UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 2055 Gt- 7-

April 14, 1978

T Power Reactor Licensees

Gentlemen:

Enclosed for your 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

document provides (1) additional guidance for the type and extent

of information needed by the NRC 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 authorizing such modifications. This includes the

information needed to make the findings called for by the Commission

in the Federal Register Notice dated September 16, 1975 (copy enclosed)

with regard to authorization of fuel pool modifications prior to the

completion of the Generic Environmental Impact Statement, "Handling

and Storage of Spent Fuel from Light Water Nuclear Power Reactors".

The overall design objectives of a fuel storage facility at a reactor

complex are governed by various Regulatory Guides, the Standard

Review Plan (NUREG-75/087), and various industry standards. This

guidance provides a compilation in a single document of the pertinent

portions of these applicable references that are needed in addressing

spent fuel pool modifications. No additional regulatory requirements

are imposed or implied by this document.

Based on a review of license applications to date requesting authorization

to increase spent fuel storage capacity, the staff has had to request

additional information 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

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

to process a fuel pool modification amendment.

Sincerely,

Brian K. Grimes, Assistant Director

for Engineering and Projects

Division of Operating Reactors

Enclosures:

1. NRC Guidance / { j

2. Notice
ENCLOSURE NO. 1

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

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 generic environmental impact statement on handling

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 impact statement.

The Commission directed that in the consideration of any such proposed

licensing action, an environmental impact 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 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 applicable.

I-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.

Nuclear and thermal-hydraulic aspects of the review include the potential 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.

Mechanical, 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 earthquakes, 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

and radiological releases from the facility under normal as well as

accident conditions, the occupational radiation exposures, the generation of radioactive waste, the need for expansion, the commitment of

material and nonmaterial resources, realistic accidents, alternatives

to the proposed action and the cost-benefit balance.

The information related to nuclear and thermal-hydraulic type of

analyses is discussed in Section III.

The mechanical, material, and structural related aspects of information are discussed in Section IV.

The information required to complete an environmental impact assessment, including the five factors specified by the Commission, is

provided in Section V.

II-1
III. NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS

1. *Neutron Multiplication Factor

To include all credible conditions, the licensee shall calculate

the effective neutron multiplication factor, k f in the fuel

storage pool undef the following sets of assumed conditions:

1.1 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 life.

b. The moderator shall be assumed to be pure water 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 specifically for

neutron absorption, provided a means of inspection is established (refer to Section 1.5).

1.2 Postulated Accidents

The double contingency principle of ANSI N 16.1-1975 shall be

applied. It shall require two unlikely, independent, concurrent

events to produce a criticality accident.

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

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

1"NoncoOntairined" burnable poison is that which is not an integral part of

the fuel assembly.

**It should be noted that under certain conditions concrete may be a more

effective reflector than water.

III-1
postulated accidents shall include: (1) 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 tipping of the fuel cask or other heavy objects into the fuel pool;

(3) effect of tornado or earthquake on the deformation and relative position of the fuel racks; and (4) loss of all cooling

systems 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

"cancellation of error" in the calculations. So far as practicable the ability to correctly account for heterogeneities (e.g.,

thin slabs of absorber between storage locations) shall be

demonstrated.

A calculational bias, including the effect of wide spacing between

assemblies shall be determined from the comparison between calculation and experiment. A calculation uncertainity shall be

determined such that the true multiplication factor will be less

than the calculated value with a 95 percent probability at a 95

percent confidence level. The total uncertainity factor on k ff

shall be obtained by a statistical combination of the calculag

tional and mechanical uncertainties. The k value for the

racks shall be obtained by summing the calc8tlted value, the

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

thickness and type of stainless steel used in the storage

racks and the thermal (.025 ev) macroscopic neutron absorption cross section that is used in the calculation method

for this stainless steel;

(c) Also, for the H 0 + stainless steel flux trap lattices, the

change of the calculated neutron multiplication factor of

TTT-9
infinitely long fuel assemblies in infinitely large arrays

in the storage rack (i.e., the 9 of the nominal fuel storage

lattice cell and the changed g) for:

(1) A change in fuel loading in grams of U235, or equivalent, per axial centimeter of fuel assembly where it is

assumed that this change is made by increasing the

enrichment of the U2 35 ; and,

(2) A change 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 absorbers provide:

(1) The effective areal density of the boron-ten atoms

(i.e., B10 atoms/cm2 or the equivalent number of boronten atoms for other neutron absorbers) between fuel

assemblies.

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

the storage lattice cell k to:

(a) The fuel loading in grams of U23 5, or equivalent,

per axial centimeter of fuel assembly,

(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 conditions

(1) 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 sufficient 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 sample size on a

III-3
periodic basis throughout the life of the racks to verify

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 removed by the spent fuel pool cooling system

shall be made in accordance with Branch Technical Position

APCSB 9-2 entitled, "Residual Decay Energy for Light Water

Reactors for Long Term Cooling." 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. The maximum void fraction in the fuel

assembly and between fuel assemblies should also be calculated.

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

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

certain amount of cooling in the reactor vessel after reactor

shutdown prior to moving fuel assemblies into the spent fuel

pool. The bases for the analyses should include the established 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 H20

flux trap storage lattice exists if, somehow, the water is

kept out or forced out of the space between the fuel assemblies, conceivably by trapped air or steam. For this reason,

it is necessary 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 them. Also, in some cases, direct

gamma heating of the fuel storage cell walls and of the

intercell 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 significant physical 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.

III-4
(4) Potential Fuel 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

fuel handling procedures than that described in the Final

Safety Analysis Report, the differences should be discussed.

If potential fuel and rack handling accidents occur, the

neutron multiplication factor 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 specifications are needed on fuel storage in high density racks:

1. 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 number of grams of

uranium-235, or equivalent, put in the plant's technical specifications shall preclude criticality in the

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 specified spent fuel pool

temperatures, consideration shall be given to incorporating

a technical specification limit on the pool water temperature 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.l.d of the Standard Review Plan is applicable for

the maximum heat load with normal cooling systems in

operation.

III-5
IV. MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS

(1) Description of the Spent Fuel Pool and Racks

Descriptive information including plans 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 spent fuel 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 spent fuel racks should be provided indicating the methods of transferring the loads on the racks to

the fuel pool wall and the foundation slab. All gaps

(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 spectra, 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 radioactive water of the pool should be indicated.

(b) Fuel Handling: Postulation of a drop accident, and quantification 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. Integrity of the racks and the fuel pool due to a postulated

fuel handling accident is reviewed under the mechanical,

material, and structural disciplines. Sketches and sufficient details of the fuel handling system should be provided

to facilitate this review.

IV-1
I

(2) Applicable Codes, Standards and Specifications

Construction materials should conform to Section III, Subsection NF of the ASME* Code. All Materials should be selected to

be compatible with the fuel pool environment to minimize corrosion 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 metal may be obtained from the

Section III of the ASME B&PV Code, and the design stresses defined in the AISC specifications as percentages of the yield

stress may be used. Permissible stresses for stainless steel

welds used in accordance 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 reviewed on a case by case basis.

(3) Seismic and Impact Loads

For plants where dynamic input data such as floor response spectra 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

1.60 and 1.61 respectively. For plants where dynamic data are

available, e.g., ground response spectra for a fuel pool supported 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 reactor building, the design and analysis of the

new rack system may be performed by using either the existing

input parameters including the old damping values or new parameters 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 directions should be

imposed simultaneously for the design of the new rack system.

*American Society of Mechanical Engineers Boiler and Pressure Vessel

Codes, Latest Edition.

**American Institute of Steel Construction, Latest Edition.

IV-2
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

horizontal response spectra may be applied along the other horizontal 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 uEffective 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

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 supporting 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 acceptable, 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:

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 the thermal load due to the effect of the

maximum temperature distribution through the pool walls and base

IV-3
slab. Temperature gradient across the rack structure due to

differential heating effect between a full and an empty cell

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,

if applicable.

The specific loads and load combinations are acceptable if they

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 including 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 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 effective damping of the fuel bundle and the fuel racks.

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 loads

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,

two separate analyses should be performed as indicated below:

(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 the lower and higher elevations; 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.

IV-4
In order to 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

"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 corresponding 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

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.II.5 of the Standard Review

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

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

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 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 shown by detailed nonlinear dynamic 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 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 clearances is incorporated.

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

The materials, quality control procedures, and any special construction techniques should be described. The sequence of installation of the new fuel racks, and a description of the precautions to be taken to prevent damage to the stored fuel during

IV-5
TABLE

Load Combination

Elastic Analysis

0 + L

D + L + E.

Acceptance Limit

Normal limits of NF 3231.1a

Normal limits of NF 3231.1a

0 + L + To 1.5 times normal limits

lesser of 2 Sy and Su

or the

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

Faulted condition limits of

NF 3231.1c D + L + Ta +

Limit Analysis

1.7 (0 + L)

1.7 (0 + L + E)

Limits of XVII-4000 of Appendix XVII

of ASME Code Section III

1.3 (D + L + To)

1.3 (0 + 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.

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

Combinations for Class 1 Linear-Type Component Supports."

IV-6
the construction phase should be provided. Methods for structural 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 and the available chlorides, fluorides, boron, heavy

metals should be indicated so that the long-term integrity 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 upon the results of the qualification program

supported by test data and/or analytical procedures.

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 applicable code.

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 Inservice Surveillance

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 surveillance necessary to assure long-term safety and integrity of the

pool and the fuel rack system.

IV-1
V. COST/BENEFIT ASSESSMENT

1. Following is a list of information needed for the environmental

Cost/Benefit Assessment:

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 fuelstorage or fuel-reprocessing facilities,

(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,

(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 period that spent fuel 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.

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 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 for providing

replacement power either from within or outside the licensee's

generating system.

V-1
1.4 Discuss whether the commitment of material resources (e.g.,

stainless steel, boral, B C, etc.) would tend to significantly

foreclose the alternativet 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.

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 cooling water

systems and whether there will 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 generated by the SFP purification system. Discuss the expected

increase in solid wastes which will result from the expansion of

the capacity of the SFP.

2.2. Data regarding krypton-85 measured from the fuel building ventilation 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 concentrations 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 radionuclides and their respective concentrations.

(b) The models used to determine the external dose equivalent

rate from these radionuclides. Consider the dose equivalent rate at some 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 area and at the site boundary.

V -L
(e) An estimate of the increase in the annual man-rem burden

from more frequent changing of the demineralizer resin and

filter media.

(f) The buildup of crud (e.g., 5sCo, 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 low as

reasonably achievable.

(g) 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 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 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

(b) evaluation of the overhead handling system with respect to

Regulatory Guide 1.104.

3.2 If the accident aspects of review do not establish acceptability

with respect to either (a) or (b) above, then technical specifications 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

required that:

(1) define cask transfer path including control of

(a) cask height during transfer, and

(b) cask lateral position during transfer

(2) indicate the minimum age of fuel in pool sections 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.

V-3
3.4 If the cask drop/tip analysis as in 3.1(a) above is promised for

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 of 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 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

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

issued. This will involve checking realistic as well as conservative accident analyses.

V-4
VI. REFERENCES

1. Regulatory Guides

1.13 - Design Objectives for Light Water Reactor Spent 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 Nuclear Power

Plants

1.76 - Design Basis Tornado for Nuclear Power Plants

1.92 - Combining Modal Responses and Spatial Components in

Seismic Response Analysis

1.104 - Overhead Crane Handling Systems for Nuclear Power

Plants

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 Pressure Vessel Code Section III, 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-T6 and 6067-T6

VI-1
5. The Aluminium Association, Specification for Aluminium

Structures

4

VI-2
ENCLOSURE NO. 2 j *OCES -

oed Ginreial. Nuclear &ervics'

(AGNS) proposed :plant in * Brnwell

-'Bouth Carolina, is- under conostrction

and Is the subject of pending proceedings

t before- the Commission -regarding the

%continnuttlon, modification or.suspension - btf the construction permit from amen-

-vironmental protection 'standpoint. and

-.the possible Issuance of an operating' i-

-cse (docket' no. 50-332), as well as a

related matter (docket no. W01729).- -

-On May 8, 1975, the Nuclear 'Regulatory Commission published a notice hi7

the ftDER L RzcrsTr= settlng forth Its.

-provisional views that, sublect to con-'

-sideratlon of comments, 1() a cost;-

benefit analysis of alternative safeguards - ; SPENT FUEL STORAGE programs should be -ppae and -set

'forth in 'draft sand final environmnental

.Intent To PreparieGeneric Environmental impact statements before a- Commission

* Impact Statement on : Handling and decision is reached on wide-scale use of

= .Storage of Spent Light Water Power Re- mixed oxide (recycle plutonlum) fuels

actor Fuel .- .-. .- . in light 'water nuclear power reactors, -

.Prom the earlr days of the nuclear (2) therc should be no additional llcenses

*.-power Industry in this Country, electric granted for .use of mixed oiide fliel In

,-utlitles planning-to construct and oper- -light water nuclear power reactors esx-

;.,te .light water nuclear power reactors cept for exprlmerital purposes, (3) with

icontemplated that the used or spent fuel 'respect to light water. nuclear w dowepn

discharged from the reactors would be reactor fue cycle activities which depend -

chemically reprocessed to recover the for their Justification on wide-scale use

*remaining quantities of fissile and fer- of mixed oxide fuel In light water nu-'

t2le materials (uranium and plutonium), -clear power reactors, there should be no'

and -that the materials so recovered additional licenses granted which would

*Pwould be recycled back Into fresh reactor -foreclose future .safeguards options .or

ueL It was contemplated by the nuclear '7sesult In unnecessary."grandfather

industry that sPent fuel would be dis- and (4) the granting of licenses would

.charged periodically from operating re- not be-preluded for fue eycle activities

actors stored tn onsite fue storage pools for.experidiental and/or tecbnlcal feast-,

,Ior a period of time to permit decay of bility purposes. -a; ' .. '.....'-

.rulioactive'materials contained within -.. In light "of the status of Lb thee

'the Iue nd to *eool. and periodically -planned commercial repro plants.

'.-hpd ouisite for reprocessing. Typicai- in the United States, as outlined above,

y,' space rwas provided In orlsite storage the earliest that spent -fuel reprocessing

pools for about one and one-third nu- could begin on a commercial basis, If auclearreactorcores.Assumingafour-year thorized, would be late 1976. This as-:

-Teact6r fuel ':reload cycle, suchb onsite .-sumes that: the pending - licensing

torage .ools were planned to hold an proceedings are completed and llcenses

average rf one year's discharge with suf- issued by this date. However. the Spent

Sclent remaining capacity to hold a com- fuel pools at a number 'of reactors may

plete core -should unloading of all of the soon be filled, -and still other reactors

fuel Srom the reactor be necessary or will have their pools filled before the end

*desirable because of. operational dimicul- 'of 1978. Accordingly, even If limited reties. Under normal operating conditions. processing should begin In late 1976, there

-an average of five years' discharge could would stil be a-shortage In spent fuel

*be accommodated before the pools were storage capacity., . .

; ;.. - .. The- existing pools at the GE and.

w-persons planning to-conduct commnrer- NFS reprocessing. plants -have.some re-

,jaI-reprocessing of 'spent 'reactor fuels -malning marginal licensed -storage ca-'

>jrovided'sufflcient storage capachy Tor pacity which may be ablejp accommothe spent fuels at their facilities to allow "date the fuel discharges from- aome

-some operational ftexibility. TYPIcallY, reactors; any Increases planned at these':

.space has been-prOvIded or Plened for plants may not be sufficlent for Industry'

"several- spent -fuel Tfe-. reloads. ThThe in the future. Consequently, there -is the

commercal" reprocessing -plants 'have possibility of a future-shortage in 11-.

been planned for operation in the United censed spent fuel capacity regardless of

S.tates. The only such-plant that' hss. the outcome of the proceedings on the

'actually operated. Nuclear Fuel Services -May 8th notice.- ' :- ,-. - :

(N ) plant at West Valley. New Y;, -. The Commission has not promulgated

was shut down In 1972 for extensive- any regulation which speclfies a given

aIterations and expanslon. There 1i a size for on-site reactor sPent- fuel pools;

pending proceeding before the Nuclear however, proposals 'by reactor licensees

regulatory '-CommIsslon (CommI0s0on)- to significantly change the manner of

on N7'8s applicstion for a permit to spent fuel storage or spent fud pool size

construct these Alterations and expan- would be subject to licensing review by

sion. '(docket no. 50-201). The' secod -the Commission. In the event that a

'plant, General Electric Comipany's Mid- o-iesntfl olaod

west Fuel Recovery Plant at Morris, n- particular on-site spent fue pool should

linoda, has never operated and Is ln a' become filled, and no alternative form

-decommlssiened condition. The --third of spent fuel storage -could be found.

.4I

FEDERAL REGISTER. VOL 40, NO. 1S0-TUESDAY, SIPTEMBEP. 16, 1975
42802 oNOTICES

the reactor would be eventually forced mennt as a suitable vehele 'for such Ln Comiion had tao' bslc. *bi

to shut-down and "store" the last spent.' eva-lnatoio.-Notlce is hereby given that mind: on the one hand, the generic 1m4

.reactor fuel in-the reactor pressure ves- 'a geric environmental Impact. state- :.pact statementashould not serve ami jus-

`seL :wilne no serlous..dverse, conse- mrent.on the handling and storage of *tlcation.S-or. a -falt accoml ,-je-

%que&esrto th publc health and safety, spent light water power reactor fuels wL other hand, the public Interest consld-,

.the comrmon. deensa and security, or be prepared by tie -Commlsslon..-The erations associated with such a defezru * the environment would likely result, the statement will focus on, the lm period .shouldbe carefulby weghed..The Co* :t

i'reactor shutdown would, 'of. course, re- . between now and the mid 1980's and, will missionuhas concluded that tt h

',movotheoplantfrom servlce.andthis In.-addres: -. - * . ' -- be no such general deferral, .tbt

'.turn could adversely affect the electric --U() The- mgniltude of 'the-possible these related licensg actnsa con-

:-ut11t7 ability to- meet electrical energ . shortage o spent fuel storage capacity; . tinue during. the period reed-for -.needs, or force the utility to ope zte other - ' (2) The alternatives for dealing with .preparation of. the, gener state t

>lants that are less economical to operate the problemTs Including. -but not neces- subject to certain cohdtioci'Xn e

sor'which have greater environmental im- sarilylimnitedto: -- * -. ' % ..- _., -Ing this conclusion, the Co on

nthereby thact, adversely affect the (a) Permitting the expnslon of spent considered the following speift

"public. interest. fud 'i. stoqge capacity at power reactors; (1) it is likely that each 1ndvldu'lli.,:

-;There.ppeat be a number of ps- (b) Permitting the expansion of spent - ceastng acton 6 ts type WC h-

%ilbe- aliternatlves for lncreasing.spent fuel storage capacityat reprocessing autilhtythatlslndependentoftheutltyr fuel storage capacity Including, among plants; - - ' . . of.other licensing ations of thl5- type;,

othezrtlnpgs, increasing the storage-ca- (c) Licensing oLfIndependent spent (2) It i not likely that the taking (if

;paclty at present reactor sites, and con- -fuel storage facilities; . . . - any partlcularjlcesrig action ot this

struction of Independent spent fuel . -. (d) Storage' of spent fuel from one type during the time rame under constorage facilities. The shortage of spent . or more reactors *tzthe storage pools of dideration would constitute a commit:.

-fuel storage capacity wi occur at lndl- . ,other reactors; - , . . ment of resources- that- would tend *to'

"Vidual reactors.-and the Commlssion. (e).Orderingthat generatoof spn t -gnifcantl foreose altmatlves

could adequatl addre issues In-; fuud. (reactor operatipn) be stoped or -"availale' with s espect to ayother rolredt on'a case-by-caseabasls- iithin restrlcted; .-' . a-'fdl'idualflcnsigation'of ths type; ',

bthi context of Individual 'licensing re- (3) A cost-benefit analjsis of the al- '. (3) St is lely that any enrisnmentil'

5iewsT Indeed, the ComRtrion has not, -ternatlves listed In (2). along with any impacts assoclated with eny Individuali

Ito dati, found It necessary,'In the dis- other reasonably -easlble alternatives licensing action of this type would le,

charge Of Its licensing and related regu- including:. such that they could adequately be ad_

'latory-functions, to develop any overall (a) Impacts on- public heaIth and drcse wit the context of the -:

Jropamof action.to deal with the prob- -aety. and. thc common defense ,and .vidual license application. without over- - !em!h Commlsslon does, howeverhave .security; '- -. -i - .: - - looking any cumulative environmentalr

4thi dlscretion to deal with Issues of this - (b) Environmental .social, and eco- impacts; - ; a : -. -

4ype-= a generic basis through the ex- onoml~'costs and benefits;- -- - -- _ 4Yr It 13 'lily that' any technicalercIse o its.rulemaking authority and/ . * (c) Commlltments of resources;-, .-*..ssues that may arise In the course of a"'

-or the Issuance of a "generic" environ-. (d) Implications regarding oPtions review of an Individual lIcense appulca- -

m'entar mpact 'statment. lRulemalng available for the Intermediate andilong- tlion can be resolved within 'that com-'

Jproceedlnzs. and/or the Issuance of a 'term storage of nuclear waste materials; -text; and :-- - -

generic environmental impact statement Ce) .Relationship between local shorSt (5) A deferrseiere restrition on, . t.as -appropriate, serve as the con- -knn uses of the envfronmnt and long- licensing actions of this type would re-:-.

-xtrSor the promulgation of nore de- -term productlvlty;z- - - gft In su bstaidaI hr -to the publlc I&Rlve 'criteria regardIng size and de- -. (4) 7he impacts of possible additional Interestt As Indicated, auc a t uestrict10

"sidgn of spent fuel pools and/or the - transportation of spent fuel that may or deferral could result In reactor shut--

:censdngofindependentspentfuelstorage -be required should one or more of the downsasexlstlnjspentfuelpoolsbecome Bfacilitles. and for. consideration of pos-`-- alternatives be adopted; - . -' filled. It now appeasn tat. .the.. apent

; zible-revision of the fuel cycle environ- '(5) 3fy deve standards and cvi: fuel pools- of 'as many as ten .reactors'

.:mSntal, impacts set forth ln .10 Et -terla to govern the-licensing of one or could be filled by mid-197B. These teni :i 5t (e)n light of additional spent fudl more of the alternatives for dealing with reactors represent a total of about a mQ- * storae and attendant transportation. polemn and .- - ., lion kloatts of eectrlcal energen*' ~~Asthe possible ImPlicatiorts .of In- 'theprbe;ad-inkawtsoelcialnrg n7

creased tpenmt fuel storage on the optios '(a) Possilbe aaendments to 10.'CFR rtn aait.Termvlof -- "Pse =;= P 15120(~~~~~~~~e *.ace,,; to-rs C, freoin fsamservice ctou~#ld reduc tbi; ivsaallble for Intermediate and long-term , 1.20(e-- - ' '- - - -' :'- rlatorsf serv ice co r tvhr

stoge:of nuclear 'waste materials could , 2 approprlakt, rulng iroceen i st

0 rdal e xmndwti'ti -o tm 5 n S itdaav . reliable service -would be In Jeopardy. ay-

- l . "'.- other oeat.Issues related to the halndl .hi c then albr mto rely mformn; -oie: groii~-ofuterested 3'rgiim~utomistorage of spent reactor tul-wl &n- 1 cnmclo oepollutingfom %aS Defense- Cou'nc, :Utedpnor U of > o-f; g eneratlon that- would imposeecoF *t-S~ara:Club, and Buslneniiefor the-- bf the dritgn envr:me tal te-reaseenrnmnl o enaleon consumers a. in'Y , PubMc Interest) has requested the Corn- pact ttatement. n , - , . , . o n Ia,

*-;isson to prepare a generic environmen- - The Comzsnsslon &a also given careful - censlng action 1ntenpec toat'nyd l Impact statement on the hanng and conidTeCon 'to. -the gquestion whetfler- g f - t. mestorage oftspent reacthor fue and related licensinlg actions "Intded to unde amelioratedrod" to 

-mnatters- (letter to L. V. Gosslck -from ,possible shortage Of spent fuel storage 'would beaccompnived by an environ-'-- Anthony Z. RoismandatedMay20.1975, -capacity, Including emh actiosa as the' mental lniiact statement.(OC' 1-iL5 - cop COrOD fle at.the Commissionfs Public- ssuance- of operatine license .men&d&-W) O zact appraisal (10 CFR I 5L5 1

sDocument Room, 1717 H Street, NW. montetopermit ncraseasinhe Cc)) tailored to the facts of the Ease;

washlngt, DC.G)., . . - - ,! -'-..capacity of reactor spentjfel pool s ce the ComolSrcn'S general concl3-

-While the Commission believes, as ear' reprocessing. plant spent-fuel storge os rspect to the ve facto, 1eir indcatedt that the matter of spent -pools, or the licensing of Independent set with respet to t A factr2

uel storage capacity can adequately be spent fuel storage facilities, should be set forth above,.may' not fit the factual

addessd o a as-bycas baiswithin deferred pending completion of the ge- circmnstances of particular licensing so- -

* addressed on a case-by-case basis oions, the 'five factors will be applied''' the context of Individual licensing re-. neric environmental Impact statement weighed and balanced withln the con-

,'vlews, lt also believes that. from the Such's deferral was requested in th text of these statements or appraisals 13

standpoint of longer range policy, this letter on behalf of Naturl Resources texoh ese determlnatlons In

matter can profitably be examined In a Defense Council, Sierra Club. and Busl-. -- r i n d.. 

broader context. It views the preparation nessmen for the Public Interest noted Dated at Washington, D.C. this 10th; of a generic environmental lmpact state-- above. In considering thls matter, the day of SeptemberL975.

FEDERAL MISTER, VOL 40, NO. 180-TUESDAY, SEPTEMBER 16, 1975
Consolidated Edison Company of - 2 -

New York, Inc.

cc: White Plains Public Library

100 Martine Avenue

White Plains, New York 10601

Leonard M. Trosten, Esquire

LeBoeuf, Lamb, Leiby & MacRae

1757 N Street, NW.

Washington, D.C. 20036

Anthony Z. Roisman

Natural Resources Defense Council

917 - 15th Street, NW

Washington, D.C. 20005 o

Paul S. Shemin, Esquire

Assistant Attorney General

State of New York

Department of Law

Two World Trade Center

New York, New York 10047

Sarah Chasis, Esquire

Natural Resources Defense Council

122 East 42nd Street

New York, New York 10017