ML20197G618

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Reducing Size of Spent Fuel Rack Assembly N3 from 8x13 Cells to 8x12 Cells & Deleting Proposed Inclusion of Fuel Pool Water Level Inadvertent Drainage Into Amend
ML20197G618
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 12/03/1998
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20138L977 List:
References
NUDOCS 9812090122
Download: ML20197G618 (17)
v • d • e


Text

- . . - .. _ _ . . . -

, 1 l

Attachment 1 to JAFP-98-0385 -

l l

REVISED TECHNICAL SPECIFICATION PAGE PROPOSED TECHNICAL SPECIFICATION j

_ CHANGE REGARDING DESIGN FEATURES j i

I l

i i

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

DPR-59 9812090122 9812037 l DR ADOCK 0500

. l l

l

O.

JAFNPP 5.5 FUEL STORAGE 5.5.1 Criticajity 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum k.of 1.32 in the normal reactor core configuration at cold conditions (200C);
b. k,, < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.3 of the FSAR; and
c. A nominal center to center distance between fuel assemblies placed in the storage racks as described in Section 9.3 of the FSAR.

5.5.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum k.of 1.31 in the normal reactor core configuration at cold condition:, (200C);
b. k,, s 0.90 if dry;
c. k,, s 0.95 if fully flooded with unborated water; and
d. A nominal 6.625 inch center to center distance between fuel assemblies placed in storage racks.

5.5.2 . Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3239 fuel assemblies.

5.6 SEISMIC DESIGN The reactor building and all engineered safeguards are designed on a basis of dynamic analysis using acceleration response spectrum curves which are normalized to a ground motion of 0.08 g for the Operating Basis Earthquake and 0.15 g for the Design Basis Earthquake.

Amendmelt No. 5 , S3, 9,101,174, 246 1

Attachment 2 to JAFP-98-0385 REVISED SAFETY EVALUATION PAGES PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES 4

)

New York Power Authority JAMES A. FITA5'"31CK NUCLEAR POWER PLANT docket No. 50-333 DPR-59

Attachment 2 ta JAFP-98-0385 REVISION 1 TO SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES

1. DESCRIPTION OF THE PROPOSED CHANGES The following are proposed changes to the James A. FitzPatrick Technical Specifications. A revised infinite lattice multiplication factor for individual fuel bundles is proposed to ensure that the effective neutron multiplication factor of fuel stored in the spent fuel pool (SFP) will be maintained less than 0.95. The maximum number of stored assemblies has been raised to allow installation of additional storage racks to extend the time available before independent spent fue! storage is required. The page layout is changed from landscape to portrait format and l

Specification numbering is revised to be consistent with STS. Also an editorial correction is made to Specification 5.1.1 (old 5.1.A).

Paae 245 Specification 5.1.A has been renumbered 5.1.1 and the word " country" has been replaced with " county."

Specifications 5.1.B, 5.2.A, 5.2.B, 5.4.A, 5.4.B and 5.4.C have been respectively renumbered to 5.1.2, 5.2.1, 5.2.2, 5.4.1, 5.4.2 and 5.4.3.

Specification 5.5 has been revised to be consistent with STS 4.3 as follows:

Specification 5.5.A has been replaced with (the specification regarding the new fuel storage facility is relocated to Specification 5.5.1.2):

"5.5.1 Criticality 5.5.1.1 The spent fuel storage racks are der.igned and shall be maintained with:

a. Fuel assemblies having a maximum k. of 1.32 in the normal reactor core configuration at cold conditions (20 C);"

Paae 246 ,

Replace Specification 5.5.B (the spent fuel storage facility specifications are relocated to Specifications 5.5.1.1, 5.5.2 and 5.5.3) with:

(5.5.1.1.b) "b. k,,, < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.3 of the FSAR: and Page 1 of 9

Atta hment 2 to JAFP-98-0385

! REVISION 1 TO SAFETY EVALUATION l PROPOSED TECHNICAL SPECIFICATION CHANGES l

REGARDING DESIGN FEATURES l

c. A nominal center to center distance between fuel i

assemblies placed in the storage racks as described in l Section 9.3 of the FSAR."

l Add Specification 5.5.1.2 (analogous to old Specification 5.5.A):

"5.5.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum k. of 1.31 in the normal reactor core configuration at cold conditions (20 C).;
b. k,,,10.90 if dry; l
c. k,,,10.95 if fully flooded with unborated water; and
d. A nominal 6.625 inch center to center distance between fuel assemblies placed in storage racks."

l r

Specification 5.5.2 provides the maximum number of stored spent fuel assemblies (raised from 2797 to 3239 and relocated from old Specification 5.5.8) to be j consistent with STS: i l l l "5.5.2 Capacity l

l The spent fuel storage poolis designed and shall be maintained with a storage capacity limited to no more than 3239 fuel assemblies."

l '

l l Delete page 246a containing Bases section 5.5.B. STS do not contain Bases sections for the Design Features Specifications.

II. PURPOSE OF THE PROPOSED CHANGES l

The proposed changes provide a limit on fuel bundle reactivity which ensures that '

l criticality margin for the SFP is maintained; provide limitations on the number of j assemblies loaded in the pool to assure tbrmal and structural design requirements

! are met; correct an editorial error; and make this section of Technical Specifications consistent with STS.

l l

1 l

Page 2 of 9 l

Attachment 2 to JAFP-98-0385 REVISION 1 TO SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES

  • lil. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES The respective renumbering of Specifications 5.5.A, 5.5.B, 5.2.A, 5.2.B, 5.4.A, 5.4.B and 5.4.C to 5.1.1, 5.1.2, 5.2.1, 5.2.2, 5.4.1, 5.4.2, and 5.4.3 is an editorial change which has no effect on content or safety.

The correction of the word " Country" in Specification 5.5.1 to " County" is likewise an editorial change with no impact on safety.

The changes to Specification 5.5 are a combination of editorial changes, additional specifications and revised requirements.

The editorial changes associated with making the section consistent with STS are:

The specification of new fuel storage facility K.,, (wet and dry) has been relocated from Specification 5.5.A to Specifications 5.5.1.2.b and 5.5.1.2.c.

The effective neutron multiplication factor limit for the spent fuel storege pool has been relocated from Specification 5.5.B to 5.5.1.1.b. The corresponding fuel assembly infinite neutron multiplication factor in reactor core geometry has been relocated from Specification 5.5.8 to 5.5.1.1.a.

The limitation on the number of stored fuel assemblies has been relocated from Specification 5.5.B to 5.5.2.

l None of these changes have an effect on safety.

Changes adding new limitations to the Technical Specificatioris are:

Specification of spent fuel rack center to center spacing (Specification 5.5.1.1.c) and new fuel rack center to center spacing (Specification 5.5.1.2.c) add details presently described in the FSAR to the Technical Specifications. l

- Addition of a limit on fuel assembly infinite neutron multiplication factor in reactor core geometry applicable to the new fuel racks provides the design limitation which ensures compliance with Specifications 5.5.1.2.b and l 5.5.1.2.c for FitzPatrick (Reference 2). Compliance with this limit is analogous to the old Specification 5.5.A requirement that " Compliance shall be verified prior to the introduction of any new fuel design to this facility."

Page 3 of 9

Attachm';nt 2 ts JAFP-98-0385 REVISION 1 TO SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES  ;

These changes add limitations presently described in other design documents to the Technical Specifications (no new design requirements are established). Therefore, these changes do not affect safety.

I The change in fuel assembly infinite neutron multiplication factor in reactor core j geometry for fuel stored in the spent fuel pool contained in Specification 5.5.1.1.a I (old Specification 5.5.B) is a revised requirement. The basis of this change is as follows:

The presently installed spent fuel storage racks were licensed based on analyses of 8x8 retrofit fuel loaded with 3.3 weight percent U-235, taking no credit for bumable poison, which demonstrated compliance with the limit of k,,,10.95 (References 3,4). The present limit of k.11.36 on maximum, exposure dependent, infinite lattice multiplication was obtained by analyzing that same fuel bundle in a reactor core geometry and adjusting the result to account for calculation and model uncertainties (Reference 5). This allowed loading higher enrichment fuel bundles in the SFP, because when the burnable poison in the bundle is considered, the fuel reactivity limit (and consequently the SFP criticality limit) is met.

As fuel bundle design has evolved to use higher enrichments to support longer, higher energy fuel cycles; the correlation between bundle reactivity in the SFP and in the reactor core has changed. Specifically, higher enrichment i fuel generates more thermal fissions and results in a harder neutron spectrum. Since boron is a 1/v absorber, when high enrichment fuel is introduced in the lattice it reduces the effectiveness of boron absorption.

Therefore, a lower value of infinite lattice neutron multiplication factor in  !

reactor core geometry is required to maintain limits on effective neutron multiplication factor for the spent fuel pool.

To determine the infinite lattice k. required for higher enrichment fuel bundles, the infinite lattice multiplication factors for a GE12 fuel bundle with i an uniform enrichment of 4.6 weight percent U-235 and six 3.0 weight percent gadolinia rods were determined for reactor core and in-rack geometries (Attachment IV). Demonstrating that this design basis fuel bundle meets limits on neutron multiplication for the fuel storage racks allows use of the in-core k. of this bundle to establish the limit on reactivity for fuel stored in the SFP. These calculations were performed with the Monte Carlo N-Particle Transport Code, Version 4A (MCNP), using continuous energy cross sections. The GE12 fuel lattice was chosen for the design basis bundle since it has the highest reactivity for a given enrichment and gadolinia loading (Reference 6).

Page 4 of 9

Attachment 2 t3 JAFP-98-0385 REVISION 1 TO SAFETY EVALUATION j PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES l The calculated minimum k. of the design basis fuel assembly in the uncontrolled reactor lattice geometry at 20"C, corrected for bias and l uncertainty is 1.3207 (95 % probability at the 95% confidence level,

) 95%/95%). The maximum k, of the same bundle in a nominal storage cell at l 4*C, corrected for bias and uncertainty is 0.9189 (95%/95%), which satisfies the storage rack design limit of k.u < 0.95. The limiting temperature with respect to reactivity for spent fuel pool temperatures between 0 C and 120'C is 4 C. The infinite multiplication factor of the design basis fuel bundle stored in the aluminum storage racks is bounded by the result for the stainless steel racks (Reference 6). This is due to the larger center to center spacing of fuel bundles when stored in the aluminum racks, as well as to the higher boron loading of the boral plates used in their construction.

Compliance with the Technical Specification maximum limit of 1.32 on fuel bundle infinite multiplication frector in the uncontrolled reactor geometry at 20 C ensures that SFP effective neutron multiplication factor will be maintained less than 0.95.

Compliance with this revised limit ensures criticality limits will be met, as higher enrichment fuel assemblies are loaded into the spent fuel pool. The analysis supporting this change considered the new racks being installed to raise SFP capacity to 3239 assemblies.

l Reports of additional analyses performed to support the addition of storage racks to raise SFP capacity to 3239 assemblies are contained in Attachments IV and V. Brief l

descriptions of the thermal-hydraulic and structural analyses follow (criticality analysis was discussed above). j Thermal-Hydraulic Considerations As described in Attachment V, three systems are available to remove decay heat from fuel stored in the SFP: Fuel Pool Cooling and Cleanup system (FPCC), Residual Heat Removal system in the Fuel Pool Cooling Assist mode (RHR) and the Decay Heat Removal system (DHR).

The nominal heat removal capabilities of each of these systems are:

System Capability (106 Stu/Hr)

DHR (maximum) 45 DHR (minimum) 30 RHR Assist + FPCC (1 pump,1 Hx) 24 FPCC (2 pumps, 2 Hx =s) 10 FPCC (1 pump,1 Hx) 6.3 l

Page 5 of 9

-- - . .- = - . .

Attachment 2 to JAFP-98-0385 j i REVISION 1 TO SAFETY EVALUATION I

PROPOSED TECHNICAL SPECIFICATION CHANGES i REGARDING DESIGN FEATURES Decay heat load of fuel to be discharged is nominally: ,

Davs After Shutdown Decav Heat (10' Btu /Hr) '

1 47.9 2 39.2 l 3 33.6 I 4 29.7 l 5 26.9 <

6 24.8 l 7 23.2 10 20.0 20 15.1 30 12.4 40 10.6 Additionally, the fuel stored in the pool at the beginning of the limiting outage has a heat load of 2.OE6 Btu /hr.

Sufficient cooling systems shall be maintained available to ensure SFP bulk temperature will be maintained less than or equal to 140'F, assuming an active failure of any single component in the available systems.

When DHR is available. To operate in its maximum cooling mode and the decay heat ,

removat requirements of the previous paragraph are met, there is no limitation on  !

the time of initiation or rate of fue! movement, since the ability of DHR to remove I decay heat is independent of the location of the fuel (see Attachment V).

Otherwise, fuel movement from the core to the pool shall not begin until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> i after shutdown and the rate of movement shall be restricted to a net addition of four irradiated assemblies per hour to the SFP.

Structural Analysis The new rack modules are seismic class I in accordance with the FitzPatrick plant

Updated Final Safety Analysis Report (UFSAR). Seismic analyses of the spent fuel l storage racks were performed, as described in Section 6.0 of Attachment IV, to determine the rack behavior and to ensure no loss of function resulting from either an operating basis earthquake (OBE) or a design basis earthquake (DBE). The l

existing and new racks for the FitzPatrick plant SFP are freestanding and self-supporting structures. The seismic analyses demonstrated the racks sustain minimal kinematic displacements during postulated earthquakes. Thus, no rack-to-rack and rack-to-wallimpacts occurred under any of the dynamic conditions simulated. The analyses reported in Attachment IV demonstrate that the racks meet design requirements during seismic events so that no loss of function will result from an l earthquake.

l l

l Page 6 of 9

Attachment 2 to JAFP-98-0385 '

REVISION 1 TO SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES An evaluation of pool structural loading is described in Section 8.0 of Attachment IV. This evaluation concludes that the loads following addition of the proposed racks will be bounded by the load analyzed in the structural evaluation supporting Technical Specification amendment 175. Therefore, the safety function of the pool )

structure will be maintained with the addition of 442 storage locations, providing a '

SFP capacity of 3239 bundles.

Bases section 5.5.B is deleted to conform to tne STS. STS do not contain bases for -

the design features section. Removal of this material does not impact safety I because the information contained is in the FSAR and other design documents.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION i

Operation of the FitzPatrick plant in accordance with the proposed Amendment l would not involve a significant hazards consideration as defined in 10 CFR 50.92, i since it would not:

(

1. involve a significant increase in the probability or consequences of an accident previously evaluated because:

A change in the infinite lattice neutron multiplication factor for a fue: bundle in the reactor core geometry which enwres the criticality limit for fuel in the spent fuel pool geometry is met does not affect initiation of any accident.

Operation in accordance with the revised limit ensures the consequences of previously analyzed accidents are not changed. Storage of additional fuel assemblies in the pool does not affect the probability or consequences of dropping a fuel assembly, since this accident is localized to a small area of ,

the storage array. Likewise, addition of specifications containing details I presently in plant design documents and editorial changes do not change the probability or consequences of a previously analyzed accident. l

2. create the possibility of a new or different kind of accident from any accident previously evaluated because:

A change in the infinite lattice neutron multiplication factor for a fuel bundle in tha reactor core geometry which ensures the criticality limit for fuel in the spent fuel pool geometry is met does not affect the types of reactivity accidents which may occur. Therefore changing the limit will not initiate a new or different type of accident. Maintenance of available decay heat removal systems ensures that no new type of loss of cooling accident associated with the SFP will occur as a result of storing additionalirradiated fuel assemblies. Likewise, addition of specifications containing details presently in plant design documents and editorial changes do not create the possibility of a new or different type of accident.

I Page 7 of 9

I

  • l l Attachment 2 to JAFP-98-0385 I REVISION 1 TO SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES
3. involve a significant reduction in a margin of safety because:

The revised limit on infinite lattice neutron multiplication factor for a fuel bundle in the reactor core geometry ensures maintenance of the same margin of safety with respect to criticality as presently exists for storage of fuel in the SFP. Storing additional irradiated fuel assemblies in the pool does I

not affect the margin of safety with regard to pool cooling since sufficient

! heat removal systems will be maintained available to ensure maintenance of acceptable pool temperatures. Addition of specifications containing details presently in other design documents and editorial changes have no effect on the margin of safety.

V. IMPLEMENTATION OF THE PROPOSED CHANGES i

This amendment request meets the eligibility criteria for categorical exclusion set l- forth in 10 CFR 51.22(c)(9) as follows:

1 (i) the amendment involves no significant hazards consideration.

l As demonstrated in Section IV of this evaluation, the proposed change >

involves no significant hazards consideration.

l l (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

l The revised limitation on fuel assembly reactivity ensures the margin of i safety presently assumed for prevention of spent fuel pool criticality is l maintained. This change has no effect on effluents that may be released offsite. Storage of additional fuel assemblies in the SFP will not result in a significant change in the amount of waste generated in the Fuel Pool Cooling and Cleanup system. Therefore there is no significant change in the amount of effluents which may be released offsite. r (iii) there is no significant increase in individual or cumulative occupational l radiation exposure. l The revised limitation on fuel assembly reactivity ensures the margin of l safety presently assumed for prevention of spent fuel pool criticality is maintained. This has no effect on occupational radiation exposure. Sufficient shielding is provided by the SFP design that storage of additional fuel I assemblies has no significant affect on occupational exposure. Therefore,

! there will be no change in individual or cumulative radiation exposure.

a Page 8 of 9

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

- Attachment 2 to JAFP-98-0385 REVISION 1 TO SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DESIGN FEATURES Based on the above, it is concluded that there will be no impact on the environment resulting from the proposed change and the proposed change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.21 relative to requiring a speiific environmental assessment by the ,

Commission. I i

Additionally, implementation of the proposed change will not adversely affect the l Fire Protection Program at the FitzPatrick plant. '

VI. CONCLUSION Based on the discussions above, implementation of a maximum, infinite lattice l multiplication factor for fuel stored in the spent fuel pool does not involve a significant hazards considerati on, or an unreviewed safety question, and will not endanger the health and safety of the public. Similarly, raising the number of fuel assemblies that may be stored :n the SFP does not involve a significant hazards consideration, or an unreviewed safety question, and wil! not endanger the health and safety of the public. The Plant Operating Review Committee and Safety Review i Committee have reviewed this preposed Technical Specification change and agree i with this conclusion. i

. Vll. REFERENCES

1. " Standard Technical Specifications General Electric Plants, BWR/4,"

NUREG-1433, April 1995.

2. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A-13, August 1996.

3. Nuclear Associates International Corp. Report, " Nuclear Criticality Analysis l for the Spent Fuel Racks of the FitzPatrick Power Plant," NAl 78-12,  :

February 1978.

4. Holtec International Report, " Licensing Report for increased Storage Capacity for FitzPatrick Spent Fuel Pool," HI-89399, February 1989.
5. GE Letter, P. van Diemen to G. Rorke (NYPA), "FitzPatrick Fuel Storage
K-infinity Conversion, Revision 1," July 10,1986.
6. Evaluation of the FitzPatrick High-Density Storage Rack K. Criterion, JAF-RPT-MISC-02494R1, August 1997 i

4 i

Page 9 of 9 l

i Attachment 3 to JAFP-98-0385 6

REVISED MARKUP OF TECHNICAL SPECIFICATION PAGE CHANGES PROPOSED TECHNICAL SPECIFICATION l CHANGES REGARDING DESIGN FEATURES 1

1 i

i 1  :

New York Power Authority l

l' JAMES A. FITZPATRICK NUCLEAR POWER PLANT

j. Docket No. 50-333 DPR-59

. . _ _ _ . . _ . . _ _ . . . _ _ . _ -_ - _ . _ .. _ _ _ _ ._. _ ... _ .._ . _ _ . - . _ _ .. _ _ _ - ~. _ ... _ _. _ ..- __ - . ._..._m_ _ _ .

t t

JAFNPP t

5.0 DESIGN FEATURES 5.3 REACTOR PRESSUREVESSEL l 5.t SITE C%d1 The reactor pressure M is as W in Table 42-1 W k" PASNY JamesportenA. FN Nuclear Power Plant 'Point of thshSSe located on site, the r-;-cm;iTC."; 3,000 M Wcodes areWin

  • Unit 1. The NPP-R. east of the Nine hAs Ptitti Nuclear %^^1.

JAF siteis onI.ake Ontarioin C:: : -

^

5'4 . New York CONTAINMENT

_ 4-7-- plant islocated '54' a't coonsnate' 7 miles northeast m, on of M.r

' ' t s norti 4A19,545812 m, east 386,968.945 The pnncipal desagn parameters and charactenstics for the l

the Universal Transverse Mercator Syalem. pnmary contanment are given in Table 52-1 of the FSAR. ,

> 5I1 ,. s.g.3 B: ,& The secondary containment is as desenbod in Section 5.3 and The nearest point on the property line from the reactor building .

and any points of potenelsf gaseous eiEuents, with the exception the apphcable codes are as described in Secten 12.4 of the of the lake shoreans,is located at the norW comer of the pg

' a ,. s.g. 3 property. This datance is approidmetely 3,200 ft. and is the .

e:) Penetrations of the primary contamment and pipeng passing j ratSus of the excluelon areas as doened in 10 CFR 100.3. through such penetratens are designed in accordance with  !

standards set forth in Section 5.2 of the FSAR.

52 REACTOR O w-~

,,3 3 g ..

. - . _ _-- n n. ,

Ad The reactor core consasts of not more than 560 fuel assembbes. -

Each assembly shen conesst of a matrix of Zircaloy clad fuel rods ^

"= :::: L^^' -'.;;,_ : ^i ":1._ . ;.-L.L. ; ; O .; ^ -_i . ; .",,

wth an intial wT-- ^^* -i of slighdy ennched uranium dioxide ci :0.00 -./"::f:I :0 M 0;.. f-.e. C.C b ._~:f p._

% ) as fuel material. Fuel assemblies shas be limited to those O ; a " f -- . ;; e. ., . .

' " ^ - ' , ^. 7 ;. '- ~~, .

. . /

designs ---c -d by the NRC stas for use in BWRs.

& The reactor core contains 137 cruciform-shaped control rods as  %

described in Secton 3.4 of the FSAR. -

i l i

Amenoment no. ao. *. *.e4. w.14. w9. i>< is2 -

245

)' *, l

_ . _ _ . . _ _ . , _ . - - - - = - = +

.aM )

r JAFNPP n_..,

.J \%!%!9 Es %E]

.. ,m_____-_._._-____ _.u >- u,. .

.s. s e su p su asses essa, y y we www y. - .v- _ _ a ._ . u :. i.,g - . - . . . - - _ . . - .

n.n_e

. _ _ .a__ . u_ _.._ i,. . u.._

. _ . . . %_.: v.,_ 1ngef ,

. Q

=_ _e_ - t # u,.,,. , in rr - - - - - - " " ' " " ": _ w*-- ___

.. _, _. m._.

m. - u.

..,~-,m,,, ..T- > - - - ..

_ - - -__ _..,_---___.2_'a.

".n~- ..u._u__._..2 ,

_y __

_. _n- .'n. --.~-a_-__ e a= -

/

Q .

__ .. . - _ . . _ _ . . m

__ . ,_,m._,,m__-_..

u.___m__

8 -

.i u_ ,..

.. . w_ . % - _, ..

_ _ , - _._.n._,___..

_______.m___._m_.__. .___2.,..o, 5.6 SEISMIC DE - % s.

The reactor N=Ar%g and all enginoored safeguards are desagned on a basis of dynamic analysis using mmalarapon response spectrum curves which are normaEzed to a ground motion of 0.08g for the Operating Basis Earthquake and 0.15g for the Desip Basis Earequake.

Amendment No.K,Ji6,%)8T, 175 ,

246

.au

I t

t t

JAFMPP 'f v-

.a assas The spent fuel pool density fuel storage rocks are I structures store up to 2,797 fuel

. The storage ese doesgned to maintain a configuradet W e multiplication factor (k,)

less 0.95 for at passtle aperadonal and abnormal c.cl:11- The numinar estelsehty analyses for the Spent Fusi Rocks esences 1 and al conclude that fresh fuel hundes with 3 wh U-235 meet the 0.95 k, limit. This ' '

design beeis was roerWad to detennine its inhnite lettice _^^';"::^'_- factor, k .'when in a reactor core t O ICT ~)CtCgC, I

geometry . This k,,was obtained under U conservative -:=": _" ^Jr' eseumptions and reduced by 2.33

'1' times the standard '- - in the *W resulting in the Technical Specincetion of 1.36.

i'

]

References:

% Il incrossed Spent Fuel Storage 'fication, Stone &

f f Webster Engineering C ;- _^J: ., on, Mass. March 15,1978.

L

2) General Electric letter, P. Van Deemen t . Rorke, FitaPetrek Fust Sternes K-infinity Conver , Rowesen 1, deced July 10,1988.

t 3) incrossed Storage Capocoty for Fit Patrick Spent Pool, Holtec Intemononal, Mount Laurel, New Jet , ,

February,1989.

. s

.~ .s Amendment No. Idl. If 5 190 246a M)l

~ _ . . _ . _ . _ _ _ _ _ _ _ . _ _ ._ ___ _ _ . _ _. _

l .

l Attrchment 3 to JAFP-98-0385 (

MARKUP OF TECHNICAL SPECIFICATION PAGE CHANGES I

Insert a: '

t l 5.5 FUEL STORAGE 5.5.1 Criticality ,

l 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum k. of 1.32 in the normal reactor core configuration at cold conditions (20 C);
b. k,,, < 0.95 if fully flooded with unborated water, which l

includes an allowance for uncertainties as described in Section 9.3 of the FSAR; and l

c. A nominal center to center distance between fuel assemblies ,

i placed in the storage racks as described in Section 9.3 of the '

l' FSAR.

5.5.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum k. of 1.31 in the normal l reactor core configuration at cold conditions (20 C), ,

1

b. k,,,10.90 if dry; l
c. k,,,10.95 if fully flooded with unborated water; and  :
d. A nominal 6.625 inch center to center distance between fuel I assemblies placed in storage racks.

5.5.2 Capacity The spent fuel storage poolis designed and shall be meintained with a storage l capacity limited to no more than 3239 fuel assemblies.  !

l l

l r

, - - - _ . , - ,,

Retrieved from "https://kanterella.com/w/index.php?title=ML20197G618&oldid=3011022"