ML19280A177
| ML19280A177 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, West Valley Demonstration Project, 07001729, Barnwell |
| Issue date: | 04/14/1978 |
| From: | Grimes B Office of Nuclear Reactor Regulation |
| To: | FLORIDA POWER CORP. |
| Shared Package | |
| ML19280A178 | List: |
| References | |
| RTR-NUREG-75-087, RTR-NUREG-75-87 NUDOCS 8003040729 | |
| Download: ML19280A177 (2) | |
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50-G69 i
'o UNITED STATES
[ g, NUCLEAR REGULATORY COMMtsslON I
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WASHINGTON. D. C. 20555
\\6 f' April 14, 1978 i
To Al's Power Reactor Licensees Gentlemen:,
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Enclosed for your information and possible future use is the NRC guidance on rpent fuel pool modifications, entitled Review and A :eptance 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 I
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 Comission 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 comoilation in a single document of the pertinent portions of these appitcable references that are needed in aodressing 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 infomation that could have been included in an adequately documented initial submittal.
If in the future you find it necessary to apply for authorization to modify onsite spent fuel storage 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, lw c W
c#G g-), cl Brian K. Grimes, Assistant Director (JDN " 7s
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Division of Operating Reactors for Engineering and Projects
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Enclosures:
1.
NRC Guidance 2.
Notice
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ENCLOSURE NO. 1
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I OT POSITION FOR REVIEW AND ACCEPTANCE OF SPENT FUEL STORAGE AND HANDLING APPLICATIONS I.
BACKGROUND l.
I Prior to 1S75, low density spent fuel storage racks were designed with l
a large pit:h, to prevent fuel pool criticality even if the pool I
contained the highest enrichment uranium in the light water reactor fuel assemblics.
Due to in in:reased demand on storage space for spent fuel assemblies, tne 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
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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 r.ot 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 whic.h five specific factors in 4
addition to the normal cost / benefit balance and environmental stresses should be appli'ed, 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 criteric where applicable.
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,s II.
REVIEW DISCIPLINES l
The objective of the staff review is to prepare (1) Safety Evaluttion l
Report, and (2) Environmental Impact Appraisal.
The broad staff e
d': ciplines 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 norrral stort;e and handling of the spert f uel, v.nd the consequences of credible accidents with respect to crit'.:ality ind the abili y of the heat removal system to maintain suf ficier,t 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 earth-quakes, tornadoes, flood, effects of external and internal missiles, thermal loading, and also.other service loading conditions.
The environmental aspects of the review concern the increased thermal and radiological releases from the facility under normal as well as accident conditions, the occupational radiation exposures, the genera-tion of radioactive waste, the need for expansion, the commitment of material and nonmaterial resources, realistic acciderits, alternatives to the proposed action and the cost-benefit, balance.
The information rebated to nuclear and thermal-hydraulic type of analyses is discussed in Section III.
The mechanical, material, and structural related aspects of informa-tion are discussed in Section IV.
The information required to complete an environmental impact assess-ment, including the five factors specified by the Commission, is provided in Sec+1on V.
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III. NUCLEAR AND THERKAL-HYDRAULIC CONSIDERATIONS 1.
Neutron Multiplication Factor To include all credi51e conditions, t e licensee shall calculate i
the effective neutro: multiplication fa-tor, k in the feel assumNf,er4itions:
storage pool uncet the follo. ting sets 01 c
1.1 Notmal 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 laterr1 extent or to be surrounded by an infinitely thick water ret.ector and thick concrete,** as appropriate to the design.
d.
Mechanical u0 certainties may be treated by assuming " worst case" condit kns or by performing sensitivity studies and obttining appropriate uncertainties.
Credit may be taken for the neutron absorption in structural e.
materials and in solid materials added specifically for neutron absorption, provided a means of inspection is estab-lished (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
^"Noncontained" 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.
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4 postulated accidents shall include:
(1) dropping of a fu61 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; I
(3) effect of tornado or earthquake on the deformation and rela-tive 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 nethods 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 j
" cancellation of error" in the calculations.
So far as practi-i cable 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 calcu-lation and experiment.
A calculation uncertainity shall be determined such that the true multiplication factor will be less i
than the calculated value with a 95 percent probability at a 95 percent confidence level.
The total uncertainity factor on kEff shall be obtained by a statistical combination of the calcula tional and mechanical uncertainties.
The k value for the i
racks shall be obtained by summing the calc 6f$ted 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 thickhess and type of stainless steel used in the storage racks and the thermal.(.025 ev) macroscopic neutron absorp-tion cross section that is used in the calculation method for this stainless steel; (c) Also, for the H 0 + stainless steel flux trap latticas, the p
change of the calculated neutron multiplication factor of TTT-?
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I iniinitely long fuel assemblies in infinitely large arrays inthestoragerack(i.e.,thegofthenominalfuelstorage lattice cell and the changed g) for:
I (1) A change in fuel loading in grams of U ss, or equiva-2 lent, per axial cent heter of fuel assembly where it is assumed that this chtnge is made by increasing the enrichnent of the U2ss; ano, (2) A change in the thickness of stainless steel in the
'torage racks assuming that a decrease in stainless l
steel thickness is taken up by an increase in water thicknessandviceversa; (d) For lattices which use boron or other strong neutron absorb-ers provide:
(1) The effective areal density of the boron-ten atoms (i. e., B10 atoms /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 thestoragelatticecellgto:
(a) The fuel loading in grams of U23s"'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 l
1ess 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 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 sample size on a III-3
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u perio'dic 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.
j (2) Decay Heat Calculations for the Spent Fuel The calculations for the amount of thermal energy that will
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have to be removed by the spent fuel pool cooling system shall be cade in accordance with Branch Technical Position I
APCSB 9-2 entitled, " Residual Decay Energy for Light Water Reactors for Long #erm Cooling." This Branch Technical Position is part oc the Standard Review Plan (NUREG 75/087).
(3) Thermal-Hydraulic Analyses for Spent Fuel Cooling l
Conservative methods should be used to calculate the maximum l
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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 l
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 estab-lished cooling times for both the usual refueling case and the full core off load case.
A potential fot a large increase in the reactivity in an H O 2
flux trap storage lattice exists if, somehow, the water is kept out or forced out of the space between the fuel assem-blies, 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 l
gamma heating of the fuel storage cell walls and of the l
intercell water may be significant.
It is necessary to consider direct gamma heating of the fuel storage cell walls snd of the intercell water to show that boiling will not j
occur in the water channels between the fuel assemblies.
Under postulated accident conditions where all non-Category l
I spent fuel pool cooling systems become inoperative, it is l
necessary to show that there is an alternate method for cooling the spent pool water. When this alternative method l
requires the installation of alternate components or signifi-cant 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 expe:ted heat up rate of this water assuming loss of all cooling systems shall be specified.
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l (4) Potential Fuel and Rack 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 l'
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 truitiplication facter in the fuel pool shall not exceed 0.95.
These postulated accidents shall not be the cause cf 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:
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 irito the high density racks should be limited.
The number of grams of uranium-235, or equivalent, put in the plant's tech-nical specifications shall preclude criticality in the 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 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
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Reactor Spent Fuel Storage Facilities at Nuclear Power Stations," except that the requirements of the Section 9.1.3.III.1.d of the Standard Review Plan is applicable for the maximum heat load with normal cooling systems in operation.
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I IV.
MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS i
(1)
Description of the Spent Fuel Pool.and Racks Descriptive information including plans and sections showing the spent f;cl poci in relation to other plant st uctures shall be provided in order to define the primary struc ;ral 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 tne 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 indi-cating 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 interfaci.pg between the new rack system and the old fuel pool walls and base slab should be discussed, i.e., interf ace 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.
Postulated drop accidents must include a straight drop on the top of a rack, a straight drop through an individual cell all the way to the bottom of the rack, and an inclined drop on the top of a rack.
In-tegrity of. the racks and the fuel pool due to a postulated fuel handling accident is 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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1 (2)
Applicable Codes, Standards and Specifications Construction materials should conform to Section III, S'ubsec-tion NF of the ASME* Code.
All Materials should be selected to be compatible with the fuel pool environment to minimize corro-sion and galvanic effects.
Design, fabrication, and insta11atio'n 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 A WE 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 tase netal may be obtained from the Section III of the ASME BSPV Code, and the design stresses de-l fined in the AISC specifications as percentages of the yield p
stress may be used.
Permissible stresses for stainless steel i.
welds used in accordance with the AISC Code may be obtained from I
Table NF-3292.1-1 of ASME Section III Code.
Other materials, design procedures, and fabrication techniques will be reviewed on a case by case basis.
(3)
Seismic and Impact Loads For plants where dynamic input data such as floor response spec-tra or ground response spectra are not available, necessary dynamic analyses may be performed using the criteria described in Section 3.7 of the Standard Review Plan. The ground response spectra and damping values should correspond to Regulatory Guide l
1.60 and 1.61 respectively.
For plants where dynamic data are available, e.g., ground response spectra for a fuel pool sup-ported by the ground, floor response spectra for fuel pools l
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 i
input parameters including the.old damping values or new param-l eters in accordance with Regulatory Guide 1.60 and 1.61.
The use l
of existing input with new damping values in Regulatory Guide l
1.61 is not acceptable.
Seismic excitation along three orthogonal directions should be I
imposed simultaneously for the design of the new rack system.
"Aterican Society of Mechanical Engineers Boiler and Pressure Vessel Codes, L:atest Edition.
- American Institute of Steel Construction, Latest Edition.
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l The peak response from each direction should be combined by squar a root of the sum of the squares.
If response spectra are I
available for a vertical and horizontal directions only, the same l
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 acss of the fuel racks has been under study by the NRC.
Submeroence in s ner may intuduce damning fr r two sou s, (a) viscous drag, an (b) radiation of enerc,,y away from the abmerged l
bocy in those cases where the confining boundaries are f ar 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 configurati:n, vave reflections' occur at the boundaries so that no additional damping should be taken into account.
A report on the NRC study is to be published shortly under the title " Effective Mass and Damping of Submerged Structures (UCRL-52342)," by R. G. Dong.
The recomendations 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 enerQy 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-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 erergy.
(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
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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 acceptat,le if they are in conformity with the applicable portions of Section 3.8.4-II.3 of the Standard Review Plan.
(5)
Design cnd Analysis Procedures Details of the mathematical model including a description of how the icportant parameters are obtained should be provided includ-ing the following:
the methods used to incorporate any gaps I
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 vater on the pool walls; and, the effect of submergence on the mass, the mass distribution and the effec-l tive damping of the fuel bundle and the fuel racks.
The design and analysis procedures in accordance with Section i
3.8.4-11.4 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.
i When pool walls are utilized to provide lateral restraint at I
I 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.
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In order to determine the flexibility of the pool wall it is acceptable for the licensee to use equivalent mass and stiffness p'roperties 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 mey be assumed that the response of the pool wall and the corres-ponding lateral support to the new rack system are identical to those of the base slab, for wnich appropriate floor response spectra or ground response Epcctra may already exist.
(5)
Structural Acceptance Criteria Phen AISC Cedi: procedures are adopted, the structural acceptuice i
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 3.8.4.11.5 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 Stand-ard Review Plan.. This position on factors of safety against sliding and tilting need not be met provided any one of the following conditions is 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 nodule and the pool walls is prevented provided that the factors of safety against tilting are within the values permitted by Section 3.8.5.11.5 of the Standard Review Plan.
(b) it can be shown that any sliding and tilting motion will be contained within suitable geometric constraints. such as thermal clearances, and that any impact due to the clear-ances is incorporated.
(7), Materials, Quality Control, and Special Construction Techniques:
The materials, quality control procedures, and any special con-struction techniques should be described.
The sequence of in-stallation of the new fuel racks, and a description of the pre-cautions to be taken to prevent damage to the stored fuel during IV-5
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TABLE Lead Cor.bination Elastic Analysis Acceotance Limit D+L Norca.1 limits of NF 3231.la D + L + E.
Nomal limits of NF 3231.la 0 + L + To 1.5 times normal limits or the lesser of 2 Sy and Su 1.5 times normal limits or the D + L + To + E leser of P. Sy and Su 4
D + L + Ta + E 1_.6 times nomal limits or the lesser of 2 Sy or Su
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l 0 + L + Ta + E Faulted condition limits of NF 3231.lc 9
, Limit Analysis 1.7 (D + L)
Limits of XVII-4000 of Appendix XVII of ASME 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)
Notes:
1.
The abbreviations in the table abovt 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 amended by the requirements of the paragraphs c.2, 3, and 4 of the Regulatory Guide 1.124 entitled " Design Limits and Load Combinations for Class 1 Linear-Type Component Supports."
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l the' construction phase should be provided. 'Hethods for struc-I tural qualification of special poison materials utilized to absorb neutron radiation should be described.
The material for the fuel rack is reviewed for comp,atibility 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-tcro 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 appli-cable 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 Meth'ods for verjfication of long-term material stability and mechanical integrity of special poison material utilized for neutron absorption should include act.ual tests.
Inservice surveillance requirements for the fuel racks and the poison material, if applicable, are dependent on specific design features.
These features will be reviewed on a case by case basis to determine the type and the extent of inservice surveil-lance necessary to assure long-term safety and integrity of the pool and the fuel rack system.
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V.
COST / BENEFIT ASSESSMENT 1.
Following is a list of information needed for the environmental Cost / Benefit Assessment:
l 1.1 What are the specific needs that require increased storage ccpacity in the. spent fuel pool (SFP)?
Include in the response:
(e) rtatus of contract ual arrangements, if any, with fuel-
- orage or fuel-reprocessing f acilities, (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 preposed modification, including engineering, capital costs (direct and indirect) and allowances for funds used during construction.
1.3 Discuss the alternative to increasing the stcrage 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 f or providing replacement power either from within cr outside the licensee's generating system.
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t Discuss whether the commitment of material resources (e."g.,
1.4 stainless steel, boral, B C, etc.) would tend to significantly foreclose the alternativek 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 vill be any significant increase in l
the amount of beat released to the environment.
V.2. RADIOLMICAL EVALUATION l'"
2.
Following is a list of infomation needed for radiological evaluation:
2.1 The present annual quantity of solid radioactive wastes gen-ersted by the SFP purification system.
Discuss the expected i:
increase in solid wastes which will result from the expansion of l'
the capacity of the SFP.
l
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2.2. Data regarding kr'ypton-85 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.
b i
Theincreasesinkhedosestopersonnelfromradionuclidecon-2.3 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.
l-(b) The models used to determine the external dose equivalent rate from these radionuclides.
Consider the dose equiva-1ent 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 d' termine the e
principal airborne radionuclides ar.d their respective concentrations in the SFP area.
The model and assumptions used to determine the increase, (d) if any, in dose rate from the radionuclides. identified in (c) above in the SFP area and at the site boundary.
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(e) An estimate of the increase in the annual m'an-rem burden from more frequent changing of the demineralizer resin and filter media.
-(f) The buildup of crud (e.g., ssCo, 80Co) along the sides of the pool and the removar methods that will be used to reduce radiation levels at the pool edge to as low as reasonably achievabie.
(g) The txpected totql man-rem to be received by personnel occupying the fuel pool area based on all toerations in that arca including the doses resulting fron. (e) and (f) above.
~
A discussion of the radiation protection program as it affects
)
(a) through (g) should be provided.
l.
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 FVAWATION
_. E 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 specifica-tions may be required that prohibit cask movement in the spent fuel building.
3.3 If the accident review does not establish acceptability with respect to (b) above, then technical specifications may be required that:
l
(,1) define cask transfer path including contiel of (a) cask height during transfer, and (b) cask lateral position during transfer
..s (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
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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 i
negative declaration with an Environmental Impact Appraisal (rather than a Draft and Final Environmental Statement) can be i
issued.
This will involve checking realistic as well as con-servative accident analyses.
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V-4
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VI.
REFERENCES 1.
Regulatory Guides Design Objectives for Light Water Reactor Spent Fuel 1.13 Storage Facilities at Nu: lear Power Staticns Seismic Design Classification 1.29 Design Response Spectra for Seismic Design of Nuclear 1.00 Power Plants Damping Values for Selsmic Design of Nuclear Power 1.61 Plants Design Basis Tornado for Nuclear Power P.lants 1.76 Combining Modal Responses and Spatial Components in 1.92 Seismic Response Analysis 1.104 -
Overhead Crane Handling Systems for Nuclear Power Plants 1.124 -
Design Limits and Loading Corbinations for Class 1 L,inear-Type Components Supports
~
2.
Standard Review Plan Seismic Design 3.7 3.8.4 -
Other Category I Structures Fuel' Storage and Handling 9.1 9.5.1 -
Fire Protection System 3.
Industry Codes and Standards 1.
American Society of Mechanical Engineers, Boiler and Pres-sure Vessel Code Section III, 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 6057.-T6 VI-1
(
'1 The Aluminium Association, Specification for h1uminium 1
5.
Structures e
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a EtiCLOSURE NO. 2 7)CTICES
[CS01
. - - -.- v -.,
plant, Anled. Genera.1. Nuclear Services' i
.MAGNS) proposed. plant.in 'Barnwtn,
- Socth Carchnal is-under construction
' and is the subject of pendiag proceedings g e!cre the Commission-regsniing the b
econtinuation, mod 1Scation or suspensjon Tof the construction permit from an~ en -
- vironmental protection standpoint, and
- the pcssible issuance of an operating"11-
' cense (do:
no. 50 '.',32), as ven a.s a related ca. r (docket no. 70-1729 W *.,
~ On May I;,1775, the Nuclear Regula-tor 7 Comm!ssics pub 1hhed a notice.irb
^ithe TurP.AI.T. !strR Se".dg forth ita prordon1 views thst, cut 3ect'to con ?
- siders. tics cf commen'.s,T(1). a cost;
, beneSt e.nalysi. of alternative safeguarcs Trogra=s z.bould be.prepced s.nd set
' ' ', - SPEOT FUE!. STORAGr
-f or:3. in draft and inal environmentaJ
'J
,In'ent To P ep *e Cenetic EnArenmental 'impac* ttatements before aT-Mon
- irrpa ct !.uternent on ? HanE.9g and decisic. is rtiched en Mde-c e use of Morage of Spent LJght Water Power Re-mixed c 2 (recycle plu:.r.tum) IucIs '
, j. cor fuel
~....;..
r. in light wv.er nuclear mwer reactors,-
w
-. Prom the etr17-days of the n'uclear (2) there should be no addjtionalljeenses-grantad for,use of mixed oxide fliel in Ipower industry in this country, electric.hght water nuclear power reactors e.x-f
- utHities DWnr in construct and oper--f fate. light water nuclear power reactors ceptior exp,e.rimentalpurposes. (3) with l contemplated that the used or spent fuel yespect* to light water / nuclear power' discharged -frotn the reactors would be reactor fut! cycle actirtties which depend
- chem 3eany' reprocessed to' recover the for their just!Scation.on Mdt-scale use remintnr quantities of fissue and ferr of raised oxide fuel in light water nu '
. tHe materials (uranium and plutonium),
- clear power reactors, there should be no'
/and that the materials to recovered additional 11 censes granted which would.
,would be recycled back into !resh reactor. foreclose future safeguards options. or Jfuel. It,was contemplated by the nuclear
- Tesult in unnecessaryd'grandf athering"..
industry that spent frel would be dis-, and (4) the granting of licenses'would
.chs.rged periodicany from operating re-.not be pre 9 uded for fuel cycle activities 1
I.for a period of time to permit decay ofactors, stored in onsite fuel storage pools '.
e b!11ty purposes.,. c.Q*ayW.W-M.l; tradioactive materiali contained within UCIn lightlo! the status' of the'ihree:
~
}the fuel and to ; cool,' and periodically' planned commercialreprodeising plants.
Tshipped 05 site for reprocessing. Typical- 'in the United States, as cutlined above,*i
~
.ly, tpace was provided in onsite storage :the earliest that spent fuel reprocessing
' pools for about one and one-third nu 'could begin on a commercial basis,if au~ ;
clear reactor cores. Assuming a f our-year thorized, would be late 1976. This as'-;
reactor fuel reload cycle, such ons!!e :sumes that, the ' pending.11 censing
' storage pools were planned to hold an proceedings are completed and lleenses average df one year's discharge with suf-issued by this date. HoweverJthe spent ficient remaining. capacity to hold a com-fuel pools at a number ~cf reactors snay, plete core-should unloading of au c! the soon be flued..and still other reactors inel from the reactor be necessary or wi.1 have their pools Sued befort the'end
^
desirable because of operational di5eul
'of 1978. Accordinglyleven if limited re -
ties.Under normal operati.ng conditions, processing should begin iniste 1976, there
'an aversge of five years
- discharge could would stiU, be a. shortage'in"rpent, fuel
'be~ accommodated before the pools, were,storagecapacityy,3.'
pf g.-i ;.
" N ' The. e.xisting pools,at,'the' GE7and'.
' fined.~.
3.
c..
Cf. Persons" planning to conduct cocnmer. NFS reprocessing. plants.have.some re 7'T
', cia!' reprocessing of spent; reactor fuels dn*tng maisinal licensed storage 'em-'
p:ovided'su5clent stcrage er.pacity -for
. the spent fuels at their f acilities to aHow 'pacity which may be'.able.1o accomnio-i date the fuel discharges.from. some some operational flexibility, Typically. reactors 1 any increases p!r.nned at these-
'several* spent fuel core reloads. Three, plants may not be suSelent for industry
' space has been provided or planned for plants "have '.in the future. Consequently, there is the
'commerciali reprocessing been planned for operation in the United possibDity of a future'. shortage in 11.
censed spent fuel capacity regardless of
't States. He only such ' plant that'ha3 the outcome of the. proceedings on the actuany operated. Nuclear Fuel Serdees,Giay Eth notice.1-
.4/4-W w ", W GTS) plant at West Valley, New Terk. '. TM Commission he's no't ' pro:nukaidd was shut down in 1972 for ertensive. any regulation which sxc!Ses a >'
. alterations and expansion. nere is a si:e for on-site reactor sxnt, fuel pools
- pending proceedjng before the Nuclear however, preposals by reactor licensee Regulatory -Com=1ssion (Co-*lon). to signiScantly change the mer of on NTS's applicattion for a permit to spent fuel storage or spent fuel pool size construct these alterations and expan-would be subject to licensing review by sion. (docket no.59-201). The'second
' plant. General Electric Company's Liid.the Commission. In the event that a west Fuel Recovery Plant at Morris, II. particule.r on-site spent fuel pool should linois, has never operated and is in a~ beco:ne filled, and. no alternative form decommissicned conditio9. The -third of sxnt fuel storage sould be foun:1, o.
.~
ptettM t!GtSTit. VOL. 40, No.1Go -TUI5 cay, SEPTEMtic 16, 1975
JCTIC!.S 4 2602 mtnt as a suitsble veh!de'for such in.Co-%n had tro' bas!c chjectises'N tr e res ctor sw!d be eventually fcreed mind; en the coe hand, the ger.eri: im.:
. to ahut down sad " store" ine last spent..ezambation. Sot':e is hereby sten thata ger.eric e:Mrtnmental i=W:t state pa.ct stLte. ment abeuld no* acne as a jus.7 l
'.res.ctor f cel in the reacter pressure ves- ' ment on the handibg and storage of tis:atinn. fer. ~a f ait acco:npli: en. thei l
~sch.While no serious <dserse cense-
- c;ue5cas to the put!!: her.lth and s.fety.., spent Ught wtter po,s er reactor f uels will other hand,'the putu: interent con:id-I the ' common. defense e.nd accudt7, cr be prepared by the Cor: mission.. The crations a:.sce. lated with sud a defenal "the envitcn=ent wculd likely result, the statement will focus on the time period abou:d be carefully weighed. The Cc:n'i i
7 tea:tcr thutdown would, cf course, re-betw*en now r.nd the mid ISP.Fs and will mis:ic:: has cencluded that there should be no such general defenal, and.tp.ati i
', move the plant frc= service and this in.. address:
--(1) The magrJtude of the -poss!ble theu related licensing a: Mons may con- >
l m
. turn could adtersely a:Tect the elec*.rie shortage cf spent hel store.ce ca;Kity'.tbue during the period twuired ~foij j
-utitty's abitty to meet elecidcAl energy
'.nceds, or f orce the utinty to ope,ste other.
~(2) The alternatives for'denung with 3.regration cf. the, generk statement,.
subject to certain cond!'.'.ocs"In reath-
- yants that see less econo =1:a1 to cperate. the pecablem-including..but n:t neces-ing this cccdurJc the C0"+n has' f
cewhich have greater endren=entalim-r.ar'lylimited 2'
'; set. and thereby adic.rsely tae:t the..
(a) Permitting the expansion of spent considered be tcD: wing speci'. f a:tnru~
'pubbe inte.est.
c fud ste: age ca;scity at ;cwe' reacters, (1).It is litdy that each intridual 11-:
- . "'here a; pear to be a number cf p%- ~
(b) Permittmg the expa.nsion of sp"nt eensing action of this' type vonld have; sible ahernative: for increasing spert f uti stcrage capacity, at reproccMing a u:Chty that is inde;r*t cd the utO ty" f: : sacra;e cauc!:y inclutng, a=cng plante;
.of. ote: ll e.nsing actions of this' type; ctLcir_thtngs, increasbg the storage ca.
(c) Licensing of. Inde;c. dent spent C) It is not natly that tie IAhns c:t
-p *ity at pre.sent reactor &ites, eJ d C:n.
- f utl storage IaCilltie2; aLy pat 11 u.J.r li P+y a *.*:n Cf this
's: ceti:n of independent spent fuel (d) Etcrue' of spent fuel fic:n one type duttng the time frame under.ccs-storage f acitties. The shcitage cf sp nt. cr more reactors at,the storage ;oc.ls c! s;de ation would ccnstitute a commit.
' fuel storage capacity Mil occur at ind!.cther reseters; -
ment c! resources that' would tend -to the Ccmmiss.!cn.. (ehCrrdedng that generation of spent signiS:antly forecl:se "the s.!ternatives' Ividual reacters. s.nd could adequatdy address the issues in; " fuel (reactc; operati,en).be stcyped or *av&Dahle with respect to any other-in-S
' ;'... *dividnal Usepstng atten of this type:*)E
.~volved on 'a case-by-case 'hasis. Mihin festricted; Ithe context of indvidual licens!ng re-(3) A cost benedt anal'vsis of the al-(3) It is Ukely that any envir'onmentil; 7tiewi Indeed. the CommWien has not, *tematives Ested in (2), along with any impacts associated with any individuali to dAte,.found it necessary/in the dis-other reasonably -feasible 11ternatves, licensing action of this type would 'be' charge of its licensing and related tegu-including:
such that they could adequately be adC latory functions, to deve30p any overall * (a) Impacts on public health.and drened within the ecnte.rt of the indi-)
- program of action to deal with the prob.r.afety and. the common defense.and 'vidual Ucense application without overe Gem.The Commissien does, however, have. secudty;
- I..
- J.
. ~ '
loci :g any cumdative environmenta]*
Sthe dberet!:n to deal with issues of this..(b) Envitenmental social, and eco* impacts; A. ';4'.
r.,".% _, Q
((; n g3 gety; that' any t'echnist-gercise cf jts.ruJer:ating authodty and/
-(c)y ecsts and benefits; itype on-2 gened: basis through the ex-noml options. issues that tnay arise in the course of a~
C -Mtments of resources; er the Assuante of a *' generic'* entron-
-(d) 1.=plications regarding review of an indvidual license appiita -
t mental impact ' statement. Rulemaking availtble for the intermediate antions* Uon can be resolved within that cond 3
- proceedings and/or the issuance of a ' term storage of nuclear waste materials;. text; and,'
jgeneri: environmental i= pact state =ent
('e) Relationship between local short-(5) A deferr'al'cis'eSer'e' restrEdo pntght,as appropriate, serve as the ces -term uses of th environment and long. licensing acti0ns of this type would reS ftext'fer the prcmulgation of snora de-term productivity;c
. - sult in ubstantal htrm to the pubUcI
,y.n!tive ' criteria reguding str.e and de-(4) ne impacts of pce.stble add!tional interest. As indicated, such a restriction-%
- sign of spent fuel pools and/cr the 11- - transportauon cf spent.tuel that may or deferral cott!d result in reactor e. hut "
- c e+ r of independent spett fuel stera ge rbe required abould one or more of the den.s as exirting spent fuel pools become
. filled. It now appears that the spent (fs.cibtles, and for consideration cf pes-J' alternatives be adopted:-
. eible revisicn of the fuel cycle environ-
-(5)-More definitive standards and cri-~ fuel pools of 'as many as ten reactors' mental.1=pe.ets set f, orth in.10 CTR teria to govern thelicensing of cne or ~ could be filled by'm!d-1978. nese ten' rea:ters represent a tctal of about 6 mi!J
.4 51.:Ote) in light of adwtions.1 spent fuel f
ytorage and attendant transpcrtaticn. more of the alternat!ves for d ahng with lion tiowatts of electrical energy gen -
the problem; and..
"Praung capacitR *De rmoval of hi
, Also, the pos.sible implicatiens of in-g
- creased spent fuel storage on the eptions
,.j.,.c
- 4. v i reacters from serv!:e could reduce the
tfhties'se cE I:2.
to a point where*
s e uld " 'If apprhp'ifte.'ITEbng proc ~eediss r
se ce wo e n Mpardy, eq of nu et as ma within this. cn items '(5) and (6) listed above, or cet.J. o ce me uMMes to uly Incre heartly en)
N* eong,tablyi g. 4xamined fi be e J. other issues related to.the handling and k;iOne gro0Fof interes'ed organizatio:d 7 storage of spent reactor fuel.%in be ini, [
CC[ ftb bP L
2g, :-
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d
. Distural" Resources' Defense.;Conntil,r-tiated on o-about that time of issuance norrh penaltu ~en consumers and 1:fI
%erra :Clubf and Businessmen f or the. of the dro.Teneric,tn{ircnmental im-t.
- crease environme'nts.limpacts. '
- M rW Pub 11: Interest) has requested the Com-ptet statement. -
.w.-
"Ite Commission expects that a.ny b *
/mWien to prepare a generi: environme.u,. -EThe Commission his aho given careful. censing. action intended to amtHorate.a3 Jtalimpact statenent on the handling and h
. ettrage of spent reacter iurl and related_ oonsideradon.to the question whet er..poss!ble shortage cf spent f pe:1od2 matters fletter to L. V. Gessick from _ licensing actions" intended to ameliorate egp,eg;y. durinU this. ints
]
capacity, including such" actions t.s the. vould be accornpinled b) a possit:e shortage of spent fuel storage mental inbet statement (10 CTRU15 %
d nthpsy Z. Roisman. dated Mayy0.19{l4 5
A ais.nl (10 CTR I 515 i i
isst.ance cf operating Heense amend 'Na)) c; impact a~bh M d W h '
- copy on file at the Commissions Pub ments to permit increases in the ste-age g
l
~ Document Room,1717 R Street. hT..
Washington. D.C.)
- c. espacity ni reacter spent fuel pools or 'zg ggSin:e the Cn=mbden's general con:1nN WhSe the Comm!nics beneves. as er. reprocessing plant spent-fuel storage sions t,th respect to the f.ve fa: tors.*s'.
_ ter ind!cated; that the matter of spent -pools, cr the liceNng of iodependent set forth above. Inay not tt the factual apent fuel storage facilities, should be f
ti ular U:ensing ac.
_ fuel storage capacity can t.dequately benddreced on a case-by-case basts within. deferr-d pending cor:pletion of ti the centext of individual 1!: ens =g It-ne.ric endronmental impact state =ent.
weighed and balanced within the cen4,
. views, it also believes that, from the Such "a deferral was requested in the. tut o' the m'm*a or sm-dsah b -
stand;dnt of longer range Mbey, this letter en behalf of Natural Resourtes reaching licensing deter =.inatio~~ h t
~
ma.tter enn prodtably be exam %d in a Defense CouncH. S:erra Club. and Bus!-
broader centext It views the prepa.ratJon nessmen for the Pubhe Interest noted
. Dated at Weahington, D.C. this 10th-of a genene endronmentali= pact state-. secte. In considert g thi.s matter, the day of September 1975.
F(Oit A L etC'5112. VOL 4 0. NO I te--TU!50 AY 5t PT!*f tt 16. 1975
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7 UNITED ST A1LC l
NUCLE AR 8LECULATOR Y COMMffi.f 0N '
W ALHINGTON. D. C.
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