ASTM C1004-84(1992)e1
(Guide)Matrix for LWR Spent Fuel Receiving and Storage (Withdrawn 1999)
Matrix for LWR Spent Fuel Receiving and Storage (Withdrawn 1999)
SCOPE
1.1 This standard makes use of a matrix to identify existing standards and potentially needed standards for the receiving and storage of spent fuel.
1.2 This standard pertains to a facility for the receiving and interim storage of LWR spent fuel in either a wet or dry environment. While it is the intent of this standard matrix to apply to an independent storage facility, this standard is also applicable to a storage facility located at a reactor or reprocessing site.
1.3 This matrix is defined as an array of receiving and storage components as the horizontal axis and the functional activities as the vertical axis.
1.4 The terminology used for the components and functional activities of the matrix apply specifically to this matrix and are not intended to be universal definitions. Refer to Section 3.
1.5 Subcommittee E10.12 on Nuclear Safeguards, Task Group C26.13.08 on Waste Handling, and Subcommittee E10.03 on Decommissioning of Nuclear Facilities will deal in more detail with their respective subjects than is appropriate in this standard. These items are included in this matrix simply to preclude their oversight by a user of this standard.
1.6 While it is recognized that federal directives and guidelines are not national consensus standards, they have, nevertheless, been appended as source information in this standard. This is in keeping with the primary objective of the matrix, to identify function/component intersections for which there is a need for a consensus standard.
General Information
Standards Content (Sample)
ASTM Cl1004 84 m 0759530 0516223 979 -
AMERICAN S0C~El-f MR TESTlNG AND MATERIALS
Designation: C 1004 - 84 (Reapproved 1992)E’.
1916 Race St Philadelphia, Pa 19103
Reprinted from the Annual Book of ASTM Standards. Copyright ASTM
If not lkted in the current combined index, will appear in the neti editon.
Standard Matrix for
LWR Spent Fuel Receiving and Storage’
This standard is issued under the fixed designation C 1004: the number immediately following the designation indicates the year of
original adoption or. in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (3 indicates an editorial change since the last revision or reapproval.
tt NOTE-f%XtiOn 7, Keywords, was added editorially in November 1992.
1. Scope 3.1.1.2 building and sfr!rctures-facilities for housing the
primary and supportive equipment used in the receipt and
1.1 This standard makes use of a matrix to identify
storage of spent fuel.
existing standards and potentially needed standards for the
3.1.1.3 cask handling systems-structures and devices
receiving and storage of spent fuel.
used in lifting, transporting, cooling, cleaning, preparing, or
1.2 This standard pertains to a facility for the receiving
servicing spent fuel transportation packages,
and interim storage of LWR spent fuel in either a wet or dry
3.1.1.4 cask (transportation package) fleet servicing-the
environment. While it is the intent of this standard matrix to
physical activity, techniques, hardware, and software associ-
apply to an independent storage facility, this standard is also
ated with preventative and corrective maintenance (in-
applicable to a storage facility located at a reactor or
cluding transportation package compliance testing) required
reprocessing site.
for each spent fuel transportation package and associated
1.3 This matrix is defined as an array of receiving and
storage components as the horizontal axis and the functional ancillary equipment.
activities as the vertical axis. 3.1.1.5 cooldown or flush-the removal of heat or con-
1.4 The terminology used for the components and func- taminants from the transportation package internals in a
tional activities of the matrix apply specifically to this matrix manner that will permit the unloading of spent fuel in a wet
and are not intended to be universal definitions. Refer to system without detrimental effect on that system.
Section 3. 3.1-l .6 decontamination-the removal of radioactive ma-
1.5 Subcommittee ElO. 12 on Nuclear Safeguards, Task
terial from the surface of an object.
Group C26.13.08 on Waste Handling, and Subcommittee
3.1.1.7 disassembly and consolidation-the technique
E10.03 on Decommissioning of Nuclear Facilities will deal
and hardware used to remove fuel rods from a fuel assembly
in more detail with their respective subjects than is appro-
and to reconstitute those rods in some close-packed configu-
priate in this standard. These items are included in this
ration within a container.
matrix simply to preclude their oversight by a user of this
3.1.1.8 eJJltlent control-the provision of the necessary
standard.
equipment, technical expertise, and ongoing surveillance to
1.6 While it is recognized that federal directives and
limit each and every discharge from the facility into the
guidelines are not national consensus standards, they have,
environment to accepted, predetermined quantities and
nevertheless, been appended as source information in this
levels.
standard. This is in keeping with the primary objective of the
3.1.1.9 encapsulation-placing of an impermeable enve-
matrix, to identify function/component intersections for
lope around some subject item.
which there is a need for a consensus standard.
3.1.1.10 equipment (decontamination)-the equipment
and techniques for removing radioactive material from
2. Referenced Documents
surfaces, crevices, interstices, etc., of hardware and buildings
‘or structures.
2. I ASTM Standards:
3.1.1.11 Jirel assembly-a complete entity of rods, end
E 669 Master Matrix for Nuclear Fuel Cycle Standards2
fittings, spacers, etc., as discharged from a nuclear reactor.
E 909 Master Matrix for Development of Standards to
Assist in Safeguarding of Nuclear Materials2 3.1.1.12 fuel handling systems-devices used in lifting,
transporting, shipping, examining, restructuring, or encapsu-
3. Terminology
lating spent fuel.
3.1.1.13 fuel in cans-a fuel assembly or the product of
3.1 Descriptions of Terms SpeciJic to This Standard:
one or more disassembled fuel assemblies contained in a
3.1.1 Components:
metal envelope.
3.1.1.1 allxiliary systems-the physical installations that
3.1.1.14 handling devices-the hardware used for listing
complement fuel-storage pools.
casks, lids, etc.
3,l. 1.15 heat removal-the exchange of heat from the
t This matrix is under the jurisdiction of ASTM Committee C-26 on Nuclear
spent fuel to the pool water and then to a second medium
Fuel Cycle and is the direct responsibility of Subcommittee C26.13 on Standards
from which that heat can be discarded.
Matrix-Nuclear Fuel Cycles,
3.1.1.16 in/out preparation-the physical activity, tech-
Current edition approved Jan. 5, 1984. Published March 1984.
2 Annual Book ojASTM Standards, Vol 12.01,
niques, and hardware involved in readying a transportation
ASTM Cl1004 BLt W 0759530 0516224 805 M
package for its next operation. application to surfaces of coatings or fabrication finishes to
3.1.1.17 inspcjclion-the techniques of visual and physical achieve conditions of contamination control, corrosion pro-
examinations, the comparison of findings to specified re- tection, thermal absorption or emissivity, or any combina-
quirements. and the evaluating and reporting of the results tion of these criteria.
thcrcof. 3.1.2.7 nutwul phenomenu and lo,v-probabilir!) events-
the rare but severe occurrences of nature or acts of man that
3.1.1.18 instruments and controls-devices used to
monitor and to regulate facility operations. As applied to impose extreme conditions for which a facility must bc
designed with respect to safety.
radiation levels, these devices, when appropriate, shall cause
3.1.2.8 nuclear criticality analysis-the examination of
signals or alarms to be generated denoting the need for
the design and operation of the facility for the purpose of
corrective action.
3. I. I, 19 intrasite transport-the physical movement of reducing the probability of a nuclear criticality incident to
spent fuel assemblies between storage locations within the acceptable levels.
3.1.2.9 operating basis-those factors that must be con-
boundaries of a single site, such as from one storage location
sidered in the overall performance phase of a facility.
to another in a multiple-facility complex.
3.1.1.20 personnel (decontamination)-the equipment 3.1.2.10 personnel exposure protection-considerations
intended to limit or minimize the duration of radiation
and techniques for removing radioactive material from the
exposure to operators, or to maximize the distance or the
skin or clothing of personnel.
amount of shielding between operators and radiation
3.1.1.2 1 pools-water-filled storage basins for spent fuel
sources.
assemblies.
3.1.2. I 1 process (operating&a series of actions or opera-
3.1.1.22 racks-structures designed, built, and used exclu-
tions, the ultimate end of which is a product or service.
sivcly for supporting spent fuel assemblies during storage.
3.1.2.12 quality assurance-all those planned or system-
3.1.1.23 remote devices-equipment used with inter-
atic actions necessary to provide adequate confidence that a
vening radiation shielding between the operator and the
structure, system, or component will perform satisfactorily as
functioning location of the equipment.
designed.
3.1.1.24 safeguards-a system for nuclear material con-
3.1.2.13 shielding (structural)-those portions of the fa-
trol, accounting, physical protection including containment/
cility structure that are designed to provide radiation protec-
surveillance, and designs to allow inspection for compliance
tion to personnel and equipment.
to international/national or subnational codes, or both. (See
3.1.2.14 status monitoring (operating)-the continual
Matrix E 909.)
evaluation of safety system performance and the corrective
3.1.1.25 site-a physical location which can be described
measures appropriate to the elimination of nonconformance.
or characterized by factors such as local topography, hy-
3.1.2.15 surveillance
inspection (and maintenance)
drology, demography, geology, meterology, etc.
AURA-the ongoing process of evaluating the systems and
3.1.1.26 storage systems-facilities which are designed,
techniques controlling radiological exposure to person&
built, and used to contain, in a safe and secure manner, spent
and environs, toward the end of maintaining exposure as low
fuel assemblies.
as reasonably achievable.
3. I. 1.27 rvasle hand/ing-the identification, collection,
3.1.2.16 thermal analysis-examination of the systems in
assaying, isolation, preparation, transportation, and disposal
the facility for the purpose of assuring adequate temperature
of radioactive wastes.
control and heat dissipation.
3. I, I .28 rvu/cr klrr$cution-the removal (from pool
3.1.2.17 training and administration (operating)-a for-
water) of particulate matter and ions, especially those that
malized program for education of employees and execution
are radioactive.
of duties for control of the operations.
3.1.2 Functions:
3.1.2.1 decommissioning-those actions required to re-
4. Significance and Use
move a facility from service and reduce residual radioactivity
4.1 In the nuclear fuel cycle, spent fuel eventually must
to a level that permits release of the property for unrestricted
leave the reactor in which it was used and be transported to
use and termination of license.
an installation for interim storage in either a wet or dry
3.1.2.2 design basis-those factors that must be consid-
environment. Considerations for proper receiving and stor-
ered in the overall planning, construction, operation, main-
age must be addressed. Many consensus standards exist that
tenance, and decommissioning of the facility.
are useful in this respect. This matrix is intended to identify
3.1.2.3 fabrication/construction-the physical assembly of
needed standards.
the buildings and equipment that functions as an entity in
the performance of a service or the production of goods.
,5. Matrix for LWR Spent Fuel Receiving and Storage
3, I .2.4 getwral layotrr and arrangement-the design phase
5.1 Table I is the matrix developed for LWR Spent Fuel
that governs personnel placement and travel and material
Receiving and Storage as defined in Master Matrix E 669
flow with respect to considerations such as minimizing
and is designed to identify needed standards (Table 2) as
radiation exposure and general operational efficiencies.
described in Section 6.
3.1.2.5 inte&ces-areas of overlap or junction between
5.2 Needed standards identified in Table 2 are considered
those systems designed as a part of the facility, and the
to be of equal priority and warrant early action.
transportation packages and fuel systems which will be
6. Summary of the Matrix
supplied by others.
3.1.2.6 material coatings and finishes-the selection and
6.1 The matrix, Table 1, was completed to achieve the
ASTM CL004 BY - 0759510 0536225 74L -
4l# Cl004
TABLE 1 LWR Spent Fuel Receiving and Storage
-
1 .o 2.0 3.0 4.0 5 0 Storage 6.0
STATUS
Cask Handlino Swlems Fuel Handling Svslems Syslams Auxiliary Syslems
r
rl NOI Applicable
r
n Standards Exlsl E
Slandards Needed E
cl
ExlslfNee&l i
FUNCTIONS
a. Interlaces
b. Nucledr Crllicalily
Analysis
c. Shiilding
d. Personnel Exposure
e. Thermal Analysis
I. Malerials Coalings
Finishes
g. Nalural Phenomewm
and low Probabilily
h. Gewmmissloning
i. General Layout
and Arrangement
B. FablConstruclion
c. 1
a. Pmcess
id
b. lfaminy and
g Admmislfalion
-
e
X c. Slalus
Momlormg
- I.-- .-.--- _._^_____
II U~~dhlyAssu~.mcu
t Sulvallldnce lnspecllall
alld Mamlelldnce (ALARA)
ASTM Cl1004 84 - 0759530 0536226 688 -
4g-l Cl004
TABLE 2 Needed Standards
Table 1 NfXd Suggested
Title
Scope
Coordinates Criteria Secretariat
AJ3.5 Cask Fleet Spent fuel shippino casks routinely are serviced in facilities not soecificallv desianed ANSI N.14
1,4,5
Servicing for this work. This may result h inefficfencles in the servlclng’ operation. A more
logical location for servicing would be at a fuel receiving Installation, to which
relatively large numbers of casks woukf be sent during routine use. Such a
facility could provide special equipment end trained personnel for servicing casks
more quickly, more reliably, and at less cost than present servicing provides.
The facility, equipment needs, records system, personnel dlsclplines, and the
scope of testing procedures need to be set forth In a standard for cask fleet
servicing.
A./43 Spent Fuel The advantage of spent fuel disassembly and consolidation can be seen In terms of ANS-55
1, 2, 4, 5
Disassembly and doubled storage density, increased transportation payloads potential for cleaner
Consolidation storage pools, etc.; however, the techniques of these operations are not widely
known. Standards are needed for disassembly (both underwater and in hot cells)
and for consolidation in forms that will interface with storage pools,
transportation casks, and terminal storage facilities.
A.144 Encapsulation Spent fuel that is destined for terminal storage should be encapsulated in ANS-55
1, 2, 4, 5
accordance with the needs of the particular storage medium. These media need
to be defined and the conditions of encapsulation specified as a function of
media and time.
A.146 lntrasite Transport This need is unique to storage facilities which are colocated with existing reactor ANSI N.14
1.4,5
sites. Casks are needed to transport spent fuel from the reactor storage basins
to the longer term storage facility that is separate from the reactor but still within
the same site boundary. Commercial spent fuel shlppfng casks, licensed for off-
site use, can be used for this purpose. A cask more simple to handle and with a
greater payload to gross-weight ratlo should satisfy this need and still satisfy 10
CFR Part 50. A standard for casks of this category would be useful.
C.bJ3.0 and 4.0 Personnel Uniform industry standards defining the personnel quaflficatlon requirements ANS-3
1, 2, 3
Qualification associated with cask- and fuel-handling system operation are not curr
...
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