.1 Racentar, to Rack Welds and cell-to-cell Welde Section NF permits, for the DBE condition, an allowable weld stress t =.42 S, = 28,600 psi.

Based on the worst case of all runs reported, the maximum weld stress for the baseplate to rack welds is 10,387 psi. This value occurs using a fuel weight of 1303 lbs, per cell.

l l

l l

Holtec Repon HI 971661 Page 6-28 i

i

The weld between baseplate and support leg is checked using limit analysis techniques. The structural weld at that location is considered safe if the interaction curve satisfies F/P, + M,/M, < 1 where F,, M, are the limit load and moment under direct load only and direct moment only. F.

M, are the absolute values of the actual peak force and moments applied to the weld section. This is a much more conservative relation than the actual interaction curve. For the worst case simulation, this criterion gives F/F, + M /M, s.409 for the support leg to baseplate weld.

The critical area that must be considered for fuel tube to fuel tube welds is the weld between the fuel tubes. This weld is discontinuous as we proceed along the tube length.

Stresses in the fuel tube to fuel tube welds develop along the length of each fuel tube due to fuel assembly impact with the tube wall. This occurs if fuel assemblies in adjacent tubes are moving out of phase with one another so that impact loads in two adjacent tubes are in opposite directions which would tend to separate the channel from the tube at the weld. The critical load that can be transferred in this weld region is calculated as 5056 lbs. at every fuel tube connection to adjacent tubes. An upper bound to the load required to be transferred is

/2 x 302 x 2 =854 lbs.

where we have used a maximum impact load of 302 lbs (from Table 6.5), assumed two impact i

locations are supported by each weld region, and have increased the load by /2 to account for 3-D effects.

I i

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Hohec Repon HI-971661 Page 6-29 l

6.11.2 Heatine of an Isolated Cell Weld stresses due to heating of an isolated hot cell are also computed. The assumption used i

is that a single cell is heated, over its entire length, to a temperature above the value associated with all surrounding cells. No thermal gradient in the vertical direction is assumed so that the results are conservative. Using the temperatures associated with this unit, analysis shows that the weld stresses along the. entire cell length do not exceed the allowable value for a thermal loading condition. Section 7 reports a value for this thermal stress.

6.12 Definition ofTerms Used in Section 6.0 S1,S2,S3,S4 Support designations p,

Absolute degree-of-freedom number i l

q, Relative degree-of-freedom number i l

l p

Coefficient of friction 4

U, Pool floor slab displacement time history in the I-th direction x,y coordinates horizontal directions z coordinate vertical direction K

Impact spring between fuel assemblies and cell i

Kr Linear component of friction spring Ks Axial spring at support leg locations N

Compression load in a support foot Subscript i When used with U or X indicates direction (i = 1 x-direction, i = 2 y-direction, i = 3 z-direction) i i

llottec Report 111-971661 Page 6-30

TABLE 6.5 (CONT'D)

Dynamic Simulation Results Summary ~

Comparison of New Rack N3 Results with Other Runs Result Category Previous 8x13 Maximums New 8x12 Rack N3 Run From All Rack N3 Run Maximums Previous Runs Maximums Total Vert. Pedestal load 158132 158132 144518 l

Single Pedestal Vert. Imad 62169 62169 60947 l

Single Pedestal Shear Imad 10685 17192 11259 l

Fuel-Cell Impact lead 225 302 261 l

Top corner X-displacement

.1943

.2184

.2137 l

5 Baseplate corner X-displacement

.0112

.0337

.0122 l

l Top corner Y-displacement

.0979

.1695

.0933 l

Baseplate corner Y-displacement

.0056

.0193

.0053 l

R6 Stress Factor in Cell Wall

.184

.240

.200 l

R6 Stress Factor in Pedestal

.307

.585

.309 l

Page 648 Holtec Report HI-971661 i

J.A.

F t tzPo tr tek 0.30 E~lAIlu M & lll R L

~

ilJHHiHMhk i'~l-lillElH HWP" ppppppgr

-e. 2g g,,,,,,,,,,,,,,,,,,,,.,,,,,,gg.ee g

g g

g

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ol lb

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- 0. 2e _

i i

'M ' ' ' ' '1 ed'eb ' ' ' ' '1 di'eb ' ' ' ' '2ebe.'e'

-e.3 e

e

'l m G (SOC.

X ' 00)

Figure 6.8

J.A.

F t tz Pe tr te k Spent Fuoi Pool Ttmo Hts tory Ac c e l e r o g r a m Z d tree t ton Boundtng Sp e c tr e

( 1 % De mp tn g )

0.15_

ll l

l 01 0.10:

t

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BR E

  • y i

5 l

t I

l

-'-0.00h 'l i

, e, ln l

'I I,

p ~t -

e

[. -

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7-j g

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i 0-0.10; ji

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{

-0.15 :

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[

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-0.20~

ii

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iiiiiiiiii 0.00 500.00 1000.00 1500.00 2000.00 i

tme Csec.

X

' 00) g, g i

I compressive allowable of 2890 psi which is well above the calculated stress. Therefore, the concrete remains adequate to withstand the worst case dynamic impact loading.

The results of the previously performed analysis was thoroughly discussed in the previously submitted license amendment prepared for reracking campaign II. Table 8.1 provides a competison of the proposed storage rack loading and the loading previously considered. The increased loading to the pool structure represented by the new high density storage racks remains below the loading considered in the previous analysis. Therefore, the pool structure remains adequate.

8.1 Referencec l

I

[8.1] HEA Inc. Report No. 8603-1, " Structural Capacity of the J.A. Fitzpatrick Spent Fuel l

Pool," December 29,1986.

l l

i h

i s

Holtec Report HI-971661 Page 8-2 t'

1

... - _ _ _. - _ ~.

r to JAFP-98-0385 l

J. A. FITZPATRICK PROJECTED ALARA ESTIMATES FOR INSTALLATION OF FUEL RACK ASSEMBLY N3 l

l i

1 I

l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

ALARA Review #: 98-039 Revision: 1 Level: 2 Tracking Information:

Modification Ntimber(s):

N/A Work Request Number (s):

95-02411 RWP No.(s):

98-0046 Building:

Reactor Elevation:

369' System:

19 Component:

Spent Fuel Poal Previous ALARA Review Numbers (s):

97-014, 93 431 Job

Contact:

Personnel Contacted:

A. Pal, R. Miller Reason for evaluation:

Work in the Spent Fuel Pool Rad Survey #(s):

Various Total dose projection:

0.575 Rem 4

Shielding Authorization #:

N/A Job Scope:

Install N3 rack in the northeast corner of the spent fuel N.

i Revision 1:

Dose projection revised to reflect difficulties with interference removal.

Completed By:

{Qpy Technician:

R. Murray Planner:

None Date:

5/11/98

I

~

ALARA Review #: 98-039.1 COPY Approvals:

Level N/2AI [mMeA/ytn (1rl>Jt/

Date: 5/n/Q?

(1,2,3)

A Supervisor:

(

sed A Date: 6-M'N (1,2,3)

Radiation Protection Supervisor-

+/

s Date:

/-8 (1,2,3)

Responsible First Line Supervisor-

/ < 4/[

[a> d12 Date: 5 /2'58 (1,2,3) s -

Heg Physics General Supervisor./

/ MW' Sh2/f7 Date:

(2,33 RES Mana r:

/

[ 2/!fcI(2,3)

/

/ -e - - -

Date:

/R nsibfe D6 ed Manager:

NN/

Date:

/2/A (3)

ALARA Cornmittee Chairman:

4 l

?

l

b ALARA Review #: 98-039 EXPOSURE PROJECTION STEP DESCRIPTION TASK MAN-EFETCTIVE MAN MO.

HOURS DOSE RATE REM 1

Remove bracket, drag test and install N3 rack 3.1 93 0.001 0.098 l

2 Supervision 1.1 102 0.0003 0.031 t

3 Remove Tri-Nuc from pool 3.1 15 0.003 0.045 Revision 1 5/11/98 i

Actual to date 1

Remove bracket, drag test and install N3 rack 3.1 392 0.0007 0.282 2

Supervision 1.1 141 0.0004 0.055 l

3 Remove Tri-Nuc from pool (Task deleted) 3.1 0

0 0

Projected for conclusion 1

Remove remainining interferences, drag test and install N3 rack 3.1 240 0.001 0.216

[

2 Supervision 1.1 64 0.0003 0.022 f

i TOTAI.S 837 0.575 References ALARA Rev. No.s: 97-014,93-031 RWP #s:

98-0046 Survey No.s: Various Other: N/A Prepared by:

R.

Murray Date: 5/11/98 i

6.0 PRE-JOB PREPARATION 6.1 All personnel involved in work within the Spent Fuel Pool shall attend an ALARA Pre-Plan meet 6.2 A pre-job briefing is required by Rad Protection prior to the start of each shift.

l 6.3 Workers should maintain housekeeping and cleanliness controls per AP-17.03 and AP-05.06 while working in the Spent Fuel Pool. Each person is responsible for logging equipment into/out of the FME zone.

i 7.0 WORKER PREPARATION 7.1 All workers working in the Spent Fuel Pool should be familiar with NRC Information Notice 90-33 titled: " SOURCES OF UNEXPECTED OCCUPATIONAL RADIATION EXPOSURES AT S FUEL STORAGE POOLS."

l 7.2 All personnel shall be cognizant not to pick up, handle, retrieve, or touch any unknown objects or debris from the Spent Fuel Pool or the floor of the Equipment Storage Pit Radiation Protection shall be contacted immediately upon observing an) such items.

7.3 WIEN handling irradiated components or tools that have come in contact with irradiated components, high levels of contammation and high contact beta dose rates should be expected.

7.4 Wor'kers should be aware of the possibility of floating HOT PARTICLES emitting radiation levels in excess of 1000 R/hr at 1" producing 3 to 8 R/hr dose rates at the Gent Fuel Pool hand rail. It is possible to " stir up" these particles in the Spent Fuel Pool.

7.5 James A. FitzPatrick Deviation Event Report No. 93-0932 Before lifting equipment from the Spent Fuel Pool, assure that component is not bound on surrounding equipment. While lifting a shroud head bolt for dose profiling, a second shroud head l bolt was inadvenently lifted and dropped as it cleared the spent fuel rack.

7.6 Be alert for the presence of high activity metal chips or fragments from a missing control rod roller !

guide ball that could result in significant skin doses in a relatively short time. Fragments from control I rod roller guide balls are high dose rate Cobalt-60 sources that can exceed 10,000 Rad /hr.

8.0 WORK AREA PREPARATION 8.1

  • Portable Area Radiation Monitors (RM-16 or equivalent) shall be installed at SFP rails adjacent to {

the work area for early detection of floating hot particles and monitoring of work area. Alarm j setpoint to be 100 mR/hr above background or as specified by the Radiological Supervisor. In the i event of an unplanned ARM, AMS-3(CAM), or RM-16/DCA 3090 (portable ARM) alarm, the following actions should be taken:

a.

Work shall stop and evolution that caused the alarm shall be reversed, if possible (all personnel).

i

\\

b.

Retreat to an area as directed by Rad Protection Technician (all personnel).

Immediately contact Shift Manager, Refuel Floor Suoervisor, and Rad Protection Sup c.

Work will not restart without a recovery plan and authorization from the above individuals.

8.2 Functional tests are required at the start of each shift for all installed /ponable ARMS when work is to be performed in the Spent Fuel Pool. A portable ARM shall be located in close proximity (approximately 6') to crews working in the Spent Fuel Pool at fixed locations.

8.3 The bridge installed or portable ARM shall be used on the refuel bridge when work is to be performed in the Spent Fuel Pool and the bridge is to be occupied.

8.4 Establish adequate underwater lighting for work in the Spent Fuel Pool. Turn lights off when work in pool is finished.

9.0 AUDIO / VISUAL CONBiUNICATION 9.1 CCTV is stationed on the Refuel Floor and is available for use.

9.2 Telex and hard wired communication systems are available for use.

9.3 An underwater camera is available on the refuel floor for inspections. This camera is controlled by the Reactor Analyst Group.

4 10.0 ALARA/ RADIOLOGICAL CONTROLS 10.1 Anticipated Radiological Conditions l

Component / Area / Activity:

Refuel Floor (Clean Area)

Location:

RB 369' Refuel Floor Radiation:

0-15 mR/hr General Area Contamination:

< 1,000 dpm/100cm2 Airborne:

<25 % DAC Component / Area / Activity:

Refuel Bridge Location:

RB 369' Refuel Floor Radiation:

Bridge 2-7 mR/hr General Area j

Refuel Mast 15 mR/hr - 80 mrad /hr 8 Contact 5 mR/hr - 8 mrad /hr B' General Area Contamination:

1,200-8,000 dpm/100cm2 Airborne:

< 25 % DAC l

I 10.2 Man-Hour Reduction Controls 10.2.1 The Refuel Floor Supervisor shall be responsible for minmuzing the number of personnel on the Refuel Floor. Unnecessary personnel shall not be in the vicinity of the Spent Fuel Pool while work is in progress.

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10.3' Dose Rate Reduction Controls 1

10.3.1 Radiation Protection personnel shall perform periodic Hot Particle monitoring for pe removmg/ handling material from the SFP. Hot particle controls will be in effect for all work in the Spent Fuel Pool or contact with components removed from the Spent Fuel Pool if Ho Particles are detected.

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10.3.2 Personnel should minmuze time spent at the SFP railings and curb. When not performin pool side task personnel should move to a low dose waiting area.

10.3.3 Hollow poles for underwater work shall have holes in them to allow filling with water to elinunate potential radiation streammg.

10.3.4 An operational teletector (or equivalent extendable instrument with high range capability) shI be available on the Refuel Floor durmg work removing aquipment from the water. A teletector (or equivalent) should be used to monitor items as they are being removed from the SFP.

10.3.5 Dose rates can build up rapidly on rags used for wiping down highly contanunated components.

Rad Protection should frequently monitor used rags and work area trash receptacles for dose buildup. Consider the use of disposable rags vs. launderable rags if high

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levels of contammation are encountered.

10.4 Cantaminarian Controls l

l 10.4.11 All materialskeg-wnts that come in contact with trradiated components 'should be l

washed / rinsed during removal from the water and deconned to <100,000 dpm prior to j

l storage. IF decon is not practical, THEN the object should be wrapped or contamed in such a i

i manner that airborne conenmutation will not be generased.

l 10.4.2 Multiple gloves shall be worn when handling material being removed from the SFP.

l Periodically change gloves at the direction of the Radiation Protection Technician to mmunize potential exposure from Hot Pamcles.

10.4.3 The Tri-Nuc 260 GPM underwater vacuum cleaner will be used to capture the metal filings j

i generated t,y cutting. If EDM is performed in the pooliit is important to capture as much "schwarf" as possilbe. This fine material plates out on SFP piping and increases general area dose rates.

10.5 Airborne Radioactivity Controls 10.5.1 Reactor Building ventilation lineup /flowpath should be considered before removing highly contaminated components / materials from the SFP.

10.6 Personnel Monitoring / Protective Requirements I

10.6.1 Alarming dosimeters are required for all personnel working in the vicinity of the SFP or the refuel bridge.

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10.6.2 Hot panicle surveys shall be performed in accordance with RP-OPS-202.

1 10.7 Incident Prevention 10.7.1 Rad Protection shall be present on the Refuel Floor for all fuel movement or Spent Fuel Po work.

10.7.2 NO OBJECTIS TO BE REMOVED FROM THE WATER UNLESS RAD PROTECT 15 PRESENTAND READY TO SURVEY.

10.7.3 While removing equipment from the SFP, items with unanticipated dose rates > 1 R/hr will require that the lift be stopped and consultations made with the RP Supervisor and/or RP i Chief Technician.

10.7.4 Conspicuously mark all handling tools (poles, ropes and cables) at 8 feet from the end of the tool to prevent inadvertently bringing highly irradiated components to the surface of the water.

10.7.5 Maintain the Tri-Nuc vacuum hose at a level of at least 8' underwater during all vacuunung.

This will prevent the elevation of general area dose rates above the water from highly irradiated material passing through the hose.

10.7.6 Keep the Tri-Nuc vacuum hose away from highly irradiated components when not in use.

This will help to prevent hose disintigration.

10.7.7 The Tri Nuc vacuum shall not be used for vacuuming of the Spent Fuel Pool floor at any time l during the work governed by this ALARA review.

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- ALARA Review #: 98-039 Revision: 0 Level: 2 t

Tracking Infonnation:

t Modification Number (s):

N/A Work Request Number (s):

95-02411 RWP No.(s):

98-0046 Building:

Reactor Elevation:

369' System:

19 l

Component:

Spent Fuel Pool Previous ALARA Review Numbers (s):

97-014,93-031 Job

Contact:

Personnel Contacted:

A. Pal, R. Miller Reason for evaluation:

Work in the Spent Fuel Pool Rad Survey #(s):

Various Total dose projection:

0.174 Rem Shielding Authorization #:

N/A Job Scope:

Install N3 rack in the nortt. east corner of the spent fuel pool.

Completed By:

i Technician:

R. Murray Planner:

None Date:

1/5/98 l

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ALARA Review #: 98-039 i

Approvals:

Ievel 2

Date: /!8 8 (1,2,3)

AEARA'Superbr:

<Z/

Date:/5 98 (1,2,3)

Radiation Praicten Supervisd:

,/

b

/

Date: /~8-98 (1,2,3)

ReMible First IS S'upervisor:

i M,

d_

Date://68I (1,2,3) su Hea Physics General Supervif

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Date: [ !b l

(2,3) m u._- _ __.

f M<

Date: /

5 (2,3)

RAnnnch-@ Manager:

h Date:

(3)

AIARA Committee Chairman:

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