ML17150A128

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Ltr to All Power Reactor Licensees Forwarding Guidance on Spent Fuel Pool Mod Re Review & Acceptance of Spent Fuel Storage & Handling Applications.Certificate of Svc Encl
ML17150A128
Person / Time
Site: Dresden  
Issue date: 04/14/1978
From: Grimes B
Office of Nuclear Reactor Regulation
To:
AFFILIATION NOT ASSIGNED
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NUDOCS 8102060391
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To All Power Reactor Licensees Gentlemen:

EIiclosed for your inforIIIation and possible future use is the HRC guidance on spent fuel pool rrrnNficat;ons, ent'itled "Review and Acceptance of Spent Fuel Storage and Handling Applications".

This document provides (i) additional gr idarrue for the type and extent of information needed by the NRC Staff to perforIII the review of licensee proposed modifications of an operating reactOr Spent. fi.'el storage poo; and {2) the acceptance criteria to be used by the NRC Stat'f'n aulfrorizing such rrodifications.

This includes the information needed to make the findings called for by the Commission in the Federal Register Notice dated September

'l6, 1S15 (copy enclosed) with regard to authorization of fuel pool modifications prior to the completinn of fhe Generic EnvironIIIental Impact Statement, "Handling and Storage of Spent Fuel from Light Rater Nuclear power Reacturs".

Sincerely.

The overall ri~sign objectives of a fuel storage facility at a reactor complex are gover'ned by various Regulatory GuMes, the 5t ndard Review Plan

{NURE6-75i087), and various industry standards.

This guidance provides a compilation in a single document of the pertinent portion~

n these app1icable references that are needed in addressing spent fuel poo'i modif'ications.

No additional regulatory requirer.rents Rro 1mpoeoct nr implied by this document V

Sased on a review of license applications to date requesting authorization to incroase spent fuel storage capacity.

the staff has had to request additional information that could have been included in an adequately do~.umented initial subIIIittal. If in the future you find it necessary 4o opply for authoÃzat)on tn IIndify onsite soent fuel storage

capacity, the enclosed guidance provides the necessary information and acceptance criteria utilized by the NRC staff in evaluating these app1icntions.

Providing the information needed to evaluate the matters covered by this document would likely avoid the necessity for HRC questions and thus significantly shorten the time required to process a fur:1 pool modification amAndment.

Enclosures:

Brian K. Grimes, Assistant Director for Engineering and Projects Division nf Operating Reactors

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ATTACHMENT C UNITED STATES NUCLEAR REGULATOnY COM~.11SSION WASHINGTON, 0. C. 20SSS April 14, 1978 To All Power Reactor Licensees Gentlemen:

Enclosed for yo~r information and possible future use is the NRC guidance on spent fuel pool modifications, entitled "Review and Acceptance of Spent Fuel Storage and Handling Applications". This docu~~nt provides (1) additional guidance for the type and extent of information needed by the NP.C Staff to perform 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 authorizinQ such modif~c.ntions. Th1s includes the information needed to make the findings called for by the Corrrnission in the Federal Register :~otice. da~cd September 16~ 1975 (copy enc1osed}

with regard to authorization of fuel pool modifications prior to the completion of the Generic Environ~ntai Impact Staten~nt, "Handling and Storage of Spent Fuel* fro::i L isht Water Nuclear Po~*1er Reactors".

The overall design objectives of a fuel storage facility at a reactor complex are governed by various Regulatory Guides, the Stand~rd Review Plan (NUREG-75/087), and various industry standards. ihis guidance provides a compilation in e single docurrcnt of the pertinr.nt portions of these applicable references that arc needed in addressing spent fuel pool modifications.

No additional regulatory requirf:fuents are imposed or im?lied by this document.

Based on a review of license applications to date requesting authorization to increase spent fuel storage cap~city, the staff has had to request additional infonr.ation that could have been included in an adequately documented initial submittal. If in the future ycu find it n:?cessery to apply for authorization to modify onsite spent fue1 stor~g!

capacity, the enclosed guidance provides the necessary infor.nation and acceptance criteria utilized by the NRC staff in ev~luat1ng !hese applications.

Providing th~ infor;r.ation needi?d to evaluate the matters covered by this documc~t would likely avoid th~ necessity for NRC quest~ons and thl!s s1snif1cantly short~n the tirr.c required to process a fuel pool mod~ficaticn a~ndment.

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Enclosures:

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NRC Guidance

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Brian K.

Gri~~s. Assistant Director for Eng;neering !nd Projects D1vi~ion of Operating Reactors

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BACKGROUND ENCLOSURE NO. 1 OT POSITION FOR REVIEW AND ACCEPTANCE OF SPENT FUEL STORAGE AND HANDLING APPLICATIONS Prior to 1975, low density spent fuel storage racks were designed with 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 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 prepare a ceneric environmental impact statement on h~ndling and storage of spent fuel from light water power reactors.

In this notice, the Commission also announced its conclusion that it would not 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 im?act statement.

The Commission directed that in the consideration of any such proposed licensing action, an environmental impact statement or environmental impact appraisal sha11 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, t.he Standard R~view Plan, and industry standards which are listed in the reference section.

Based on the reviews of such applications tp date it is obvious that the staff had to request additional information that 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 appllcable.

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

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

Nuclear and thermal-hydraulic aspects of the review include the poten-tial for inadvertant criticality in the normal storage and handling of t.he spent fuel, and the consequences of credible accidents with respect to criticality and the ability of the* heat removal system to maintain sufficient cooling.

Mechanical, materiai 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 enviro~mental aspects of the review concern the increased thermal and radiological releases from the facility under normal as \\*.'ell as accident conditions, the occupational radiation exposures, the genera-tion of radioactive waste, the need for expansion, the co~~itment of material and nonmaterial resources, realistic accidents, alternatives

  • to the proposed action and the cost-benefit palance.

The.information related to nuclear and thermal-hydraulic type of analyses is discussed in Section Ill.

The mechanical, material, and structural related aspects of ~nforma tion are discussed in Section IV.

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

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l III. ~UCLEAR AND THER~AL-HYDRAULIC CONSIDERATIONS I *

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Neutron Multiplication Factor

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  • 2 To include all credible conditions, the licensee shall calculate the effective neutron multiplication factor, k ff' in the f~el storage pool under the following sets of assumeo cor_:jitions:

Normal Storage

a.

The racks shall be designed to contain the most reactive fuel authorized to be stored in the facility without any control rods or any noncontained* burnable poison and the fuel shall be assumed to be at the most reactive point in its 1 ife.

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b.

The moderator shall be assumed to be pure ~ater at the temperature within the fuel pool limits which yields the largest reactivity.

c.

The array shall be assumed to be infinite in lateral extent or to be surrounded by an infinitely thick water reflector and thick concrete,~~ as appropriate to the design.

d.

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

e.

Credit may be taken for the neutron absorption in structural materials and in solid materials added sp~cifically f6r neutron absorption, provided a means of inspection is estab-lished (refer to Section 1.5).

Postulated Accidents The double contingency princtple of ANSI N 16.1-1975 shall be

  • applied.. It shall reQuire two unlikely, independent, concurrent events to produce a criticality accident.

Rea~istic initial conditions (e.g., the presence of soluble boron) may be assumed for the fuel pool and fuel assemblies.

The 11111Noncontained 11 burnable poison i's that which is not an integral part of the fue1 a?semb1y.

    • It should be noted that under certain conditions concrete may be a more effective reflector than water.
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,,.3,e postulated accidents shall include: (l) dropping of a fuel element on top of the racks and any other achievable abnormal location of a fuel assembly in the pool; (2) a dropping or tip- --*

ping of the fuel cask or other heavy objects into the fuel pool; (3) effect of tornado or earthquake on the deformation and rela-tive position of the fuel racks; and (4) loss of all cooling syste~s or flow under the accident conditions, unless the cooling system is single failure proof.

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 configurations shall be calculated to render improbable the

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cancellation of error" in the calculations.

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

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

A calculational bias, including the effect of wide spacing betwoen asserr;blies shall be determined frorn the co:::parison bct\\l*een calcu-lation and experiment.

A calculation uncertainity shall be determined such thc:t the true multiplication factor will be less than the calculated value with a 95 percent probability at a 95

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percent confidence level.

The total uncertainity factor on keff shall be obtained by a statistical combination of the culcula-tional and mechanical uncertainties.

The k value for the racks shall be obtained by summing the calco)~ted value, the

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calculational bias, and the total uncertainty.

  • 1.4 Rack Modification For modification to existing racks in operating reactors, the following information should be provided in order to expedite the review:

(a) The overall 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 nominal storage lattice cell; (b)* For H 0 + stainless steel flux trap lattices; the nominal thickfiess 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 for this stainless steel; (c) Also, for the H70 + stainless steel flux trap lattices, the change of the calculated neutron multiplication factor of I

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infinite1y long fue~ assemblies in infinite1y large arrays in the storage rack (i.e., the ~of the nominal fuel storage lattice cell and the changed~) for:

(l) A change in fuel loading in grams of U23 S, or equiva-lent, per axial centimeter of fuel assembly where it is assumed that this change is made by increasing the enrichment of the u23s; and, (2) A chnnge in the thickness of stainless steel* in the storage racks assuming that a decrease in stainless steel thickness is taken up by an increase in water thickness and vice versa; (d) For lattices which use boron or other strong neutron absorb-ers provide:

(l) The effective areal density of the boron-ten atoms (i.e., 810 ato:ns/cm2 or the equivalent number of boron-ten atoms for other neutron absorbers) between fuel assemblies.

(2) Similar to Item C, above; provide the sensitivity of the storage lattice cell ~ to: -

(a) The fuel i*oading in grams of u2 3s, or equivalent,*

}C per axial centimeter of fuel as~cmbly~

(b) The storage lattice pitch; and,.,,,,

(c) The areal density of the boron-ten atoms between fuel assemblies.

1.5 Acceptance Criteria for Criticality The neutron multiplication factor in spent fuel pools shall be less than or equal.to 0.95, including all uncertainties, under all conditicms (l) For those facilities which employ a strong neutron absorbing 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 within 95 percent confidence limits that there is a suffi-cient amount of neutron absorber in the racks to maintain the neutron multiplication factor at or below 0.95.

In addition, coupon or other type of surveillance testing shall be performed on a statistically acceptable san1ple size on a

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periodic basis throughout the life of the racks to ~erify the continued presence of a sufficient amount of neutron absorber in the racks to maintain the neutron multiplication factor at or below 0.95.

(2) Decay Heat Calculations for the Spent Fuel The calculations for the amount of thermal energy that will have to be remove~ by the spent fuel pool cooling syste~

shall be rn~de in accord?nce with Branch Technical Position APCSB 9-2 entitled, "Residual Decay Energy for Light ~ater Reactors for Long Term Cooling.

11 This Branch Technical Position is part of the Standard Review Plan {NUREG 75/087).

(3) Thermal-Hydraulic Analyses for Spent Fuel Cooling Conservative methods should be used to calculate the maximum fuel temperature and the increase in temperature of the water in the pool.

Ths._r:iaximum void fraction in the fuel assembly and between fuel assembl1es should also be calculated.

Ordinarily, in order not to exceed the design heat load for the spent fuel cooling ~ystem it will be necessary to do a certain ~mount of cooling in the reactor vessel after reactor shutdown prior to ~oving fuel assemblies into the spent fuel pool.

The bases for the analyses should include the estab-lished cooling times for both the usual refueling case and the full core off load case.

A potential for a large increase in the reactivity in an H 0

  • flux trap storage lattice e~ists if, somehow, the water is2 kept out or forced out of the space between the fuel assem-blies, conceivably by tra?ped air or steam.

For this reason, it is necessary to show that the design of the storage ~ack is such that th~s will not occur and that these spaces will always have water in them.

Also, in some cases, direct gam:na.heating of. the fuel storage cell walls and of the interce11 water may be significant. It is necessary to consider direct gamma heating of the fuel storage cell walls and of the intercell water to-show that boiling will not occur in the water channels between the fuel assemblies.

Under postulated accident conditions where all non-Category I spent fuel pool cooling systems become inoperative, it is necessary to show that there is an alternate method for cooling the spent pool water:

When this alternative method requires the installation of alternate components or signifi-cant physic~l alteration of the cooling system, the detailed steps shall be described, along with the time required for each.

Also, the average amount of water 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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{4)

Potential Fu~l and Rack Handling 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 fue 1 hand1 i ng procedures than that described in the Fina 1 -~* *.

Safety Analysis Report, the differences should be discussed.**

If potential fuel and rack handling accidents occur, the neutron multiplication fac~or in the fuel pool shall not exceed 0.95.

These postulated accidents shall not be* the cause of the loss of cooling for either the spent fuel or the reactor..

(5) Technical Specifications To insure against criticality, the following technical speci-fications are needed on fuel storage in high density racks:'.

l.

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

2.

The fuel loading (i.e., grams of uranium-235, or equivalent, per axial centimeter of assembly) in fuel assemblies that are to be loaded into the high density racks should be limited.

The nurnber of grarns of

  • uranium-235, or equivalent, put* in the plant's tech-nical specifications shall prC?clude criticality in the fuel poo 1.

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 specified spent fuel pool ternperatures, consideration shall be given to incorporating a technical specification limit on the pool water tempera-ture that would resolve the concerns described above.

For limiting values of pool water temperatures refer to ANSI-N210-1976 entitled, "Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations, except that the requirements of the Section 9.1.3.III.~.d of the Standard Review Plan is applicable for the maximum heat load with normal cooling systems in operation.

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MECHANlCAL, ~ATERIAL, ANO STRUCTURAL CONSIDERATIONS (1) Description of the Spent Fuc1 Poo1 and Racks Descriptive information inc1uding p1ans and sections showing the spent fuel pool in relation to other plant structures shall be provided in order to define the primary structural aspects and 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 s~ent fue1 assemblies in a safe configuration through all environmental and abnormal loadings, such as earthquake, and impact due to spent fuel cask drop, drop of a spent fuel assembly, 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 horizontal and the vertical supports to the sp~nt fuel racks should be provided indi-cating the methods of transferring the loads on the racks to the fuel pool wa 11 and the foundation s 1 ab.

A 11 gaps (clearance or expansion allowance) and s1iding contacts should be indicated.

The extent of interfacing between the new rack system and the o1d 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 and to the side walls of the pool such that the pool liner may be perforated, the provisions for avoiding leakage of radio-active 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.

Postuiated drop accidents must include a straight drop on the top of a rack, a straight drop through an individual cell all the ~ay to the b~ttom of the rack, and an inclined drop on the tcp of a rack.. In-tegrity of the racks and the fuel po~l auc to a postulated fuel handling accident is reviewed under the mechanical, material, and structural disciplines. Sketches and suffi-cient details of the fuel handling system should _be provided to facilitate this review.

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,G"deJ (2) Applicable Coces, Standards.and Specifications Construction materials should conform to Section III, Subsec-tion NF of the ASHE~ Code.

All Haterials should be selected to*

be compatible with the fuel pool envi_ronment 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 provision~ must be followed in entirety.

W~n the AISC specification procedures are adopted, the yield stress values for stainless steel base metal may be obtained from the Section III of the ASME B&PV Code, and the design stresses de-fined in the AISC sp~cifications as percer.tages of the yield stress may be used.

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

Other materials, design procedures, and fabrication techniques will be revie¥i*ed on a case by case basis.

(3) Seiscic 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 Section 3.7 of the Standard Review Plan.

The ground response spectra and damping values should correspond to Regulatory Guide l.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 tor a fuel pool supported by the reactor building, the design and analysis of the new r~ck syste~ may be performed by using either the existing input para~eters including the old dampi_ng values or new param-eters in accordance with Regulatory Guide 1.60 and 1.61.

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

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

~American Society of Mechanical Engineers Boiler and Pressure Vessel Codes, latest Edition.

ti:~American Ins.titute of Steel Construction, latest Edition.

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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 ~ertical and horizontal directions only, the same

  • 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) 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 damping is generally negligible.

Based upon the findings of this current study for a typical high density rack configuration, wave 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 Hass and Damping of Subr.1erged Structures (UCRL-52342),

11 by R. G. Dong.

1he recomr.iendations provided in'*

this report on the added cass effect provide an acceptable basis for.the staff review.

Increased damping due to submergence in water is not acceptabl~ without applicable test data and/or detailed analytical results.

Due to gaps between fuel assemblies and the walls of the guide 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 the kinetic energy of the fuel assembly.

The maximum velocity of the fuel assembly may be estimated to be the spectral velocity*

associated with the natural frequency of the submerged fuel assembly.

Loads thus generated should be considered for~local as well as overall effects on the walls of the rack and the sup-porting framework.

It should be demonstrated that the consequent loads on the fuel assembly do not lead to a damage of the fuel.

Loads generated from other postulated impact events may be accept-

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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.

Loads and Load Combinations:

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

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Temper-ature gradient across the rack structure due to differential heating effect between a full and an e~pty cell shoul~ 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, if applicable.

The specific loads and load combinations are acceptable if they are in conformity with the applicable portions of Section 3.8.4-Il.3 of the Standard Review Plan.

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

the methods used to incorporate any gaps between the support systems and gaps between the fuel bundles and the guide tubes; the methods used to lu~p the mass~s of the fuel bundles and the guide tubes; the ~ethods 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 r~ckS.

The design and analysis procedures in ec.cordance with Section 3.8.4-Il.4 of the Standard Rcvie'ft' Plan are acceptable.

The effect on gaps, sloshing water, and increase of effective cass and damping due to sub~crgence in water should be qunntified.

When pool walls are utilized to provide lateral restraint at higher elevations, a deterQination of the flexibility of the pool walls and the capability of the walls to sustain such loads should be provided.

If the pool walls are flexible (having a f~nda~ental frequency less than 33 Hertz), the floor response spectra corresponding to the lateral restraint point at the higher elevation are likely to be great~r than those a~ tne base of the pcol.. In such a case using the response spectr~~ approach, two separate analyses should be pcrforoed ts indicated below:

(a) A spectru~ 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 the lower and high~r elevation~; and, (b) A static analysis of the rack system by subjecting it to the.

maximum relative support displacement.

The resulting stresses from the two analyses above should be combined by the absolute sum method.

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In order to determine the flexibility of the pool wall it~*

acc~ptable for the licensee to use equivalent mass and sti fness prop!?-\\ties obtained from calculations similar to those* d~ cribed tn *

"lntro*quction to Structural Dynamics" by J. M. Biggs pu~.1ished by McGraw Hill Book Company.

Should the fundamental freqyency of the pool\\wall model be higher than or equal to 33 Her~~ *. it may be assumed that the response of the pool wall and th.e' corres-ponding laieral support to the new rack system are ~dentical to those of the base slab, for which appropriate floor response spectra or ground response spectra may already ex.:ist.

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Structural Acceptance Criteria

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When AISC Code procedures are adopted, the structural acceptance criteria are those given in Section 3.B.4.II.5 of the Standard Review Plan for steel and concrete structu~es. For stainless steel the acceptance criteria expressed as a percentage of yield stress should satisfy Section 3.8.4.ll..5 of the Standard Review Plan.

When subsection NF, Section 111: of the ASME B&PV Code is used for the racks, the structural acceptance criteria are those given in the Table below.

When buckling loads are considered in the design, the structural acceptance criteria shall be limited by the requirements of Appendix XVII-2110(b) of the ASME Boiler and P~essure Vessel Code.

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

When.~onsidering 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.S~II~S of the Standard Review Plan. *This position on factors of safety.against sliding and tilting need not be met provided any one of the following conditions is met:

(a) it can be show; by detailed nonlinear d~namic analyses that the amplitudes of sliding motion are minimal, and impact between adjacent rack modules or between a rack module and the pool walls is prevented provided that the factors of safety against tnting are within the values *permitted by Section 3.8.5.II.5 of the Standard Review Plan~

I

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i (b) it can'be shown that any sliding and tilting motfon wil1 be contained within suitable geometric constraints such as thermal clearances, and that any impact due to the clear-a~ces is incorporated.

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Materfal s, Quality Control, and Special Construction Techniq'u.es:

The,1~aterials, 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-cautions to be taken to prevent damage to the stored fuel during IV-5

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Load Combination Elastic Analysis 0 + L D + L + E 0 + l + To D + l + To + E D + l + Ta + E 0 + L + Ta + El Limit Analysis TABLE Accep ance Limit Normal limits of~ 3231.la Normal limits o~NF 3231.la I

lesser of 2Sy br Su stress range l'

lesser of*2Sy or Su stress range I*

/

Lesser1~f 2Sy or Su stress range I

Fault~d condition lim~ts of NF 3231.1 c 1.7 {D + L)

I

,I Limits of XVII-4000 of Appendix XVII I

of ASHE Code Section III 1.7 (D + L + E)

1. 3 ( D +* l + To) 1.3 {D + l + E +To) 1.1 (D + L +Ta+ E)

/

I Notes: 1. The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan where ea.ch term is defined.except for Ta which is defined as the highest temperature associated with the postulated abnormal design conditions.

2.

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

I

3.

The1provisions of NF 3231.1 shall be amended by the require~ents of the paragraphs c.2, 3, and 4 of the Regulatory Guide 1.124 entitled "Design Limits and Load

/ombinations for Class l Linear-Type Component Supports."

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_ _J the* construct ion phase should be provided.

Methods for strLic-*

tural qualification of special poison materials utilized to absorb neutron radiation should be described.

The material for the fuel rack is reviewed for compatibility inside the fuel pool environment.

The quality of the fuel pool water in terms of the pH value aAd the available chlorides, fluorides, boron, heavy meta 1 s *should be indicated so that the 1 ong-term i ntcgri ty of the rack structure, fuel assembly, and the pool liner can be evaluated.

Acceptance criteria for special materials such as poison materials

  • should be based u~on the results of the qualification program supported by test data and/or analytical procedures.
  • If connections between the rack and the pool liner are ~ade by welding, the welder as well as the welding procedure for the welding assembly shall be qualified in accordance with the appli-cable code.

c-,

If precipitation hardened stainless 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 addition, the surface film resulting from the heat treatment should be removed from each piece to assure adequate corrosion resistance. *

(8) Testing and Inscrvice Surveillance

~~ethods 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 ~nd 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.

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COST/DENEFIT ASSESSMENT

1.

Fo11owing is a 1ist of information needed for the environmental Cost/Benefit Assessment:

l. l 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-storage or fuel-reprocessing facilities, (b) proposed refueling schedu1e, including the expected number of fuel assemblies that will be transferred into the SFP at each refueling until the total existing capacity is reached, (c) number of spent fuel assemblies presently stored in the

SFP, (d) control rod assemblies or other components stored in the SFP, and (e) the additional time p~riod that spent fue1 assemblies would.

be stored onsite as a result of the proposed expansion, and (f) the estimated date that the SFP will be filled with the proposed increase in storage capacity.

l.2 *Discuss the total construction associated with the proposed modification, including ~ngineering, capitul 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 independent 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.

The discussion of (d) should include the cost 1or providing replacement power either from within or outside the licensee's generating system.

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1. 4 Discuss whether the com:r.i tment of material resources (e.g..,

stzinless steel, boral, B~C, etc.) would tend to significantly foreclo$e the alternatives available with respect to any other licensing actions designed to arn~liorate a pO$Sible shortage of spent fuel storage clp~city. Describe the material resources that would be consumed by the proposed modificati"on.

1.5 Discuss the additior1al 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 cooling water systems and wpethcr there wil 1 be any significant increase in the amount of heat released to the environment.

V.2. RADIOLOGICAL EVALUATION

2.

Following is a list of information needed for radiological evaluation:

2. 1 The present annual quantity of solid radioactive wastes gen-erated by the SFP purification system.

Discuss the expected incrca~c in $Olid wastes which will result from the expansion of the capacity of the SFP.

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

If data are not available from the fuel building ventilation system, provide this data for the ventilation release which includes this system.

2.3 The-increases in the doses to personnel from radionuclide 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.

(b) The models used to determine the external dose equivalent rate from these radionuclides.

Consider the dose equiva-lent rate at so~e distance above the center and edge of the pool respectively.

(Use relevant experience if necessary).

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

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(e)

  • (f)

(g)

An estimate of the increase in the annual man-rem burden from more frequent changing of the demineralizer resin and filter media.

The buildup of crud (e.g., ssco, 60Co) along the sides of

  • the pool and the removal methods that will be used to reduce radiation levels at the pool edge to as lo.w as reasonably achievable.
  • The expected total man-rem to be received by personnel occupying the fuel pool area based on all operations in that area including the doses resulting from (e) and (f) above.

A discussion of the r~diation 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 re~oved from 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 3.2 3.3 (b) evaluation of the overhead handling syste~ with respect to Regulatory Guide 1. 104 *.

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

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

(J) define cask transfer path including control of (a) cask height during transfer, and (b) cark lateral position during transfer (2) indicate the minimum age of fuel in pool secti~ns during movement of heavy loads near the pool.

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

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future submittal, the staff evaluation will include a conclusion*

on the feasibility of a specification of minimum age of fuel based on previous evaluations.

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

A technical specification will be required to this effect.

3.6 Conclusions that determination of previous Safety Evaluation Reports an~ Final Environmental Statements have not changed significantly or impacts are not signific2nt are made so that a negative declaration with an Environmental

!~pact Appraisal (rather than a Draft and Final Envircnmental Statement) can be issued.

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

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REFERENCES

1.

Regu1 atory Gui des 1.1~

Design Objectives for Light ~ater Reactor Sp~nt Fuel Storage Facilities at Nuclear Power Stations 1.29 -

Seismic Design Classification 1.60 Design Response Spectra for Seismic Design of Nuclear Power Plants 1.61 -

Damping Values for Seismic Design of Huc1ear Pa.-er Plants 1.76 -

Design Basis Tornado for Nuclear Power Plants

  • 1.92 -

Combining.Modal Re~ponses and Spatial Components in Seismic Response Analysis 1.104 -

Overhead Crane Handling Systems for Nuclear Power Plants

1. 124 -

Design Limits and Loading Col:'lbinations 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. l -

Fire Protection System

3.

Industry Codes.and Standards

1.

American Society of Hechanica1 Engineers, Boiler and Pres-sure Vessel Code Section Ill, Division 1

2.

American Institute of *steel Construction Specifications

3.

American National Standards Institute, N210-76

4.

American Society of Civil Engineers, Suggested Specification for Structures of Aluminium Alloys 6061-16 and 6067-16 VI*l

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The A1~~in1u~ Association, Specification for* Aluminium Structures VI-2

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ENCLOSURE rw. 2

~TICES

~l

  • ~*-:,..

-pli~~. Anl~ ~i:en.l Nucl~ r.cm:ec'

.* HAONS>

J;>~r>O"ee plr.:i~ 1n :aa..'"tl'"!!,

      • Sc>a~ Ca.ro!lr-"\\, 1.: UTlc:!e':'. ~::iJC!..!('tl

-. and l.s the su::>Je-.:t or J)en~r ;:>:-oeeecint'

~before

  • tho Cocnm~!:>n rr~~t:ir '9"1c

,.*.:conunua.tlon. inoC!!lc:a.Uo: or a~~:::ulon

  • '.or Cle corut.ructic:i pennlt fro~ u.:i en-*
  • Tirt>ru:ien~l J;>!"OU-:t!:>n suu:~~:.n~. Llld

~ t.be ~\\!tile 1.s.si:.i:.nce o! an open. t!r.i ll*

  • cense (dockc: Do. 50-!J:?>. n.s well &S a

. relat.ed ruAtW:r <docket no. '70-17:9>.

~-*.

    • On Ma>> 2, 1975, t.'le Nuclea: I'.~b-
tory Cc.:nm!.uion f\\llbl!.shed a not.ice Ui *

.".the :f'EDtP.\\:. Rtc:1Tu set:.!::c :'o:""..t: lLS

      • prov'l.slona.J vie~*s U-..at.. ~ub3ec~ to con-'
  • "a.lder:lt!on of comments,"<ll. a* cc.s~

. be.ne:lL anal.ysis o! a.l~mAUTe SJL!e~l"IU

  • -
  • _.**, ** SPENT FUEL STORAGE *. * *..
  • J>l't';r.u:'U.s!:lould be
  • P~~e<l. l.:ld aei
  • torl!I tn d.-'&!t L"ld fln~I eDraonme::it&.l
.tnlenl To Prepare-Ceneric Environmental
  • lmpnct s!.l~ment.s btfo:-t a COl:1!:l.i~lon Impact Stalemen: on Handlinr and decision 1.: re~hed on w!:!r~ ~ o:
~.Storar.e of Spent Licht W:iter ~wcr Re-mix~d o:tide ere-cycle plui.:>ruu:n > ru:.!3
'
actor Fuel.
    • ** ~* * **.
  • 1n lli;bt T:iter nuc:leu PJ*er r~:.0:-3,
  • 0 P'ro!D tbe ea:lT* dl\\ys o! the nuc:le:ir

<:> t..'lcre s~ould t-e n:> a.~c!.:t:O!\\AI liee:ui:s *

.~Powt: i!ldust..7 i.:l t.":.1: c::-~t:-r, elecln:

trantl:d !or uu o! m~"l'.ed o:t:~e !i:el !.::i

.*.utilities ~la.~ to c
c.::s:..-uc:t !lllC: o:xr- * -:.isr.: v.-:it.er. n::c:~r ;>e>o;;er re~~:s u- *

~~a.\\.e.Utllt ti:ctcr nuc:len: powrr fC!l.C:tOl'S te:>: !o: expe..""i:e:'lbl PU..-;>Os:S. (Jl 'tlMtll

."c6Dte::::i;:>la:.Cd ~!l.t t!le \\!.SeC: o: Sl>e!l: f1.1t.l rc;::;ec: LO light

~a:.e:. tl:.>cl~ ~er*

  • dhc:ha.rtcd fro:::i the re:ic:\\.C.':"S Yould be
  • renc:i.o: fu~ cr:!e ~c~h*!tl!S *l"J:b de:--.;d
  • c.beml~ly reproc:e:;sed to. re<:o\\*c: Ute 1or t..'leir jwt!!ic:a.:.:on C:l ~~e-set.Je \\:3C

.ftml\\.lning Qua.nll:.!C$ o! fis.s!le n.:>C: fer* o: Ql:W:Cd O:tiC:t fueJ* tn llt,.'l.: 'IVB.l.C:' nu-

.. t:le ma.t.cr1a!..s cu:an.:~ a.nd p:u~r.h:!ll>. *clur ix>11o*er re:ic::O:"S, t.'lere ~'lou.ld !:.e :c.o *

ntid \\ha.t \\.'le

!TIP.~:i!W so recove:-ed aC:~!.iO~!\\l lic~~c.s ~~

T!i;C:~ wou.la

.,-;;ould be reircled b:lck i!lLO fr~ re.ac~r

  • 1on-clcr;e !utu:e s&!c~:i:~ o;;:\\ons o~

.Surl. lt wa..s con:.e:;..ln:.CC: by t!1e :i:cl::lr

  • resal: ~ '.Ul:-.ece!..U':7 '."ir:in'1a~e:'.:r*. *
  • 1Dd~:7 Ula.~ s;>e=-i: fuel *wouJC: be dis*

a:'lC: (4) t.'le i:-:-a:i~t o: ~::.ses w~-:1

  • cha.reed ~rlod!ca.:Jr !ro:ti opi:r.iti.!:6 re** no~ be~r~lut!et! !o: fuel crcle atti\\~~~.

.*~cl.o!:$ sl.Ortt! 1n o~ite fuel sto::.o;e J'OOI' :tor erpe::.."nent.:Ll i.nd/o: ~:.:ii !e&.s::..

~jor a. Period or ::me to ~rm!: Cl.'CAY o:

b!!l:Y pu:;>oses..

. ." *. -,.,. *;:-*':'.'!:**:. :_*~

  • TMiioa.ct:ve
  • uut.e:i~ls cont.'\\inee 'llllt.'un * "*l:l licM of the s:.:itus or t.he
  • t.~~
\\he 1uel atid LO. cool. llJld pc!iodJei.la *pt:i:---ied c:ornce::i~ rc;i:tlCe~~ ;:>lu-.:.s
  • &hlp;"Cd oo*slte for rt:proc:essinc. n":>:ca:-

1n the O~t.ee S::.tes. :i.s oi::.l:nct! r.~on,

  • b' spa.ct wu provide~ l.n orui:e sto:tr.e *1.1ie ur!1es: t~: s~en~ :ucl rt:;lrOCes:.!~*;
    • ~ls for about onC' end o:ir*t..'Urd DU* *r:ou!d t>rtit> or. a c.:>n:i1:1er::a.J ba:i.s, l! au:.*

clcnr renc:t.o: ct>res. ~~1ng a !ou:-1er.r thonico. woult! be bte 1 n&. This n.s-;

  • nact.or fuel rclo:id eyc:Je, such o:sli.e

~u:ne.s tt'..1t. the Pl":'lc:.i:J

  • llc:en.s!:ii*

'atorr.r:e pools v.*ere pla.:1oed to beild an procetd.incs ue.:c.m;:>!etc:! and liC'C!l:.eS aven.te o: or.e ye:ir's C:isc:bl\\rr.e,.,-i:.'l su!* wucd by trus da:e. Ho"e*.-e:. the $;>~:it 1\\c:lc.nt rcme.lrunc cap:icity to ho!t! :i com-fuc:I poob n.: a nwnbu *o: rea.etors ni017 plel.t core sho:l!C: u!'Jeia.Giot? or ~*o! t.'le soon ~ ~!.ied. nnr:: s:IU o:::ier rue:.c:s.

fuel from the re~l.(I: be r.ec~a17 or wi!l have thl'~ p:>:)!s fi!led bt!ort t!:le end de.si.ra.!:>le because o! o;:>e:a.tlor.al &..cul* o! l9':8. A:ccrdir.:lr: even !: l!m.l:.ed re-*

tle.s. Uoder norca.l o~?':l.l.inc c:oucltio:u. proces.sinc s!iould bt'!:~ ln li.te 1!1";"'3, the~

  • i.ti anru~e or fh*c n:i.:-:;' c!!.sc:t.ar;:e could

~ol.I!:! stl!.! be a. sbo!"'~Se 1D SJ)CI)t fuel

  • be e.ecoin:n~:i:.ed bcfo:e Cle pools were si.o:-~~c C3paclt:r *.. **.
  • *.~.,.~.:. : *.. *.

'Nled.. *,.. *--.*

* **. *' *.... The L'tistill: pools *a~"* Cie.*CE. *arid*~ -.
  • ~~:Pcnons 'planning to CO:'lduct eommu-

?n:"S rc;>ro:em:ic p!:i.:'ltS -have so:llt re-*:.

  • cJLJ *reprotC.\\S::Og 0: S;>c:l':t l't:\\.C:.Or :u~!.s m~ni.'lg rr.k:~.. "lnl !Jcen."ltd 11.ornl"C ca::

)>ronded

  • au.=:elc?t sr.o::~e
  • tnJ:.lCJ~ !o: pa:ily ~?t;eh zuar be n~le.\\o acc;mmo-**

. tbe s~ent tuoii.S a. 1.hcir -~tlUl.l~ to allo'l7 c!a:e the !uel c1isch:i:tc.s from ao::ne

,aome operaUoD.AI tlP.ll.i~illty. T)"plc~Jy, ruelo~; &:'lJ" L"l:reascs p!nn.n'd 11t tt:.eu.

,.apace ba.s bee:i p:-o*.1d~ or pl."\\!l.'l~ lo'!'

~~3n!.& m:iy not be s~cl~nt for l.ndwt'J'7.*

  • several 1r-ent *fuel core rclow. Three.,,!n \\.'le tu:ure. Co:ue~\\:~llt!r. t.'le:e ls tt:.e *
  • ~mmercllLI rerirotC$slnt pla:1u bne * ~-o!sib!.ll:y of a tut\\!J'e at:ortn!e l.:l 1.t-.

been plJIJ:lned fo:- o:oeraU01:1 l:i l."l' 'O:::lt~ c-er.:t-d s:>ent fuel ta:NlC!ty rer;a.rdlrs.s o!

Stt.l.e6. The onl.Y suc:b pla~t Ul:i.\\ hu lhe ou!.r.ome o: the procttdi:::' on tbe act\\:a.U, opc~ud. :Nuclear Fu~! Services. ?J3.J 8tb noUce.

  • .* ;.... ; ~ :.~. : *:-

~

<NFS> plAnt a.t West va.,!ler. ~ew Yo:-t. *. The Commluion bu not proan~.lcated va.s ahut down 1D 197. for enens1ve

  • aiiy re£'\\llaUo:::a ~n.tc:b SPli'C~ts a r.ven

.Llt.e~tJoN n.nd,u:>;.i..~lou. The!"'! !:I a s:.:e tor o::i*s!te rc::\\c:tor s;:>e::i\\. fuel pools:

pending p~i:~L.?:; -~~c.ort the ~u.r 1lea.r bo;i:v~r. i>ropo~i.!s b; r'!ACt.o: l:c:e~c:s

~)a.~:7 Co:ti ** ;..... o:'l (COin..""1!."W.On) to ~il.'UJ~COl.'1tlf dum:e \\ht DU!1r.er or on Nf"S * *P:>1Jc:l\\._1*>n for a J>C"rm.lt to

>e~t tud st.i
-:i.:e or s;>e:'lt fuel pool sl:r cons\\.n.IC\\ Oles.I!' al ~:r~.. u"ns 11nd c~!ln* v.-ould be si:t>Jcei. to ll:e~nr rcl"iew by alon. <doc'kc\\ 110...'>-.01 >. The ~'Ond C
  • pla.nt. Ocn~!"al f;J~trlc Comriruiy's MiC:*

the. 011:m1ss1on. In the event that a west Fuel ~covoery Pl:i.nt :at :.~ol":'"-'. n-1>a.rt1cul:ar on-slle asient fuel pool shou?d llnols, h~ never "'~r:>.ted and ls tn a

  • 1't1:oaic filled, :and Do altcnu~lve torm dtc*onu:nh.,trned Mr.t11tlo"'. The U.ird of arcnt fuel 1Lora1e could be to~C:.

\\...

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  • ~!5T.l~t:~:~ JO lcp iq'l *~1.m ITt."1 J~;i~ao~ at *a.\\~ *-~n.11'"11111 rinaallnlOJi,\\ln ;)JnQU V JO
    • not srt:i *.:ra "t!O'Jr.:N~.,,M i" ~9<1 *

~oa,~ar )l:Clltd *tn I"l u;nn~a aon11.z,.c:a111 ai., c.... a~\\ H °'~~acn.ra'7'10..q

~omr:rmao:~' :tnr.znrr Jt:':*P*~.

.~ ~m.ri:c:1 10 n:;,.;:.:.1.1';1.-:r aui.,, ;~ lX~
  • ::o:> ;i;."1 1.0rt;Wl ~.*:."t';1:q pail ~!;-'13A

-~->>1::de aq =:1-a szc-:JeJ 9.lt; :11.11

  • ~o:i
  • 1sna paw *~n1.:J tJ1a:g "ft~ar.n:::> xua1~,. tr'!

?ol<::;a::w~*.-1.<;::;... '":~OJC: i:~.:aiTTUI

~J..,,Owl:f [~U;1:~' JO

~11~;.i U->

~!**t 'ltn..(J11"4 iliU'IU

~.ll'U<>t ;o \\t:!~P<:-c;t:

  • l(t u1 p.111=.)r.l~J. ~'*'* 1~;;;a~ " q:ins: a:n coa..is *1v111 \\;;,\\Jl!*1 o~ 11
  • ~µ._'.

1UJCJ"1~S,~~*~IT [l.'1~.1W01ll.\\Ui ;)µaa

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of COMMONWEALTH EDISON COMPANY (Dresden Station, Units 2 & 3)

Docket Nos. 50-237-SP 50-249-SP (Spent Fuel Pool Modification)

CERTIFICATE OF SERVICE I, Philip P. Steptoe, one of the attorneys for CoITU1lonwealth Edison Company, certify that copies of Applicant's Supplemental Testimony on Fuel Channell Bowing and Affidavit of Kin Wong of Quadrex/NSC have been served in the above-captioned matter on the following by depositing the same in the United States mail, first class postage prepaid, this 30th day of January, 1981:

Mr. Richard Hubbard MHB Technical Associates 1723 Hamilton Avenue Suite K San Jose, California 95125 Dr. Forrest J. Remick 305 East Hamilton Avenue State College, Pennsylvania 16801 Docketing and Service U.S. Nuclear Regulatory CoITU1lission Washington, *D.C~

20555 Richard Goddard Off ice of Executive Legal Director U.S. Nuclear Regulatory CoITU1lission Washington, D.C.

20555 John F. Wolf, Esq.

3409 Shephard Street Chevy Chase, Maryland 20015 Dr. Linda w. Little 5000 Hermitage Drive Raleigh, North Carolina 27612 Atomic Safety and Licensing Board.Panel U.S. Nuclear Regulatory CoITU1lission Washington, D.C.

20555 Mary Jo Murray Assistant Attqrney General Environmental Contrbl Division 188 West Randolph Street Suite 2315 Chicago, Illinois 60601

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