[Federal Register Volume 68, Number 179 (Tuesday, September 16, 2003)]
[Rules and Regulations]
[Pages 54123-54143]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-23554]
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Rules and Regulations
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Federal Register / Vol. 68, No. 179 / Tuesday, September 16, 2003 /
Rules and Regulations
[[Page 54123]]
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 50 and 52
RIN 3150-AG76
Combustible Gas Control in Containment
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations for combustible gas control in power reactors applicable to
current licensees and is consolidating combustible gas control
regulations for future reactor applicants and licensees. The final rule
eliminates the requirements for hydrogen recombiners and hydrogen purge
systems, and relaxes the requirements for hydrogen and oxygen
monitoring equipment to make them commensurate with their risk
significance. This action stems from the NRC's ongoing effort to risk-
inform its regulations, and is intended to reduce the regulatory burden
on present and future reactor licensees. Additionally, the final rule
grants in part and denies in part a petition for rulemaking (PRM-50-68)
submitted by Mr. Bob Christie. This notice constitutes final NRC action
on PRM-50-68. The final rule also denies part of a petition for
rulemaking (PRM-50-71) submitted by the Nuclear Energy Institute. The
remaining issue in PRM-50-71 that is not addressed by this final rule
will be evaluated in a separate NRC action. The NRC has updated a
guidance document, ``Control of Combustible Gas Concentrations in
Containment'' to address changes in the rule. A draft regulatory guide
containing the revisions was published for comment with the proposed
rule.
EFFECTIVE DATE: October 16, 2003.
FOR FURTHER INFORMATION CONTACT: Richard Dudley, Office of Nuclear
Reactor Regulation, Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone (301) 415-1116; e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. Rulemaking Initiation
III. Final Action
A. Retention of Inerting, BWR Mark III and PWR Ice Condenser
Hydrogen Control Systems, Mixed Atmosphere Requirements, and
Associated Analysis Requirements
B. Elimination of Design-Basis LOCA Hydrogen Release
C. Oxygen Monitoring Requirements
D. Hydrogen Monitoring Requirements
E. Technical Specifications for Hydrogen and Oxygen Monitors
F. Combustible Gas Control Requirements for Future Applicants
G. Clarification and Relocation of High Point Vent Requirements
From 10 CFR 50.44 to 10 CFR 50.46a
H. Elimination of Post-Accident Inerting
IV. Comments and Resolution on Proposed Rule and Draft Regulatory
Guide Topics
A. General Comments
B. General Clarifications
C. Monitoring Systems
D. Purge
E. Station Blackout/Generic Safety Issue 189
F. Containment Structural Uncertainties
G. PRA/Accident Analysis
H. Passive Autocatalytic Recombiners
I. Reactor Venting
J. Design Basis Accident Hydrogen Source Term
K. Requested Minor Modifications
L. Atmosphere Mixing
M. Current Versus Future Reactor Facilities
N. Equipment Qualification/Survivability
V. Petition for Rulemaking, PRM-50-68
VI. Petition for Rulemaking, PRM-50-71
VII. Section-by-Section Analysis of Substantive Changes
VIII. Availability of Documents
IX. Voluntary Consensus Standards
X. Finding of No Significant Environmental Impact: Environmental
Assessment
XI. Paperwork Reduction Act Statement
XII. Public Protection Notification
XIII. Regulatory Analysis
XIV. Regulatory Flexibility Certification
XV. Backfit Analysis
XVI. Small Business Regulatory Enforcement Fairness Act
I. Background
On October 27, 1978 (43 FR 50162), the NRC adopted a new rule, 10
CFR 50.44, specifying the standards for combustible gas control
systems. The rule required the applicant or licensee to show that
during the time period following a postulated loss-of-coolant accident
(LOCA), but prior to effective operation of the combustible gas control
system, either: (1) An uncontrolled hydrogen-oxygen recombination would
not take place in the containment, or (2) the plant could withstand the
consequences of an uncontrolled hydrogen-oxygen recombination without
loss of safety function. If neither of these conditions could be shown,
the rule required that the containment be provided with an inerted
atmosphere to provide protection against hydrogen burning and
explosion. The rule defined a release of hydrogen involving up to 5
percent oxidation of the fuel cladding as the amount of hydrogen to be
assumed in determining compliance with the rule's provisions. This
design-basis hydrogen release was based on the design-basis LOCA
postulated by 10 CFR 50.46 and was multiplied by a factor of five for
added conservatism to address possible further degradation of emergency
core cooling.
The accident at Three Mile Island, Unit 2 involved oxidation of
approximately 45 percent of the fuel cladding [NUREG/CR-6197, dated
March 1994] with hydrogen generation well in excess of the amounts
required to be considered for design purposes by Sec. 50.44.
Subsequently, the NRC reevaluated the adequacy of the regulations
related to hydrogen control to provide greater protection in the event
of accidents more severe than design-basis LOCAs. The NRC reassessed
the vulnerability of various containment designs to hydrogen burning,
which resulted in additional hydrogen control requirements adopted as
amendments to Sec. 50.44. The 1981 amendment, which added paragraphs
(c)(3)(i), (c)(3)(ii), and (c)(3)(iii) to the rule, imposed the
following requirements:
(1) An inerted atmosphere for boiling water reactor (BWR) Mark I
and Mark II containments,
(2) installation of recombiners for light water reactors that rely
on a purge or repressurization system as a primary means of controlling
combustible gases following a LOCA, and
(3) installation of high point vents to relieve noncondensible
gases from the reactor vessel (46 FR 58484; December 2, 1981).
[[Page 54124]]
On January 25, 1985 (50 FR 3498), the NRC published another
amendment to Sec. 50.44. This amendment, which added paragraph
(c)(3)(iv), required a hydrogen control system justified by a suitable
program of experiment and analysis for BWRs with Mark III containments
and pressurized water reactors (PWRs) with ice condenser containments.
In addition, plants with these containment designs must have systems
and components to establish and maintain safe shutdown and containment
integrity. These systems must be able to function in an environment
after burning and detonation of hydrogen unless it is shown that these
events are unlikely to occur. The control system must handle an amount
of hydrogen equivalent to that generated from a metal-water reaction
involving 75 percent of the fuel cladding surrounding the active fuel
region.
When Sec. 50.44 was amended in 1985, the NRC recognized that an
improved understanding of the behavior of accidents involving severe
core damage was needed. During the 1980s and 1990s, the NRC sponsored a
severe accident research program to improve the understanding of core
melt phenomena, combustible gas generation, transport and combustion,
and to develop improved models to predict the progression of severe
accidents. The results of this research have been incorporated into
various studies (e.g., NUREG-1150 and probabilistic risk assessments
performed as part of the Individual Plant Examination (IPE) program) to
quantify the risk posed by severe accidents for light water reactors.
The result of these studies has been an improved understanding of
combustible gas behavior during severe accidents and confirmation that
the hydrogen release postulated from a design-basis LOCA was not risk-
significant because it was not large enough to lead to early
containment failure, and that the risk associated with hydrogen
combustion was from beyond design-basis (e.g., severe) accidents. These
studies also confirmed the assessment of vulnerabilities that went into
the 1981 and 1985 amendments that required additional hydrogen control
measures for some containment designs.
II. Rulemaking Initiation
In a June 8, 1999, Staff Requirements Memorandum (SRM) on SECY-98-
300, Options for Risk-informed Revisions to 10 CFR Part 50--``Domestic
Licensing of Production and Utilization Facilities,'' the NRC approved
proceeding with a study of risk-informing the technical requirements of
10 CFR Part 50. The NRC staff provided its plan and schedule for the
study phase of its work to risk-inform the technical requirements of 10
CFR Part 50 in SECY-99-264, ``Proposed Staff Plan for Risk-Informing
Technical Requirements in 10 CFR Part 50,'' dated November 8, 1999. The
NRC approved proceeding with the plan for risk-informing the Part 50
technical requirements in a February 3, 2000, SRM. Section 50.44 was
selected as a test case for piloting the process of risk-informing 10
CFR Part 50 in SECY-00-0086, ``Status Report on Risk-Informing the
Technical Requirements of 10 CFR Part 50 (Option 3).''
Mr. Christie of Performance Technology, Inc. submitted letters,
dated October 7 and November 9, 1999, that requested changes to the
regulations in Sec. 50.44. He requested that the regulations be
amended to:
1. Retain the existing requirement in Sec. 50.44(b)(2)(i) for
inerting the atmosphere of existing Mark I and Mark II containments.
2. Retain the existing requirement in Sec. 50.44(b)(2)(ii) for
hydrogen control systems in existing Mark III and PWR ice condenser
containments to be capable of handling hydrogen generated by a metal/
water reaction involving 75 percent of the fuel cladding.
3. Require all future light water reactors to postulate a 75
percent metal/water reaction (instead of the 100 percent required by
the current rule) for analyses undertaken pursuant to Sec. 50.44(c).
4. Retain the existing requirements in Sec. 50.44 for high point
vents.
5. Eliminate the existing requirement in Sec. 50.44(b)(2) to
insure a mixed atmosphere in containment.
6. Eliminate the existing requirement for hydrogen releases during
design basis accidents of an amount equal to that produced by a metal/
water reaction of 5 percent of the cladding.
7. Eliminate the requirement for hydrogen recombiners or purge in
LWR containments.
8. Eliminate the existing requirements for hydrogen and oxygen
monitoring in LWR containments.
9. Revise GDC 41--Containment Atmosphere Cleanup--to require
systems to control fission products and other substances that may be
released into the reactor containment for accidents only where there is
a high probability that fission products will be released to the
reactor containment.
These letters have been treated by the NRC as a petition for
rulemaking and assigned Docket No. PRM-50-68. The NRC published a
document requesting comment on the petition in the Federal Register on
January 12, 2000 (65 FR 1829). The issues associated with Sec. 50.44
raised by the petitioner were discussed in SECY-00-0198, ``Status
Report on Study of Risk-Informed Changes to the Technical Requirements
of 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed
Changes to 10 CFR 50.44 (Combustible Gas Control).'' The final rule and
the petition are consistent in many areas, but differ regarding the
functional requirements for hydrogen and oxygen monitoring, the
requirement for ensuring a mixed atmosphere, the source term of
hydrogen for water-cooled reactors to analyze in order to ensure
containment integrity, and the need to revise GDC-41. The NRC's
detailed basis for including these requirements in the rule is
addressed in a subsequent section of this supplementary information.
The NRC also received a petition for rulemaking filed by the
Nuclear Energy Institute. The petition was docketed on April 12, 2000,
and has been assigned Docket No. PRM-50-71. The petitioner requests
that the NRC amend its regulations to allow nuclear power plant
licensees to use zirconium-based cladding materials other than zircaloy
or ZIRLO, provided the cladding materials meet the requirements for
fuel cladding performance and have received approval by the NRC staff.
The petitioner believes the proposed amendment would improve the
efficiency of the regulatory process by eliminating the need for
individual licensees to obtain exemptions to use advanced cladding
materials that have already been approved by the NRC. The change would
remove the language in 10 CFR 50.44 regarding the use of zirconium-
based cladding materials other than Zircaloy or ZIRLO. The NRC
published a document requesting comment on the petition in the Federal
Register on May 30, 2000 (65 FR 34599). The requested change is
unrelated to the risk-informing of 10 CFR 50.44. The NRC addressed the
NEI petition in this rulemaking for effective use of resources.
Although the final rule does not contain the rule language changes
requested by the petitioner, in its revision to 10 CFR 50.44, the NRC
eliminated the old language referring to various types of fuel
cladding. Thus, the final rule resolves the petitioner's concern
regarding Sec. 50.44. The NRC's detailed basis for this decision is
addressed in a subsequent section of this supplementary information.
In SECY-00-0198, dated September 14, 2000, the NRC staff proposed a
risk-informed voluntary alternative to the current Sec. 50.44.
Attachment 2 to that
[[Page 54125]]
paper, hereafter referred to as the Feasibility Study, used the
framework described in Attachment 1 to the paper and risk insights from
NUREG-1150 and the IPE programs to evaluate the requirements in Sec.
50.44. The Feasibility Study found that combustible gas generated from
design-basis accidents was not risk-significant for any containment
type, given intrinsic design capabilities or installed mitigative
features. The Feasibility Study also concluded that combustible gas
generated from severe accidents was not risk significant for: (1) Mark
I and II containments, provided that the required inerted atmosphere
was maintained; (2) Mark III and ice condenser containments, provided
that the required igniter systems were maintained and operational, and
(3) large, dry and sub-atmospheric containments because of the large
volumes, high failure pressures, and likelihood of random ignition to
help prevent the build-up of detonable hydrogen concentrations.
The Feasibility Study did conclude that the above requirements for
combustible gas mitigative features were risk-significant and must be
retained. Additionally, the Feasibility Study also indicated that some
mitigative features may need to be enhanced beyond current
requirements. This concern was identified as Generic Safety Issue-189
(GI-189). The resolution of GI-189 will assess the costs and benefits
of improvements to safety which can be achieved by enhancing
combustible gas control requirements for Mark III and ice condenser
containment designs. The resolution of GI-189 is proceeding
independently of this rulemaking. In an SRM dated January 19, 2001, the
NRC directed the NRC staff to proceed expeditiously with rulemaking on
the risk-informed alternative to Sec. 50.44.
In SECY-01-0162, ``Staff Plans for Proceeding with the Risk-
Informed Alternative to the Standards for Combustible Gas Control
Systems in Light-Water-Cooled Power Reactors in 10 CFR 50.44,'' dated
August 23, 2001, the NRC staff recommended a revised approach to the
rulemaking effort. This revised approach recognized that risk-informing
Part 50, Option 3 was based on a realistic reevaluation of the basis of
a regulation and the application of realistic risk analyses to
determine the need for and relative value of regulations that address a
design-basis issue. The result of this process necessitates a
fundamental reevaluation or ``rebaselining'' of the existing
regulation, rather than the development of a voluntary alternative
approach to rulemaking. On November 14, 2001, in response to NRC
direction in an SRM dated August 2, 2001, the NRC staff published draft
rule language on the NRC Web site for stakeholder review and comment.
In an SRM dated December 31, 2001, the NRC directed the staff to
proceed with the revision to the existing Sec. 50.44 regulations.
III. Final Action
The NRC is retaining existing requirements for ensuring a mixed
atmosphere, inerting Mark I and II containments, and hydrogen control
systems capable of accommodating an amount of hydrogen generated from a
metal-water reaction involving 75 percent of the fuel cladding
surrounding the active fuel region in Mark III and ice condenser
containments. The NRC is eliminating the design-basis LOCA hydrogen
release from Sec. 50.44 and consolidating the requirements for
hydrogen and oxygen monitoring into Sec. 50.44 while relaxing safety
classifications and licensee commitments to certain design and
qualification criteria. The NRC is also relocating and rewording
without materially changing the hydrogen control requirements in Sec.
50.34(f) to Sec. 50.44. The high point vent requirements are being
relocated from Sec. 50.44 to a new Sec. 50.46a with a change that
eliminates a requirement prohibiting venting the reactor coolant system
if it could ``aggravate'' the challenge to containment.
Substantive issues are addressed in the following sections.
A. Retention of Inerting, BWR Mark III and PWR Ice Condenser Hydrogen
Control Systems, Mixed Atmosphere Requirements, and Associated Analysis
Requirements
The final rule retains the existing requirement in Sec.
50.44(c)(3)(i) to inert Mark I and II type containments. Given the
relatively small volume and large zirconium inventory, these
containments, without inerting, would have a high likelihood of failure
from hydrogen combustion due to the potentially large concentration of
hydrogen that a severe accident could cause. Retaining the requirement
maintains the current level of public protection, as discussed in
Section 4.3.2 of the Feasibility Study.
The final rule retains the existing requirements in Sec.
50.44(c)(3)(iv), (v), and (vi) that BWRs with Mark III containments and
PWRs with ice condenser containments provide a hydrogen control system
justified by a suitable program of experiment and analysis. The amount
of hydrogen to be considered is that generated from a metal-water
reaction involving 75 percent of the fuel cladding surrounding the
active fuel region (excluding the cladding surrounding the plenum
volume). The analyses must demonstrate that the structures, systems and
components necessary for safe shutdown and maintaining containment
integrity will perform their functions during and after exposure to the
conditions created by the burning hydrogen. Environmental conditions
caused by local detonations of hydrogen must be included, unless such
detonations can be shown unlikely to occur. A significant beyond
design-basis accident generating significant amounts of hydrogen (on
the order of Three Mile Island, Unit 2, accident or a metal water
reaction involving 75 percent of the fuel cladding surrounding the
active fuel region) would pose a severe threat to the integrity of
these containment types in the absence of the installed igniter
systems. Section 4.3.3 of the Feasibility Study concluded that hydrogen
combustion is not risk-significant, in terms of the framework
document's quantitative guidelines, when igniter systems installed to
meet Sec. 50.44(c)(3)(iv), (v), and (vi) are available and operable.
The NRC retains these requirements. Previously reviewed and approved
licensee analyses to meet the existing regulations constitute
compliance with this section. The results of these analyses must
continue to be documented in the plant's Updated Final Safety Analysis
Report in accordance with Sec. 50.71(e).
The final rule also retains the Sec. 50.44(b)(2) requirement that
containments for all currently-licensed nuclear power plants ensure a
mixed atmosphere. A mixed containment atmosphere prevents local
accumulation of combustible or detonable gases that could threaten
containment integrity or equipment operating in a local compartment.
B. Elimination of Design-Basis LOCA Hydrogen Release
The final rule removes the existing definition of a design-basis
LOCA hydrogen release and eliminates requirements for hydrogen control
systems to mitigate such a release at currently-licensed nuclear power
plants. The installation of recombiners and/or vent and purge systems
previously required by Sec. 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The NRC finds that this hydrogen release is
not risk-significant. This finding is based on the Feasibility Study
which found that the design-basis LOCA
[[Page 54126]]
hydrogen release did not contribute to the conditional probability of a
large release up to approximately 24 hours after the onset of core
damage. The requirements for combustible gas control that were
developed after the Three Mile Island Unit 2 accident were intended to
minimize potential additional challenges to containment due to long
term residual or radiolytically-generated hydrogen. The NRC found that
containment loadings associated with long term hydrogen concentrations
are no worse than those considered in the first 24 hours and therefore,
are not risk-significant. The NRC believes that accumulation of
combustible gases beyond 24 hours can be managed by licensee
implementation of the severe accident management guidelines (SAMGs) or
other ad hoc actions because of the long period of time available to
take such action. Therefore, the NRC eliminates the hydrogen release
associated with a design-basis LOCA from Sec. 50.44 and the associated
requirements that necessitated the need for the hydrogen recombiners
and the backup hydrogen vent and purge systems.
In plants with Mark I and II containments, the containment
atmosphere is required to be maintained with a low concentration of
oxygen, rendering it inert to combustion. Mark I and II containments
can be challenged beyond 24 hours by the long-term generation of oxygen
through radiolysis. The regulatory analysis for this proposed
rulemaking found the cost of maintaining the recombiners exceeded the
benefit of retaining them to prevent containment failure sequences that
progress to the very late time frame. The NRC believes that this
conclusion would also be true for the backup hydrogen purge system even
though the cost of the hydrogen purge system would be much lower
because the system also is needed to inert the containment.
The NRC continues to view severe accident management guidelines as
an important part of the severe accident closure process. Severe
accident management guidelines are part of a voluntary industry
initiative to address accidents beyond the design basis and emergency
operating instructions. In November 1994, current nuclear power plant
licensees committed to implement severe accident management at their
plants by December 31, 1998, using the guidance contained in NEI 91-04,
Revision 1, ``Severe Accident Issue Closure Guidelines.'' Generic
severe accident management guidelines developed by each nuclear steam
system supplier owners group includes either purging and venting or
venting the containment to address combustible gas control. On the
basis of the industry-wide commitment, the NRC is not requiring such
capabilities, but continues to view purging and/or controlled venting
of all containment types to be an important combustible gas control
strategy that should be considered in a plant's severe accident
management guidelines.
C. Oxygen Monitoring Requirements
The final rule amends Sec. 50.44 to codify the existing regulatory
practice of monitoring oxygen in currently-licensed nuclear power plant
containments that use an inerted atmosphere for combustible gas
control. Standard technical specifications and licensee technical
specifications currently require oxygen monitoring to verify the
inerted condition in containment. Combustible gases produced by beyond
design-basis accidents involving both fuel-cladding oxidation and core-
concrete interaction would be risk-significant for plants with Mark I
and II containments if not for the inerted containment atmosphere. If
an inerted containment was to become de-inerted during a significant
beyond design-basis accident, then other severe accident management
strategies, such as purging and venting, would need to be considered.
The oxygen monitoring is needed to implement these severe accident
management strategies, in plant emergency operating procedures, and as
an input in emergency response decision making.
The final rule reclassifies oxygen monitors as non safety-related
components. Currently, as recommended by the NRC's Regulatory Guide
(RG) 1.97, oxygen monitors are classified as Category 1. Category 1 is
defined as applying to instrumentation designed for monitoring
variables that most directly indicate the accomplishment of a safety
function for design-basis events. By eliminating the design-basis LOCA
hydrogen release, the oxygen monitors are no longer required to
mitigate design-basis accidents. The NRC finds that Category 2, defined
in RG 1.97, as applying to instrumentation designated for indicating
system operating status, to be the more appropriate categorization for
the oxygen monitors, because the monitors will still continue to be
required to verify the status of the inerted containment. Further, the
NRC believes that sufficient reliability of oxygen monitoring,
commensurate with its risk-significance, will be achieved by the
guidance associated with the Category 2 classification. Because of the
various regulatory means, such as orders, that were used to implement
post-TMI requirements, this relaxation may require a license amendment
at some facilities. Licensees would also need to update their final
safety analysis report to reflect the new classification and RG 1.97
categorization of the monitors in accordance with 10 CFR 50.71(e).
D. Hydrogen Monitoring Requirements
The final rule maintains the existing requirement in Sec.
50.44(b)(1) for monitoring hydrogen in the containment atmosphere for
all currently-licensed nuclear power plants. Section 50.44(b)(1),
standard technical specifications and licensee technical specifications
currently contain requirements for monitoring hydrogen, including
operability and surveillance requirements for the monitoring systems.
Licensees have made commitments to comply with design and qualification
criteria for hydrogen monitors specified in NUREG-0737, Item II.F.1,
Attachment 6 and in RG 1.97. The hydrogen monitors are required to
assess the degree of core damage during a beyond design-basis accident
and confirm that random or deliberate ignition has taken place.
Hydrogen monitors are also used, in conjunction with oxygen monitors in
inerted containments, to guide response to emergency operating
procedures. Hydrogen monitors are also used in emergency operating
procedures of BWR Mark III facilities. If an explosive mixture that
could threaten containment integrity exists, then other severe accident
management strategies, such as purging and/or venting, would need to be
considered. The hydrogen monitors are needed to implement these severe
accident management strategies.
The final rule reclassifies the hydrogen monitors as non safety-
related components for currently-licensed nuclear power plants. With
the elimination of the design-basis LOCA hydrogen release (see Item B.
earlier), the hydrogen monitors are no longer required to support
mitigation of design-basis accidents. Therefore, the hydrogen monitors
do not meet the definition of a safety-related component as defined in
Sec. 50.2. This is consistent with the NRC's determination that oxygen
monitors that are used for beyond-design basis accidents need not be
safety grade.
Currently, RG 1.97 recommends classifying the hydrogen monitors in
Category 1, defined as applying to instrumentation designed for
monitoring key variables that most directly indicate the accomplishment
of a safety function for design-basis
[[Page 54127]]
accident events. Because the hydrogen monitors no longer meet the
definition of Category 1 in RG 1.97, the NRC believes that licensees'
current commitments are unnecessarily burdensome. The NRC believes that
Category 3, as defined in RG 1.97, is an appropriate categorization for
the hydrogen monitors because the monitors are required to diagnose the
course of significant beyond design-basis accidents. Category 3 applies
to high-quality, off-the-shelf backup and diagnostic instrumentation.
As with the revision to oxygen monitoring, this relaxation may also
require a license amendment at some facilities. Licensees will also
need to update their final safety analysis report to reflect the new
classification and RG 1.97 categorization of the monitors in accordance
with 10 CFR 50.71(e).
E. Technical Specifications for Hydrogen and Oxygen Monitors
As discussed in III.C and III.D above, the amended rule requires
equipment for monitoring hydrogen in all containments and for
monitoring oxygen in containments that use an inerted atmosphere. The
rule also requires that this equipment must be functional, reliable,
and capable of continuously measuring the concentration of oxygen and/
or hydrogen in containment atmosphere following a beyond design-basis
accident for combustible gas control and severe accident management,
including emergency planning. Because of the importance of these
monitors for the management of severe accidents, the NRC staff
evaluated whether operability and surveillance requirements for these
monitors should be included in the technical specifications.
In order to be retained in the technical specifications, the
monitors must meet one of the four criteria set forth by 10 CFR 50.36.
These criteria are as follows:
1. Installed instrumentation that is used to detect, and indicate
in the control room, a significant abnormal degradation of the reactor
coolant pressure boundary.
2. A process variable, design feature, or operating restriction
that is an initial condition of a design basis accident or transient
analysis that either assumes the failure of or presents a challenge to
the integrity of a fission product barrier.
3. A structure, system, or component that is part of the primary
success path and that functions or actuates to mitigate a design basis
accident or transient that either assumes the failure of or presents a
challenge to the integrity of a fission product barrier.
4. A structure, system or component that operating experience or
probabilistic risk assessment has shown to be significant to public
health and safety.
As stated in the Federal Register notice (60 FR 36953) for the
final rule for technical specifications, these criteria were
established to address a ``trend toward including in technical
specifications not only those requirements derived from the analyses
and evaluations included in the safety analysis report but also
essentially all other Commission requirements governing the operation
of nuclear power plants. This extensive use of technical specifications
is due in part to a lack of well-defined criteria (in either the body
of the rule or in some other regulatory document) for what should be
included in technical specifications.'' As such, the NRC has decided,
and established by rule, not to duplicate regulatory requirements in
the technical specifications.
Hydrogen and oxygen monitors do not meet criteria 1, 2, or 3 of 10
CFR 50.36 described above. In addition, the Feasibility Study performed
by the NRC, and documented in section 4 of Attachment 2 of SECY-00-
0198, concluded that the requirement to provide a system to measure the
hydrogen concentration in containment does not contribute to the risk
estimates for core melt accidents for large dry containments; is not
risk significant during the early stages of core melt accidents for
Mark I and Mark II containments; and is not risk significant in terms
of dealing with the combustion threat of a core melt accident (except
for those conditions when the igniters are not operable, e.g., Station
Blackout) for Mark III and ice condenser containments. These
conclusions were based on the assumptions that Mark I and Mark II
containments are inert and hydrogen igniters are operable for Mark III
and ice condenser containments. It should be noted that the existing
technical specification requirements for hydrogen igniters and for
maintaining primary containment oxygen concentration below 4 percent by
volume (i.e., inerted), are not being removed; therefore, the
conclusions in the Feasibility Study on the risk significance of the
hydrogen monitors remain valid. On this basis, the NRC has concluded
that hydrogen monitors do not meet criterion 4 of 10 CFR 50.36.
Oxygen monitoring is not the primary means of indicating a
significant abnormal degradation of the reactor coolant pressure
boundary. Oxygen monitors are used to determine the primary containment
oxygen concentration in boiling water reactors. As stated above, the
limit for primary containment oxygen concentration for Mark I and II
containments will remain in technical specifications; therefore, a
technical specification requirement for oxygen monitors would be
redundant. In addition, technical specifications for hydrogen igniters
for Mark III containments will remain. The oxygen monitors have been
shown by probabilistic risk assessment to not be risk-significant. On
this basis, the NRC has concluded that oxygen monitors do not meet
criterion 4 of 10 CFR 50.36.
The NRC has several precedents regarding not duplicating regulatory
requirements for severe accidents in the technical specifications. The
Anticipated Transients Without Scram (ATWS) rule, (10 CFR 50.62)
requires each pressurized water reactor to have equipment from sensor
output to final actuation device, diverse from the reactor trip system,
to automatically initiate the auxiliary (or emergency) feedwater system
and initiate a turbine trip under conditions indicative of an ATWS.
This equipment is required to be designed to perform its function in a
reliable manner and has no associated requirements incorporated in the
technical specifications. The Station Blackout (SBO) rule, (10 CFR
50.63) requires that each light water reactor must be able to withstand
and/or recover from a station blackout event. Section 50.63 also states
that an alternate ac power source will constitute acceptable capability
to withstand station blackout provided an analysis is performed that
demonstrates that the plant has this capability from onset of the
station blackout until the alternate ac source and required shutdown
equipment are started and lined up to operate. Again, no requirements
for the alternate ac source are required to be in technical
specifications.
NRC experience with implementation of the above regulations for non
safety-related equipment has shown that reliability commensurate with
severe accident assumptions is assured without including such equipment
in technical specifications. According to the ``Final Report--
Regulatory Effectiveness of the Station Blackout Rule'' (ADAMS
ACCESSION NUMBER: ML003741781), the reliability of the alternate ac
power source has improved after implementation of the SBO rule. It
states:
``Before the SBO rule was issued, only 11 of 78 plants surveyed had
a formal EDG reliability program, 11 of 78 plants had a unit average
EDG reliability less that 0.95, and 2 of 78 had a unit average EDG
reliability of less that 0.90. Since
[[Page 54128]]
the SBO rule was issued, all plants have established an EDG reliability
program that has improved EDG reliability. A report shows that only 3
of 102 operating plants have a unit average EDG reliability less than
0.95 and above 0.90 considering actual performance on demand, and
maintenance (and testing) out of service (MOOS) with the reactor at
power.''
Therefore, the NRC staff has concluded that requirements for
hydrogen and oxygen monitors can be removed from technical
specifications. The basis for this conclusion is:
1. These monitors do not meet the criteria of 10 CFR 50.36,
2. The amended 10 CFR 50.44 requires hydrogen and oxygen monitors
to be maintained reliable and functional, and
3. The regulatory precedents set by the treatment of other
equipment for severe accidents required by 10 CFR 50.62 and 50.63.
F. Combustible Gas Control Requirements for Future Applicants
Section 50.44(c) of the final rule sets forth combustible gas
control requirements for all future water-cooled nuclear power reactor
designs with characteristics (e.g. type and quantity of cladding
materials) such that the potential for production of combustible gases
is comparable to currently-licensed light-water reactor designs. The
NRC's requirements for future reactors previously specified in Sec.
50.34(f)(2)(ix) have been reworded for conciseness but without material
change and relocated to Sec. 50.44(c)(2) to consolidate the
combustible gas control requirements in Sec. 50.44 for easier
reference. This sub-paragraph requires a system for hydrogen control
that can safely accommodate hydrogen generated by the equivalent of a
100 percent fuel clad metal-water reaction and must be capable of
precluding uniformly distributed concentrations of hydrogen from
exceeding 10 percent (by volume). If these conditions cannot be
satisfied, an inerted atmosphere must be provided within the
containment. The requirements specified in amended Sec. 50.44(c)(2)
are applicable to future water-cooled reactors with the same potential
for the production of combustible gas as currently-licensed light-water
reactor designs and are consistent with the criteria currently
contained in Sec. 50.34(f)(2)(ix) to preclude local concentrations of
hydrogen collecting in areas where unintended combustion or detonation
could cause loss of containment integrity or loss of appropriate
accident mitigating features. Additional advantages of providing
hydrogen control mitigation features (rather than reliance on random
ignition of richer mixtures) include the lessening of pressure and
temperature loadings on the containment and essential equipment. These
requirements reflect the Commission's expectation that future designs
will achieve a higher standard of severe accident performance (50 FR
32138; August 8, 1985).
Section 50.44(d) applies to non-water-cooled reactors and water-
cooled reactors that have different characteristics regarding the
production of combustible gases from current light-water reactors.
Because the specific details of the designs and construction materials
used in such future reactors cannot now be known, paragraph (d)
specifies a general performance-based requirement that future
applicants submit information to the NRC indicating how the safety
impacts of combustible gases generated during design-basis and
significant beyond design-basis accidents are addressed to ensure
adequate protection of public health and safety and common defense and
security. This information must be based in part upon a design-specific
probabilistic risk assessment. The Commission has endorsed the use of
PRAs as a tool in regulatory decisionmaking, see Use of Probabilistic
Risk Assessment Methods in Nuclear Activities: Final Policy Statement
(60 FR 42622, August 16, 1995), and is currently using PRAs as one
element in evaluating proposed changes to licensing bases for currently
licensed nuclear power plants, see Regulatory Guide 1.174, An Approach
for Using Probabilistic Risk Assessment in Risk-Informed
Decisionmaking: General Guidance (July 1998) and Standard Review Plan,
Chapter 19, ``Use of Probabilistic Risk Assessment in Plant-Specific,
Risk Informed Decisionmaking: General Guidance,'' NUREG-0800 (July
1998). The use of PRA methodologies in determining whether severe
accidents involving combustible gas must be addressed by future non-
water-cooled reactor designs (and water-cooled designs which have
different combustible gas generation characteristics as compared with
the current fleet of light-water-cooled reactors) is a logical
extension of the NRC's efforts to expand the use of PRAs in regulatory
decisionmaking.
At this time, the NRC is not able to set forth a detailed
description of, or specific criteria for defining a ``significant''
beyond design-basis accident for these future reactor designs, because
the fuel and vessel design, cladding material, coolant type, and
containment strategy for these reactor designs are unknown at the time
of this final rulemaking. Based in part upon the design-specific PRA,
the NRC will determine: (i) What type of accident is considered
``significant'' for each future reactor design, (ii) whether
combustible gas control measures are necessary, and if so, (iii)
whether the combustible gas control measures proposed for each design
provide adequate protection to public health and safety and common
defense and security. Although it is impossible at this time to provide
a detailed description or criteria for determining what constitutes a
``significant'' beyond design-basis accident for the future reactors
that are subject to this provision, the NRC nonetheless believes that
the concept of ``significant'' with respect to severe accidents has
regulatory precedent which will guide the NRC staff's evaluation of the
PRA information for future plants. Section 50.34(f)(2)(ix) of the NRC's
current regulations already defines what is in essence the significant
beyond design-basis accident which future reactor designs comparable to
current light-water reactor designs must be capable of addressing,
viz., an accident comparable to a degraded core accident at a current
light-water reactor in which a metal-water reaction occurs involving
100 percent of the fuel cladding surrounding the active fuel region
(excluding the cladding surrounding the plenum volume). With respect to
other ``beyond design-basis'' accidents, the Commission has addressed
anticipated transients without scram (ATWS), and station blackout,
which are currently regarded as ``beyond design-basis accidents.'' The
nuclear power industry, at the behest of the NRC, has developed severe
accident management guidelines to provide for a systematized approach
for responding to severe accidents during operations. Finally, the
Commission has required all nuclear power plant licensees to implement
emergency preparedness planning to address the potential for offsite
releases of radiation in excess of 10 CFR Part 100 limits. A careful
review of these regulatory efforts discloses a common thread:
regulatory actions addressing ``beyond design-basis'' accidents have
generally been determined based upon a consideration of probability of
the accident, together with consideration of the potential scope and
seriousness of the health and property value impacts to the general
public. Thus, it is possible to set forth a high-level conceptual
description of a ``significant'' beyond design-basis accident involving
combustible gas for which the
[[Page 54129]]
Commission intends for future non-water-cooled reactor designers to
address. First, such an accident would have relatively low probability
of occurrence, based upon the PRA, but would not be so small that the
accident would be deemed incredible. Second, a ``significant'' beyond
design-basis accident involving combustible gas would have serious
offsite consequences for the public, involving the potential for death
or significant acute or chronic health effects to the general public
and/or significant radioactive contamination of offsite property which
could result in permanent or long-term commitment of property to
nuclear use. Such accidents would typically call for activation of
offsite emergency preparedness measures in order to mitigate the
adverse effects on public health and safety.
The NRC is currently preparing a Draft Regulatory Guide DG-1122 for
public comment, in which the terms, ``significant sequences'' and
``significant contributors'' are expected to be addressed. In addition,
as part of the proposed rulemaking for risk-informing 10 CFR Sec.
50.46 the Commission has instructed the NRC staff to develop suitable
metrics for determining the appropriate risk cutoff for defining the
maximum LOCA size. The metrics are to take into account the
uncertainties inherent in development of PRAs. The NRC expects that
these regulatory activities will ultimately result in more detailed
examples of the ``significant beyond design-basis'' concept to assist a
potential applicant in developing the design for a future non-water-
cooled reactor (and water-cooled reactor designs which are
significantly different in concept from current light-water-cooled
reactors), and to guide the NRC's review of an application involving
such a design.
G. Clarification and Relocation of High Point Vent Requirements From 10
CFR 50.44 to 10 CFR 50.46a
The final rule removes the current requirements for high point
vents from Sec. 50.44 and transfers them to a new Sec. 50.46a. The
NRC is relocating these requirements because high point vents are
relevant to emergency core cooling system (ECCS) performance during
severe accidents, and the final Sec. 50.44 does not address ECCS
performance. The requirement to install high point vents was adopted in
the 1981 amendment to Sec. 50.44. This requirement permitted venting
of noncondensible gases that may interfere with the natural circulation
pattern in the reactor coolant system. This process is regarded as an
important safety feature in accident sequences that credit natural
circulation of the reactor coolant system. In other sequences, the
pockets of noncondensible gases may interfere with pump operation. The
high point vents could be instrumental for terminating a core damage
accident if ECCS operation is restored. Under these circumstances,
venting noncondensible gases from the vessel allows emergency core
cooling flow to reach the damaged reactor core and thus, prevents
further accident progression.
The final rule amends the language in Sec. 50.44(c)(3)(iii) by
deleting the statement, ``the use of these vents during and following
an accident must not aggravate the challenge to the containment or the
course of the accident.'' For certain severe accident sequences, the
use of reactor coolant system high point vents is intended to reduce
the amount of core damage by providing an opportunity to restore
reactor core cooling. Although the release of noncondensible and
combustible gases from the reactor coolant system will, in the short
term, ``aggravate'' the challenge to containment, the use of these
vents will positively affect the overall course of the accident. The
release of any combustible gases from the reactor coolant system has
been considered in the containment design and mitigative features that
are required for combustible gas control. Any reactor coolant system
venting is highly unlikely to affect containment integrity; however,
such venting will reduce the likelihood of further core damage. Because
overall plant safety is increased by venting through high point vents,
the final rule does not include this statement in Sec. 50.46a.
H. Elimination of Post-Accident Inerting
The final rule no longer provides an option to use post-accident
inerting as a means of combustible gas control. Although post-accident
inerting systems were permitted as a possible alternative for
mitigating combustible gas concerns after the accident at Three Mile
Island, Unit 2, no licensee has implemented such a system to date.
Concerns with a post-accident inerting system include increase in
containment pressure with use, limitations on emergency response
personnel access, and cost. Sections 50.44(c)(3)(iv)(D) and
50.34(f)(ix)(D) of the former rule were adopted to address these
concerns. On November 14, 2001, draft rule language was made available
to elicit comment from interested stakeholders. The draft rule language
recommended eliminating the option to use post-accident inerting as a
means of combustible gas control and asked stakeholders if there was a
need to retain these requirements. Stakeholder feedback supported
elimination of the post-accident inerting option and indicated that
licensees do not intend to convert existing plants to use post-accident
inerting. Because there is no need for the regulations to support an
approach that is unlikely to be used, the NRC has decided to eliminate
post-accident inerting requirements in the final rule.
IV. Comments and Resolution on Proposed Rule and Draft Regulatory Guide
The 60-day comment period for the proposed rule closed on October
16, 2002. The NRC received 14 letters, from 14 commenters, containing
approximately 43 comments on the proposed rule and draft regulatory
guide. Seven of the commenters were licensees, two were vendors, two
were representatives of utility groups (the Nuclear Energy Institute
and the Nuclear Utility Group on Equipment Qualification), two were
private citizens, and one was a citizen group, Nuclear Information and
Resource Service. All comments were considered in formulating the final
rule. Copies of the letters are available for public inspection and
copying for a fee at the Commission's Public Document Room, located at
11555 Rockville Pike, Room O-1 F23, Rockville, Maryland 20852.
Documents created or received at the NRC after October 16, 2002,
are also available electronically at the NRC's Public Electronic
Reading Room on the Internet at http://www.nrc.gov/reading-rm.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. These same documents also may be
viewed and downloaded electronically via the interactive rulemaking Web
site established by NRC for this rulemaking at http://ruleforum.llnl.gov.
The following sections set forth the resolution of the public
comments.
A. General Comments
Many commenters expressed strong support for the rule to improve
the regulations in Sec. 50.44 and ``commend[ed] the NRC for developing
a rule based on risk-informed and performance-based insights that would
eliminate unnecessary regulatory requirements.'' One industry commenter
indicated that this rule will enhance public health and safety because
it increases the reliability of the hydrogen and oxygen monitoring
systems. The Advisory Committee on Reactor
[[Page 54130]]
Safeguards (ACRS) stated that the draft proposed rulemaking for risk-
informed revisions to 10 CFR 50.44 will provide more effective and
efficient regulation to deal with combustible gases in containments.
The NRC also received feedback on several issues for which comments
were specifically requested in the draft rule language. The existing
rule provides detailed, prescriptive instructions using American
Society of Mechanical Engineers (ASME) references for analyzing the
performance of boiling water reactor (BWR) Mark III and pressurized
water reactor (PWR) ice condenser containments. In the final rule, the
NRC has provided an option for a more performance-based approach, which
received positive public comment. Based upon stakeholder input, the
final rule eliminates the existing references to ASME standards and
other prescriptive requirements. The regulatory guide attached to this
paper includes the ASME approach as one in which the intent of the
regulations could be satisfied.
One private citizen questioned why the NRC was considering relaxing
requirements that provide protection against some of the uncertainties
and hazards of nuclear power. A citizen group opposed the changes by
contending that eliminating the design-basis accident release, relaxing
safety classifications, and relaxing licensee commitments to certain
design and qualification criteria only benefits the money interests of
the licensees. This group also stated its belief that the NRC's
reliance on limited Three Mile Island (TMI) data points was
insufficient to relax requirements solely to accommodate industry cost
cutting strategies.
The NRC is moving to risk-informed, performance-based regulation
that takes into account the benefits and consequences of actions by
licensees and the NRC. One of the benefits of risk-informed regulation
is that it concentrates resources on areas that are more important and
minimizes resource allocation on areas that are shown to be less
significant. As part of the basis for deciding the level of importance
of various areas, during the 1980s and 1990s, the NRC sponsored a
severe accident research program to improve the understanding of core
melt phenomena, combustible gas generation, transport, and combustion,
and to develop improved models to predict the progression of severe
accidents. The results of this research have been incorporated into
various studies (e.g., NUREG-1150 and probabilistic risk assessments
performed as part of the Individual Plant Examination (IPE) program) to
quantify the risk posed by severe accidents for light water reactors.
The result of these studies has been an improved understanding of
combustible gas behavior during severe accidents and confirmation that
the combustible gas release postulated from a design-basis LOCA was not
risk-significant because it would not lead to early containment
failure, and that the risk associated with gas combustion was from
beyond-design-basis (e.g., severe) accidents.
In making its regulatory decisions, the NRC first considers public
safety, then other issues such as public confidence and reducing
unnecessary regulatory burden. Based upon the results of significant
research into design-basis and beyond design-basis accidents, the NRC
has determined that a design-basis combustible gas release is not risk-
significant and certain beyond design-basis combustible gas releases
are risk-significant. Therefore, the NRC is removing the requirements
for combustible gas control systems that mitigate consequences of non-
risk-significant design-basis accidents which are also not effective in
reducing the risk from combustible gas releases in beyond-design-basis
accidents.
The citizen group also contended that because GSI-191, ``Assessment
of Debris Accumulation on PWR Sump Pump Performance'', is not resolved,
removing the hydrogen recombiner requirements and relaxing the hydrogen
and oxygen monitoring requirements are premature and constitute a
dangerous trend towards risk ``misinformed'' regulation.
The NRC disagrees with the commenter's contention. The NRC's
philosophy on all GSIs is to first determine whether the existing
situation provides adequate protection of public health and safety, and
if there is sufficient margin to allow continued safe operation of the
affected plants while seeking a final resolution of the GSI. For GSI-
191, the NRC concluded that even though uncertainties remained
regarding the debris accumulation issue, adequate protection of public
health and safety was maintained. Accordingly, the fact that GSI-191
has not reached final resolution does not present an impediment to the
revision to Sec. 50.44.
An industry group requested that the terms ``safety-significant''
and ``industrial'' instead of high and low safety/risk significance be
used in this rule and regulatory guide. The NRC disagrees. The terms
``high and low safety/risk significance'' were not included in the
proposed rule and are not in the final rule. The term ``safety-
significant'', when used in supporting documentation, is used to
identify systems, structures, and components (SSCs) that contribute to
safety. The term does not confer the level of significance on the SSC.
Additionally, the term ``risk significant'' is used to identify those
conditions that contribute to risk. Again, no level of significance is
assigned by the use of this term. Additionally, the change in
terminology requested by the commenter would be inconsistent with the
supporting NRC documents and reports. Changing terminology could cause
unnecessary confusion on the part of licensees and the public.
B. General Clarifications
One commenter questioned if the draft regulatory guide would become
Regulatory Guide 1.7, Revision 3. When the NRC resolves the comments on
DG-1117, the guidance will be published as Regulatory Guide 1.7,
Revision 3.
A licensee requested that the first sentence of Item 3 of the
fourth paragraph of section B of the draft regulatory guide be revised
to read: ``The following requirements apply to all construction permits
or operating licenses under 10 CFR Part 50, and to all design
approvals, design certifications, or combined licenses under 10 CFR
Part 52, any of which are issued after the effective date of the
rule.'' The NRC agrees that the commenter's request represents a
clearer way of expressing the NRC's intent. In addition, the term
``manufacturing licenses'' has been added to make clear that the
revised requirements apply to applicants for manufacturing licensees,
which was inadvertently omitted from the proposed rule. These changes
have been included in both the regulatory guide and in the final rule.
The licensee also requested that the NRC reword the statement in
section 5 of the draft regulatory guide to read: ``For future
applicants and licensees as defined in Part 50.44(c), the analysis must
address an accident that releases hydrogen generated from 100 percent
fuel clad-coolant reaction accompanied by hydrogen burning.'' Another
licensee requested that section C.5, ``Containment Integrity'', should
state that it does not apply to currently licensed plants. The NRC
disagrees with these requests. Section 5 of DG-1117 was intended to
apply to current and future plants. However, the wording was not clear
and inadvertently caused some confusion on the applicability of the
section. To clarify that section 5 applies to current and future
plants, its wording has been revised to more closely reflect the rule
intent. This revision removes the following
[[Page 54131]]
statements from the draft regulatory guide: ``The analysis must address
an accident that releases hydrogen generated from 100 percent fuel
clad-coolant reaction accompanied by hydrogen burning. Systems
necessary to ensure containment integrity must also be demonstrated to
perform their function under these conditions.'' The above changes
remove the misleading language and clarify the applicability of the
section.
C. Monitoring Systems
A private citizen expressed concern about the adequacy and
survivability of non safety-related hydrogen and oxygen monitors for
assessing hydrogen and oxygen levels after an accident. A reactor
licensee stated that the changes to the requirements for hydrogen and
oxygen monitoring would actually increase the reliability of hydrogen
and oxygen monitoring equipment. A monitor vendor indicated that high-
quality commercial grade hydrogen monitors may be susceptible to
radiation-induced calibration degradation. The vendor also indicated
that these monitors are susceptible to damage from aerosols released
during the accident. The vendor believes that commercial grade
detectors located inside containment would probably not function in a
post-accident environment without verification testing and test-based
modifications. The vendor claimed the more severe the accident, the
less likely the sensors would properly operate due to increased
radiation exposure and increased aerosol loading. In addition, the
vendor believes that remote sampling lines for monitors located outside
of containment are susceptible to clogging from high-solid aerosols.
The vendor suggests it is prudent to retain the safety-related status
of hydrogen monitors to ensure comprehensive qualification testing.
The NRC believes that the changes to the requirements for hydrogen
and oxygen monitors will continue to ensure acceptable monitor
performance. If the changes result in a decrease in monitor
reliability, it will not be significant and will not affect public
health and safety because the functions served by the monitoring
systems are not risk-significant for core melt accident sequences. This
conclusion is supported by studies documented in the Feasibility Study
(Attachment 2 to SECY-00-0198) which indicate the relatively low risk
significance of monitoring systems. Because large, dry and sub-
atmospheric containments are robust enough to withstand the effects of
hydrogen combustion during full core melt accident sequences, hydrogen
monitoring is not risk-significant for these containment designs. For
BWR Mark I and Mark II containments, hydrogen monitoring systems are
not risk-significant in the early stages of a core melt accident
because these containments are inerted. For control of combustible
gases generated by radiolysis in the late stage of a core melt
accident, oxygen monitors are more important than hydrogen monitors for
these designs. For this reason, the design and qualification
requirements for oxygen monitors are more stringent than they are for
hydrogen monitors. During core melt accidents in BWR Mark III and ice
condenser containments, the hydrogen igniter systems are initiated by
high containment pressure. Because hydrogen monitors are not needed to
initiate or activate any mitigative features during these accidents,
they are not risk-significant for reducing the combustible gas threat
as long as the hydrogen igniters are operable. If the igniters are not
operating (such as during station blackout) hydrogen monitoring does
not reduce risk since the containment cannot be purged or vented
without electrical power. Nevertheless, the amended rule requires
licensees to retain hydrogen monitors (and oxygen monitors in Mark I
and Mark II BWRs) for their containments because they are useful in
implementing emergency planning and severe accident management
mitigative actions for beyond design basis accidents.
As noted in sections III C. and D. of this Supplementary
Information, as a consequence of eliminating the design-basis LOCA
hydrogen release, the oxygen and hydrogen monitors are no longer
required to mitigate potential consequences of combustible gases during
design-basis LOCA accidents; thus the monitors are not required to be
safety-related and need not meet the procurement, quality assurance,
and environmental qualification requirements for safety-related
components. Even though amended Sec. 50.44 reclassifies requirements
for monitoring systems, the hydrogen and oxygen monitoring systems are
still required by the rule to be functional, reliable, and capable of
continuously measuring the appropriate parameter in the beyond-design-
basis accident environment. Thus, licensees must consider the effects
of radiation exposure and high-solid aerosols on monitor performance if
they will be present in the post-accident environment for the specific
type of facility and monitoring system design. The change made by the
amended rule is that licensees are no longer required to use only
safety-grade monitoring equipment. For a particular facility and
monitoring system design, licensees will, in many cases, be able to
select appropriate, high quality, commercial-grade monitors that will
meet the performance requirements in the rule. In other cases, if no
suitable commercial-grade monitors are available, safety-grade monitors
may still be necessary. Also, because there are more types and designs
of commercial-grade monitors available than there are safety-grade, the
ability to use commercial-grade equipment may make it possible for
licensees to select a better-suited monitor for their particular
application. For example, it is stated in Attachment 2 to SECY-00-0198
that existing safety-grade hydrogen monitors have a limited hydrogen
concentration range and are not the optimum choice. Commercial-grade
monitors have the ability to monitor a wider range of hydrogen
concentration and could be a better solution.
Because the amended rule implements a performance-based requirement
for hydrogen and oxygen monitors to be functional, reliable, and
capable of continuously measuring the appropriate parameter in the
beyond-design-basis accident environment, licensees will have to ensure
that their procurement and quality assurance processes for such
equipment address equipment reliability and operability in the beyond
design basis accident environmental conditions for the specific
facility and monitoring system design. Licensees who do not consider
reliability and operability in appropriate environmental conditions
when designing and procuring monitoring equipment could be found by NRC
inspectors to be in violation of the amended rule.
Another vendor asked if additional requirements beyond commercial
grade will be imposed on the monitor's pressure retaining components
because the analyzer loop forms part of the containment boundary. The
monitor's pressure retaining components must meet current regulations
concerning containment penetrations. This vendor also asked if their
conclusion that grab samples cannot replace continuous monitoring is
correct. The NRC has determined that grab samples cannot replace
continuous monitoring. However, grab samples may be taken to verify
hydrogen concentrations in the latter stages of the accident response.
A vendor asked if two trains of equipment would be an appropriate
solution for ensuring analyzer availability. The NRC cannot respond to
[[Page 54132]]
such a question without more information about the reliability of each
individual train. Licensees are required to meet the requirements of
the rule. Individual licensees may determine how they will meet the
functionality, reliability, and capability requirements of the rule,
using appropriate guidance such as the regulatory guide, and subject to
NRC review and inspection.
A licensee requested that section C.2.2 of the draft regulatory
guide indicate that oxygen monitors are only required for plants that
inerted containments. The NRC agrees with the commenter that oxygen
monitors are only required for inerted containments, but disagrees with
the suggested addition. The first sentence of section C.2.2 already
states: ``The proposed Section 50.44 would require that equipment be
provided for monitoring oxygen in containments that use an inerted
atmosphere for combustible gas control.'' The final version of the
regulatory guide continues to indicate that oxygen monitoring is only
necessary for facilities that have inerted containments. Thus, the NRC
believes that the existing guidance is sufficient. This licensee also
requested that another statement in section C.2.2 of the draft
regulatory guide regarding existing oxygen monitoring commitments be
clarified to show that these systems meet the intent of the rule. The
NRC agrees with the need for clarification. The statement has been
revised to read: ``Existing oxygen monitoring systems approved by the
NRC prior to the effective date of the rule are sufficient to meet this
criterion.''
D. Purge
A licensee stated that the (model) safety evaluation (SE) should
address the acceptability of eliminating containment purge as the
design basis method for post-LOCA hydrogen control. The NRC disagrees.
The NRC model SE only addresses requirements in the standard technical
specifications or licensee technical specifications (TS). In this case,
the NRC model SE is for the elimination of the requirements of hydrogen
recombiners, and hydrogen and oxygen monitors from the TS. Because
containment purging requirements are not in the standard technical
specifications or licensees' technical specifications, the NRC model SE
does not make conclusions regarding the acceptability of eliminating
containment purging as the design basis method for post-LOCA hydrogen
control. However, the following statement from the Statements of
Considerations was added to the model SE to address the comment: ``. .
. the NRC eliminated the hydrogen release associated with a design-
basis LOCA from Sec. 50.44 and the associated requirements that
necessitated the need for the hydrogen recombiners and the backup
hydrogen vent and purge systems.''
E. Station Blackout/Generic Safety Issue 189
The citizens group stated that the proposed Sec. 50.44 should
require the deliberate ignition systems in Mark III and ice condenser
containments to be available during station blackout. This comment
pertains to resolution of GSI-189. The NRC disagrees with the
commenter. The evaluation and resolution of GSI-189 is ongoing and
proceeding independently of the rule as noted in Section II of this
Supplementary Information.
F. Containment Structural Uncertainties
The citizens group argues that the NRC does not have an adequate
non-destructive tool to eliminate concerns that containments were built
with voids in their walls, that all steel reinforcement bar was
improperly installed during construction to ensure uniform structural
integrity of containment walls, and that the concrete used in
containment walls is of sufficient quality that leaching of containment
walls has not weakened the structure. The commenter states that without
such non-destructive tools, it is unreasonable to reduce the defense-
in-depth strategy with the proposed rule. The commenter provided no
technical basis or information to support the assertion that
containments were inadequately constructed. The commenter also asserts
that the proposed rule creates an undue risk to the public health and
safety to solely accommodate the financial interest of the regulated
industry. Again, no technical basis was provided to support the
assertion of increased risk.
The NRC disagrees with the commenter. The NRC relies on several
layers of protection to prevent, detect, and repair defects discovered
during construction of concrete containments, including voids,
improperly installed reinforcement bar, and low quality concrete. These
layers of protection include:
(1) The implementation by the licensee of their NRC-approved 10 CFR
Part 50, Appendix B, Quality Assurance (QA) program and the licensee's
Quality Control (QC) program;
(2) The requirements of 10 CFR 50.55(e) that holders of
Construction Permits identify, evaluate, and report defects and
failures to comply with NRC requirements associated with substantial
safety hazards to the NRC in a timely manner, generally within 60 days;
and
(3) The verification by NRC inspectors as defined by the NRC's
construction inspection program contained in NRC Inspection Manual
Chapter 2512 that the construction is in accordance with approved
design documents, that the licensee is properly and effectively
implementing their QA/QC program, that construction defects are
reported to NRC as required by 10 CFR 50.55(e), and that appropriate
corrective actions are taken by the licensee.
Whenever there is a doubt about the proper locations of reinforcing
bars, or voids in a concrete containment structure, appropriate non
destructive examination methods and conservative analysis are used by
the licensees to demonstrate that the containment and its vital
components are able to perform their intended functions.
In addition, the pre-operational performance of the Structural
Integrity Test (SIT) provides an added assurance by physically
demonstrating the overall structural capability of a concrete
containment. Also, 10 CFR 50.65, the maintenance rule, requires
licensees to monitor the performance or condition of certain structures
to provide reasonable assurance that the structures are capable of
fulfilling their intended function throughout the life of the plant.
Licensees must also periodically inspect and test their containments in
accordance with the ASME Boiler and Pressure Vessel Code, Section XI,
Subsection IWL, and Appendix J to 10 CFR Part 50. Finally, at plants
that have renewed their licenses, aging management programs are in
effect to monitor containment structures to ensure that aging does not
significantly degrade their functional capability.
G. PRA/Accident Analysis
An individual submitted questions in three areas. First, the
commenter asked why the 30-minute initiation time for initiating
hydrogen monitoring was overly burdensome and suggested that the
proposed 90-minute initiation time was arbitrary. The NRC disagrees
with the commenter. The 30-minute initiation time was developed
following the TMI-2 accident based on engineering judgement on the time
within which the hydrogen monitors needed to be made functional.
Putting this equipment into service within 30 minutes, as directed in
NUREG-0737, was found by some utilities during severe accident training
(e.g., on nuclear power plant simulators) to be unnecessarily
distracting to operators,
[[Page 54133]]
because it took them away from more important tasks that needed to be
implemented in the near term while the monitoring did not need to be
initiated for a longer period. The NRC has determined that performance-
based functional requirements rather than prescriptive requirements
achieve the desired goal of hydrogen monitor functionality while giving
licensees an opportunity to better use operators' time during an
accident. The noted 90 minutes come from the time licensees found was
needed to get the monitors running in a manner that still met the goal
of monitoring hydrogen levels and allowed sufficient time for other
operator actions based on severe accident emergency operating
procedures. Thus, the 90 minute time period was a result of changing to
a performance-based approach and was not arbitrarily specified as the
time within which the operators had to act.
The individual also stated that the proposed rule was reducing
``defense in depth'' and that if a utility cannot afford to operate and
maintain its nuclear power reactors with the requisite caution and
oversight, then the utility should not operate them at all. The NRC
disagrees with the commenter's assertion that the amended regulations
do not provide adequate defense-in-depth. Defense-in-depth continues to
be a prime consideration in NRC decision making. The NRC makes its
decisions considering public safety first. Only after public safety is
ensured are other issues such as public confidence and reduction of
unnecessary burden considered. Defense-in-depth is an element of the
NRC's safety philosophy that employs successive measures to prevent
accidents or mitigate damage if a malfunction, accident, or naturally
caused event occurs at a nuclear facility. It provides redundancy as
well as the philosophy of a multiple-barrier approach against fission
product releases. Defense-in-depth does not mean that equipment
installed in a nuclear power plant never should be removed. Adequate
defense-in-depth may be achieved through multiple means or paths.
The commenter also questioned whether the NRC staff has adequate
data to demonstrate that the amount of residual and radiolytically-
generated combustible gases generated during a design-basis LOCA would
not be risk-significant--especially if the LOCA occurred in a plant
with older fuel and SSCs than were present during the accident at Three
Mile Island, Unit 2. The NRC disagrees with the commenter's assertion
that insufficient information is known about hydrogen generation to
support amending the current regulations. The amount of hydrogen
generated during a design-basis LOCA is not affected by the relative
age or vintage of reactor fuel or SSCs. The NRC has developed
significant data and insights on the behavior of design-basis and
severe accidents after the TMI-2 accident. In amending Sec. 50.44 in
1985, the NRC recognized that an improved understanding of the behavior
of accidents involving severe core damage was needed. During the 1980s
and 1990s, the NRC devoted significant resources and sponsored a severe
accident research program to improve the understanding of core melt
phenomena; combustible gas generation, transport, and combustion; and
to develop improved models to predict the progression of severe
accidents. The results of this research have been incorporated into
various studies (e.g., NUREG-1150 and probabilistic risk assessments
performed as part of the Individual Plant Examination (IPE) program) to
quantify the risk posed by severe accidents for light water reactors.
The result of these studies has been an improved understanding of
combustible gas behavior during severe accidents. One of the insights
from these studies is confirmation that the hydrogen release postulated
from a design-basis LOCA was not risk-significant because it would not
lead to early containment failure. In addition, it was found that the
vast majority of the risk associated with hydrogen combustion was from
beyond design-basis (e.g., severe) accidents. The amended requirements
are based on the NRC's careful consideration of the post-Three Mile
Island information.
H. Passive Autocatalytic Recombiners
An individual questioned why the United States was allowing the
removal of recombiners while the French are requiring the installation
of passive autocatalytic recombiners in their reactors. The NRC has
determined that passive autocatalytic recombiners (PARs) do not need to
be considered for U.S. PWRs with large-dry containments or sub-
atmospheric containments. This conclusion was drawn after applying the
quantitative and qualitative criteria in the form of a framework for
risk-informed changes to technical requirements of 10 CFR Part 50 (See
attachment 1, SECY-00-0198). The NRC found that hydrogen combustion is
not a significant threat to the integrity of large, dry containments or
sub-atmospheric containments when compared to the 0.1 conditional large
release probability of the framework document. In SECY-00-0198, the NRC
also concluded that additional combustible gas control requirements for
currently licensed large-dry and sub-atmospheric containments were
unwarranted.
I. Reactor Venting
An individual expressed concern for the elimination of the
requirement prohibiting venting the reactor coolant system if it would
aggravate the challenge to containment. According to the comment, the
venting could cause an increase in the radiological effluents released
off site and an increase in public exposure. The NRC disagrees with the
individual's conclusion. As noted in section III.F of this
Supplementary Information, the requirement to install high point vents
was imposed by the 1981 amendment to Sec. 50.44. This requirement
permitted venting of noncondensible gases that may interfere with the
natural circulation pattern in the reactor coolant system. This process
is regarded as an important safety feature in accident sequences that
credit natural circulation of the reactor coolant system. In other
sequences, the pockets of noncondensible gases may interfere with pump
operation. The high point vents could be instrumental for terminating a
core damage accident if ECCS operation is restored. Under these
circumstances, venting noncondensible gases from the vessel allows
emergency core cooling flow to reach the damaged reactor core and thus,
prevents further accident progression.
For certain severe accident sequences, the use of reactor coolant
system high point vents is intended to reduce the amount of core damage
by providing an opportunity to restore reactor core cooling. Although
the release of noncondensible and combustible gases from the reactor
coolant system could, in the short term, ``aggravate'' the challenge to
containment, the use of these vents will positively affect the overall
course of the accident. The release of combustible gases from the
reactor coolant system has been considered in the containment design
and mitigative features that are required for combustible gas control.
Any venting is highly unlikely to affect containment integrity or cause
an increase in the radiological effluents released off site that could
potentially increase public radiation exposure. However, such venting
may reduce the likelihood of further core damage. The reduction in core
damage would reduce both the generation of combustible gases and the
magnitude of the radiological source term that could be released, thus
[[Page 54134]]
reducing the potential for public exposure.
An industry organization requested a revision in a statement in
section III.F in the statement of considerations (SOC) concerning the
purposes of the high point vents from: `` * * * venting noncondensible
gases from the vessel allows emergency core cooling flow to reach the
damaged core and thus prevents further accident progression'' to `` * *
* the purpose of the high point venting is to ensure that natural
circulation cooling is an option for maintaining a long term safe
stable state following a core damage accident in which significant
amounts of noncondensible gases, such as hydrogen might be generated
and retained in the reactor coolant system.'' The NRC disagrees with
the comment and believes the current wording is adequate. Other
information in section III.F adequately defines the purpose of high
point vents by acknowledging their usefulness both for forced
circulation scenarios and in the natural circulation mode.
J. Design Basis Accident Hydrogen Source Term
A private citizen questioned that because an unexpected hydrogen
bubble and an unexpected hydrogen burn occurred during the accident at
Three Mile Island, should hydrogen buildup be considered a known risk
for which licensees should try to monitor and control as thoroughly as
possible? The NRC agrees with the commenter that hydrogen generation
during severe accidents is an expected phenomenon. After the TMI
accident, the NRC has sponsored an extensive research program on the
behavior of severe accidents. This program was designed improve the
understanding of core melt phenomena; combustible gas generation,
transport, and combustion; and to develop improved models to predict
the progression of severe accidents. The results of this research have
been incorporated into various studies (e.g., NUREG-1150 and
probabilistic risk assessments performed as part of the Individual
Plant Examination (IPE) program) to quantify the risk posed by severe
accidents for water-cooled reactors.
The result of these studies has been an improved understanding of
combustible gas behavior during severe accidents and confirmation that
the combustible gas release postulated from a design-basis LOCA was not
risk-significant because it would not lead to early containment
failure, and that the risk associated with gas combustion was from
beyond-design-basis (e.g., severe) accidents. Thus, the requirements
for control and monitoring of combustible gases are being reduced for
the non-risk-significant design-basis accident scenarios. The amended
regulations are entirely consistent with and justified by the findings
of the post-TMI studies.
K. Requested Minor Modifications
An industry group requested that the last paragraph of Section B of
the draft regulatory guide be changed to read: ``The treatment
requirements for the safety-significant components in the combustible
gas control systems, the atmospheric mixing systems and the provisions
for measuring and sampling are delineated in Section C, Regulatory
Position.'' The NRC disagrees with the requested change. Section 50.44
is being revised to eliminate unnecessary requirements relating to
combustible gas control in containment. The remaining requirements have
been determined by the NRC to be necessary to mitigate the risk
associated with combustible gas generation. The regulatory guide
provides recommended treatments for all structures, systems, and
components credited for meeting those requirements. Because the
regulatory guide is only guidance, licensees are free to devise their
own treatments for these structures, systems, and components, subject
to NRC review and inspection.
L. Atmosphere Mixing
A private citizen suggested adding criteria to the regulatory guide
to assess the adequacy of the performance of atmosphere mixing systems.
The NRC disagrees with the commenter that these criteria are needed.
The NRC has already evaluated the adequacy of atmosphere mixing at
currently operating pressurized and boiling water reactors. However,
for future water-cooled reactor designs, the NRC has decided to specify
that containments must have the capability for ensuring a mixed
atmosphere during ``design-basis and significant beyond design-basis
accidents''. Other guidance on determining the adequacy of atmosphere
mixing systems is also provided in the rule and the regulatory guide.
An industry group requested that the SOC and regulatory guide be
revised to only impose requirements on safety-significant hydrogen
(atmospheric) mixing systems. They contend that some large dry
containments have hydrogen mixing systems in addition to containment
fan cooler units. The fan cooler units are supposedly the prime mode of
ensuring a mixed atmosphere; therefore, the hydrogen mixing systems are
classified as low safety-significance. The industry group believes that
regulatory requirements should not be imposed on low safety-significant
equipment. The NRC disagrees with the requested change. Section 50.44
is being revised to eliminate unnecessary requirements relating to
combustible gas control in containment. The remaining requirements have
been determined by the NRC to be necessary to mitigate the risk
associated with combustible gas generation. The regulatory guide
provides recommended treatments for all structures, systems, and
components credited for meeting those requirements. Because the
regulatory guide only provides guidance, licensees are free to devise
their own treatments for these structures, systems, and components,
subject to NRC review and inspection.
M. Current Versus Future Reactor Facilities
An industry group requested that Sec. 50.44(c) be amended to
clarify that its requirements relate only to light-water reactors. The
NRC acknowledges that the proposed requirements in Sec. 50.44(c) were
largely patterned after light-water reactor requirements and might not
be specifically applicable to all types of future light-water and non
light-water reactor designs. Therefore, the NRC has modified Sec.
50.44(c) to apply only to future water-cooled reactors with
characteristics such that the potential for production of combustible
gases during design-basis and significant beyond design-basis accidents
is comparable to current light-water reactor designs. In addition, the
NRC has added a new paragraph (d) that specifies combustible gas
control information to be provided by applicants for future reactor
designs when the potential for the production of combustible gases is
not comparable to current light-water reactor designs. The purpose of
this information is to determine if combustible gas generation is
technically relevant to the proposed design; and, if so, to demonstrate
that safety impacts of combustible gases generated during design-basis
and significant beyond design-basis accidents have been addressed in
the design of the facility to ensure adequate protection of public
health and safety and common defense and security.
The industry group also commented that the regulatory guide is
unclear on what parts are applicable to existing reactors and what
parts are applicable to future reactors. The Introduction and section B
do not agree. The NRC agrees. The regulatory guide has been modified to
clarify the applicability of the revised Sec. 50.44 to present and
future water-
[[Page 54135]]
cooled and non water-cooled reactors. The industry group also noted
that the proposed language, the draft regulatory guide, and the
proposed change to the Standard Review Plan incorrectly assume that all
new reactor designs will be light-water reactors and will present the
same combustible gas hazard. Future reactors, whether light-water or
non-light-water may use different materials, cooling, or moderating
mediums that may not result in the production of the same combustible
gases, or quantities of combustible gas as the current light-water
reactor designs. The NRC agrees. For the reasons given above, the final
rule, the regulatory guide, and the standard review plan have all been
modified to clarify their applicability to future reactor designs.
N. Equipment Qualification/Survivability
A licensee suggested adding a clarifying statement to the SOC
concerning equipment survivability for Mark III and ice condenser
plants. The commenter requested a statement clearly stating that no new
equipment survivability requirements are being imposed and that
existing equipment survivability and environmental analyses remain
valid for compliance with the revised rule. The NRC agrees with
commenter that the rule does not impose any additional equipment
survivability requirements on licensees; existing equipment
survivability and environmental analyses remain valid. The hydrogen and
oxygen monitoring systems are required by the rule to be functional,
reliable, and capable of continuously measuring the appropriate
parameter in the beyond design-basis accident environment.
This licensee also noted that, due to the reclassification of the
hydrogen and oxygen monitors from RG 1.97 Category I to lower
categories, these monitors no longer have to be qualified in accordance
with 10 CFR 50.49. The NRC agrees that the monitoring equipment need
not be qualified in accordance with Sec. 50.49. The hydrogen and
oxygen monitoring systems are still required by the rule to be
functional, reliable, and capable of continuously measuring the
appropriate parameter in the beyond design-basis accident environment.
The licensee suggested that the NRC clarify that the revised rule
will not affect the requirements or environmental conditions used by
licensees to demonstrate compliance with Sec. 50.49. The NRC agrees
with the commenter that existing licensee analyses and environmental
conditions used to establish compliance with 10 CFR 50.49 will not be
affected by the amended rule and that no new analyses or environmental
conditions are imposed by these amendments to Sec. 50.44.
V. Petitions for Rulemaking--PRM-50-68
The NRC received a petition for rulemaking submitted by Bob
Christie of Performance Technology, Knoxville, Tennessee, in the form
of two letters dated October 7, 1999, and November 9, 1999. The
petition requested that the NRC amend its regulations concerning
hydrogen control systems at nuclear power plants. The petitioner
believes that the current regulations on hydrogen control systems at
some nuclear power plants are detrimental and present a health risk to
the public. The petitioner believes that similar detrimental situations
may apply to other systems as well (such as the requirement for a 10-
second diesel start time). The petitioner believes his proposed
amendments would eliminate those situations associated with hydrogen
control systems that present adverse conditions at nuclear power
plants. The petition was docketed as PRM-50-68 on November 15, 1999. On
January 12, 2000 (65 FR 1829), the NRC published a notice of receipt of
this petition in the Federal Register that summarized the issues it
contains.
Specifically, the petitioner performed a detailed review of the San
Onofre Task Zero Safety Evaluation Report (Pilot Program for Risk-
Informed Performance-Based Regulation) conducted by the NRC staff and
dated September 3, 1998, concerning that plant's hydrogen control
system. The petitioner requested that the NRC:
1. Retain the existing requirement in Sec. 50.44(b)(2)(i) for
inerting the atmosphere of existing Mark I and Mark II containments.
2. Retain the existing requirement in Sec. 50.44(b)(2)(ii) for
hydrogen control systems in existing Mark III and PWR ice condenser
containments to be capable of handling hydrogen generated by a metal/
water reaction involving 75 percent of the fuel cladding.
3. Require all future light water reactors to postulate a 75
percent metal/water reaction (instead of the 100 percent required by
the current rule) for analyses undertaken pursuant to Sec. 50.44(c).
4. Retain the existing requirements in Sec. 50.44 for high point
vents.
5. Eliminate the existing requirement in Sec. 50.44(b)(2) for a
mixed atmosphere in containment.
6. Eliminate the existing requirement for hydrogen releases during
design basis accidents of an amount equal to that produced by a metal/
water reaction of 5 percent of the cladding.
7. Eliminate the requirement for hydrogen recombiners or purge in
LWR containments.
8. Eliminate the existing requirements for hydrogen and oxygen
monitoring in LWR containments.
9. Revise GDC 41--Containment Atmosphere Cleanup--to require
systems to control fission products and other substances that may be
released into the reactor containment for accidents only where there is
a high probability that fission products will be released to the
reactor containment.
10. Issue an interim policy statement applicable to all NRC staff
to ensure that the NRC Executive Director for Operations was promptly
notified whenever staff discovered cases where compliance with design-
basis accident requirements was detrimental to public health.
The NRC received five comment letters on PRM-50-68. The commenters
included two nuclear power plant licensees, a nuclear reactor vendor, a
nuclear power plant owners group, and the Nuclear Energy Institute
(NEI). Copies of the public comments on PRM-50-68 are available for
review in the NRC Public Document Room, 11555 Rockville Pike,
Rockville, Maryland. All commenters were supportive of some of the
issues raised by the petition. One of the reactor licensees commented
that analytical and risk bases exist to support the proposed changes
for Mark I Boiling Water Reactor containments. The other licensee
endorsed the comments submitted by NEI. The reactor vendor commented
that the petitioner's proposal simplifies the language and requirements
of the regulation while retaining an equivalent level of safety.
However, the vendor also noted that the proposal does not appear to
address the structural integrity of the containment as in the existing
language at Sec. 50.44(c)(3)(iv). The owner's group commented that the
changes requested by the petitioner for large, dry containments were
also applicable to ice condenser containments and suggested that the
requirement for all hydrogen control measures in Sec. 50.44 be
reexamined and made ``consistent with many other portions of plant
operation and maintenance.'' The NEI agreed with the petitioner that
the San Onofre hydrogen control licensing actions could be applied
generically for pressurized water reactors with large, dry (including
subatmospheric) containments. One licensee, the reactor vendor and the
NEI disagreed with the petitioner's position that an interim policy
statement is necessary to instruct
[[Page 54136]]
the NRC staff how to proceed in instances when ``adherence to design
basis requirements would be detrimental to public health.'' The other
commenters were silent regarding the request for an interim policy
statement.
The NRC has evaluated the technical issues and the associated
public comments and has determined that the specific issues contained
in PRM-50-68 should be granted in part and denied in part as discussed
in the following paragraphs.
Issue 1: Retain the existing requirement for inerting the
atmosphere of existing Mark I and Mark II containments.
Resolution of Issue 1: Consistent with the petitioner's request,
Sec. 50.44(b)(2)(i) of the final rule retains the current requirement
for inerting of existing Mark I and Mark II containments. The NRC's
basis for this decision is provided in section III A. of this document.
Issue 2: Retain the existing requirement for hydrogen control
systems in existing Mark III and PWR ice condenser containments to be
capable of handling hydrogen generated by a metal/water reaction
involving 75 percent of the fuel cladding.
Resolution of Issue 2: Consistent with the petitioner's request,
Sec. 50.44(b)(2)(ii) of the final rule retains the above requirement
for hydrogen control systems in existing Mark III and PWR ice condenser
containments to be capable of handling hydrogen generated by a metal/
water reaction involving 75 percent of the fuel cladding. The NRC's
basis for this decision is provided in section III A. of this document.
Issue 3: Require all future light water reactors to postulate a 75
percent metal/water reaction (instead of the 100 percent required by
the current rule) for analyses under Sec. 50.44(c).
Resolution of Issue 3: The NRC declines to adopt this request. For
future water-cooled reactors, the final rule retains the previous
requirement to postulate hydrogen generation by a 100 percent metal/
water reaction when performing structural analyses of reactor
containments under accident conditions. Future containments that cannot
structurally withstand the consequences of this amount of hydrogen must
be inerted or must be equipped with equipment to reduce the
concentration of hydrogen during and following an accident. The NRC's
basis for this decision is provided in section III E. of this document.
Issue 4: Retain the existing requirements for high point vents.
Resolution of Issue 4: Consistent with the petitioner's request,
the requirements for high point vents in former 10 CFR 50.44(c)(3)(iii)
have been retained in the final rule, but have been modified slightly
to clarify the acceptable use of these vents during and following an
accident. Because the need for high point vents is relevant to ECCS
performance during severe accidents and is not pertinent to combustible
gas control, these high point venting requirements have been removed
from 10 CFR 50.44 and relocated to 10 CFR 50.46a where the remaining
requirements for ECCS are located. The basis for this decision is
provided in section III F. of this document.
Issue 5 Eliminate the existing requirement in Sec. 50.44(b)(2) to
ensure a mixed atmosphere in containment.
Resolution of Issue 5: The NRC declines to adopt this request. The
final rule retains the requirement for all containments to ensure a
mixed atmosphere to prevent local accumulation of combustible or
detonable gasses that could threaten containment integrity or equipment
operating in a local compartment. The NRC's basis for retaining this
requirement is provided in section III A. of this document.
Issue 6: Eliminate the existing requirement for postulating design
basis accident hydrogen releases of an amount equal to that produced by
a metal/water reaction of 5 percent of the cladding.
Resolution of Issue 6: The NRC grants this request. The NRC has
determined that hydrogen release during design basis accidents is not
risk-significant because it does not contribute to the conditional
probability of a large release of radionuclides up to approximately 24
hours after the onset of core damage. The NRC believes that
accumulation of combustible gases beyond 24 hours can be managed by
implementation of severe accident management guidelines. The NRC's
technical basis for eliminating this requirement is discussed in
greater detail in section III B. of this document.
Issue 7: Eliminate the requirement for hydrogen recombiners or
purge in light-water reactor containments.
Resolution of Issue 7: The NRC grants this request. As noted in
Issue 6 above, the NRC has determined that hydrogen release during
design basis accidents is not risk-significant because it does not
contribute to the conditional probability of a large release of
radionuclides up to approximately 24 hours after the onset of core
damage. The NRC believes that accumulation of combustible gases beyond
24 hours can be managed by implementation of severe accident management
guidelines. Thus, hydrogen recombiners and hydrogen vent and purge
systems are not required. The NRC's basis for eliminating these
requirements is discussed in greater detail in section III B. of this
document.
Issue 8: Eliminate the existing requirements for hydrogen and
oxygen monitoring in light-water reactor containments.
Resolution of Issue 8: The NRC declines to adopt this request. The
final rule retains the existing requirement for monitoring hydrogen in
the containment atmosphere for all plant designs. Hydrogen monitors are
required to assess the degree of core damage during beyond design-basis
accidents. Hydrogen monitors are also used in conjunction with oxygen
monitors to guide licensees in implementation of severe accident
management strategies. Also, the NRC has decided to codify the existing
regulatory practice of monitoring oxygen in containments that use an
inerted atmosphere for combustible gas control. If an inerted
containment became de-inerted during a beyond design-basis accident,
other severe accident management strategies, such as purging and
venting, would need to be considered. Monitoring of both hydrogen and
oxygen is necessary to implement these strategies. The NRC's bases for
these requirements are discussed in greater detail in sections III C.
and III D. of this document.
Issue 9: Revise GDC 41--Containment Atmosphere Cleanup--to require
systems to control fission products and other substances that may be
released into the reactor containment for accidents only when there is
a high probability that fission products will be released to the
reactor containment.
Resolution of Issue 9: The NRC declines to adopt the petitioner's
request on this issue. The NRC believes that the amended rule
alleviates the need to revise Criterion 41. In a December 4, 2001,
letter from the petitioner to the NRC, the petitioner inferred that the
intent of the proposed change was to focus Criterion 41 on the
containment capability when a severe accident occurs. This concern is
addressed in the final Sec. 50.44 that establishes the design criteria
for reactor containment and associated equipment for controlling
combustible gas released during a postulated severe accident. The
General Design Criteria in Appendix A of 10 CFR Part 50 were
established to set the minimum requirements for the principal design
criteria for water-cooled nuclear power plants. The postulated
accidents used in the development of these minimum design criteria are
normally design-basis accidents. The NRC believes it is not
[[Page 54137]]
appropriate to address severe accident design requirements in the
General Design Criteria.
Issue 10: The petitioner requested the NRC to issue an interim
policy statement applicable to the NRC staff to ensure that the NRC
Executive Director for Operations was promptly notified whenever the
staff discovered cases where compliance with design-basis accident
requirements was detrimental to public health.
Resolution of Issue 10: The petitioner's additional request for an
interim policy statement is not part of the petition for rulemaking.
Nevertheless, the NRC has evaluated the request and associated public
comments and has concluded that hydrogen control requirements
referenced by the petitioner have been modified in the final rule so
that design basis requirements ensure adequate protection of public
health and safety. The NRC also believes that if NRC staff members
discover future situations when design basis requirements detract from
safety, the staff will elevate these issues for management review;
thus, no NRC staff guidance in this area is necessary.
Petition for Rulemaking--PRM-50-71
The NRC also received a petition for rulemaking submitted by NEI.
The petition, dated April 12, 2000, was published in the Federal
Register for public comment on May 31, 2000 (65 FR 34599). The
petitioner requested that the NRC amend its regulations to allow
nuclear power plant licensees to use zirconium-based cladding materials
other than Zircaloy or ZIRLO, provided the cladding materials meet the
requirements for fuel cladding performance and have been approved by
the NRC staff. The petitioner believes the proposed amendment would
improve the efficiency of the regulatory process by eliminating the
need for individual licensees to obtain exemptions to use advanced
cladding materials that have already been approved by the NRC.
Specifically, the petitioner states that the NRC's current
regulations require uranium oxide fuel pellets, used in commercial
reactor fuel, to be contained in cladding material made of Zircaloy or
ZIRLO. The petitioner indicates that the requirement to use either of
these materials is stated in Sec. 50.44 and Sec. 50.46. The
petitioner notes that subsequent to promulgation of these regulations,
commercial nuclear fuel vendors have developed and continue to develop
materials other than Zircaloy or ZIRLO that the NRC reviews and
approves for use in commercial power reactor fuel. Each of these
approvals requires the NRC to grant an exemption to the licensee that
requests to use fuel with these cladding materials. The petitioner
requests that the NRC amend its regulations to allow licensees
discretion to use zirconium-based cladding materials other than
Zircaloy or ZIRLO, provided that the cladding materials meet the fuel
cladding performance requirements and have been reviewed and approved
by the NRC staff. The petitioner notes that during the past nine years
there have been at least eight requests for exemptions and that each
exemption has cost more than $50,000. The petitioner states that the
requests for exemptions have become increasingly more frequent, causing
significant administrative confusion and having a potentially adverse
effect on efficient and effective use of NRC, licensee, and vendor
resources.
The petitioner believes the NRC should amend Sec. 50.44 and Sec.
50.46 to allow the use of other zirconium-based alloys in addition to
those specified in the current regulations. The petitioner states that
the stated goal of the existing regulations is to ensure adequate
cooling for reactor fuel in case of a design-basis accident. However,
the petitioner asserts that the proposed amendment does not degrade the
ability to meet that goal. The petitioner believes it removes an
unwarranted licensing burden without increasing risk to public health
and safety.
The NRC received 11 comment letters on PRM 50-71. Seven comments
were from nuclear reactor licensees, two from individual members of the
public, one from a nuclear reactor vendor and one from a nuclear
industry trade association (NEI). Five of the nuclear reactor licensees
were supportive of the petition and endorsed the comments and positions
provided by NEI in their comments on the petition. One licensee stated
that the proposed rule should note that if a fuel vendor's cladding has
met the requirements for use on a generic basis, a process for the
implementing utility to use that fuel under their existing license
already exists. Another licensee agreed that industry needs relief on
use of zirconium-based cladding, but because cladding is a critical
safety barrier, the basis for relief should come from proven, in-
reactor performance. A better approach would be to update the approved
list of allowed fuel rod cladding materials as more products
demonstrate reliable, in-reactor performance.
Two comments were received from individuals. One individual opposed
the petition because it did not contain the specific review and
acceptance criteria that NRC would utilize when reviewing and approving
future cladding materials under the proposed rule. The commenter also
opposed the practice of allowing lead fuel assembly tests to
demonstrate performance of new materials in commercial reactors before
NRC approval, but also stated that long term performance testing of
materials was necessary, must take into account any differences at
individual utilities, and must consider future performance in dry cask
storage systems. Another individual commented that the petition should
be denied because the evaluations of cladding materials do not account
for the realities of plant operation under normal conditions and the
loss of coolant accident environment. This commenter stated that NRC
approval of materials whose properties fell ``within'' acceptance
criteria was unacceptable because an approval might be issued for a
material whose properties were ``right to the limit'' without an
adequate margin of safety. With respect to hydrogen generation, the
commenter opposed generic approvals of new materials because site-
specific material variations might yield unexpected results.
The nuclear reactor vendor supported adoption of the proposed rule
changes published in the Federal Register and agreed with the suggested
revision of Sec. 50.46(e) proposed by NEI in its comments on the
document. The vendor also recommended consideration of a direct final
rule process to implement the petition. The NEI provided revised
wording for proposed language in Sec. 50.46(e) and urged the NRC to
promulgate the revision as a direct final rule.
After evaluating the petition and public comments, the NRC has
determined that the petition should be denied in part. The final Sec.
50.44 rule has been written so that it does not refer to specific types
of zirconium cladding; instead, the rule applies to all boiling and
pressurized water reactors. When the NRC approves the use of boiling or
pressurized water reactor fuel with other types of cladding, no
exemptions from Sec. 50.44 will be needed. Thus, even though the final
rule does not contain the language specifically requested to be added
by the petitioner, the rule accomplishes the petitioner's intended
purpose with respect to Sec. 50.44. Also, the NRC did not utilize the
direct final rulemaking process because the other provisions being
amended in Sec. 50.44 were too complex to allow the promulgation of a
direct final rule. The NRC is making no decision at this time
[[Page 54138]]
on the part of the petition regarding the request to amend the
regulations in Sec. 50.44 to allow the use of other zirconium-based
alloys in addition to those specified in the current regulations. The
NRC will evaluate that portion of the NEI petition in a separate
action.
VII. Section-by-Section Analysis of Substantive Changes
Section 50.34--Contents of Applications; Technical Information
Paragraph (a)(4) on ECCS performance is revised to reference the
reactor coolant system high point venting requirements located in Sec.
50.46a. These requirements were relocated to Sec. 50.46a from Sec.
50.44.
Paragraph (g) is redesignated as paragraph (h) and a new paragraph
(g) is added, that requires applications for future reactors to include
the analyses and descriptions of the equipment and systems required by
Sec. 50.44.
Section 50.44--Combustible Gas Control in Containment
Paragraph (a), Definitions. Paragraph (a) adds definitions for two
previously undefined terms, ``mixed atmosphere,'' and ``inerted
atmosphere.''
Paragraph (b), Requirements for currently-licensed reactors. This
paragraph sets forth the requirements for control of combustible gas in
containment for currently-licensed reactors. All BWRs with Mark I and
II type containments are required to have an inerted containment
atmosphere, and all BWR Mark III type containments and PWRs with ice
condenser type containments are required to include a capability for
controlling combustible gas generated from a metal water reaction
involving 75 percent of the fuel cladding surrounding the active fuel
region (excluding the cladding surrounding the plenum volume) so that
there is no loss of containment integrity. Current requirements in
Sec. 50.44(c)(i), (iv), (v), and (vi) are incorporated in to the
amended regulation without substantial change. Previously reviewed and
installed combustible gas control mitigation features to meet the
existing regulations are considered to be sufficient to meet this
section. Because these requirements address beyond design-basis
combustible gas control, it is acceptable for structures, systems, and
components provided to meet these requirements to be non safety-related
and may be procured as commercial grade items.
Paragraph (b)(1), Mixed atmosphere. The requirement for capability
ensuring a mixed atmosphere in all containments is consistent with the
current requirement in Sec. 50.44(b)(2) and does not require further
analysis or modifications by current licensees. The intent of this
requirement is to maintain those plant design features (e.g.,
availability of active mixing systems or open compartments) that
promote atmospheric mixing. The requirement may be met with active or
passive systems. Active systems may include a fan, a fan cooler, or
containment spray. Passive capability may be demonstrated by evaluating
the containment for susceptibility to local hydrogen concentration.
These evaluations have been conducted for currently licensed reactors
as part of the IPE program.
Paragraph (b)(3) retains the existing requirements for BWR Mark III
and PWR ice condenser facilities that do not use inerting to establish
and maintain safe shutdown and containment structural integrity to use
structures, systems, and components capable of performing their
functions during and after exposure to hydrogen combustion.
Paragraph (b)(4)(i) codifies the existing regulatory practice of
monitoring oxygen in containments that use an inerted atmosphere for
combustible gas control. The rule does not require further analysis or
modifications by current licensees but certain design and qualification
criteria are relaxed. The rule requires that equipment for monitoring
oxygen be functional, reliable and capable of continuously measuring
the concentration of oxygen in the containment atmosphere following a
beyond design-basis accident. Equipment for monitoring oxygen must
perform in the environment anticipated in the severe accident
management guidance. The oxygen monitors are expected to be of high-
quality and may be procured as commercial grade items. Existing oxygen
monitoring commitments for currently licensed plants are sufficient to
meet this rule.
Paragraph (b)(4)(ii) retains the requirement in Sec. 50.44(b)(1)
for measuring the hydrogen concentration in the containment. The rule
does not require further analysis or modifications by current licensees
but certain design and qualification criteria are relaxed. The rule
requires that equipment for monitoring hydrogen be functional, reliable
and capable of continuously measuring the concentration of hydrogen in
the containment atmosphere following a significant beyond design-basis
accident of comparable severity to the accident at Three Mile Island.
Equipment for monitoring hydrogen must perform in the environment
anticipated in the severe accident management guidance. The hydrogen
monitors may be procured as commercial grade items. Existing hydrogen
monitoring commitments for currently licensed plants are sufficient to
meet this rule.
Paragraph (b)(5) retains the current analytical requirements in
Sec. 50.44(c)(3)(iv) that BWR Mark III and PWR ice condenser
containments be provided with a hydrogen control system justified by a
suitable program of experiment and analysis that can handle without
loss of containment integrity an amount of hydrogen equivalent to that
generated by a metal-water reaction involving 75 percent of the fuel
cladding surrounding the active fuel. Existing licensee hydrogen
control systems and analyses are expected to be sufficient to
demonstrate compliance with this requirement.
Paragraph (c), Requirements for future water-cooled reactor
applicants and licensees. Paragraph (c) promulgates requirements for
combustible gas control in containment for all future water-cooled
reactor construction permits or operating licenses under Part 50 and
for all water-cooled reactor design approvals, design certifications,
combined licenses, or manufacturing licenses under Part 52, whose
reactor designs have comparable potential for the production of
combustible gases as current light water reactor designs. The current
requirements in Sec. 50.34(f)(2)(ix) and (f)(3)(v) are retained
without material change, but have been consolidated and reworded to be
more concise. Paragraph (c)(1) requires a mixed containment atmosphere
during design-basis and significant beyond design-basis accidents. This
wording was chosen to specify a mixed atmosphere requirement during
important accident scenarios similar to the current requirements for
PWR and BWR containments. Paragraph (c)(2) requires all containments to
have an inerted atmosphere or limit hydrogen concentrations in
containment during and following an accident that releases an
equivalent amount of hydrogen as would be generated from a 100 percent
fuel-clad coolant reaction, uniformly distributed, to less than 10
percent and maintain containment structural integrity and appropriate
accident mitigating features. Structures, systems, and components
(SSCs) provided to meet this requirement must be designed to provide
reasonable assurance that they will operate in the severe accident
environment for which they are intended and over the time span for
which they are needed. Equipment survivability expectations under
severe accident conditions should consider the
[[Page 54139]]
circumstances of applicable initiating events (such as station blackout
\1\ or earthquakes) and the environment (including pressure,
temperature, and radiation) in which the equipment is relied upon to
function. The required system performance criteria will be based on the
results of design-specific reviews which include probabilistic risk-
assessment as required by Sec. 52.47(a)(1)(v). Because these
requirements address beyond design-basis combustible gas control, SSCs
provided to meet these requirements need not be subject to the
environmental qualification requirements of Sec. 50.49; quality
assurance requirements of 10 CFR Part 50, Appendix B; and redundancy/
diversity requirements of 10 CFR Part 50, Appendix A. Guidance such as
that found in Appendices A and B of RG 1.155, ``Station Blackout,'' is
appropriate for equipment used to mitigate the consequences of severe
accidents. Paragraph (c) also promulgates requirements for ensuring a
mixed atmosphere and monitoring oxygen and hydrogen in containment,
consistent with the requirements for current plants set forth in
paragraphs (b)(1), and (b)(4)(i) and (ii).
---------------------------------------------------------------------------
\1\ Section 50.44 does not require the deliberate ignition
systems used by BWRs with Mark III type containments and PWRs with
ice condenser type containments to be available during station
blackout events. The deliberate ignition systems should be available
upon the restoration of power. Additional guidance concerning the
availability of deliberate ignition systems during station blackout
sequences is being developed as part of the NRC review of Generic
Safety Issue 189: ``Susceptibility of Ice Condenser and Mark III
Containments to Early Failure from Hydrogen Combustion During a
Severe Accident.''
---------------------------------------------------------------------------
Paragraph (d), Requirements for future non water-cooled reactor
applicants and licensees and certain water-cooled reactor applicants
and licensees. A new paragraph (d) is added to specify information that
must be submitted by future reactor applicants to determine if
combustible gas generation is technically relevant to the proposed
design. If combustible gas generation is technically relevant, the
applicant must submit additional information to demonstrate that safety
impacts of combustible gases generated during design-basis and
significant beyond-design-basis accidents have been addressed in the
design of the facility to ensure adequate protection of public health
and safety and common defense and security. Paragraph (d) is applicable
to non water-cooled reactors and water-cooled reactors that have
different characteristics regarding the production of combustible gases
from current light water reactors. The information must address the
potential for producing combustible gases during design basis accidents
and significant beyond design-basis accidents comparable to accident
scenarios that were evaluated for combustible gas generation at current
light water reactors.
Section 50.46a--Acceptance Criteria for Reactor Coolant System Venting
Systems
Section 50.46a is a new section that contains the relocated
requirements for high point vents currently contained in Sec. 50.44.
The amendment includes a change that eliminates a requirement
prohibiting venting the reactor coolant system if it could
``aggravate'' the challenge to containment. Any venting is highly
unlikely to affect containment integrity; however, such venting will
reduce the likelihood of further core damage. The NRC continues to view
use of the high point vents as an important strategy that should be
considered in a plant's severe accident management guidelines.
Section 52.47--Contents of Applications
Section 52.47 is amended to eliminate the reference to paragraphs
within Sec. 50.34(f) for technically relevant requirements for
combustible gas control in containment for future design
certifications. Under the final rule, the technical requirements for
combustible gas control will be set forth in Sec. 50.44, rather than
in Sec. 50.34(f).
VIII. Availability of Documents
The NRC is making the documents identified below available to
interested persons through one or more of the following methods as
indicated.
Public Document Room (PDR). The NRC Public Document Room is located
at One White Flint North, Public File Area O 1F21, 11555 Rockville
Pike, Rockville, Maryland.
Rulemaking Web site (Web). The NRC's interactive rulemaking Web
site is located at http://ruleforum.llnl.gov. These documents may be
viewed and downloaded electronically via this Web site.
NRC's Electronic Reading Room (ERR). The NRC's public electronic
reading room is located at http://www.nrc.gov/NRC/ADAMS/index.html.
(Provide accession number for each document.)
The NRC staff contact (NRC Staff). Richard Dudley, Office of
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone (301) 415-1116; e-mail
[email protected].
----------------------------------------------------------------------------------------------------------------
Document PDR Web ERR NRC staff
----------------------------------------------------------------------------------------------------------------
Comments received........................ X X X............................. ...........
Regulatory Analysis...................... X X ML031640482................... ...........
RG 1.7, Rev. 3........................... X X ML031640498................... X
Rev. SRP, Section 6.2.5.................. X X ML031640518................... X
----------------------------------------------------------------------------------------------------------------
A free single copy of Regulatory Guide 1.7 may be obtained by
writing to the Office of the Chief Information Officer, Reproduction
and Distribution Services Section, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or E-mail: [email protected] or
Facsimile: (301) 415-2289.
Copies of NUREGS may be purchased from The Superintendent of
Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington,
DC 20402-0001; Internet: [email protected]; (202) 512-1800. Copies are
also available from the National Technical Information Service,
Springfield, VA 22161-0002; http://www.ntis.gov; 1-800-533-6847 or,
locally, (703) 605-6000. Some publications in the NUREG series are
posted at NRC's technical document Web site http://www.nrc.gov/NRC/NUREGS/indexnum.html.
IX. 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
using such a standard is inconsistent with applicable law or is
otherwise impractical. In this final rule, the NRC is using the
following Government-unique standard: 10 CFR 50.44, U.S.
[[Page 54140]]
Nuclear Regulatory Commission, October 27, 1978 (43 FR 50163), as
amended. No voluntary consensus standard has been identified that could
be used instead of the Government-unique standard.
X. Finding of No Significant Environmental Impact: Environmental
Assessment
The NRC has determined under the National Environmental Policy Act
of 1969, as amended, and the Commission'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 reads as follows:
This action endorses existing requirements and establishes
regulations that reduce regulatory burdens for current and future
licensees and consolidates combustible gas control regulations for
future reactor applicants and licensees. This action stems from the
NRC's ongoing effort to risk-inform its regulations. The final rule
reduces the regulatory burdens on present and future power reactor
licensees by eliminating the LOCA design-basis accident as a
combustible gas control concern. This change eliminates the
requirements for hydrogen recombiners and hydrogen purge systems and
relaxes the requirements for hydrogen and oxygen monitoring equipment
to make them commensurate with their safety and risk significance.
This action does not significantly increase the probability or
consequences of an accident. No changes are being made in the types or
quantities of radiological effluents that may be released off site, and
there is no significant increase in public radiation exposure because
there is no change to facility operations that could create a new or
affect a previously analyzed accident or release path. There may be a
reduction of occupational radiation exposure since personnel will no
longer be required to maintain or operate, if necessary, the hydrogen
recombiner systems which are located in or near radiologically
controlled areas.
With regard to non-radiological impacts, no changes are being made
to non-radiological plant effluents and there are no changes in
activities that would adversely affect the environment. Therefore,
there are no significant non-radiological impacts associated with the
proposed action.
The primary alternative to this action would be the no action
alternative. The no action alternative would continue to impose
unwarranted regulatory burdens for which there would be little or no
safety, risk, or environmental benefit.
The determination of this environmental assessment is that there is
no significant offsite impact to the public from this action.
The NRC requested the views of the States on the environmental
assessment for this rule. No comments were received.
XI. Paperwork Reduction Act Statement
This final rule decreases the burden on new applicants to complete
the hydrogen control analysis required to be submitted in a license
application, as required by sections 50.34 or 52.47. The public burden
reduction for this information collection is estimated to average 720
hours per request. Because the burden for this information collection
is insignificant, Office of Management and Budget (OMB) clearance is
not required. Existing requirements were approved by the Office of
Management and Budget, approval numbers 3150-0011 and 3150-0151.
XII. 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.
XIII. Regulatory Analysis
The NRC has prepared a regulatory analysis on this regulation. The
analysis examines the costs and benefits of the alternatives considered
by the NRC. The regulatory analysis is available as indicated under the
Availability of Documents heading of the Supplementary Information
section.
XIV. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act (5 U.S.C.
605(b)), the Commission certifies that this rule does not have a
significant economic impact on a substantial number of small entities.
This 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 size standards established by the NRC
(10 CFR 2.810).
XV. Backfit Analysis
The NRC has determined that the backfit rule does not apply to this
final rule; and therefore, a backfit analysis is not required for this
final rule because these amendments do not impose more stringent safety
requirements on 10 CFR Part 50 licensees. For current licensees, the
amendments either maintain without substantive change existing
requirements or provide voluntary relaxations to current regulatory
requirements. Voluntary relaxations (i.e., relaxations that are not
mandatory) are not considered backfitting as defined in 10 CFR
50.109(a)(1). For future applicants and future licensees, the
amendments also do not involve backfitting as defined in 10 CFR
50.109(a)(1) because the changes have only a prospective effect on
future design approval and design certification applicants and future
applicants for licensees under 10 CFR Part 50 and 52. As the Commission
has indicated in other rulemakings, sec., e.g., 54 FR 15372, April 18,
1989 (Final Part 52 Rule), the expectations of future applicants are
not protected by the Backfit Rule. Therefore, the NRC has not prepared
a backfit analysis for this final rule.
XVI. Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness 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 of OMB.
List of Subjects
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
record keeping requirements.
10 CFR Part 52
Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Inspection, Limited work authorization, Nuclear power plants and
reactors, Probabilistic risk assessment, Prototype, Reactor siting
criteria, Redress of site, Reporting and record keeping requirements,
Standard design, Standard design certification.
0
For the reasons set out 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
[[Page 54141]]
NRC is adopting the following amendments to 10 CFR Parts 50 and 52.
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, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 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, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242,
as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951, as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42
U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68
Stat. 936, 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 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. In Sec. 50.34, paragraph (a)(4) is revised, paragraph (g) is
redesignated as paragraph (h), and a new paragraph (g) is added to read
as follows:
Sec. 50.34 Contents of applications; technical information.
(a) * * *
(4) A preliminary analysis and evaluation of the design and
performance of structures, systems, and components of the facility with
the objective of assessing the risk to public health and safety
resulting from operation of the facility and including determination of
the margins of safety during normal operations and transient conditions
anticipated during the life of the facility, and the adequacy of
structures, systems, and components provided for the prevention of
accidents and the mitigation of the consequences of accidents. Analysis
and evaluation of ECCS cooling performance and the need for high point
vents following postulated loss-of-coolant accidents must be performed
in accordance with the requirements of Sec. 50.46 and Sec. 50.46a of
this part for facilities for which construction permits may be issued
after December 28, 1974.
* * * * *
(g) Combustible gas control. All applicants for a reactor
construction permit or operating license under this part, and all
applicants for a reactor design approval, design certification, or
license under part 52 of this chapter, whose application was submitted
after October 16, 2003, shall include the analyses, and the
descriptions of the equipment and systems required by Sec. 50.44 as a
part of their application.
* * * * *
0
3. Section 50.44 is revised to read as follows:
Sec. 50.44 Combustible gas control for nuclear power reactors.
(a) Definitions.
(1) Inerted atmosphere means a containment atmosphere with less
than 4 percent oxygen by volume.
(2) Mixed atmosphere means that the concentration of combustible
gases in any part of the containment is below a level that supports
combustion or detonation that could cause loss of containment
integrity.
(b) Requirements for currently-licensed reactors. Each boiling or
pressurized water nuclear power reactor with an operating license on
October 16, 2003, except for those facilities for which the
certifications required under Sec. 50.82(a)(1) have been submitted,
must comply with the following requirements, as applicable:
(1) Mixed atmosphere. All containments must have a capability for
ensuring a mixed atmosphere.
(2) Combustible gas control. (i) All boiling water reactors with
Mark I or Mark II type containments must have an inerted atmosphere.
(ii) All boiling water reactors with Mark III type containments and
all pressurized water reactors with ice condenser containments must
have the capability for controlling combustible gas generated from a
metal-water reaction involving 75 percent of the fuel cladding
surrounding the active fuel region (excluding the cladding surrounding
the plenum volume) so that there is no loss of containment structural
integrity.
(3) Equipment Survivability. All boiling water reactors with Mark
III containments and all pressurized water reactors with ice condenser
containments that do not rely upon an inerted atmosphere inside
containment to control combustible gases must be able to establish and
maintain safe shutdown and containment structural integrity with
systems and components capable of performing their functions during and
after exposure to the environmental conditions created by the burning
of hydrogen. Environmental conditions caused by local detonations of
hydrogen must also be included, unless such detonations can be shown
unlikely to occur. The amount of hydrogen to be considered must be
equivalent to that generated from a metal-water reaction involving 75
percent of the fuel cladding surrounding the active fuel region
(excluding the cladding surrounding the plenum volume).
(4) Monitoring. (i) Equipment must be provided for monitoring
oxygen in containments that use an inerted atmosphere for combustible
gas control. Equipment for monitoring oxygen must be functional,
reliable, and capable of continuously measuring the concentration of
oxygen in the containment atmosphere following a significant beyond
design-basis accident for combustible gas control and accident
management, including emergency planning.
(ii) Equipment must be provided for monitoring hydrogen in the
containment. Equipment for monitoring hydrogen must be functional,
reliable, and capable of continuously measuring the concentration of
hydrogen in the containment atmosphere following a significant beyond
design-basis accident for accident management, including emergency
planning.
(5) Analyses. Each holder of an operating license for a boiling
water reactor with a Mark III type of containment or for a pressurized
water reactor with an ice condenser type of containment, shall perform
an analysis that:
(i) Provides an evaluation of the consequences of large amounts of
hydrogen generated after the start of an accident (hydrogen resulting
from the metal-water reaction of up to and including 75 percent of the
fuel cladding surrounding the active fuel region, excluding the
cladding surrounding the plenum volume) and include consideration of
hydrogen control measures as appropriate;
(ii) Includes the period of recovery from the degraded condition;
(iii) Uses accident scenarios that are accepted by the NRC staff.
These scenarios must be accompanied by sufficient supporting
justification to show that they describe the behavior of the reactor
system during and following an accident resulting in a degraded core.
(iv) Supports the design of the hydrogen control system selected to
meet the requirements of this section; and,
(v) Demonstrates, for those reactors that do not rely upon an
inerted atmosphere to comply with paragraph (b)(2)(ii) of this section,
that:
[[Page 54142]]
(A) Containment structural integrity is maintained. Containment
structural integrity must be demonstrated by use of an analytical
technique that is accepted by the NRC staff in accordance with Sec.
50.90. This demonstration must include sufficient supporting
justification to show that the technique describes the containment
response to the structural loads involved. This method could include
the use of actual material properties with suitable margins to account
for uncertainties in modeling, in material properties, in construction
tolerances, and so on; and
(B) Systems and components necessary to establish and maintain safe
shutdown and to maintain containment integrity will be capable of
performing their functions during and after exposure to the
environmental conditions created by the burning of hydrogen, including
local detonations, unless such detonations can be shown unlikely to
occur.
(c) Requirements for future water-cooled reactor applicants and
licensees.\2\ The requirements in this paragraph apply to all water-
cooled reactor construction permits or operating licenses under this
part, and to all water-cooled reactor design approvals, design
certifications, combined licenses or manufacturing licenses under part
52 of this chapter, any of which are issued after October 16, 2003.
---------------------------------------------------------------------------
\2\ The requirements of this paragraph apply only to water-
cooled reactor designs with characteristics (e.g., type and quantity
of cladding materials) such that the potential for production of
combustible gases is comparable to light water reactor designs
licensed as of October 16, 2003.
---------------------------------------------------------------------------
(1) Mixed atmosphere. All containments must have a capability for
ensuring a mixed atmosphere during design-basis and significant beyond
design-basis accidents.
(2) Combustible gas control. All containments must have an inerted
atmosphere, or must limit hydrogen concentrations in containment during
and following an accident that releases an equivalent amount of
hydrogen as would be generated from a 100 percent fuel clad-coolant
reaction, uniformly distributed, to less than 10 percent (by volume)
and maintain containment structural integrity and appropriate accident
mitigating features.
(3) Equipment Survivability. Containments that do not rely upon an
inerted atmosphere to control combustible gases must be able to
establish and maintain safe shutdown and containment structural
integrity with systems and components capable of performing their
functions during and after exposure to the environmental conditions
created by the burning of hydrogen. Environmental conditions caused by
local detonations of hydrogen must also be included, unless such
detonations can be shown unlikely to occur. The amount of hydrogen to
be considered must be equivalent to that generated from a fuel clad-
coolant reaction involving 100 percent of the fuel cladding surrounding
the active fuel region.
(4) Monitoring. (i) Equipment must be provided for monitoring
oxygen in containments that use an inerted atmosphere for combustible
gas control. Equipment for monitoring oxygen must be functional,
reliable, and capable of continuously measuring the concentration of
oxygen in the containment atmosphere following a significant beyond
design-basis accident for combustible gas control and accident
management, including emergency planning.
(ii) Equipment must be provided for monitoring hydrogen in the
containment. Equipment for monitoring hydrogen must be functional,
reliable, and capable of continuously measuring the concentration of
hydrogen in the containment atmosphere following a significant beyond
design-basis accident for accident management, including emergency
planning.
(5) Structural analysis. An applicant must perform an analysis that
demonstrates containment structural integrity. This demonstration must
use an analytical technique that is accepted by the NRC and include
sufficient supporting justification to show that the technique
describes the containment response to the structural loads involved.
The analysis must address an accident that releases hydrogen generated
from 100 percent fuel clad-coolant reaction accompanied by hydrogen
burning. Systems necessary to ensure containment integrity must also be
demonstrated to perform their function under these conditions.
(d) Requirements for future non water-cooled reactor applicants and
licensees and certain water-cooled reactor applicants and licensees.
The requirements in this paragraph apply to all construction permits
and operating licenses under this part, and to all design approvals,
design certifications, combined licenses, or manufacturing licenses
under part 52 of this chapter, for non water-cooled reactors and water-
cooled reactors that do not fall within the description in paragraph
(c), footnote 1 of this section, any of which are issued after October
16, 2003. Applications subject to this paragraph must include:
(1) Information addressing whether accidents involving combustible
gases are technically relevant for their design, and
(2) If accidents involving combustible gases are found to be
technically relevant, information (including a design-specific
probabilistic risk assessment) demonstrating that the safety impacts of
combustible gases during design-basis and significant beyond design-
basis accidents have been addressed to ensure adequate protection of
public health and safety and common defense and security.
0
4. Section 50.46a is added to read as follows:
Sec. 50.46a Acceptance criteria for reactor coolant system venting
systems.
Each nuclear power reactor must be provided with high point vents
for the reactor coolant system, for the reactor vessel head, and for
other systems required to maintain adequate core cooling if the
accumulation of noncondensible gases would cause the loss of function
of these systems. High point vents are not required for the tubes in U-
tube steam generators. Acceptable venting systems must meet the
following criteria:
(a) The high point vents must be remotely operated from the control
room.
(b) The design of the vents and associated controls, instruments
and power sources must conform to appendix A and appendix B of this
part.
(c) The vent system must be designed to ensure that:
(1) The vents will perform their safety functions; and
(2) There would not be inadvertent or irreversible actuation of a
vent.
PART 52--EARLY SITE PERMITS; STANDARD DESIGN CERTIFICATIONS; AND
COMBINED LICENSES FOR NUCLEAR POWER PLANTS
0
5. The authority citation for Part 52 continues to read as follows:
Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat.
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as
amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282); secs.
201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C.
5841, 5842, 5846).
0
6. In Sec. 52.47, paragraph (a)(1)(ii) is revised to read as follows:
Sec. 52.47 Contents of applications.
(a) * * *
(1) * * *
(ii) Demonstration of compliance with any technically relevant
portions of the
[[Page 54143]]
Three Mile Island requirements set forth in 10 CFR 50.34(f) except
paragraphs (f)(1)(xii), (f)(2)(ix) and (f)(3)(v);
* * * * *
Dated at Rockville, Maryland, this 10th day of September 2003.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 03-23554 Filed 9-15-03; 8:45 am]
BILLING CODE 7590-01-P