[Federal Register Volume 71, Number 221 (Thursday, November 16, 2006)]
[Rules and Regulations]
[Pages 66648-66657]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-19372]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AH95
Criticality Control of Fuel Within Dry Storage Casks or
Transportation Packages in a Spent Fuel Pool
AGENCY: Nuclear Regulatory Commission.
ACTION: Direct final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations that govern domestic licensing of production and
utilization facilities so that the requirements governing criticality
control for spent fuel pool storage racks do not apply to the fuel
within a spent fuel transportation package or storage cask when a
package or cask is in a spent fuel pool. These packages and casks are
subject to separate criticality control requirements. This action is
necessary to avoid applying two different sets of criticality control
requirements to fuel within a package or cask in a spent fuel pool.
DATES: Effective Date: The final rule will become effective January 30,
2007, unless significant adverse comments are received by December 18,
2006. A significant adverse comment is a comment where the commenter
explains why the rule would be inappropriate, including challenges to
the rule's underlying premise or approach, or would be ineffective or
unacceptable without a change (refer to ``Procedural Background'' in
the Supplementary Information section of this document for further
details). If the rule is withdrawn, timely notice will be published in
the Federal Register. Comments received after December 18, 2006 will be
considered if it is practical to do so, but the NRC is able to ensure
only that comments received on or before this date will be considered.
ADDRESSES: You may submit comments by any one of the following methods.
Please include the following number RIN 3150-AH95 in the subject line
of your comments. Comments on rulemakings submitted in writing or in
electronic form will be made available for public inspection. Because
your comments will not be edited to remove any identifying or contact
information, the NRC cautions you against including personal
information such as social security numbers and birth dates in your
submission.
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
E-mail comments to: [email protected]. If you do not receive a reply e-
mail confirming that we have received your comments, contact us
directly at (301) 415-1966. You may also submit comments via the NRC's
rulemaking Web site at http://ruleforum.llnl.gov. Address questions
about our rulemaking website to Carol Gallagher at (301) 415-5905; e-
mail [email protected]. Comments can also be submitted via the Federal
eRulemaking Portal http://www.regulations.gov.
Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland
20852, between 7:30 a.m. and 4:15 p.m. Federal workdays [telephone
(301) 415-1966].
Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at
(301) 415-1101.
Publicly available documents related to this rulemaking may be
viewed electronically on the public computers located at the NRC's
Public Document
[[Page 66649]]
Room (PDR), O-1F21, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852. The PDR reproduction contractor will copy
documents for a fee. Selected documents, including comments, can be
viewed and downloaded electronically via the NRC rulemaking Web site at
http://ruleforum.llnl.gov.
Publicly available documents created or received at the NRC after
November 1, 1999, are available electronically at the NRC's Electronic
Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this
site, the public can gain entry into the NRC's Agencywide Document
Access and Management System (ADAMS), which provides text and image
files of NRC's public documents. If you do not have access to ADAMS or
if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1-800-397-4209, 301-415-4737, or by
e-mail to [email protected].
FOR FURTHER INFORMATION CONTACT: George M. Tartal, Project Manager,
Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, telephone (301) 415-0016, e-mail
[email protected].
SUPPLEMENTARY INFORMATION:
I. Background
Storage of spent fuel can be done safely in a water filled spent
fuel pool under 10 CFR Part 50, a transportation package under 10 CFR
Part 71, or a dry storage cask under 10 CFR Part 72. The primary
technical challenges involve removing the heat generated by the spent
fuel (decay heat), storing the fuel in an arrangement that avoids an
accidental criticality, and providing radiation shielding. Removing the
decay heat keeps the spent fuel from becoming damaged due to excessive
heatup. Transportation packages and dry storage casks are designed to
be capable of removing the decay heat generated by the fuel when filled
with water or when dry without the need for active heat removal
systems. Avoiding an accidental criticality is important to preclude
the possibility of overheating the spent fuel and damaging the fuel.
When dry, transportation packages and dry storage casks are subcritical
by the absence of water as a neutron moderator, as well as by geometric
design, and through the use of neutron poison materials such as boral
and poison plates. When the packages and casks are flooded with water,
they may also rely on soluble boron to maintain the subcritical
condition. Therefore, a boron dilution event is the scenario that could
result in an accidental criticality with the possibility of excessive
fuel temperature and subsequent fuel damage. Radiation shielding,
provided by the water in a spent fuel pool or the container material in
a transportation package or dry storage cask, is important to protect
people that may be near the spent fuel from unacceptable exposure to
radiation. The NRC has promulgated regulations governing the capability
of both spent fuel pools (10 CFR Parts 50 and 70), dry storage casks
(10 CFR Part 72) and transportation packages (10 CFR Part 71) to
address these technical challenges for the protection of public health
and safety.
10 CFR 50.68 requires that spent fuel pools remain subcritical in
an unborated, maximum moderation condition. Implementation of this
regulation also allows credit for the operating history of the fuel
(fuel burnup) when analyzing the storage configuration of the spent
fuel. 10 CFR Parts 71 and 72 approve the use of spent fuel
transportation packages and storage casks, respectively. 10 CFR Part 71
requires that transportation packages be designed assuming they can be
flooded with fresh water (unborated), and thus are already analyzed in
a manner that complies with the 10 CFR 50.68 assumption. However, 10
CFR Part 72 was, in part, predicated on the assumption that spent fuel
(without any burnup) would remain subcritical when stored dry in a cask
and remain subcritical when placed in a cask in a spent fuel pool at a
commercial power reactor. Implementation of 10 CFR Part 72 relies on
soluble boron, rather than on burnup, to assure subcriticality when the
fuel is in a cask in a spent fuel pool.
On March 23, 2005, the NRC issued Regulatory Issue Summary (RIS)
2005-05 addressing spent fuel criticality analyses for spent fuel pools
under 10 CFR 50.68 and Independent Spent Fuel Storage Installations
(ISFSI) under 10 CFR Part 72. The intent of the RIS was to advise
reactor licensees that they must meet both the requirements of 10 CFR
50.68 and 10 CFR Part 72 with respect to subcriticality during storage
cask loading in spent fuel pools. The need to meet both regulations and
the differences in the assumptions described above create an additional
burden on licensees to show that credit for soluble boron is not
required to preclude an accidental criticality in a water-filled, high-
density dry storage cask used for storing fuel. In order to satisfy
both of these requirements, a site-specific analysis that demonstrates
that the casks would remain subcritical for the specific irradiated
fuel loading planned, without credit for soluble boron, as described in
10 CFR 50.68 is required. This analysis relies on the fuel burnup to
determine the margin to criticality for the specific cask loading. The
analysis is similar to that conducted for the spent fuel pool itself,
but takes into account the unique design features of the cask when
determining the minimum burnup required for spent fuel storage in the
specific cask. This issue only applies to pressurized water reactors
(PWR) because boiling water reactor (BWR) spent fuel pools do not
contain soluble boron and the casks that are used to load BWR fuel do
not rely on soluble boron to maintain subcriticality.
The regulations, as currently written, create an unnecessary burden
for both industry and the NRC, of performing two different analyses
with two different sets of assumptions for the purpose of preventing a
criticality accident, with no associated safety benefit. This burden is
considered unnecessary because the conditions which could dilute the
boron concentration within a transportation package or dry storage cask
(hereinafter ``package or cask'') in a spent fuel pool, and cause fuel
damage with the release of radioactive material, are highly unlikely.
The NRC evaluated the two scenarios in which a boron dilution could
occur: (1) A rapid drain down and subsequent reflood of the spent fuel
pool, or (2) a slow boron dilution of the spent fuel pool. The result
of the NRC evaluation is that the possibility of each scenario is
highly unlikely (see Appendix A for additional details). Therefore,
there is no safety benefit from requiring the licensee to conduct a
site specific analysis to comply with 10 CFR 50.68(b) while fuel is
within a package or cask in a spent fuel pool.
As a result, a revision to the Commission's regulations is
necessary to eliminate the requirement for separate criticality
analyses using different methodologies and acceptance criteria for fuel
within a package or cask in a spent fuel pool. This direct final rule
will eliminate the need to comply with the criticality control
requirements in Sec. 50.68 if fuel is within a package or cask in a
spent fuel pool. Instead, the criticality requirements of 10 CFR Parts
71 and 72, as applicable, would apply to fuel within packages and casks
in a spent fuel pool. For fuel in the spent fuel pool but outside the
package or cask, the criticality requirements of 10 CFR 50.68 would
apply.
[[Page 66650]]
II. Section-by-Section Analysis of Substantive Changes
Section 50.68 Criticality Accident Requirements
Section 50.68 describes the requirements for maintaining
subcriticality of fuel assemblies in the spent fuel pool. New paragraph
(c) of this section states that the criticality accident requirements
of 10 CFR 50.68(b) do not apply to fuel within a package or cask in a
spent fuel pool. Rather, the criticality accident requirements of 10
CFR Part 71 or 72, as applicable, apply to fuel within a package or
cask in a spent fuel pool. This new paragraph provides the regulatory
boundary between Sec. 50.68(b) and 10 CFR Part 71 or 72 for performing
criticality analyses. A licensee moving fuel between the spent fuel
pool and a package or cask need only analyze fuel within the package or
cask according to 10 CFR Part 71 or 72, as applicable, and is not
required to analyze fuel within the package or cask using Sec.
50.68(b) requirements.
For the purpose of this paragraph, any package or cask that is in
contact with the water in a spent fuel pool is considered ``in'' the
spent fuel pool. Also, once any portion of the fuel (fuel assembly,
fuel bundle, fuel pin, or other device containing fuel) enters the
physical boundary of the package or cask, that fuel is considered
``within'' that package or cask. When a package or cask is in a spent
fuel pool, the criticality requirements of 10 CFR Part 71 or 72, as
applicable, and the requirements of the Certificate of Compliance for
that package or cask, apply to the fuel within that package or cask.
Criticality analysis for the fuel in that package or cask in accordance
with Sec. 50.68(b) is not required. For fuel in the spent fuel pool
and not within a package or cask, the criticality requirements of Sec.
50.68(b) apply.
III. Procedural Background
The NRC is using the ``direct final rule procedure'' to issue this
amendment because it is not expected to be controversial. The amendment
to the rule will become effective on January 30, 2007. However, if the
NRC receives significant adverse comments by December 18, 2006, then
the NRC will publish a document that withdraws this action. In that
event, the comments received in response to this amendment would then
be considered as comments on the companion proposed rule published
elsewhere in this Federal Register, and the comments will be addressed
in a later final rule based on that proposed rule. Unless the
modifications to the proposed rule are significant enough to require
that it be republished as a proposed rule, the NRC will not initiate a
second comment period on this action.
A significant adverse comment is a comment where the commenter
explains why the rule would be inappropriate, including challenges to
the rule's underlying premise or approach, or would be ineffective or
unacceptable without a change. A comment is adverse and significant if:
(1) The comment opposes the rule and provides a reason sufficient
to require a substantive response in a notice-and-comment process. For
example, a substantive response is required when:
(a) The comment causes the NRC to reevaluate (or reconsider) its
position or conduct additional analysis;
(b) The comment raises an issue serious enough to warrant a
substantive response to clarify or complete the record; or
(c) The comment raises a relevant issue that was not previously
addressed or considered by the NRC.
(2) The comment proposes a change or an addition to the rule, and
it is apparent that the rule would be ineffective or unacceptable
without incorporation of the change or addition.
(3) The comment causes the NRC to make a change (other than
editorial) to the rule.
IV. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995 (Pub.
L. 104-113) requires that Federal agencies use technical standards that
are developed or adopted by voluntary consensus standards bodies unless
the use of such a standard is inconsistent with applicable law or
otherwise impractical. This direct final rule eliminates duplication of
criticality control requirements for fuel within a package or cask in
the spent fuel pool. These packages and casks have separate
requirements for criticality control during loading, storage and
unloading operations. This rulemaking does not involve the
establishment or use of technical standards, and hence this act does
not apply to this direct final rule.
V. Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs'' approved by the NRC on June 30, 1997, and
published in the Federal Register on September 3, 1997 (62 FR 46517),
this rule is classified as Compatibility Category ``NRC.''
Compatibility is not required for Category ``NRC'' regulations. The NRC
program elements in this category are those that relate directly to
areas of regulation reserved to the NRC by the Atomic Energy Act of
1954, as amended (AEA), or the provisions of Title 10 of the Code of
Federal Regulations. Although an Agreement State may not adopt program
elements reserved to NRC, it may wish to inform its licensees of
certain requirements via a mechanism that is consistent with the
particular State's administrative procedure laws but does not confer
regulatory authority on the State.
VI. Plain Language
The Presidential Memorandum dated June 1, 1998, entitled ``Plain
Language in Government Writing,'' directed that the Government's
writing be in plain language. The NRC requests comments on this direct
final rule specifically with respect to the clarity and effectiveness
of the language used. Comments should be sent to the address listed
under the heading ADDRESSES above.
VII. Finding of No Significant Environmental Impact: Environmental
Assessment
The NRC has determined under the National Environmental Policy Act
of 1969, as amended, and the NRC's regulations in Subpart A of 10 CFR
Part 51, that this rule is not a major Federal action significantly
affecting the quality of the human environment and, therefore, an
environmental impact statement is not required. The basis for this
determination is set forth below.
This direct final rule eliminates duplication of criticality
control requirements for fuel within a package or cask in the spent
fuel pool. These packages and casks are required to meet the licensing
requirements, defined in 10 CFR Part 71 or 72, as applicable, and the
applicable Certificate of Compliance (CoC), which currently provide
criticality control requirements for fuel loading, storage and
unloading. This rulemaking will preclude the necessity for nuclear
power plant licensees to meet the criticality control requirements for
both regulations (for 10 CFR Part 50 and for 10 CFR Part 71 or 72)
while fuel is within a package or cask in a spent fuel pool. The
regulations in 10 CFR Parts 71 and 72, as applicable, coupled with the
package or cask CoC, provide adequate assurance that there are no
inadvertent criticality events while fuel is within a package or cask
in a spent fuel pool. Experience over 20 years has demonstrated that
the regulations in 10 CFR Parts 71 and 72 have been effective in
preventing inadvertent criticality events, and the NRC concludes that
as a matter of regulatory efficiency, there is
[[Page 66651]]
no purpose to requiring licensees to apply for and obtain exemptions
from requirements of Sec. 50.68(b) if they adhere to the regulations
in 10 CFR Part 71 or 72 as applicable. Since the regulations in 10 CFR
Parts 71 and 72 and the CoC provide safe and effective methods for
preventing inadvertent criticality events in nuclear power plants, the
NRC concludes that this direct final rule will not have any significant
impact on the quality of the human environment. Therefore, an
environmental impact statement has not been prepared for this direct
final rule.
The foregoing constitutes the environmental assessment for this
direct final rule.
VIII. Paperwork Reduction Act Statement
This direct final rule does not contain a new or amended
information collection requirement subject to the Paperwork Reduction
Act of 1995 (44 U.S.C. 3501 et seq.). Existing requirements were
approved by the Office of Management and Budget, Approval Number 3150-
0011, 3150-0008 and 3150-0132.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
IX. Regulatory Analysis
Statement of the Problem and Objectives
As described in the Background section of this document, the need
to meet the criticality accident requirements of 10 CFR 50.68 and of 10
CFR Part 71 or 72, and the differences in their assumptions, create an
additional burden on licensees to show that credit for soluble boron is
not required to preclude an accidental criticality in a water-filled
package for transporting fuel or a water-filled, high-density dry
storage cask used for storing fuel. In order to satisfy both of these
requirements, a site-specific analysis that demonstrates that the fuel
in the package or cask would remain subcritical for the specific
irradiated fuel loading planned, without credit for soluble boron,
would be required. In the Sec. 50.68 analysis, the licensee would rely
on the fuel burnup to determine the margin to criticality for the
specific package or cask loading. The Sec. 50.68 analysis would be
similar to that conducted for the spent fuel pool itself, but would
take into account the unique design features of the package or cask
when determining the minimum burnup required for spent fuel storage in
the specific package or cask. This issue only applies to PWRs because
BWR spent fuel pools do not contain soluble boron and the packages and
casks that are used to load BWR fuel do not rely on soluble boron to
maintain subcriticality. As currently written, these regulations create
an unnecessary burden for both industry and the NRC with no associated
safety benefit.
The objective of this rulemaking activity is to revise 10 CFR 50.68
to eliminate the requirement for licensees to perform a separate
criticality analysis based on the requirements of 10 CFR 50.68 for fuel
within a package or cask in a spent fuel pool. As a result, any fuel
that is in the spent fuel pool and not within the physical boundary of
a package or cask remains subject to the criticality requirements of
Sec. 50.68. Once the fuel enters the physical boundary of the package
or cask, it is then subject to the criticality requirements of 10 CFR
Part 71 or 72, as applicable, and no longer subject to the criticality
requirements of Sec. 50.68.
Alternative Approaches and Their Values and Impacts
Another option to this amendment is for the NRC to make no changes
and allow the licensees to continue requesting exemptions. If no
changes are made, the licensees will continue to incur the costs of
submitting exemptions (approximately $300k) and NRC will incur the
costs of reviewing them (approximately $150k). Under this rule, an
easing of the burden on licensees results from not having to request
exemptions. Similarly, the NRC's burden will be reduced by avoiding the
need to review and evaluate these exemption requests. Another downfall
to this option is that licensees may not apply 10 CFR 50.59 to
exemptions, instead necessitating a new exemption for future
modifications to package or cask design. Furthermore, licensees would
not be in compliance with existing regulations, and that the NRC would
then be regulating by exemption rather than by rule.
A final option is for the NRC to make no change and licensees to
request a license amendment to add a Technical Specification which
restricts the burnup of spent fuel assemblies loaded into the package
or cask. This license amendment would only be required once, putting
the licensee into compliance with NRC regulations, and would then
permit licensees to make modifications using 10 CFR 50.59. However, the
burden of producing and approving an amendment on both the licensee
(approximately $300k) and the NRC (approximately $100k) is quite
significant, with no safety benefit.
Decision Rationale for the Selected Regulatory Action
Based on the evaluation of values and impacts of the alternative
approaches, the NRC has decided to revise 10 CFR 50.68 to eliminate the
requirement for licensees to perform a separate criticality analysis
based on the requirements of 10 CFR 50.68 for fuel within a package or
cask in a spent fuel pool. This rule revision is an easing of burden
action which results in increased regulatory efficiency. The rule does
not impose any additional costs on existing licensees and has no
negative impact on public health and safety. The rule will provide
savings to licensees that transfer fuel from the spent fuel pool to a
dry storage cask or transportation package. There will also be savings
in resources to the NRC as well.
X. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the
NRC certifies that this rule does not have a significant economic
impact on a substantial number of small entities. This direct final
rule affects only the licensing and operation of nuclear power plants.
The companies that own these plants do not fall within the scope of the
definition of ``small entities'' set forth in the Regulatory
Flexibility Act or the Small Business Size Standards set out in
regulations issued by the Small Business Administration at 10 CFR
2.810.
XI. Backfit Analysis
The NRC has determined that the backfit rule does not apply to this
direct final rule because this amendment does not involve any
provisions that would impose backfits as defined in 10 CFR 50.109.
Reactor licensees are currently required to meet both the requirements
of 10 CFR 50.68 and 10 CFR Part 71 or 72, as applicable, with respect
to subcriticality during package or cask loading or unloading in spent
fuel pools. The need to meet both regulations creates an additional
burden on licensees to show that credit for soluble boron is not
required to preclude an accidental criticality in a package or cask
when filled with water. In order to satisfy both of these requirements,
a site specific analysis that demonstrates that the fuel in the package
or cask would remain subcritical for the specific irradiated
[[Page 66652]]
fuel loading planned, without credit for boron, would be required. This
action amends 10 CFR 50.68 so that the criticality accident
requirements for spent fuel pool storage racks do not apply to the fuel
within a package or cask in a spent fuel pool. This rule constitutes a
voluntary relaxation of requirements, and as a result, a backfit
analysis is not required.
During the 535th meeting of the Advisory Committee for Reactor
Safeguards on September 7, 2006, a concern was raised regarding any
actions that would be required for licensees who have previously
requested and been granted either: (1) a license amendment to modify
the plant technical specifications to comply with the criticality
accident requirements of 10 CFR 50.68 for fuel in a 10 CFR Part 72
licensed cask in their spent fuel pool, or (2) an exemption from the
criticality accident requirements of 10 CFR 50.68 for fuel in a 10 CFR
Part 72 licensed cask in their spent fuel pool. The NRC position is
that this rulemaking activity does not constitute a backfit. The
following discussion in the Backfit Analysis clarify this NRC position
for the amendment or exemption cases described above.
For licensees with an approved license amendment, no action is
required by the licensee. The license amendment modified the licensee's
10 CFR Part 50 technical specifications by adding minimum fuel burnup
limits to the fuel being loaded into a licensed dry storage cask. This
direct final rule does not affect the licensee's ability to load spent
fuel into the cask in accordance with the amended technical
specifications, nor does it create any conflict with the amended
technical specifications. Therefore, a licensee may choose to continue
to comply with the requirements of their amended 10 CFR Part 50 license
and with the requirements of 10 CFR Part 71 or Part 72, as applicable,
while loading or unloading a package or cask in the spent fuel pool.
However, for those licensees who have amended their 10 CFR Part 50
license to comply with 10 CFR 50.68 and have included minimum fuel
burnup limits, and choose to take advantage of this voluntary
relaxation of requirements, they must request removal of the previously
amended portions of the 10 CFR Part 50 technical specifications as a
conforming change consistent with the amended rule.
For licensees with an approved exemption, no action is required by
the licensee. The exemption permitted licensees to be exempt from the
criticality accident requirements of 10 CFR 50.68 for fuel being loaded
into a licensed dry storage cask. These licensees can continue
operating under their approved exemption. However, a licensee may
instead choose to comply with the amended rule. Operating under the
exemption or the amended rule have effectively the same criticality
accident requirements for fuel within a package or cask in a spent fuel
pool, namely only those of 10 CFR Part 71 or Part 72, as applicable.
XII. Congressional Review Act
In accordance with the Congressional Review Act of 1996, the NRC
has determined that this action is not a major rule and has verified
this determination with the Office of Information and Regulatory
Affairs, Office of Management and Budget.
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
0
For the reasons set forth in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
1. The authority citation for part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note). Section 50.7 also
issued under Pub. L. 95-601, sec. 10, 92 Stat. 2951 (42 U.S.C.
5841). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955,
as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-190, 83
Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 50.103
also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C.
2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under
sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and
Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853
(42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec.
204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and
50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C.
2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42
U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68
Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under
sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
0
2. Section 50.68 is amended by adding a new paragraph (c) to read as
follows:
Sec. 50.68 Criticality accident requirements.
* * * * *
(c) While a spent fuel transportation package approved under Part
71 of this chapter or spent fuel storage cask approved under Part 72 of
this chapter is in the spent fuel pool:
(1) The requirements in Sec. 50.68(b) do not apply to the fuel
located within that package or cask; and
(2) The requirements in Part 71 or 72 of this chapter, as
applicable, and the requirements of the Certificate of Compliance for
that package or cask, apply to the fuel within that package or cask.
Dated at Rockville, Maryland, this 31st day of October, 2006.
For the Nuclear Regulatory Commission.
William F. Kane,
Deputy Executive Director for Reactor and Preparedness Programs Office
of the Executive Director for Operations.
Note: This Appendix will not appear in the Code of Federal
Regulations.
Appendix A: Technical Basis Document for RIN 3150-AH95 (RN 678)
I. Background
In the production of electricity from commercial power reactors,
spent fuel that is generated needs to be stored and safely managed.
As part of the design of all commercial power reactors, spent fuel
storage pools (SFP) were included to provide for the safe storage of
spent fuel for a number of years. For many years there was
sufficient room in the original spent fuel pools to continually
store spent fuel without space restrictions being an immediate
concern. In the 1960's and 1970's, when the spent fuel pools
currently in use were designed and built, it was anticipated that
the spent fuel would be moved off the reactor site for further
processing and/or permanent disposal. The planned long-term approach
is for disposal of this spent fuel in a permanent geological
repository.
As delays were encountered with the development of the permanent
geological disposal site, the spent fuel pools began to fill up and
space restrictions became a concern. Since the 1970's licensees,
with NRC approval, have increased the storage capacity of the spent
fuel pools by changing the designs of the storage racks to allow the
fuel to be safely stored closer together. This was recognized as a
short term solution, with the assumption that permanent disposal
would be made available within a reasonable period. As additional
delays were encountered with the permanent geological disposal of
the spent fuel, the nuclear power industry, in conjunction with the
NRC,
[[Page 66653]]
developed alternative storage solutions, including storing the spent
fuel in dry storage casks on their sites.
Maintaining the capacity to store spent fuel in a spent fuel
pool is important for safety. Being able to store the spent fuel in
a water filled spent fuel pool allows the fuel that is removed from
the reactor core at the start of a refueling outage to be safely
cooled at the time it is generating the greatest decay heat. Also,
the water provides shielding for the workers involved in conducting
maintenance on the various systems and components necessary to
safely operate the reactor. During a refueling outage, inspection
and maintenance activities need to be performed on the systems and
components that would normally protect the fuel from damage as a
result of the operation of the reactor. These inspections and
maintenance activities can be accomplished more effectively and
efficiently by draining the water from the reactor coolant and other
supporting systems. Placing the fuel assemblies in the spent fuel
pool during this period allows the reactor coolant and other systems
to be drained while keeping the spent fuel safe (covered with
water). Therefore, it is important to maintain the capability to
completely remove all of the fuel assemblies from the reactor vessel
during a refueling outage (full core offload capability). From an
operational perspective, additional capacity should be maintained to
accommodate a full core offload as well as the storage of new fuel
that replaces the spent fuel permanently removed from the reactor
core.
Storage of spent fuel can be done safely in a water filled spent
fuel pool under 10 CFR Part 50, a transportation package under 10
CFR Part 71, or a dry storage cask under 10 CFR Part 72. The primary
technical challenges involve removing the heat generated by the
spent fuel (decay heat), storing the fuel in an arrangement that
avoids an accidental criticality, and providing radiation shielding.
Removing the decay heat keeps the spent fuel from becoming damaged
due to excessive heatup. Dry storage casks are designed to be
capable of removing the decay heat generated by the fuel when filled
with water or when dry without the need for active heat removal
systems. Avoiding an accidental criticality is important to preclude
the possibility of overheating the spent fuel and damaging the fuel.
When dry, casks are subcritical by the absence of water as a neutron
moderator, as well as by geometric design, and for some cask designs
through the use of neutron poison materials such as boral and poison
plates. When the casks are flooded with water, they may also rely on
soluble boron to maintain the subcritical condition. Therefore, a
boron dilution event is the scenario that could result in an
accidental criticality with the possibility of excessive fuel
temperature and subsequent fuel damage. Radiation shielding,
provided by the water in a spent fuel pool or the container material
in a dry storage cask, is important to protect people that may be
near the spent fuel from unacceptable exposure to radiation. The NRC
has promulgated regulations governing the capability of both spent
fuel pools (10 CFR Parts 50 and 70), dry storage casks (10 CFR Part
72) and transportation packages (10 CFR Part 71) to address these
technical challenges for the protection of public health and safety.
Since the original design of commercial reactors included spent
fuel pools, the spent fuel is stored in these pools when it
initially comes out of the reactor. Decay heat from this spent fuel
is primarily produced by the radioactive decay of fission products
generated during the period the fuel is in the reactor core. As the
fission products decay, the amount of decay heat generated in the
spent fuel also decreases. So, over time the spent fuel becomes
cooler, requiring less heat removal capability. Since the decay heat
is higher when the spent fuel is removed from the reactor, it is
more efficient to cool the fuel in a spent fuel pool where the fuel
is surrounded by water. This allows the heat to be transferred to
the water in the pool. The spent fuel pool requires a dedicated
cooling system to maintain the temperature of the water in the pool
cool enough to prevent the water from boiling. The spent fuel is
allowed to cool down in the spent fuel pool for several years before
it is placed in a dry cask storage cask or transportation package.
When placed in a dry storage cask or transportation package, the
amount of heat generated by the spent fuel is low enough that the
fuel can be cooled by the gas surrounding the fuel with the heat
being transferred through the cask or package to the surrounding
air. Once placed in the dry storage cask or transportation package,
the fuel will remain cool enough to prevent fuel damage without the
need for an auxiliary cooling system.
Spent fuel pools, dry storage casks and transportation packages
are designed to preclude an accidental criticality primarily by
relying on the geometrical configuration of how the spent fuel is
stored. Both wet and dry storage may rely on material that absorbs
the neutrons necessary for the fission process to occur (fixed
neutron poisons, such as boral, poison plates, etc.). This material
is inserted when building the storage racks or when building the
cask/package. This material is integral to the storage racks in the
spent fuel pool and in the cask/package used to physically hold the
spent fuel in place. This establishes the geometrical configuration
of how the spent fuel is stored. Criticality is of a greater concern
when the fuel is stored in a spent fuel pool because the water used
to cool the fuel is also a very effective moderator that facilitates
the nuclear fission process. In dry storage, the spent fuel is
surrounded by a gas that does not act as a moderator, therefore,
criticality is a significantly smaller concern and the spent fuel
can be safely stored closer together than in a spent fuel pool.
Transfer of the spent fuel from the spent fuel pool to the cask/
package is performed while the cask/package is submerged in the
spent fuel pool. When the cask/package is in the spent fuel pool,
the fuel stored in the cask/package is surrounded by water, making
an accidental criticality a concern. To preclude an accidental
criticality in this circumstance, other physical processes or
systems are used, primarily by putting a neutron poison (boron) in
the water. Before any spent fuel is placed in either a spent fuel
pool or a cask/package, a detailed analysis is conducted that
demonstrates that the geometrical configuration and other physical
systems or processes provide reasonable assurance that an accidental
criticality will be prevented.
It is also possible that the spent fuel would need to be
transferred out of a dry storage cask and back in to the spent fuel
pool. This might arise in one of two situations. The first situation
is that it might be necessary to inspect the spent fuel or the dry
storage cask itself. This would necessitate transferring some or all
of the spent fuel in the dry storage cask back into the spent fuel
pool. The second and more probable situation that would require
unloading the spent fuel from the dry storage cask back into the
spent fuel pool, would be in preparation for shipment of the spent
fuel. Before the spent fuel in a dry storage cask licensed pursuant
to 10 CFR Part 72 only (not also licensed pursuant to 10 CFR Part
71) can be shipped, it must first be transferred to an approved
transportation package licensed pursuant to 10 CFR Part 71. In order
to place the spent fuel into the transportation package, it must
first be unloaded from the dry storage cask back into the spent fuel
pool. The dry storage cask is then removed from the spent fuel pool
and is replaced by the transportation package. The spent fuel is
then loaded into the transportation package.
As described in more detail below, there are sufficient
regulatory controls in place to provide reasonable assurance that
spent fuel can be safely stored both in spent fuel pools and in dry
storage casks or transportation packages. The purpose for the change
to 10 CFR 50.68 is to reduce the regulatory burden imposed on
licensees by removing a requirement for an unnecessary criticality
analysis. This change clarifies that, when loading spent fuel into a
dry storage cask or transportation package while in the spent fuel
pool, the license requirements and controls (including the physical
processes and systems) relied on by the NRC in its determination
that a specific dry storage cask or transportation package is
acceptable shall be followed and provide the basis for the NRC
concluding that public health and safety are maintained.
II. Regulatory Evaluation
The regulation at 10 CFR 50.68 requires that pressurized water
reactor (PWR) SFPs remain subcritical in an unborated, maximum
moderation condition. To demonstrate that the fuel in the SFP
remains subcritical in this condition, 10 CFR 50.68 allows credit
for the operating history of the fuel (fuel burnup) when analyzing
the storage configuration of the spent fuel. Taking the burnup of
the spent fuel into consideration reduces the reactivity of the fuel
and reduces the need for soluble boron to demonstrate
subcriticality. Meeting the unborated condition requirement provides
reasonable assurance that potential boron dilution events that could
occur during the storage period of spent fuel in the SFP would not
result in an accidental criticality. Boron dilution events could
occur due to leakage from the spent fuel pool requiring
replenishment from an unborated water source. For example, a SFP
liner rupture due
[[Page 66654]]
to an earthquake could result in a rapid drain down of the SFP as
could a rupture of the SFP cooling system. Dilution could also
result from the introduction of unborated water in the vicinity of
the SFP, such as from a fire suppression system. For the rapid drain
down scenario, the SFP might be replenished with unborated sources
of water in an effort to quickly reestablish spent fuel cooling and
to provide shielding. It is necessary to reestablish spent fuel
cooling during a rapid drain down event to preclude the possibility
of the elevated cladding temperature that could cause overheating of
the fuel and a loss of fuel cladding integrity. Because of the very
low likelihood of a rapid drain down event, it is not considered
part of the licensing basis for commercial nuclear power reactors.
Storage casks are approved for use by the NRC by the issuance of
specific and general licenses pursuant to 10 CFR Part 72.
Transportation packages for spent fuel are licensed pursuant to 10
CFR Part 71. 10 CFR Part 71 currently requires that the criticality
safety system for transportation packages be designed with the
assumption that a package can be flooded with fresh water (i.e., no
soluble boron). Therefore, the transportation packages are already
analyzed in a manner that complies with the 10 CFR 50.68 assumption.
The following discussions will then focus only on storage casks.
However, the transportation packages are included in the proposed
change in order to allow loading/unloading operation of a
transportation package into a 10 CFR Part 50 facility (i.e., spent
fuel pool) without the need for a specific license or exemption
considerations under 10 CFR Part 50.
The certificates and licenses issued by the NRC for these
storage casks and the requirements of 10 CFR Part 72 include
controls for fuel loading, storage, and unloading that provide
reasonable assurance that spent fuel cooling is maintained and an
accidental criticality is avoided. These controls are not identical
to the requirements contained in 10 CFR 50.68, but instead allow for
an alternate means of assuring safety by providing additional
requirements that are not present in 10 CFR 50.68. NRC approval of
the storage cask designs was, in part, predicated on the assumption
that unirradiated commercial nuclear fuel (fresh fuel) of no more
than 5 weight percent enrichment would remain subcritical when
stored in its dry configuration and that it would remain subcritical
with a sufficient boron concentration (if any boron was required)
when stored in a water filled configuration, such as when it is in a
SFP at a commercial power reactor. Under 10 CFR Part 72, reliance is
placed on soluble boron to assure subcriticality when the cask is
full of water, rather than relying on fuel burnup. The fresh fuel
assumption allowed the NRC to generically approve storage casks
without regard to the operating history of the fuel from a
criticality perspective by establishing a bounding case for the
various fuel types that could be stored in the approved storage
casks. If generic fuel burnup data were available, the NRC may have
been able to approve storage cask designs without the need for boron
to assure subcriticality, but would have put in place a minimum fuel
burnup requirement instead. By having the 10 CFR Part 72 controls in
place, loading, storage, and unloading of spent fuel can be
accomplished in a manner that precludes an accidental criticality
while maintaining sufficient fuel cooling capabilities.
III. Problem Statement
On March 23, 2005, the NRC issued Regulatory Issue Summary (RIS)
2005-05 addressing spent fuel criticality analyses for SFPs under 10
CFR 50.68 and Independent Spent Fuel Storage Installations (ISFSI)
under 10 CFR Part 72. The intent of the RIS was to inform reactor
licensees that they must meet both the requirements of 10 CFR 50.68
and 10 CFR Part 72 with respect to subcriticality during storage
cask loading in SFPs. Different assumptions are relied on under
these regulations to achieve the same underlying purpose, namely to
place spent fuel in a condition such that it remains cooled and to
preclude an accidental criticality.
The need to meet both regulations and the differences in the
assumptions creates an additional burden on licensees to show that
credit for boron is not required to preclude an accidental
criticality in a storage cask when filled with water. This condition
exists for NRC approved high density storage casks used for storing
PWR fuel. As permitted under 10 CFR Part 72, boron can be relied on
at PWR SFPs to maintain subcriticality during storage cask loading
or unloading. However, 10 CFR 50.68 requires that spent fuel
assemblies be subcritical with unborated water in SFPs. In order to
satisfy both of these requirements, a site specific analysis that
demonstrates that the storage casks would remain subcritical for the
specific irradiated fuel loading planned, without credit for boron,
would be required. In this analysis, the licensee would rely on the
fuel burnup to determine the margin to criticality for the specific
cask loading. The analysis would be similar to that conducted for
the SFP itself, but would take into account the unique design
features of the storage cask when determining the minimum burnup
required for spent fuel storage in the specific cask.
In a July 25, 2005, letter to the NRC, the Nuclear Energy
Institute (NEI) indicated that the implementation of the RIS
recommendations would ``create an unnecessary burden for both
industry and the NRC with no associated safety benefit for public.''
In other words, preparing an amendment application by performing a
redundant criticality analysis consistent with 10 CFR 50.68 would
cause ``an unnecessary administrative burden for licensees with no
commensurate safety benefits'' because the dry storage cask had
already been approved based on the criticality analysis and
assumptions required by 10 CFR Part 72, i.e., boron credit with no
burnup credit. NEI reiterated its position at a meeting with the NRC
staff on November 10, 2005.
Subsequent to the November 10, 2005 meeting, the NRC decided to
examine the likelihood of criticality in casks while submerged in
SFPs during loading or unloading in the event of a boron dilution in
SFPs due to natural phenomena and other scenarios. Based on the low
likelihood of such an event, NRC has determined that a revision to
10 CFR 50.68 clarifying that the requirements of 10 CFR Part 71 or
72, as appropriate, apply to transportation packages and storage
casks during loading and unloading operations while submerged in a
PWR SFP. This issue does not apply to boiling water reactors (BWR)
because BWR SFPs do not contain boron and dry storage casks that are
used to load BWR fuel do not rely on boron to maintain
subcriticality. As discussed below, there is no safety benefit from
requiring the licensee to conduct a site specific analysis to comply
with 10 CFR 50.68(b) in support of dry storage cask loading, fuel
storage, or unloading activities.
IV. Technical Evaluation
In assessing the proposed change to 10 CFR 50.68, the staff
considered what type of events could lead to damage of the fuel in a
storage cask as a result of the proposed change. Since the central
issue in the application of the regulations is whether boron is
credited as a control for avoiding an accidental criticality, events
that reduce the boron concentration in the storage cask were
considered the only events that would be affected by the proposed
change. There are two types of scenarios in which a boron dilution
could occur. A rapid drain down and subsequent reflood of the SFP or
in leakage from the SFP cooling system or from an unborated water
source in the vicinity of the SFP (i.e., fire suppression system)
that would go undetected by normal licensee activities (slow boron
dilution event). Each of these scenarios are addressed below.
a. Slow Boron Dilution Event
The possibility of a slow boron dilution event resulting in an
accidental criticality event in a storage cask in a SFP is highly
unlikely based on the requirements contained in the technical
specifications attached to the Certificate of Compliance issued
under 10 CFR Part 71 or 72 for the specific cask design.
The storage cask technical specifications require measurements
of the concentration of dissolved boron in a SFP before and during
cask loading and unloading operations. At a point a few hours prior
to insertion of the first fuel assembly into a storage cask,
independent measurements of the dissolved boron concentration in the
SFP are performed. During the loading and unloading operation, the
dissolved boron concentration in the water is confirmed at intervals
that do not exceed 72 hours. The measurements of the dissolved boron
in the SFP are performed independently by two different individuals
gathering two different samples. This redundancy reduces the
possibility of an error and increases the accuracy of the
measurement that is used to confirm that the boron concentration is
in compliance with the storage cask's technical specifications.
These measurements are continued until the storage cask is removed
from the SFP or the fuel is removed from the cask.
In addition to the storage cask technical specification boron
concentration sampling requirements, 10 CFR Part 72 also requires
criticality monitoring. As stated in 10 CFR
[[Page 66655]]
72.124(c), a criticality monitoring system is required for dry
storage cask loading, storage, or unloading operations:
``A criticality monitoring system shall be maintained in each
area where special nuclear material is handled, used, or stored
which will energize clearly audible alarm signals if accidental
criticality occurs. Underwater monitoring is not required when
special nuclear material is handled or stored beneath water
shielding. Monitoring of dry storage areas where special nuclear
material is packaged in its stored configuration under a license
issued under this subpart is not required.''
Although 10 CFR 72.124(c) states ``underwater [criticality]
monitoring is not required,'' criticality monitoring is required
when special nuclear material is handled, used, or stored at
facilities where the requirements of 10 CFR Part 72 apply. The point
being made in 10 CFR 72.124(c) is that the criticality monitors are
not required to be located under the water, but rather that
criticality monitors can be located above the water to satisfy this
requirement. The facilities to which this requirement applies
include 10 CFR Part 50 SFPs when loading, storing, or unloading fuel
in storage casks licensed under 10 CFR Part 72. The underlying
intent of 10 CFR 72.124(c) is that criticality monitors are required
under circumstances where an accidental criticality could occur as
the result of changes in the critical configuration of special
nuclear material. As such, storage cask loading and unloading
activities need to be monitored to provide reasonable assurance that
these fuel handling activities (changes in the critical
configuration) do not result in an accidental criticality.
When storing fuel in a storage cask that requires boron to
remain subcritical while submerged in the SFP, the critical
configuration can be affected by changes to the moderation
(temperature changes of the water) or boron concentration. The
primary concern during storage under these circumstances is the
dilution of the boron concentration. Therefore, to meet the
underlying intent of 10 CFR 72.124(c) either criticality monitors
are required to detect an accidental criticality or controls are
necessary to preclude a boron dilution event that could lead to an
accidental criticality. As previously discussed, periodic sampling
(at intervals no greater than 72 hours) of the boron concentration
is required when fuel is stored in storage casks in the SFP. The
requirement to periodically sample the boron concentration provides
reasonable assurance that should a slow boron dilution event occur,
it would be identified such that actions could be taken to preclude
an accidental criticality and thereby meet the underlying intent of
10 CFR 72.124(c).
A slow boron dilution event would require that an unborated
source of water be injected into the SFP and be undetected by normal
plant operational activities for sufficient duration to allow the
boron concentration to drop below the level required to maintain a
storage cask subcritical. First, consider the nature of the boron
dilution event that would be required to dilute the SFP boron
concentration from the storage cask technical specification
concentration level (typically about 2200 ppm) to the critical boron
concentration value (typically around 1800 ppm). The in-leakage rate
would have to be large enough to dilute the entire volume of the
pool between the time of the initial boron concentration sample and
the time of the subsequent boron concentration sample and yet be
small enough to remain undetected. Cask loading and unloading are
conducted by licensed operators or certified fuel handlers who are
present during any fuel movement. It is reasonable to conclude that
these operators or handlers would detect all but the smallest
increases in SFP level that would be indicative of a slow boron
dilution event. Second, consider the storage casks loading and
unloading operation frequency and duration. The frequency and
duration depend on the dry storage needs and the reactor facility
design. Based on historical average data, only a few casks (on the
order of about 5 casks) are loaded each year at an operating reactor
that is in need of dry storage. Third, consider that the time a
storage cask is actually loaded with fuel while in the SFP is
typically between 24 and 72 hours. When all of these factors are
considered, it is clear that the likelihood of an undetected slow
boron dilution event occurring during the time that a storage cask
is loaded with fuel in the SFP is very remote.
Another scenario that could result in a slow boron dilution
event is the intentional injection of unborated water into the
storage cask while loaded with fuel. A person would need access to a
source of unborated water and a means for injecting the water
directly into the cask (e.g., using a fire hose). While it is
possible that someone could intentionally inject unborated water
into the cask, it is highly unlikely that this could be done without
being promptly detected by other licensee personnel monitoring cask
loading or unloading activities. This scenario would result in a
localized dilution of boron concentration in the storage cask. As
the soluble boron concentration decreased in the storage cask, the
fuel in the cask could become critical. The inadvertent criticality
would be detected by the criticality monitors required by 10 CFR
72.124 during cask loading and unloading operations. As such, the
licensee would be notified of the inadvertent criticality and could
take action to stop the intentional injection of unborated water
into the cask, re-establish a subcritical boron concentration in the
cask, and terminate the inadvertent criticality event. This scenario
is essentially the same as any other slow boron dilution event in
that it requires an undetected injection of unborated water into a
cask that is loaded with fuel.
With the controls of the storage cask technical specifications
related to monitoring boron concentration, the requirements of 10
CFR 72.124(c) for criticality monitoring to detect and avoid an
accidental criticality, and the very remote likelihood of an
undetected slow boron dilution event occurring at the time a storage
cask is being loaded, it is reasonable to conclude that considering
a slow boron dilution event there is no safety benefit in requiring
a licensee to conduct a site specific analysis to demonstrate that a
dry storage cask will remain subcritical in an unborated condition
as required by 10 CFR 50.68(b).
b. Rapid Drain Down Event
A rapid drain down event could be postulated if there were an
event that caused a catastrophic failure of the SFP liner and
supporting concrete structure. If there were a catastrophic failure
of the SFP liner that resulted in a rapid drain down while a storage
cask was in the SFP, the borated water in the storage cask would
likely remain in the storage cask providing reasonable assurance
that the fuel would be cooled and remain subcritical. However, if
the storage cask were to become dry, the design of the storage cask
would allow the fuel to remain cooled, and without water as a
moderator the fuel in the storage cask would be significantly
subcritical.
To assess whether there is a safety benefit from requiring
licensees to conduct an analysis of storage casks assuming no boron
as the result of a rapid SFP drain down event three factors were
considered in the NRC's assessment. The first factor is the
probability that a storage cask will be in the SFP, loaded with
fuel. The second factor is whether there are credible scenarios that
could result in the rapid drain down of the SFP. The third factor is
whether a boron dilution event would occur in the storage casks if
the rapid SFP drain down event were to occur. As described below,
when taken together, it is clear that it is not necessary to require
licensees to conduct additional criticality analyses to demonstrate
that the storage casks will remain subcritical assuming no boron as
required by 10 CFR 50.68 in response to a SFP rapid drain down event
due to its highly unlikely occurrence.
For the first factor, historical data suggests that
approximately five storage casks are loaded on a annual basis at
those facilities that need dry storage. The casks are typically in
the SFP with fuel installed for as long as 72 hours. Using 72 hours
and the historical data as initial assumptions, the probability of a
storage cask loaded with spent fuel being in a SFP is about 4E-2/yr.
Licensees only have the capability of moving one storage cask at a
time into or out of the SFP. The total time it typically takes to
bring a storage cask into the SFP, load it with fuel, and remove it
from the SFP area for transport to the ISFSI is between 3 and 5
days. If a licensee were to continuously load storage casks,
assuming the shortest duration to complete the transfer cycle (24
hours to transfer the cask from outside the building into the spent
fuel pool; loading two to three assemblies per hour, or 12 hours to
load the cask to capacity; and 36 hours for removing the cask from
the spent fuel pool, sealing the cask and removing it from the
building), the licensee would be able to load approximately 120
storage casks per year. Under these assumptions, the probability of
having a storage cask loaded with fuel in the SFP would increase to
1.6E-1/year. If one assumes that it is possible to load 1 storage
cask a week (for a total of 52 casks a year) this would result in a
probability of having a cask that is loaded with fuel physically in
the pool of 4E-1/year.
For the second factor, the NRC has assessed the possibility of
rapid drain down
[[Page 66656]]
events at SFPs. From NUREG-1738, ``Technical Study of Spent Fuel
Pool Accident Risk at Decommissioning Nuclear Power Plants,''
phenomena that could cause such a catastrophic failure include a
storage cask drop (event frequency of about 2E-7/year), an aircraft
impact (event frequency of about 2.9E-9/year), a tornado missile
(event frequency of <1E-9/year) or a seismic event. A dropped
storage cask does not affect the proposed change to 10 CFR 50.68
because the dilution of boron in the cask is the issue of interest.
When moving a storage cask, it is either empty (no fuel) or has fuel
stored in it with a closure lid installed. In each case a boron
dilution event that could result in an accidental criticality in a
dry storage cask would be precluded. The aircraft impact and tornado
missile events are of such a low frequency that they do not need to
be considered within the scope of the proposed change. However, the
consequences of the aircraft and tornado events would be similar to
a SFP liner rupture due to other events (such as an earthquake).
This leaves a seismic event as the only initiating event for a rapid
drain down of a SFP that may be credible.
In Sections 3.5.1 and 3.7.2 of NUREG-1738, the NRC describes the
beyond design basis seismic event that would have to occur to result
in a rapid drain down of a SFP. Given the robust structural design
of the spent fuel pools, the NRC expects that a seismic event with a
peak spectral acceleration several times larger than the safe
shutdown earthquake (SSE) would be required to produce a
catastrophic failure of the structure.
There are two information sources that the NRC relies upon to
provide reasonable estimates of seismic event frequency: (1)
Lawrence Livermore National Laboratory (LLNL) seismic hazard curves,
published in NUREG-1488, ``Revised Livermore Seismic Hazard
Estimates for Sixty-Nine Nuclear Power Plant Sites East of the Rocky
Mountains;'' and (2) Electric Power Research Institute (EPRI)
seismic hazard curves, published in EPRI NP-4726, ``Seismic Hazard
Methodology for the Central and Eastern United States.'' Both the
LLNL and EPRI hazard estimates were developed as best estimates
based on data extrapolation and expert opinion and are considered
valid by the NRC.
In NUREG-1738, a general high confidence with a low probability
of failure (HCLPF) capacity of 1.2g peak spectral acceleration
(PSA), which is equivalent to about 0.5g peak ground acceleration
(PGA), is established for SFPs. Under 10 CFR Part 100, ``Seismic and
Geologic Siting Criteria for Nuclear Power Plants,'' the minimum SSE
seismic PGA value is 0.1g. Typical PGA values for plants east of the
Rocky Mountains range from 0.1g to 0.25g and the PGA values for
plants west of the Rocky Mountains range from 0.25g to 0.75g. Using
the LLNL seismic hazard curves, with a SFP HCLPF capacity of 1.2g
PSA, the mean frequency of a seismically-induced rapid drain down
event is estimated to be about 2E-6/year, ranging from less than 1E-
7/year to 1.4E-5/year, depending on the site-specific seismic
hazard. The EPRI seismic hazard curves provide a mean frequency of a
seismically-induced rapid drain down event of about 2E-7/year,
ranging from less than 1E-8/year to about 2E-6/year, depending on
the site-specific seismic hazard.
For sites west of the Rocky Mountains, the SFP HCLPF capacity
would be site-specific, but would be at least equal to the SSE. The
SSE for Columbia is 0.25g PGA and has an annual probability of
exceedance (APE) of 2E-4. However, it is important to note that a
seismic event capable of rupturing the SFP would have to be much
greater than the SSE. Therefore, it is reasonable to conclude that
mean frequency of a seismically-induced rapid drain down event at
Columbia is bounded by the analysis for plants East of the Rocky
Mountains.
Diablo Canyon's SSE is 0.75g PGA with an APE of 2.5E-4. San
Onofre's SSE is 0.5g PGA with an APE of 5E-4. An SSE is the
earthquake that is expected to occur that produces the maximum
ground motion for which certain structures must remain capable of
performing their safety function. SFPs are designed to remain
functional following an SSE. Further, as noted for all of the other
SFPs, the as-designed and as-built structures have significant
margin to failure and are capable of remaining functional (not
subject to a rapid drain down event) for earthquakes well above the
SSE. Both the Diablo Canyon and San Onofre SFPs were designed and
constructed in a manner that provides significant structural margin.
Therefore, it is reasonable to conclude that the probability of an
earthquake causing a rapid drain down event would be similar to the
probabilities determined for plants East of the Rocky Mountains. As
such, the NRC concluded that for these two plants, specific SFP
failure probabilities where not a factor that would have an adverse
affect on its determination with regard to the acceptability of the
proposed change to 10 CFR 50.68.
Based on the above, it would take a seismic event significantly
greater than the design basis SSE to credibly cause a SFP rapid
drain down event. Using the most conservative results for a
seismically-induced SFP rapid drain down event (1.4E-5) and the
probability of having a storage cask with fuel installed in the pool
(4E-1), the probability of having a SFP rapid drain down event when
a storage cask is in the pool would likely be significantly less
than 5.6E-6. This is a low probability of SFP failure when a dry
storage cask is in the SFP. Coupled with the fact that to reach this
low probability would require a seismic event well in excess of the
SSE, the NRC concludes there is no safety benefit from requiring the
licensee to conduct a site specific analysis in support of storage
cask loading, fuel storage, or unloading activities.
For the third factor, a rapid drain down event is considered to
be a gross, rapid loss of the water that provides cooling for the
spent fuel. This event is beyond the licensing basis for PWR plants.
Minor leakage is not considered to constitute failure. As such, a
rapid drain down event would have to exceed the makeup capability of
the normal and alternative water supplies by a significant amount to
drain the pool in a short period. The makeup capacities available to
refill the SFPs typically range from about 20 gallons per minute
(gpm) for normal makeup to around 1000 gpm for alternative makeup
supplies such as the fire suppression system. Many sites have the
capability to supply borated water to refill the spent fuel pool.
However, to assess the affect of a rapid drain down event on a boron
dilution event in a dry storage cask, the NRC assumed that the
makeup would be from an unborated water source such as a fire
suppression system. The main concern with a rapid drain down event
as it affects a dry storage cask is subsequently diluting the boron
concentration in the cask during the attempt to refill the SFP to
keep the fuel stored in the pool cooled to preclude overheating the
fuel and a loss of fuel cladding integrity. Therefore, the
assumption that a licensee would use an unborated source of water,
such as the fire suppression system, with the largest capacity
available to provide cooling water in its attempt to reflood the SFP
following a rapid drain down event is reasonable given the
importance of quickly re-establishing cooling of the fuel stored in
the SFP. The need to establish alternative means for cooling the
fuel stored in the SFP during a rapid drain down event is
independent of whether a storage cask is located in the SFP and
therefore, has no relation to the proposed change to 10 CFR 50.68.
The NRC considered four scenarios when assessing the affect of a
rapid drain down event on diluting the boron concentration in a dry
storage cask. First, the cask might drain as the SFP drains (some
older cask designs have drain ports at the bottom of the cask) and
the licensee is unable to reflood the SFP because the leak rate is
well in excess of the normal or alternate makeup capacity available
to reflood the SFP. This scenario results in the fuel stored in the
dry storage cask in essentially the same condition under which it
would be permanently stored. The geometrical configuration of the
dry storage casks are such that without the water, the fuel will
remain subcritical. Further, the dry storage cask is designed to
remove the decay heat from the fuel in this configuration, so
excessive cladding temperatures would not be reached and there would
be no fuel damage.
The second scenario involves those storage casks that do not
have drain ports at the bottom of the cask and therefore would
remain filled with water as the SFP experiences the rapid drain down
event. In this scenario, the licensee would likely use the largest
capacity, unborated source of cooling water to keep the spent fuel
in the SFP storage racks cooled. As noted before, a rapid drain down
event would significantly exceed the makeup capacity of available
water systems and the licensee would need to use an alternative
means, such as spraying the fuel stored in the SFP racks to keep the
fuel cool. In this scenario, the water that remains in the dry
storage cask would still be borated and would maintain the fuel
storage in the cask subcritical. The fuel in the cask would remain
cooled by the water surrounding it and the heat transfer through the
cask consistent with the cask design. Again, in this situation, the
fuel in the cask would be adequately cooled and maintained
[[Page 66657]]
in a subcritical configuration providing reasonable assurance that
excessive fuel cladding temperatures and subsequent fuel damage
would not occur.
The third scenario involves those dry storage casks that would
remain filled with borated water. The possibility exists for a
licensee to cause a boron dilution event in the dry storage cask
when spraying the fuel stored in the SFP racks. The location of the
dry storage cask might be close enough to the SFP storage racks that
it could inadvertently be sprayed at the same time as the SFP racks,
overfilling the dry storage cask, and eventually diluting the boron.
Under these conditions, the boron concentration would slowly
decrease and this scenario becomes very similar to a slow boron
dilution event as discussed previously. The criticality monitors
required for dry cask loading would still be available and would
provide indication of an accidental criticality. With indication of
an accidental criticality, it is reasonable to assume that the
licensee would take action to stop the boron dilution from
continuing and restore the dry storage cask to a subcritical
configuration.
Actions the licensee could take to return the dry storage cask
to a subcritical configuration could include:
1. Stop spraying unborated water into the dry storage cask and
allow the water in the cask to heat up with a subsequent reduction
in the moderation provided by the water that would eventually re-
establish a subcritical configuration at a higher water temperature.
In this condition, the temperature of the water may be high enough
that the water would eventually boil off (be higher than 212 degrees
F at atmospheric conditions). If this were to occur, the cask would
eventually become dry and the fuel would be in a subcritical
configuration and cooled consistent with the design of the cask. As
the water boiled off, it would continue to provide cooling to the
fuel such that the fuel would not experience significantly elevated
temperatures and there would be no fuel damage; or
2. Spray water into the cask from a borated water source to
increase the boron concentration, re-establishing a subcritical
configuration and keeping the fuel cooled.
In each case, the fuel would not be subject to excessive
temperatures and therefore, there would be no fuel damage that could
impact public health and safety.
Under this third scenario there is also the possibility that the
licensee might intentionally spray water into the dry storage cask
in an attempt to keep the fuel in the cask cool. Given that the cask
will already be filled with water and the importance of cooling the
fuel in the SFP storage racks (where there is no water following a
rapid drain down event), the NRC considers the possibility of the
intentional diversion of cooling water from the fuel stored in the
SFP racks to the fuel stored in the dry storage cask to be very
remote. Therefore, the NRC does not consider this as a factor that
would have an adverse affect on its determination with regard to the
acceptability of the proposed change to 10 CFR 50.68. However, even
if the licensee intentionally diverted water from cooling the fuel
in the SFP racks to the fuel in the dry storage cask, there would be
a slow boron dilution event, a slow approach to criticality, and
indication of an accidental criticality from the required
criticality monitors. As such, this case would be very similar to
the unintentional dilution case described above.
In the fourth scenario, the NRC assumed that the licensee was
able to repair the damage to the SFP and reflood the pool. In this
scenario as the licensee reflooded the SFP the dry storage cask
would either reflood as the SFP was filled (for those casks with
drain ports at the bottom); if the cask had dried out it would
reflood once the water level in the SFP reached the top of the cask
and water began spilling into the cask; or if the cask remained
flooded following the rapid drain down event, there would be a slow
dilution of the boron in the water in the cask as the SFP level
continued to rise. In each of these cases, as the cask was filled
with water or as the boron dilution of the water in the cask
occurred, the possibility increases that an accidental criticality
might occur. However, because of the relatively slow reactivity
addition that would occur during each of these cases, the approach
to criticality would be reasonably slow. As noted previously, the
licensee is required to have criticality monitors in place during
dry storage cask loading (or unloading) activities. These
criticality monitors would provide indication that an accidental
criticality had occurred. Once identified, it is reasonable that the
licensee would take action to re-establish a subcritical
configuration. However, as discussed above for the third scenario,
even if there were an accidental criticality, the likelihood of fuel
damage is very remote.
The possibility of an accidental criticality in the fourth
scenario is even less likely given the following factors:
1. Dry storage casks are typically loaded with fuel that has
significant burnup that reduces the reactivity of the assembly. As
such, it is reasonable to conclude that even in an unborated
condition, the fuel stored in the cask would remain subcritical.
2. As the licensee refilled the SFP, it is reasonable to assume
that it would be injecting borated water to re-establish the boron
concentration level required by plant technical specifications as
soon as practical.
Based on the above, even if there were an event that caused a
rapid drain down of a SFP while a dry storage cask was in the SFP,
the likelihood of a boron dilution event causing fuel damage is very
remote. Therefore, the NRC concludes there is no safety benefit from
requiring the licensee to conduct a site specific analysis in
support of dry storage cask loading, fuel storage, or unloading
activities.
V. Conclusion
As discussed above the NRC assessed the safety benefit of
requiring licensees to conduct an additional criticality analysis to
meet the requirements of 10 CFR 50.68 while loading a transportation
package or dry storage cask in the SFP. The NRC determined that the
controls required by 10 CFR Part 71 or 72 for the associated package
or cask provide reasonable assurance that a slow boron dilution
event would not result in elevated fuel temperature and subsequent
fuel damage. Therefore, for a slow boron dilution event, there is no
benefit to the additional criticality analysis. The NRC further
determined that the probability of having a rapid drain down event
result in elevated fuel temperatures and subsequent fuel damage was
highly unlikely. Based on its analysis, the NRC concludes there is
no safety benefit from requiring a licensee to conduct a site
specific analysis in support of storage cask loading, fuel storage,
or unloading activities and that the proposed rule change is
therefore acceptable.
[FR Doc. E6-19372 Filed 11-15-06; 8:45 am]
BILLING CODE 7590-01-P