[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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  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.

-----------------------------------------------------------------------

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