ML20237K167

Applicant Exhibit A-12,consisting of to All Power Reactor Licensees Forwarding Guidance on Spent Fuel Pool Mods Entitled, Review & Acceptance of Spent Fuel Storage & Handling Applications

ML20237K167
Person / Time
Site:	Diablo Canyon
Issue date:	06/17/1987
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
OLA-A-012, OLA-A-12, NUDOCS 8709040228
Download: ML20237K167 (28)

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PGandE Exhibit No. 12 i'

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1, NUCLEAR. REGULATORY COMMISSION I G 25 P3 :33 J

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U WASHINGTON. D. C. 20555 0l I

7 April 14, 1978 6w1 n

To All Power Reactor Licensees

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1 Gentlemen:

i Enclosed for your information and possible future use is the NRC

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guidance on spent fuel pool modifications, entitled " Review and Acceptance of Spent Fuel Storage and Handling Applications". This 1

document provides (1) e.dditional guidance for the type and Gxtent I

I of information needed by the NRC Staff to perfom the review of licensee proposed modifications of an operating reactor spent fuel storage pool and (2) the acceptance criteria to be used by the NRC Staff in authorizing such modifications.

This includes theE information needed to make the findings called for by the CommfysiS5 in the Federal Register Notice dated September 16,1975 (copy encitited) with regard to authorization of fuel pool modifications prior-io th Z

completion of the Generic Environmental Imoact Statement, "gtdli,dg fri and Storage of Spent Fuel from Liaht Water Nuclear Power Reiminr;da O

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, The overall design objectives of a fuel storage facility at f;reattor complex are governed by various Regulatory Guides, the Standjfrd c.7 iQ J

Review Plan (NUREG-75/087), and various industry standards.1-Thk O

guidante provides a compilation in a single document of the-}erfinent portions of these applicable references that are needed in addressing spent fuel pool modifications.,No additional regulatory requirements i

are imposed or implied by this document.

Based on a review of license applications to date requesting authorization i

to increase spent fuel storage capacity, the staff has had to request additional infomation that could have been included in an adequately documented initial submittal.

If in the future you find it necessary to apply for authorization to modify onsite spent fuel storage J

i capacity, the enclosed guidance provides the necessary information and acceptance criteria utilized by the NRC staff in evaluating these applications.

Providing the information needed to evaluate the matters covered by this document would likely avoid the necessity for NRC questions and thus significantly shorten the time required

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to process a fuel pool modification amendment.

Sincerely, i

870 040228 070ho[73 hDR A D D ";t 0 5 o

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PDR Brian K. Grimes, Assistant Director j

for Engineering and Projects Division of Operating Reactors l

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Enclosures:

o93 78 i

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NRC Guidance 2.

Notice 6

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hifCMAR RIGATORY C0MfA115101 b

k.ket No. M, b 4' E10$_O O Th 0ffd1 hti. N4 in the m.itter of J.'

rA Staff fDINTWlED RECIIVIis t

Applicant RlHCTED I&M*0f Cont's Off'r b'?

• 0 ~ 1 CATE Contractor Witness Other bYN 3 ? M M N 6":

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ENCLOSUdE NO. 1 i

i OT POSITION FOR REVIEW AND ACCEPTANCE OF SPENT FUEL STORAGE AND HANDLING APPLICATIONS I.

BACKGROUND Prior to 1975, low density spent fuel storage racks were designed with i

a large pitch, to prevent fuel pool criticality even if the pool contained the highest enrichment uranium in the light water reactor fuel assemblies.

Due to an increased demand on storage space for spent fuel assemblies, the more recent approach is to use high density r

storage racks and to better utilize available space.

In the case of operating plants the new rack system interfaces with the old fuel pool structure.

A proposal for installation of high density storage racks may involve a plant in the licensing stage or an operating plant.

The requirements of this position do not apply to spent fuel storage and handling facilities away from the nuclear reactor complex.

On September 16, 1975, the Commission announced (40 F. R. 42801) its intent to prepart a generic environmental impact statement on handling g

and storaga of spent fuel frcm light water power reactors.

In this notice, the Commission also announced its conclusion that it would not s

be in the public interest to defer all licensing actions intended to ameliorate a possible shortage of spent fuel. storage capacity pending completion of the generic environmental impact statement.

The Commission directed that in the consideration of any such proposed licensing action, an environmental % pact statement or environmental impact appraisal shall be prepared in which five specific factors in addition to the normal cost / benefit balance and environmental' stresses should be applied, balanced and weighed.

The overall design objectives of a fuel storage facility at the reactor complex are governed by various Regulatory Guides, the Standard Review Plan, and industry standards which are listed in the reference section.

Based on.the reviews of such applications to date it is obvious that the staff had to request additional information thet could be easily included in an adequately documented initial submittal.

It is the intent of this document to provide guidance for the type and extent of information needed to perform the review, and to indicate the acceptance criteria where applicable.

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1 II.

_ REVIEW DISCIPLINES The objective of the staff review is to prepare (1) Safety Evaluation Report, and (2) Environmental Impact Appraisal.

The broad staff disciplines involved are nuclear, mechanical, material, structural, and environmental.

I Nuclear and thermal-hydraulic aspects of the review include the poten-a i

tial for inadvertent criticality in the normal storage and handling of the spent fuel, and the consequences of credible accidents with respect to criticality and the ability of the heat removal system,to maintain sufficient cooling.

Hechanical, material and structural aspects of the review concern the capability of the fuel assembly, storage racks, and spent fuel pool system to withstand the effects of natural phenomena such as earth-quakes, tornadoes, flood, effects of external and internal missiles, thermal loading, and also other service loading conditions.

The environmental aspects of the review concern the increased thermal

-I and radiological releases from the facility under normal as well as I

accident com"tions, the occupational radiation exposures, the genera-q g

tio,n of radioactive waste, the need for expansion, the commitment of y

material and nonmaterial resources, realistic accidents, alternatives to the proposed action and the cost-benefit, balance.

j The information related to nuclear and thermal-hydraulic type of i

analyses is discussed in Section III.

The mechanical, material, and structur.a1 related aspects of informa-tion are discussed in Section IV.

The information required to complete an environmental impact assess-ment, including the five factors specified by the Commission, is provided in Section V.

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III. NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS 1.

Neutron Multiplication Factor To include all credible conditions, the licensee shall calculate in the fuel the effective neutron multiplication factor -k j

storagepoolundet*thefollowingsetsof.nosumI$f,ortitions:

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1.1 Normal Storage 1

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a.

The racks shall be designed to contain the most reactive l

fuel authorimi to be stored in the facility without any control rods or,any noncontained" burnable poison and the fuel shall)e assund to be at the most reactive point in its life.

1 b.

The moderator sha:11 be assumed to be pure water at the temperature within the fuel pool limits which yields the largest reactivity.

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c.

The array shall be assumed to be infinite in lateral extent

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or to be surrounded by an infinitely thick water reflector and thick concrete,** as appropriate to the design.

1 d.

Mechanical uncertainties may be treated by assuming " worst case" conditions or by performing sensitivity studies and obtaining appropriate uncertainties.

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e.

Credit may be taken for the neutron absorption in structural materials and in solid materials added specifically for!

neutron absorption, provided a means of intpection ie, enab-lished (refer to Section 1.5).

1.2 Postulated Accidents 4

The double contingency principle of ANSI N 16.1-1975 chall '.e applied.

It shall require two unlikely, independent, concurrent events to produce a criticality accident.

Realistic initial conditions (e.g., the presence of soluble I

boron) vay be assumed for the fuel pool and fuel assemblies. The

""Noncontained" burnable poison is that which is not an integral part of the fuel assembly.

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• • It should be noted that under certain conditions concrete may be a more-effective reflector than water.

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postulated accidents shall include:

(1) dropping of a fuel element on top of the racks and any other achievable abnormal loc.ation of a fuel assembly in the pool; (2) a dropping or tip-ping of the fuel cask or other heavy objects into the fuel pcol; (3) effect of tornado or earthquake on the deformation and rela-l tive position of the fuel racks; and (4) loss of all cooling i

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systems or flow under the accident conditions, unless the cooling l

system is single failure proef.

1.3 Calculation Methods The calculation method and cross-section values.shall be verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed.

Sufficiently diverse conf 4urations shall be calculated to render improbable the

" cancellation of error" in the calculations.

So far as practi-cable the ability to correctly account for heterogeneities (e.g.,

thin slabs of absorber between storage locations) shall be demonstrated.

A calculations 1 bias, including the effect of wide spacing between assemblies shall be determined from the comparison between calcu-lation and experiment.

A calculation uncertainty shall be

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determined such that the true multiplication factor will be less s.'

than the calculated value with a 95 percent probability at a 95 percent confidence level.

The total uncertainty factor on k 1

shall be obtained by a statistical combination of the calculaEf7 tional and mechanical uncertainties.

The k value for the racksshallbeobtainedbysummingthecalcN$tedvalue,the

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calculadonal bias, and the total uncertainty.

1.4 Rack Modification f

For modification to existing racks in operating reactors, the

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following information should be provided in order to expedite the review:

(a) The everall size of the fuel assembly which is to be stored in the racks and the fraction of the total cell area which represents the overall fuel assembly in the model of the

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nominal storage lattice cell; j

i (b) For H,30 + stainless steel flux trap lattices; the nominal I

thickhess and type of stainless steel used in the storage racks and the thermal.(.025 ev) macroscopic neutron absorp-tion cross section that is used in the calculation method

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(c) Also, for the H 0 + stainless steel flux trap lattices, the changeofthec$1culatedneutronmultiplicationfactorof j

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9 infinitely long fuel assemblies in infinitely large arrays inthestoragerack(i.e.,thegofthenominalfuelstorage lattice cell and the changed g) for:

(1) A change in fuel loading in grams of U2ss, or equiva-1ent, per axial centimeter of fuel assembly where it is assumed that this change is made by increasing the enrichment of the U2ss; and, (2) A change in the thickness of stainless steel.in the storage racks assuming that a decrease in stainless i

steel thickness is taken up by an increase in water thickness and vice versa; 1

(d) For lattices which use boron or other strong neutron absorb-ers provide:

1 (1) The effective areal density of the boron-ten atoms (i.e.,B10 atcms/cm2 or the equivalent number of boron-ten atoms for other neutron absorbers) between fuel I

cssemblies.

(2) Similar to Item C, above, provide the sensitivity of q

)g thestoragelatticecellgto:.

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(a) The fuel loading in grams of U2ss, or equivalent, l

per axial centimeter of fuel assembly, (b) The storage lattice pitch; and, (c) The areal density of the boron-ten atoms'between fuel assemblies.

i 1.5 Acceptance Criteria for Criticality The neutron multiplication factor in spent fuel pools shall be i

less than or equal to 0.95, includino all uncertainties, under all conditions l

(1) For those facilities which employ a strong neutron absorbing I

material to reduce the neutron multiplication factor for the storage pool, the licensee shall provide.the description of onsite tests which will be performed to confirm the presence and retention of the strong absorber in the racks.

The results of an initial, onsite verification test shall show 4

within 95 percer.t confidence limits that there is a suffi-cient amount of neutron absorber in the racks to maintain

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the neutron multiplication factor at or below 0.95.

In addition, coupon or other type of surveillance testing shall l

be performed on a statistically acceptable sample size on a l

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I pediedit basis thrcui,hout the life of the racks to verify the continued preser.ce of a sufficient amount cf neutron i

absorber in the racks to maintain the neutron multiplication l

fSctor at or below 0.95.

(2) Decay Hect Calculations for the Spent Fuel d

The esir,6tt,tions for the amount of thermal energy that will p

have te be removed by the spent fuel poe1 cooling system 1

shall be eade in accordaace with Branch Technical Position AFCM 9-2 entitles, " Residual Decay Energy for Light Water Reactors fer Long Term Cooling." This Granch Technict.1 Position is.part ef the Standard P wien "rlan (IM!EG 75/087).

(1) Thermal-tryd aulic Analyses fer Spent Feel Ccoling Conservative, methods should be usci to caledate the maximum fuel temperature and the increase in temperatur' of the s

water in the pool.

The maximum void fraction in the fuel assembly and between fuci assemblies should alsc be calculated.

Ordinarily, in order not to exceed the design heat load for i

the spent fuel cooling system it will be necessary to de a

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certair; amount of cooling in the react.or vessel af*.er nactor shutdown prior to moving fuel essemblies into the spent Tael pool.

The bases for the analyses should include the estab-lished cooling times for both the usual refueling case and q

the full core off load case.

A potential for a large increase ir.<the reactivity in an H O 2

i flux trap storage.lettice e,xists if, x mehow, the vater is J

kept out or forced out of the space be:Neen the tuel assem-blies, conceivably by trapped air or steam =

For this reason, it-is wxersary to show that the design of the storage rack is such that this will not occur and that these spaces will alWays have water in thea.

Also, in some cases, direct garmia heating of the fuel storaga cell valls end of the intArcell e,ter may be significant.

It is recessary to consider direct gamma heating of the fuel storage cell walls and of the Mtercell water te show that boiling will not-occur in the water channels "setween the fuel assemblies.

Under postulated accident conditions where all non-Category I spent feel pool cooling systems become inoperative, it is necessary to show thr;t there is an alternt.te method for h

cooling the spent pr;1, water.

When this alternative method j

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requires the installation of alternate components or signifi-i j

cant physicas alteration of the ecoling system, the detailed l

cteps shall be described, along with tt.'e time required fcr

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Aisc, the average amount of u tcr in the fuel pool and the expected heat up rate of this water assuming loss of all cooling systems shall be specified.

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,l (4) Potential Fuel and Rack Handlin0 Accidents The method for moving the racks to and from and into and out of the fuel pool, should be described.

Also, for plants where the spent fuel pool modification requires different fuel handling procedures than that described in the Final Safety Analysis hport, the differences should ba discussed.

If potential fuel and rack handling accidents occur, the neutron multiplication factor in the fuel pool shall not l

l exceed 0.S5.

These postulated accidents shall not be the l

g cause of the loss of cooling for either the spent fuel or l

L the reactor.

(5) Technical Specifications i

To insure against criticality, the following technical speci-fications are needed on fuel storage in high density racks:

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The neutron multiplication factor in the fuel pool-shall be less than or equal to 0.95 at all times.

I I) 2.

The fuel loading (i.e., grams of uranium-235, or H

equivalent, per axial centimeter of assembly) in fuel assemblies that are to be loaded into the high density, racks should be limited.

The number of grams of uranium-235, or equivalent, put in the plant's tech-nical specifications shall preclude criticality in the i

fuel pool.

Excessive pool water temperatures may lead to excessive loss of water due to evaporation and/or cause fogging.

Analyses of thermal load should consider loss of all pool cooling systems.

To avoid exceeding the specilied spent fuel pool temperatures, consideration shall be given to incorporating a technical specification limit on the pool water tempera-ture that would resolve the concerns described above.

For l

limiting values of pool water temperatures refer to ANSI-N210-1976 entitled, "Desien Objectives for Light Water l

Reactor Spent Fuel Storage Facilities at Nuclear Power i

Stations," except that the requirements of the Section j

9.1.3.III.1.d of the Standard Review Plan is applicable for the maximum heat load with normal cooling systems in operation.

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IV.

MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS (1) Description of the Spent Fuel Poo1 and Racks Descriptive information including plans and sections showing the 1

spent fuel pool in relation to other plant structures shall be provided in order to define the primary structural aspects and 1

elements relied upon to perform the safety-related functions of the pool and the racks.

The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all environmental and abnormal i

loadings, such as earthquake, and impact due to spent fuel cask drop, drop of a spent fuel as:;embly, or drop of any other heavy object during routine spent fuel handling.

The major structural elements reviewed and the extent of the descriptive information required are indicated below.

(a) Support of the Spent Fuel Racks:

The general arrangements and principal features of the hori2.ontal and the vertical supports to the spent fuel racks should be provided indi-h cating the methods of transferring the ' loads on the racks to the fuel pool wall and the foundation slab.

All gaps i

(clearance or expansion allowance) and sliding contacts

• should be indicated.

The extent of interfacing between the new rack system and the old fuel pool walls and base slab should be discussed, i.e., interface loads, response spec-tra, etc.

If connections of the racks are made to the base arid to the

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side walls of the pool such that the pool liner may be perforated, the provisions for avoiding leakage of radio-activa water of the pool should be indicated.

(b) Fuel Handling:

Postulation of a drop accident, and quanti-fication of the drop parameters are reviewed under the environmental discipline.

Postulated drop accidents must include a straight drop on the top of a rack, a straight drop through an individual cell all the way to the bottom of the rack, and an inclined drop on the top of a rack.

In-tegrity of the racks and the fuel pool due to a postulated fuel handling accident is re11ewed under the mechanical, material, and structural dis'iplines.

Sketches and suffi-c a

l cient details of the fuel handling system should be provided i

to 7acilitate this review.

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(2) Applicable Codes, Standards and Specifications Construction materials should conform to Section III, Subsec-tion NF of the ASME* Code.

All Materials should be selected to L

be compatible with the fuel pool environment to minimize corro-sion and galvanic effects.

Design, fabrication, and installation of spent fuel racks of stainless steel material may be performed based upon the AISC**

specification or Subsection NF requirements of Section III of the ASME B&PV Code for Class 3 component supports.

Once a code is chosen its provisions must be followed in entirety. When the AISC specification procedures are adopted, the yield stress values for stainless steel base matal may be obtained from the Section III of the ASME B&PV Code, and the design stresses de-fined in the AISC specifications as percentages of the yield stress may be used.

Permissible stresses for stainless steel welds used in accordance with the AI5C Ccde may be obtained from Table NF-3292.1-1 of ASME Section III Code.

Other materials, design procedures, and fabrication techniques will be reviewed on a case by case basis.

(3) Seismic and Impact Loads For plants where dynamic input data such as floor response spec-tra or ground response spectra are not available, necessary dynamic analyses may be performed using the criteria described in l

Section 3.7 of the Standard Review Plan.

The ground response i

spectra and damping values should correspond to Regulatory Guide 1.60 and 1.61 respectively.

For plants where dynamic data are available, e.g., ground response spectra for a fuel pool sup-ported by the ground, floor response spectra'for fuel pools supported on soil where soil-structure interaction was considered in the pool design or a floor response spectra for a fuel pool supported by the Teactor building, the design and analysis of the new rack sy: tem may be performed by using either the existing input parameters including the old damping values or new param-eters in accordance with Regulatory Guide 1.60 and 1.61.

The use i

of existing input with new damping values in Regulatory Guide 1.61 is not acceptable.

Seismic excitation along three orthogonal directions' should be imposed simultaneously for the design of the new rack system.

^American T6ciety of Mechanical Engineers Boiler and Pressure Vessel i

Codes, Latest Edition.

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• • American Institute of Steel Construction, Latest Edition.

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l The peak response from each direction should be combined by square root of the sum of the squares.

If response spectra are available for a vertical and horizontal directions only, the same i

horizontal response spectra may be applied along the other hori-zontal direction.

The effect of submergence of the rack system on the damping and the mass of the fuel racks has been under study by the NRC.

Submergence in water may introduce damping from two sources, (a)

J viscous drag, and (b) radiation of energy away from the submerged body in those cases where the confining boundaries are far enough away to prevent reflection of waves at the boundaries.

Viscous i

damping is generally negligible.

Based upcn the findings of this current study for a typical high density rack configuration, wave J

reflections occur at the boundaries so that no additional damping should be taken into account.

A report on the NRC study is to be published shortly under the title " Effective Mass and Damping of Submerged Structures (UCRL-52342)," by R. G. Dong.

The recommendations provided in this report on the added mass effect provide an acceptable basis for the staff review.

Increased damping due to submergence in water is not acceptable without applicable test data and/or

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p detailed analytical results.

Due to gaps betkeen fuel assemblies and the walls of the guid.e tubes, additional loads will be generated by the impact of fuel assemblies during a postulated seismic excitation.

Additional loads due to this impact effect may be determined by estimating l

l the kinetic energy of the fuel assembly. The maximum velocity of j'

the fuel assembly may be estimated to be the spectral velocity l

associated with the natural frequency of the submerged fuel assembly.

Loads thus generated should be considered for local as l

well as overall effects on the walls of the rack and the sup-i porting framework.

It should be demonstrated that the consequent i

loads on the fuel assembly do not lead to a damage of the fuel.

l Loads generated from other postulated impact events may be accept-able, if the following parameters are described in the report:

the. total mass of the impacting missile, the maximum velocity at the time of impact, and the ductility ratio of the target material utilized to absorb the kinetic energy.

(4) Loads and Load Combinations:

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Any change in the temperature distribution due to the proposed modification should be identified.

Information pertaining to the applicable design loads and various combinations thereof should I

I be provided indicating the thermal load due to the effect of the maximum temperature distribution through the pool walls and base l

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slab.

Temperature gradient across the rack structure due to I

differential heating effect between a full and an empty cell o

should be indicated and incorporated in the design of the rack structure.

Maximum uplift forces available from the crane should be indicated including the consideration of these forces in the design of the racks and the analysis of the existing pool floor, 1

if applicable.

The specific loads and load combinations are acceptable if they i

are in conformity with the applicable portions of Section 3.8.4-II.3 of the Standard Review Plan.

(5) Design and Analysis Procedures Details of the mathematical model including a description of how the important parameters are obtained should be provided includ-ing the following:

the methods used to incorporate any gaps between the support systems tind gaps between the fuel bundles j

and the guide tubes; the methods used to lump the masses of the fuel bundles and the guide tubes; the methods used to account for the effect of sloshing water on the pool walls; and, the effect of submergence on the mass, the mass distribution and the effec-tive damping of the fuel bundle and the fuel racks.

9 The design and analysis procedures in accordance with Section.

3.8.4-II.4 of the Standard Review Plan are acceptable. The effect on gaps, sloshing water, and increase of effective mass and damping due to submergence in water should be quantified.

When pool walls are utilized to provide lateral restraint at higher elevations, a determination of the flexibility of the pool walls and the capability of the~ walls to sustain such loa'ds should be provided.

If the pool walls are flexible (having a fundamental frequency less than 33 Hertz), the floor response spectra corresponding to the lateral restraint point at the higher elevation are likely to be greater than those at the base of the pool.

In such a case using the response spectrum approach, j

two separate analyses should be performed as indicated below:

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.(a) A spectrum analysis of the rack system using response spectra corresponding.to the highest support elevation provided that there is not significant peak frequency shift between the response spectra at 1M lower and higher elevations; and, (b) Astaticanalysisoftheracksystembysubjectingittothe maximum relative support displacement.

4 The resulting stresses f*om the two analyses above should be combined by the absolute sum method.

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(Q In order to determine the flexibility of the pool wall it is acceptable for the licensee to use equivalent mass and stiffness p'ropertiesobtainedfromcalculationssimilartothosedescribed Introduction to Structural Dynamics" by J. M. Biggs published by McGraw Hill Book Company.

Should the fundamental frequency of the pool wall model be higher than or equal to 33 Hertz, it may be assumed that the response of the pool wall and the corres-i I

ponding lateral support to the new rack system are identical to those of the base slab, for which appropriate floor response spectra or ground response spectra may already exist.

(6) Structural Acceptance Criteria

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When AISC Code procedures are adopted, the structural acceptance criteria are those given in Section 3.8.4.II.5 of the Standard l

Review Plan for steel and concrete structures.

For stainless steel the acceptance criteria expressed as a percentage of yield stress should satisfy Section 3.8.4.11.5 of tie Standard Review l

Plan.

When subsection NF,Section III, of the ASME B&PV Code is l

used for the racks, the structural acceptance criteria are those given in the Table below.

For impact loading the ductility ratios utilized to absorb kinetic I) energy in the tensile, flexural, compressive, and shearing modes should be quantified.

When considering the effects of seismic loads, factors of safety against gross sliding and overturning of racks and rack modules.under all probable service conditions shall be in accordance with the Section 3.8.5.II-5 of the Stand-I ard Review Plan.

This position on factors of safety against sliding and tilting need not be met provided any one of the following conditions is niet:

(a) it can be shown by detailed nonlinear dynamic analyses that the amplitudes of sliding motion are minimal, and impact between adjacent rack modales or between a rack module and the pool walls is prevented provided that the factors of safety against tilting are within the values permitted by Section 3.8.5.II.5 of the Standard Review Plan.

(b) it can be shown that any sliding and tilting motion will be contained within suitable geometric constraints such as thermal clearances, and that any impact due to the clear-ances is incorporated.

f Materials, Quality C' ntrol, and Special Construction Techniques:

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(7) o The materials, quality control procedures, and any special con-4

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struction techniques should be described.

The sequence of in-l sta11ation of the new fuel racks, and a description of the pre-1 i

cautions to be taken to prevent damage to the stored fuel during j

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q IB TABLE Load Combination Elastic Analysis Acceptance Limit f

D+L hormal limits of NF 3231.la 4

D + L + E.

Normal limits of NF 3231.la i

D + L + To 1.5 times normal limits or the lesser of 2 Sy and Su D + L + To + E 1.5 times normal limits or the leser of 2 Sy and Su 0 + L + Ta + E 1.6 times normal limits or the lesser of 2 Sy or Su l

D + L + Ta + E Faulted condition limits of 1

NF 3231.1c i

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Limit Analysis l

I 1.7 (D + L)

Limits of XVII-4000 of Appendix XVII

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of ASME Code Section III i

1.7 (D + L + E) 1.3 (D + L + To) 1.3 (D + L + E + To) 1.1 (D + L + Ta + E)

Notes:

1.

The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan where each term is defined except for Ta which is defined as the highest i

temperature associated with the postulated abnormal design conditions.

2.

Deformation limits specified by the Design Specification limits shall be satisfied, and such deformation limits should preclude damage to the fuel assemblies.

3.

The provisions of NF 3231.1 shall be ammended by the requirements of the paragraphs c.2, 3, and 4 of the Regulatory Guide 1.124 entitled " Design Limits and Load

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Combinations for Class 1 Linear-Type Component Supports."

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I the construction phase should be provided.

Methods for struc-tural qualification of special poison materials utilized to absorb neutron radiation should be described.

The material for 2

the fuel rack is reviewed for compatibility inside the f9e1 pool environment.

The quality of the fuel pool water in terms of the pH value and the available chl'orides, fluorides, horon, heavy j

metals should be indicated so that the long-term integrity of the 4

i rack structure, fuel assembly, and the pool liner can be evaluated.

f Acceptance criteria for special materials such as poison materials should be based upon the results of the qualification program supported by test data and/or analytical procedures. -

3 If connections between the rack and the pool liner are made by welding, the welder as well as the welding procedure for the welding assembly shall be qualified in accordance with the appli-cable code.

If precipitation hardened stainles's steel material is used for the construction of the spent fuel pool racks, hardness testing should be performed on each rack component of the subject material to verify that each part is heat treated properly.

In additica,

. the surface film resulting from the heat treatment should be

{

$)

removed from each piece to assure adequate corrosion resistance.

(8) Testing and Inservice Surveillance I

Methods for verification of long-term material stability and mechanical integrity of special poison material utilized for neutron absorption should include actual tests.

Inservice surveillance requirements for the fuel racks and the.

poison material, if applicable, are dependent on specific design features.

These features will be reviewed on a case by case basis to determine the type and the extent of inservice surveil-lance necessary to assure long-term safety and integrity of the pool and the fuel rack system.

i e

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1 V.

COST / BENEFIT ASSESSMENT i

l.

Following is a list of information needed for the environmental Cost / Benefit Assessment:

1 1.1 What are the specific needs that require increased storage capacity in the, spent fuel pool (SFP)?

Include in the response:

(a) status of contractual arrangements, if any, with fuel-l storage or fuel-reprocessing facilities, 1

(b) proposed refueling schedule, including the expected number of fuel assemblies that will be transferred into the SFP at each refueling until the total existing capacity is reached, i

(c) number of spent fuel assemblies presently stored in the

SFP, (d) control rod assemblies or other components stored in the SFP, and i

(e) the aeditional time period that spent fuel assemblies would be stcred onsite as a result of the proposed expansion, and I

b (f) the estimated date that the SFP will be filled with the

{

proposed increase in storage capacity.

1.2 Discuss the total construction associated with the proposed modification, including engineering, capital costs (direct and indirect) and. allowances for funds used during construction.

1.3 Discuss the alternative to increasing the storage capacity of the SFP.

The alternatives considered should include:

(a) shipment to a fuel reprocessing facility (if available),

(b) shipment to an indeper; dent spent fuel storage facility, (c) shipment to another reactor site, (d) shutting down the reactor.

The discussion of options (a), (b) and (c) should include a cost comparison in terms of dollars per Kgu stored or cost per assembly.

i The discussion of ('d) should include the cost for providing replacement power either from within or outside the licensee's generating system..

V-1 1

.)

e 1.4 Discuss whether the commitment of materini resources (e.g.,

stainless steel, boral, B,C, etc.) would tend to significantly foreclose the alternatives available with respect to any other licensing actions designed to ameliorate a possible shortage of spent fuel storage capacity.

Describe the material resources that would be consumed by the proposed modification.

a 1.5 Discuss the additional heat load and the anticipated maximum

{

temperature of water in the SFP which would result from the proposed expansion, the resulting increase in evaporation rates, the additional heat load on component and/or plant cool'ing water systems and whether there will be any significant increase in j

the amount of heat released to the environment.

i V.2. RADIOLOGICAL EVALUATION 2.

Following is a list of information needed for radiological evaluation:

i 2.1 The present annual quantity of solid radioactive wastes gen-erated by the SFP purification system.

Discuss the expected increase in solid wastes which wi.11 result from the expansion of the capacity of the SFP.

I I_)

2.2. Data regarding krypton-85 measured from the fuel building ven-tilation system by' year for the last two years.

If data are not i

available from the fuel building vs.itilation system, provide this data for the ventilation release which includes this system.

2.3 The increases in the doses to personnel from radionuclides con-centrations in the SFP due to the expansion of the capacity of the SFP, including the following:

(a) Provide a table showing the most recent gamma isotopic analysis of SFP water identifying the principal radio-nuclides and their respective concentrations.

1 (b) The r:odels used to determine the external dose equivalent rate from these radionuclides.

Consider the dose equiva-lent rate at some distance above the center and edge of the j

pool respectively.

(Use relevant experience if necessary).

J (c) A table of recent analysis performed to determine the principal airborne radionuclides and their respective concentrations in the SFP area.

(d) The model and assumptions used to determine the increase, if any, in dose rate from the radionuclides identified in

{

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(c) above in the SFP area and at the site boundary.

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(e) An estim te of the ir. crease in the annual man-rem burden i

I from mor6 frequent changing of the demineralized resin and filter media.

-(f) The buildup of crud (e.g., ssCo, 80Co) along the sides of the pool and the removal methods that will be used to reduce radiation levels at the pool edge to as icw as j

reasonably achievable.

(g) The expected total man-rem to be received by personnel i,

occupying the fuel pool area based o.n all opcrations in

~

that area including the doses rer,ulting from (e) and (f) hbove.

A discussion of the radiation protection program as it affects (a) through (g) should be provided.

2.4 Indicate the weight of the present spent fuel racks that will be removed frora the SFP due to the modification and discuss what will be done with these racks.

V.3 ACCIDENT EVALUATION 3.1 The accident review shall consider:

(a) cask drop /tip analysis, and (b) evaluation pf the overhead handling system w'ith respect to Regulatory Guide 1.104.

l 3.2 If the accident aspects of review do not establish acceptability with respect to either (a) or (b) above, then technical specifica-tions may be required that. prohibit cask movement in, the spent fuel building.

3.3 If the accident review does not establish acceptability with respect to (b) above, then technical specifications may be i

required that:

j l

(,1) define cask transfer path including control of I

(a) cask height during transfer, and (b) cask lueral position during transfer (2) indicate the minimum age of fuel in pool sections during movement of heavy loads n' ear the pool.

In special cases evaluation of consequences-limiting engineered safety features such as isolation systems and filter systems may j

be required.

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I 3.4 If the cask drop /tip analysis as in 3.1(a) above is promised for l

future submittal, the staff evaluation will include a conclusion on the feasibility of a specification of minimum age of fuel based on previous evaluations.

c 3.5 The maximum weight of loads which may be transported over spent fuel may not be substantially in excess of that of a single fuel c

J assembly.

A technical specification will be required to this effect.

I 3.6 Conclusions that determination of previous Safety Evaluation Reports and Final Environmental Statements have not changed significantly or impacts are not significant are made so that a negative declaration with an Environmental Impact Appraisal i

(rather than a Draft and Final Environmental Statement) can be 1

issued.

This will involve checking realistic as well as con-servative accident analyses.

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VI.

REFERENCES 1

l 1.

Regulatory Guides l

Design Objectives for Light Water Reactor Spent Fuel 1.13 Storage facilities at Nuclear Power Stations 1

j 1.29 Seismic Design Classification Design Response Spectra for Seismic Design of 5'aclear 1.60 i

Power Plants Damping Values for Seismic Design of Nuclear Power 1.61 Plants Design Basis Tornado for Nuclear Power Plants 1.76 Combining Modal Responses and Spatial Components in 1.92 Seismic Response Analysis 1.104 -

Overhead Crane Handling Systems for Nuclear Power Plants j

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h q'

1.124 -

Design Limits and Loading Combinations for Class 1 Linear-Type Components Supports 2.

Standard Review Plan 3.7 Seismic Design 3.8.4 -

Other Category I Structures 9.1 Fuel Storage and Handling 9.5.1 -

Fire Protection System 3.

Industry Codes and Standards 1.

American Society of Mechanical Engineers, Boiler and Pres--

sure Vessel Code Section III, Di. vision 1 2.

American Institute of Steel Construction Specifications 3.

American National Standards Institute, H210-76 4.

American Society of Civil Engineers, Suggested Specification for Structures of Aluminium Alloys 6061-T6 and 6067-T6 I

I VI-1

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5.

The Aluminium Associatfan, Specification for Aluminium Structures i

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u

.e ENCLOSURE NO. 2 0

u6ncts G

recor wet, nied Gen'eni%n~e'TJvii -

.)(AGNS) prorosed. plant in ' Barnwen, d

g

",',* South Carounn, is tmder construction 1

• and is the subject of pending proceedings

before the Commission regarding the woontinuation, modiScation or suspension Jof the co6struction permit froca an en '

'vironmental protection standpoint, and

the possible issuance of an operating 11

• cense (docket no.54-332), as weu as a

• , related matter (docket no. 'l0-1729). *..,

1. 'On May 8,1975, the Nuclear Regula-

' tory Commission published a notice id

• the FEDeRA!.'RrctsrEn sett!ng forth Its

1. provisional views that, subject 'to conn

• sideration of comments, -(1). a' costa e

beneSt analysis of alternative safernards

. ~.

• SPEST FUEL STORAGE ' ". * * * ' programs should be prepared and +ct l

J

• forth in draft and Anal environmental

,ilotent To Prepa're-* Generic Environmental

• impact statements before a Coc tnission impact Statement on Handling and decision is reached on wide y. Storare of Spent LJght Water Power Re-mixed oxide (recycle pluto'clum) fuels

.ector Fuel,

.a r

in Ught vater nuclear power reactors.-

I

.l.' From the earis-days 'of the n'uclear (2) there should be no additionally 1 censes -

jpower industry in this country, electric,11tht water, nuclear power' reactors ex.'

granted for.use of mixed oxide fuel in sutilities planning to ennstruct and oper--(

fate.Ught water nuclear power reacton cept for experimental purposes, (3) with.

contemplated that the used or spent fuel ' respect to 11ght we.ter nuclear power *'

discharged frpm the reactors wodd be reactorfue! cycle activities which depend" chemicaUy reprocessed to' recover the for their justtScation on wide-somle use remaininF quantities of Sssile and ferr of mixed oxide fuel in 1!sht water nu.

l

. tile materials (uranium and plutonium), clear power reactors, there should be no'

/and that -the materials so recovered additional 11 censes granted which vot.2d

.>would be recycled back into fresh nactor. foreclose future safeguards options or-Jfuel. It,was contemplated by the nua. lear-result in unnecessaryfgrandfathering",

42ndustry that spent fuel would be dis-,and (4) the granting of Ucenses would 4 charged periodicauy from operating re...not be pre 91uded for fuel cycle activities d

0 2 actors, stored in ensite fuel storage pools Nor experimental and/or technical feasf%

i v

rfor a period of time to permit decay of bhity purposes.. W.U.,w y.'p 5,*,g

' radioactive materials contained within M"In 11gnt.'of the stahts of the*.three'.

Ithe Tuel and to ; cool l and periodicaUy** planned commercial reprodeksing plants.

rahipped offsite for reprocessing. Typical-lin the United States, as outlined above,"

.)y, space was provided in onsite storage the earliest that spent fuel reprocessing pools for about one and ondthird nu,couli! begin on a commercial basis.if au: g clear reactor cores. Assmnics a four-year thorized, wcMd be late 1978. This as-;

• reactor fuel" reload cycle, such onsite sumes that ; the ~ pending licensing

' storage pools were planned to hold an proceedings are coropleted and IIcenses average of one year's discharge with suf-issued by this date. HoweverJthe ipent

- Scient remaining capacity to hold a com-fuel pools at a number ~of reactors may.

piete core should unloading of all of f,he soon be BUed,.and stiU other reactors fuel from' the reactor be necessary or wiu have their pools Sued before the~end

. desirable because of operational dif5cul 'of 1978. Accordingly/even if Emited re -

. ties. Under normal operating conditions, processing should begin inlate 1976, there

~?mn averase of Sve years' discharge could would stiU. be a.sbortars'in spent. fuel

,'fuled..be~ accommodated before the pools were, storage capacity.f,,.;Gfp*LJQ l

' The caisting pools at'the 'OE.and.

4~

U. Persons' planning to. conduct commer% -. NFS reprocessing plants.have sogne re P

,r.!al* reprocessor.g of spent

• reactor fuels hatning marginal Ucensed.-storage 'ca '?

lprovided'auf5clent storage, capacity for pacity which may be.able.to acconuno ':

the spent fuels at their facilities to r.Dow ' date the fuel discharges from. sotne some operational Sexibliity. Typictuy. 'reactorsl any increases planned at these '-

l space has been provided or planned for. plants may not be suf5clent fo-industry'"

• several' spent fuel core reloads. Three. fin the future. Consequently, there.is the

' commercial' reprocessing. plants have

'been planned for operation in the United. possibuity of a future shortage in-11-.

-censed spent fuel capacity regardless of States. The only such plant that'has actuaUy operated, Nuclear Fuel Services. the outcome of the". proceedings on the (NPS) plant at West VaUey, New York., May 8th notice.N + @ k.r.a k r'.* 5 was shut down in 1972 for extensive.. T!d CommissioE has no't protnulgate'd' i

4

, alterations and expansion. There is a any regulatJon which specL5es a given-pending proceedicg before the Nuclear mise for on site reactor spent fuel pools; however, proposals by reactor licensees Regulato-y Commission (Committlon). to signincanUy change the manner of on NTS's appuca, tion for a permit to spent fuel storage or spent fuel pool si:e construct these alterations' and expan-alon. (docket no. 50-201). The 'second would be subject to licensing review by

' plant. General Electric Company's Midg the Commission. In the event that a west Fuel Tlecovery Plant at Morris,21-particular on s!te spent fuel pool should 11nols, has never operated and is in a

• become filled, and no alternative form d-mm?iened condition. The thin $ of spent fuel storage ;could be found.

4-n FgottAt REGl$f te, VOL. 40, No.180--TUf1 DAY, StPffMSEg F6,1973

i l

42 % 2 50T1GS I

~

the reactor would be e$ctually forud ' ment as o aultabl2 vthtclifrr such ta.Commitoon had tro baalc objeElises'IrN

. to snut down and *sbre" the last spent ' examination. Notice is hereby given that mind: on the one hand. the generic im--

!rearlor iucj in the reactor pressure ves

a generic environmer tal impact statc pact state. ment sh se.l..Whue no serious *dverse conse " ment 'on the handling and storage of MScation. Jor. a falt accompil; on. thei fcuences to'the r>ubuc health and safety, spent Ught water po,t er reactor fuels vul other hsud, the pubMe interest consid.; I l

the ' common. defense and recurity, or t-u prepared by the Comm!ssion...The erstions associated with such a deferral.

. the environment would Ukdy res:lt, the siste: Dent wS1 focus on, the time period should be carefully weJghed. The Comn

'ste. tor shutdown would, of course, re-betvaen now and the m!d 1980's andy111 rnis:f on has concluded that there should?

',inove the plant from servicemand this in, add.ress:

- ^

be no such general de!erTA and tpat

• Ltro could adversely idect the electne - (1) *ne magnitude of the-poutble these related licensing actions may coo-1 Sumity's abLuty to meet e.lectrica.1 energy abonare of spent fuel storage capacity tinue during. the period required 100

~.taedL ar iorce the utility to ope, ate otber-

'(2) The alterr.atives for dealing with. preparation of. the. generic stdement.:

• lants that a.re less economical to operate. the problene including..but not neces-subject to certab cond!tions"In reach-p or which have gTester enurcamentat trn-sarily limited to: -

C... //-.ing this conclustoc. the Commission has*

i

• pact.'and thereby adversely afect the. (a) Permitting the expansion of vent consfdered the foucwing specif.c factarr.:-

(1).It is likely that each individual U '

's. There appear to be a number of bos.gfuel 5togage espacity at power reactors;

'pubuc interic.st.

(b) Percitting the expansion of spent. cenring action of this' type would base; "ble alternauves for increasing. spent fuel storage capacity at reprocessing a tullity thatis independent of the util!ty feel stortge capacity including, arr.ong plants;

.of.other licensing actions t.f f.his' type:'

~

other thLors. Increasha the storaAe ca-(c) Licensing of. independent spent. (2) It is not iltely that the taxing of

~

any particular }1 censing actics cf this 1

pacity at present reactor altes, and con " fuel storage faciuties; struction of independent spent fuel

'. (d) Storage' of spent fuel from, one type during the time frame under cen-storage facilities. The shortage of spent, or more reactors at.the stcrase pools of sideration wodd constitute a commit;.

rfuel storage capacity sill occur at indl.other reactors:

ment of resources

• that' would tend to N1 dual reactors.. and the ComNutm... (eLordering that ge.nuation of spent.signiacantir forecJose 'the alternsthes-

%could adequately address the issues Ina.Tuel (. reactor operation.), be stopped er 'avaDable with respect to any other in-voired' on 's case-by-case " basis. within hestricted;.'

, ~..-' F.2' 'dh'idual 15 sensing actf en 'of this type; f.;ii Ithe context of individual licensing re-(3) A cost-beneSt anal,tns of the al

(3) Itis lhely that any enMron= ental

7 views Indeed, the Coelnion has not, "ternatives listed in (2), along with any impacts a.ssociated with any individual'

~

• to date..found it necessary,'in the dis-other reasonably -fess!ble altematives. licensing action of this type would be-charge of its beensing and related regu-

~1 story functions. it.) develop any overa.U -includhg:

such that they could adecuately be ad."

(a) Impacts on. public health.and dressed within the ecnuxt of the indi ~-

program of action to deal with the prob. safety and. the. common defense.and "vidual license application trithout over-h riem.The Commtulon does. however.have. sw.trity; i.

W.. 9J

..D looktg any cumulative environmental' DthCdiscretion to deal with issues of this..(b) Environmental. social, and'eco-impacts; ' s Ukely th'at+an[;Ehnl8

r. 3:.;"T.C

• ?

Itype on-a generic basis through the ex-nom!

.}ercise of its.rulemaking author.*fy ud/ '.,,. (c)9 costs and benefits;. '".,,.i..**

~ i

'(4y It i i

Cc=mitments of retou'ces:

."?.

or the lasuance, of a generic" environ.

-(d) Implications regarding options.. issues that may'arise in t.he course of a*

review of an Individual Ucense applicar g(mental" impact ' statement. Rulemking available for the intermediate and-long-4

{ proceedings and/or the issuance of a ' term storage of nuclear waste ma*erials; ' tio text; and.. '-...

v . 7' ~

• A deferrai'cEe'v~cre're'strdtic2cI generic environmental impact statement ~ ' (e) Relationship between local short-(5) pnfght.as appropriate, serve s.s the con -term uses of toe environment and long* licenst::g actions of this type would red fer the promulgation of tnere. de term productivity;r,r,,, -

sult in substantial harm 'to the public' l

) fin!tive criteria regarding size and de.

(4) The hnpacts of possible add!tional interest. As indicated, such a restriction-

sign of spent fuel bools and/or the 11-transportation of spent fuel that may or deferral could result in reactor t.hutU

-censing of independent spent fuel htorage rf acilities, and for consideration of pes-c.;be required should one ot me,re of the ' downs'as existing spent fuel poo!.s become '

faible revision of the fuel cycle environ. alternatives be adoptedr

. fiUed. It nor appesrs that the spen *

. (5).More definitive standards and cr!- fuel pools of'as many as ten reacto:2 teria to govern the licenstag of one or

• could.be Elled by m!d 1973. These te:f, 3

$ in t Iad !

a nt

? storage. and attendant' transportation..more of the alternatives for d,ealing with reactors represent a total of about 6 mils

.1100 Mowatts of electrical energy gen-Also, the poss~ble imputations of in-the problem; and.,

.x.

~~

18) Possible arned.nents to M;CE. Wu capa@. m nmoral of these; Scnased rpent fuel storage on the options n from she could reduce the b.arnilable for intermediate and long term {51.2e)., ;, ;,* 3,',Ty r.6;.,y*:'Q",. t ties service" rec gi s to a point whered storase of nuclear vaste matenals could ; If appropriate,ruiemakinst proceedings.

within'~ this. 'on items (5) and (8) listed above, or on..ttliable se-vice wou!d be in jeopardy, or, a-r 7...* r. b ctber issues related to.1.be he.ndlins and ; force the uti}f tiet to rely more heari

1. reatably be enmined k;coutent. d Pe..". eione groWofIntere'sted 6'rganiza'tjond*.'* storage of spent reactor,ftel,-win be i EtneratJon.that would.' impose ece ',

-(Natural ^ Resources N Defense

• Cocneil, - tiated on or aboi:t that time of issuance he MWn,on consumers and tre,

bSierra : Club,' and Businessnien for the t

of the drtit generic environmentoJ im pase entronmetalimpacts. J ygg Pact statem nt.

[$iT f,

E ubuc Interest) has requested the Com-P d

• Dnission to prepare a generic t.1vironmen ' *.Ihe C'ocnmkdo$hb'ald.given'carEf E1.'"'"

E act! n intended tG stneliorate a%.

0" ' *

• ""# 0 italimpact rtatement on the handling and c$on'ideration to 'the,que; tion the*.ber %y 3e s appspM MsWages {

s (storage of spent reactor fuel and related~ 11 censing actions intended to ameUorate capacity. dutt.this intirim. pededs. !

Ematters Get',cr to L. V. Gossiek frum a possible shortage of spent iuel storage WW Waceps W an Qcn--

e r ate =ent S M W ' J FAnthony Z. Roisman, da' ?d May 20,1975 capeity, including suct sedons as the..

' copy 'on file at the Commission's Pubife tumee og operating license amend."jaD or impact appraisal (M Cm (51.5 >

"'Docurnent Room,1"17 H ' Street,.NW, ments to permitincreases in the storen Jc0 taDored to me facts of me case.; j

~'

Washington, D.C.)

' ' WhDe the Commission betieves, as earl' capacity of reactor spent. fuel pools or hee the NCen's geral coneln,

.~

reproccasing plant spent-fuel atorage s ns respect, k me N fach, as..

4

11'er indicated, that the matter of spent. pools, or the limer of independent 8'

• I fuel storage capacity can adequately be spe.ot fuel storage f aciutfes, should be

" *[*gg gg ge g,,

addres:rd oc a esse-by case bas s withtn, deferred pending cornpletion of the ge. gt'ons, the five factors will be appUed /

d i

. the context of individual ifcensing re-neric environmental impact statement,

. views, it also beueves that, from the Such 'a deferrs! was requested in the weWd and balmed en the ecm.- !

standpotet of longer range poucy, this letter en behalf of Natural Resou tes. text of these statements or sprnisals in i j

h snarter can profitably be examined in a Defense Council. Sierra Club, and Busi. reaching Mcensmg %erminadons.,A,

\\s broader contert. It views the preparation nasmen for the PubDe Interest noted. Dated at Washington, D.C. this 10thi of a generic environmental 1:npac,t state-. sLbove. In considering this matier, the day of September 1975.

I

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8 e-Fiofut MCf 5 Tit, WOL do NO.1M-TuthoAY, StPTEMsts 1s,1975 1

1

_x 1

J

O Florida Power & Light Company 2

l I

Mr. Robert Lowenstein Esquire cc:

Lowenstein, Newman, Reis & Axelrad 1025 Connecticut Avenue, NW Sui +a 1214 Washington, D.C.

20036 Environmental & Urban Affairs Library Florida International University Miami, Florida 33199 a

l Mr. Norman A. Coll, Esquire Steel Hector and Davis 1400 Southeast First National Bank Building Miami, Florida 33131 Florida Power & Light Company 1

ATTN:

Mr. Henry Yaeger Plant Manager Turkey Point Plant P. O. Box 013100 i

Miami, Florida 33101

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M UNITED 5TATES j

b NUCLEAR REGULATORY OOMMIS$10N t

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WASHINGTON, D. C. 20656 January 18, 1979 To All Power Reactor Licensees Gentlemen:

Our letter of April 14, 1978, provided NRC Guidance entitled,

" Review and Acceptance of Spent Fuel Storage and Handling Applications." Enclosed are modifications to this document for your information and use.

These involve pages IV-5 and IV-6 of the document and comprise modified rationale and corrections.

Sincerely, l.

i 7' l w,,ak L

Brian K. Grimes, Assistant Director

,\\el for Engineering and Projects s

i Division of Operating Reactors k S.,,

Enclosure:

Pages IV-5 and IV-6 cc w/ enclosure:

Service List da...,

k 4

[()

In order tc determine the flexibility of the pool wall it is acceptable for the licensee to use equivalent mass and stiffness properties obtained from calculations similar to those described in

" Introduction to Structural Dynamics" by J. M. Biggs published by McGraw Hill Book Company. Should the fundamental frequency of the pool wall model be higher than or equal to 33 Hertz, it may be cssumed that the response of the pool wall and the corres-ponding lateral support to the new rack system are identical to those of the base slab, for which appropriate floor response spectra or ground response spectra may already exist.

(6) Structural Acceptance Criteria When AISC Code procedures are adopted, the structural acceptance criteria are those given in Section 3.8.4.II.5 of the Standard i

Review Plan for steel and concrete structures.

For stainless steel the acceptance criteria expressed es a percentage of yield stress should satisfy Section 3.8.4.11.5 of the Standard Review 3,

Plan. When subsection NF,Section III, of the ASME B&PV Code is y..

used for the racks, the structural acceptance criteria are those f,

given in the Table below. When buckling loads are considered in the I

design, the structural acceptance criteria shall be limited by the a.,

requirements of Appendix XVII-2110(b) of the ASME Boiler and Pressure Vessel Code.

tb For impact loading the ductility ratios utilized to absorb kinetic d {'

energy in the tensile, flexural, compressive, and shearing modes should be quantified. When considering the effects of seismic hm.,,

loads, factors of safety against gross sliding and overturning of racks and rack modules under all probable service conditions shall 4

be in accordance with the Section 3.8.5.II-5 of the Standard Review Plan. This position on factors of safety against sliding and tilting f'

need not be met provided any one of the following conditions is met:

(a) it can be shown by detailed nonlinear dynamic analyses that 7

the amplitudes of sliding motion are minimal, and impact y.,m; j between adjacent rack modules or between a rack module and the pool walls is prevented provided that the factors of D

safety against tilting are within the values permitted by Section 3.8.5.II.5 of the Standard Review Plan.

(b) it can be shown that any sliding and tilting motion will be contained within suitable geometric constraints such as thermal clearances, and that any impact due to the clear-ances is incorporated.

(7) Materials, Quality Control, and Special Construction Techniques:

The materials, quality control procedures, and any special con-struction techniques should be described. The sequence of in-

[]

stallation of the new fuel racks, and a description of the pre-V cautions to be taken to prevent damage to the stored fuel during i

IV-5

~,

(D

\\_.)

i TABLE Load Combination Elastic Analysis Acceptance Limit Normal ifmits of NF 3231.la D+L D+L+E Normal limits of NF 3231.la c

D + L + To 7.T.

Lesser of 2Sy or Su stress range r,-

g-D + L + To i E j' Lesser of 2Sy or Su stress range D + L + Ta + E Lesser of 2Sy or Su stress range D + L + Ta + El Faulted condition limits of NF 3231.1c Limit Analysis 2

1.7 (D + L)

Limits of XVII-4000 of Appendix XVII

5.,

~

of ASME Code Section III

(

1.7 (D + L + E) k.

~ ~ ~ ~ ~ ~

i O 1.3 (D + L + To) -

1.3 (D + L + E + To) 1.1 (D + L + Ta,+ E)

Notes:

1.

The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan where each term

?

is defined except for Ta which is defined as the highest temperature associated with the postulated abnormal design conditions.

j 2.

Deformation limits specified by the Design Specification limits shall be satisfied, and such deformation limits w

should preclude damage to the fuel assemblies.

3.

The provisions of NF 3231.1 shall be amended by the requirements of the paragraphs c.2, 3, and 4 of the Regulatory Guide 1.124 entitled " Design Limits and Load Combinations for Class 1 Linear-Type Component Supports."

y d

IV-6

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_