[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 If the rule is withdrawn, timely notice will be published in 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