[Federal Register Volume 65, Number 207 (Wednesday, October 25, 2000)]
[Rules and Regulations]
[Pages 63769-63789]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-27283]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 72

RIN 3150-AF98


Reporting Requirements for Nuclear Power Reactors and Independent 
Spent Fuel Storage Installations at Power Reactor Sites

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending the event 
reporting requirements for nuclear power reactors to reduce or 
eliminate the unnecessary reporting burden associated with events of 
little or no safety significance. This final rule continues to provide 
the Commission with reporting of significant events where Commission 
action may be needed to maintain or improve reactor safety or to 
respond to heightened public concern. This final rule also better 
aligns event reporting requirements with the type of information NRC 
needs to carry out its safety mission, including revising reporting 
requirements based on importance to risk and extending the required 
reporting times consistent with the time that information is needed for 
prompt NRC action. Also, NUREG-1022, Revision 2, ``Event Reporting 
Guidelines, 10 CFR 50.72 and 50.73,'' is being made available 
concurrently with the amendments.

DATES: The final rule is effective January 23, 2001.

ADDRESSES: Documents related to this action may be examined, and/or 
copied for a fee, at the NRC's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Documents created or received at the NRC after November 1, 
1999 are also available electronically at the NRC's Public Electronic 
Reading Room on the Internet at http://www.nrc.gov/NRC/ADAMS/index.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. For further 
information contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737 or by email to <a href="/cdn-cgi/l/email-protection#69190d1b29071b0a470e061f"><span class="__cf_email__" data-cfemail="0a7a6e784a647869246d657c">[email protected]</span></a>.

FOR FURTHER INFORMATION CONTACT: Dennis P. Allison, Office of Nuclear 
Reactor Regulation, Washington, DC 20555-0001, telephone (301) 415-
1178, e-mail <a href="/cdn-cgi/l/email-protection#32564253725c40511c555d44"><span class="__cf_email__" data-cfemail="e1859180a18f9382cf868e97">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

Contents

I. Background
II. Analysis of Comments
III. Discussion
    1. Objectives
    2. Section by Section Discussion of Final Amendments
    3. Revisions to Event Reporting Guidelines in NUREG-1022
    4. Reactor Oversight
    5. Enforcement
    6. Electronic Reporting
    7. State Input
    8. Plain Language
IV. Environmental Impact: Categorical Exclusion
V. Backfit Analysis
VI. Regulatory Analysis
VII. Paperwork Reduction Act Statement
VIII. Regulatory Flexibility Act Certification
IX. Small Business Regulatory Enforcement Fairness Act
X. National Technology Transfer and Advancement Act
XI. Final Amendments

I. Background

    The reporting requirements in Sections 50.72 and 50.73 have been in 
effect, with minor modifications, since 1983. Experience has shown a 
need for change in several areas. On July 23, 1998 (63 FR 39522), the 
NRC published in the Federal Register an advance notice of proposed 
rulemaking (ANPR) to announce a contemplated rulemaking that would 
modify reporting requirements for nuclear power reactors. Among other 
things, the ANPR requested public comments on several concrete 
proposals for modification of the event reporting rules. Public 
meetings were held to discuss the ANPR at NRC Headquarters on August 
21, 1998, in Rosemont, Illinois on September 1, 1998, and at NRC 
Headquarters on November 13, 1998.
    A proposed rule was published in the Federal Register on July 6, 
1999 (64 FR 36291), including a conforming change to Section 72.216. 
Concurrently, a draft revision to the associated event reporting 
guidelines was made available for public comment (NUREG-1022, Draft 
Revision 2). A public meeting was held at NRC Headquarters on August 3, 
1999, to discuss the proposed rule and draft guidelines. Public 
comments were due on September 20, 1999. Additional public meetings 
were held on February 25, and March 22, 2000, to discuss public 
comments.

II. Analysis of Comments

    The comment period for the proposed rule expired September 20, 
1999. Twenty-seven comment letters were received, representing comments 
from 24 nuclear power plant licensees (utilities), two organizations of 
utilities, and one State agency.
    In addition to the written comments received, the proposed rule was 
the subject of a public meeting on August 3, 1999, as discussed above 
under the heading ``Background,'' and comments made at that meeting 
have also been considered.
    Most commenters expressed support for amending the rules in 
accordance with the objectives discussed in the proposed rule. However, 
they objected to some of the specific provisions. Many comments also 
provided specific recommendations for changes to the proposed rules. 
The resolution of comments is summarized below. This summary addresses 
the principal comments (i.e., comments other than those that are: minor 
or editorial in nature; supportive of the approach described in the 
proposed rules; or applicable to another area or activity outside the 
scope of sections 50.72 and 50.73).
    Comment A (Do not require reporting of degraded components): The 
proposed rule included a new component

[[Page 63770]]

reporting criterion. It would have required reporting ``Any event or 
condition that resulted in a component being in a degraded or non-
conforming condition such that the ability of the component to perform 
its specified safety function is significantly degraded and the 
condition could reasonably be expected to affect other similar 
components in the plant.'' The term ``significantly degraded'' was 
defined by providing several examples of reportable and non-reportable 
events. The stated purpose was to ensure that design basis or other 
discrepancies would continue to be reported if the capability to 
perform a specified safety function is significantly degraded and the 
condition has generic implications.
    Most commenters strongly objected to the proposed component 
reporting criterion. Among other things, they indicated:
    (1) The proposed component reporting criterion is not needed 
because, after deleting the requirement to report a condition that is 
outside the design basis of the plant, any significant events would 
still be captured by the other existing criteria.
    (2) The proposed component reporting criterion would be unclear and 
subject to widely varying interpretation with regard to the meaning of 
the term ``significantly degraded'' and the term ``could reasonably be 
expected to apply to other similar components.''
    (3) The proposed component reporting criterion would be overly 
burdensome. For example, it would become necessary to screen all single 
component failures for reportability.
    (4) The proposed component reporting criterion would be contrary to 
the stated objectives of the rulemaking. For example, it would result 
in many additional reports for events with little or no safety- or 
risk-significance.
    Response: In the final rule, the proposed reporting criterion has 
been retained, but modified to address the concerns about unnecessary 
burden and clarity expressed in the comments. It requires reporting any 
event or condition that as a result of a single cause could have 
prevented the fulfillment of a safety function for two or more trains 
or channels in different systems that are needed to:
    (1) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (2) Remove residual heat;
    (3) Control the release of radioactive material; or (4) Mitigate 
the consequences of an accident.
    Events covered by this criterion may include cases of procedural 
error, equipment failure, and/or discovery of a design, analysis, 
fabrication, construction, and/or procedural inadequacy. However, 
licensees are not required to report an event pursuant to this 
criterion if the event results from:
    (1) A shared dependency among trains or channels that is a natural 
or expected consequence of the approved plant design; or
    (2) Normal and expected wear or degradation.
    Subject to the two exclusions stated above, this criterion, as 
modified, is needed to capture those events with enough generic 
significance that a single cause could have prevented the fulfillment 
of the safety function of multiple trains or channels, but the event:
    (1) Would not be captured by Secs. 50.73(a)(2)(v) and 
50.72(b)(3)(v) [event or condition that could have prevented 
fulfillment of the safety function of structures and systems needed to 
* * *] because the affected trains or channels are in different 
systems; and
    (2) Would not be captured by Sec. 50.73(a)(2)(vii) [common cause 
inoperability of independent trains or channels] because the affected 
trains or channels are either:
    (i) Not assumed to be independent in the plant's safety analysis; 
or
    (ii) Not both considered to be inoperable.
    The criterion, as modified, would not be unclear because it uses 
the term ``could have prevented fulfillment of the safety function,'' 
which is already used in a previously existing criterion.
    The criterion, as modified, is not considered overly burdensome 
because it is estimated to result in fewer reports than the previous 
requirement to report a condition outside the design basis of the 
plant. It is not necessary to screen all single component failures for 
reportability.
    The criterion, as modified, is considered consistent with the 
objectives of the rulemaking for the same reasons.
    Comment B (Do not change the term ``any engineered safety feature 
[ESF] * * *''): In the proposed rule, the term ``any engineered safety 
feature (ESF), including the reactor protection system (RPS),'' which 
defines the systems for which actuation must be reported, would be 
replaced by a specific list of systems. It was recognized that this 
proposal to list the systems in the rule was controversial and public 
comment was specifically invited in this area. In particular, three 
principal alternatives to the proposed rule were identified for 
comment. They are:
    Alternative 1, status quo. The rule would continue to require 
reporting for actuation of ``any ESF.'' The guidance would continue to 
infer that reporting should include the systems on a list which is 
similar to the list in the proposed rule.
    Alternative 2, plant-specific list. The rule would require that 
licensees develop a plant-specific, risk-informed list.
    Alternative 3, pre-1998 practice. The rule would continue to 
require reporting actuation of ``any ESF.'' The guidance would indicate 
that this includes those systems identified as ESF's in each plant's 
final safety analysis report (FSAR).
    The comments may be summarized as follows:
    (1) Most commenters objected to the proposed rule, which would 
replace the term ``any engineered safety feature (ESF), including the 
reactor protection system (RPS)'' with a list of specific systems. The 
reasons cited by the commenters include the following:
    (a) Providing an all-inclusive list of systems in the rules is 
inappropriate.
    (b) Each facility's FSAR specifies equipment that is designated as 
ESF equipment.
    (c) Plant-specific differences exist in the safety-related status 
of their systems.
    (d) The risk-significance of a particular system can vary greatly 
between plants, due to a wide variety of design differences. An all-
inclusive list would increase the burden for some plants whose 
equipment on the list was not ESF equipment or equipment with a 
suitably high risk-significance.
    (e) There are a number of specific problems with the proposed list. 
Specific examples were provided.
    (2) Most commenters recommended in favor of Alternative 3, 
returning to the pre-1998 practice of reporting actuation for only 
those systems that are designated as ESFs in each facility's FSAR. They 
stated that this option best meets the goal of clarity and simplicity.
    (3) One commenter recommended in favor of Alternative 1 (status 
quo), where the reporting guidelines contain a list of systems similar 
to the list proposed for the rule. It stated that the facility's 
internal reporting procedures already reflect the current practice. Any 
benefit that might be obtained by returning to the pre-1998 practice 
would be so slight that it would not justify the cost of changing the 
procedures.
    (4) Some commenters indicated that there are problems with the 
status quo that need to be solved. For example, the reporting 
guidelines should exclude

[[Page 63771]]

reporting of reactor water cleanup system (RWCU) isolations that 
routinely occur during system restoration following maintenance 
outages, due to rapid pressurization following valve opening.
    (5) Most commenters objected to Alternative 2 (developing a plant-
specific, risk-informed list of systems). They stated that this would 
require a significant expenditure of resources and it is unclear as to 
whether or how it would meet the NRC's needs better than Alternative 3 
(returning to pre-1998 practice). They also noted that there is a 
separate initiative to ``risk-inform'' 10 CFR Part 50. This may result 
in development of plant-specific lists of systems based on risk 
significance. However, the commenters do not believe the necessary 
criteria have been adequately established to make that shift as part of 
this rulemaking to modify 10 CFR 50.72 and 50.73. They recommended that 
later, as part of the rule change to ``risk-inform'' Part 50, the NRC 
should evaluate whether or not it is appropriate to ``risk-inform'' ESF 
systems subject to the event reporting requirements of 10 CFR 50.72 and 
50.73.
    Response: (1) The NRC believes providing a list of systems is the 
best approach because it will obtain consistent reporting of events 
that result in actuation of highly risk-significant systems. Consistent 
reporting for such events is needed to support estimating equipment 
reliability parameters and is important to several aspects of the NRC's 
general move towards more risk-informed regulation.
    Commenters stated that the risk-significance of the systems varies 
depending on plant design. As discussed below under the headings 
``(e)(i)'' through ``(e)(vii),'' a number of items have been removed 
from the list based on specific comments. The NRC believes that these 
systems remaining on the list are of sufficient risk significance to 
warrant reporting of a system actuation. The principal reason for 
reporting an actuation of one of these systems is that it is indicative 
of an unplanned plant transient that the NRC needs to evaluate to 
determine if action is necessary to address a safety problem. In this 
context, the NRC's need to evaluate the event is independent of 
classification of the system. For example, a valid actuation of the 
auxiliary feedwater (AFW) system at a pressurized water reactor (PWR) 
means there was a transient that involved an abnormal plant parameter, 
such as low steam generator level, which initiated the actuation. This 
is the reason the NRC needs to evaluate the event, and it is 
independent of how the AFW system happens to be classified at the 
particular plant.
    The classification of systems in the FSARs has evolved over the 
years. For example, in earlier PWR designs the auxiliary feedwater 
system was not considered to be an ESF, and this is reflected in early 
FSARs. Later, although the system's function and importance did not 
change, it came to be considered an ESF, and this is reflected in later 
FSARs. Since the function and importance is the same regardless of 
classification, it does not make sense to exclude reporting for 
actuation of the auxiliary feedwater system based on its classification 
in the FSAR.
    Furthermore, this approach is estimated to result in a net 
reduction in the number of events reported under this criterion. Some 
licensees will make additional reports involving highly risk-
significant systems. However, these additional reports will be 
outweighed by the elimination of reports involving systems with lesser 
risk-significance.
    (a) Commenters indicated that providing an all-inclusive list of 
systems in the rules is inappropriate. However, the NRC does not 
believe the list is all inclusive. It contains only systems that are 
highly risk-significant and omits systems of lesser risk-significance, 
even if the systems of lesser risk-significance are designated as ESFs. 
The NRC also believes the list is appropriate because it provides 
consistent reporting of events that result in actuation of these highly 
risk-significant systems and, at the same time, a net reduction in 
reporting burden.
    (b) Commenters stated that each facility's FSAR specifies equipment 
that is designated as ESF equipment. However, the NRC believes that 
those lists are not consistent or risk-informed. For example, at 
several plants, emergency diesel generators (EDGs), which are highly 
risk-significant, are not identified as ESFs. At several pressurized 
water reactors (PWRs), the AFW system which is highly risk-significant, 
is not identified as an ESF. At most boiling water reactors (BWRs), the 
reactor core isolation cooling (RCIC) system, which is highly risk-
significant, is not identified as an ESF. On the other hand, most 
plants identify systems with lesser risk-significance, such as fuel 
building ventilation and filtration systems, as ESFs.
    (c) Commenters stated that plant-specific differences exist in the 
safety-related status of systems. However, the NRC does not believe 
that this fact bears directly on the question of which system 
actuations should be reported. There is no need to report the actuation 
of all safety related systems, and there is no reason to exclude 
reporting for the actuation of a non-safety-related system if it is 
highly risk-significant simply on the basis that it has not been 
classified by the licensee as an ESF.
    (d) Commenters stated that the risk-significance of a particular 
system can vary greatly among plants. They further stated that an all-
inclusive list would therefore increase the burden for some plants 
whose equipment on the list was not ESF equipment or equipment with a 
suitably high risk-significance. The NRC agrees with the general 
statement that the risk-significance of a particular system can vary 
greatly among plants. However, the systems on the list are virtually 
always of high risk-significance. While it is true that, as a result of 
the list, some licensees will make additional reports, any additional 
reports will involve systems that are highly risk-significant. Also, 
these additional reports will be outweighed by the elimination of 
reports involving systems with lesser risk-significance. Thus, the net 
effect is a reduction in reporting.
    (e) Commenters provided several specific examples of items they 
considered to be problems with the list. These examples are:
    (i) In the proposed rule, the feedwater coolant injection (FWCI) 
system was characterized as an example of an emergency core cooling 
system (ECCS). Commenters stated that FWCI systems are not considered 
to be ECCS. The NRC believes that clarification is warranted. In the 
final rule, FWCI is not characterized as an ECCS. However, it is 
included as a separate item in the list.
    (ii) The proposed rule would have required reporting actuations of 
the RCIC system. Commenters stated that RCIC is included in the 
Improved Standard Technical Specifications (ISTS) because it meets 
criterion 4 of 10 CFR 50.36, based on its contribution to the reduction 
of overall plant risk. They further stated that RCIC is not credited in 
the plant's safety analysis. The NRC believes that RCIC is highly risk-
significant and, therefore, it remains on the list in the final rule.
    (iii) Commenters stated that non-reportable exceptions should be 
allowed for systems that are considered to be ESFs, yet have lower 
levels of risk significance (control room ventilation systems, reactor 
building ventilation systems, fuel building ventilation systems, 
auxiliary building ventilation systems, RWCU isolations during 
restoration from maintenance, etc.). The NRC agrees. The final rule 
eliminates unnecessary reporting for systems that are considered to be 
ESFs, yet have lower levels of risk significance. It also

[[Page 63772]]

eliminates reporting for RWCU isolations during restoration from 
maintenance because they are routine and are of low risk and safety 
significance.
    (iv) Commenters stated that the list inappropriately includes 
``associated support systems'' for BWR Division 3 EDGs. The NRC agrees. 
In the final rule the term ``associated support systems'' has been 
eliminated for BWR Division 3 EDGs, and other EDGs as well.
    (v) Commenters stated that the list inappropriately includes 
station blackout diesel generators (and black start gas turbines that 
serve a similar purpose) that are not safety related. The NRC agrees. 
The final rule does not require reporting for station blackout diesel 
generators (and black start gas turbines that serve a similar purpose).
    (vi) Commenters stated that although the term ``anticipated 
transient without scram (ATWS) mitigating systems'' is clear to those 
licensees that have dedicated systems (i.e. AMSAC), a great deal of 
confusion exists for those that have no dedicated system. Due to the 
lack of clarity, it could be interpreted that any system that might be 
used during an ATWS would fall into this category (i.e. feedwater 
systems, borating systems, control rods, etc.). Extensive clarification 
would be needed to eliminate this ambiguity. The NRC agrees that 
clarification is warranted. In the final rule this item has been 
eliminated. Reporting is not needed for actuations for a system such as 
AMSAC. The reports needed for other systems are captured by other items 
on the list.
    (vii) Commenters stated that it is unclear as to whether the 
service water entry applies only to emergency service water systems 
(i.e., those that don't operate unless there is an accident) or also to 
the standby service water systems that only run to remove heat from the 
residual heat removal (RHR) heat exchangers. The NRC agrees. In the 
final rule this item has been clarified to indicate that reporting is 
required for emergency service water (ESW) systems that do not normally 
run and that serve as ultimate heat sinks. In addition, this item has 
been deleted from the list of systems for which telephone notification 
is required under section 50.72 because an ESW actuation by itself does 
not indicate the type of transient that the NRC needs to evaluate. 
However, ESW system actuations are reportable only under section 50.73 
because the information is needed to support the NRC staff's equipment 
reliability estimates.
    (2) As stated by commenters, Alternative 3 would provide clarity 
and simplicity. However, the NRC believes that adoption of the list of 
systems in the final rule also provides clarity and simplicity.
    (3) Although one commenter recommended in favor of Alternative 1, 
the NRC believes that this alternative would invite variable 
interpretation. The event reporting guidelines would contain a list of 
systems, whereas the rule would require reporting the actuation of 
``any ESF.''
    (4) Some commenters stated that, under previous requirements it was 
necessary to report reactor water cleanup system isolations that 
routinely occur during system restoration following maintenance 
outages, due to rapid pressurization following valve opening. The list 
of systems eliminates these unneeded reports because it limits the 
reporting of containment isolation signals to those that affect 
multiple systems or multiple main steam isolation valves (MSIVs).
    (5) As indicated in the comments, with respect to Alternative (2), 
the project to ``risk-inform'' 10 CFR Part 50 may, in the future, lead 
to development of plant-specific lists of systems based on importance 
to risk and, as part of that project, it may be appropriate to consider 
whether or not the applicability of this reporting criterion, as well 
as other reporting criteria, should be based on such lists. It is 
expected that at that time the criteria necessary for development of 
the list will have been adequately established.
    Comment C (Eliminate reporting for historical events): The proposed 
rule would have eliminated the requirement for a telephone 
notification, under 10 CFR 50.72, for:
    (1) ``Any event, found while the reactor is shutdown, that, had it 
been found while the reactor was in operation, would have resulted in 
the nuclear power plant, including its principal safety barriers, being 
seriously degraded or being in an unanalyzed condition that 
significantly compromises plant safety,'' and
    (2) ``Any event or condition that alone could have prevented the 
fulfillment of the safety function of structures or systems that are 
needed to: (A) Shut down the reactor and maintain it in a safe shutdown 
conditions; (B) remove residual heat; (C) Control the release of 
radioactive material; or (D) Mitigate the consequences of an accident'' 
if the condition no longer exists at the time of discovery.
    The proposed rule would also have eliminated the requirement for a 
written licensee event report (LER), under 10 CFR 50.73, for:
    (1) ``Any operation or condition prohibited by the plant's 
Technical Specifications,'' if the condition has not existed within 
three years of the date of discovery, and
    (2) ``Any event or condition that alone could have prevented the 
fulfillment of the safety function of structures or systems that are 
needed to: (A) Shut down the reactor * * *; (B) remove residual heat; 
(C) Control the release of radioactive material; or (D) Mitigate the 
consequences of an accident,'' if the condition has not existed within 
three years of the date of discovery.
    With regard to 10 CFR 50.73, public comment was specifically 
invited on whether such historical events and conditions should be 
reported (rather than being excluded from reporting, as proposed). 
Public comment was also invited on whether the three year exclusion of 
such historical events and conditions should be extended to all written 
reports required by section 50.73(a) (rather than being limited to 
these two specific reporting criteria, as proposed).
    Most commenters supported the revisions to 10 CFR 50.72 that 
eliminate reporting of historical events. They stated that no safety 
significance exists for 10 CFR 50.72 reporting of historical events.
    Most commenters also supported: (1) the elimination of written LERs 
for historical events for the two cases proposed; (2) extending the 
exclusion to all written reports required under section 50.73(a); and 
(3) using two years as a cutoff point, rather than three years. They 
stated that two years encompasses one refueling cycle of operation. 
Significant effort can be expended searching back in history for 
historical events. Reporting historical events more than two years old 
provides a low safety benefit and unnecessarily increases the reporting 
burden. It was recognized that three years is consistent with the time 
period that performance indicators are tracked under the new oversight 
process. However, most commenters stated that no safety significance 
exists for 10 CFR 50.73 reporting of historical events which occurred 
more than two years ago.
    Response: The final rule eliminates the requirement to provide a 
telephone notification or a written LER for a historical event for the 
reasons discussed above.
    The cutoff date for reporting of historical events remains at 3 
years, as was indicated in the proposed rule. The 3-year cutoff is 
necessary because the NRC staff tracks performance indicators for a 
period of 3 years, in order to include a refueling outage as well as an 
extended period of operations, which provides more stable performance

[[Page 63773]]

indicators. The additional burden of searching back for 3 years to 
determine if a condition existed within three years of the date of 
discovery, instead of only 2 years, is minimal because this type of 
event is rarely identified. Thus, it is considered justified in order 
to provide better performance indicators.
    Comment D (Time limits for reporting): The proposed rule would have 
continued to require reporting within one hour after occurrence for 
declaration of an Emergency Class, or for deviation from the plant's 
Technical Specifications authorized pursuant to 10 CFR 50.54(x). 
Reporting of other events that are reportable by telephone under 10 CFR 
50.72 would be reportable within 8 hours after occurrence, rather than 
within 1 hour or 4 hours as was previously required. Submittal of 
written LERs would be required within 60 days after discovery, rather 
than within 30 days as previously required.
    Public comment was specifically invited on the question of whether 
additional levels should be used to better correspond to particular 
types of events. For example, 10 CFR 50.72 previously required 
reporting within 4 hours for events that involve low levels of 
radioactive releases, and events related to safety or environmental 
protection that involve a press release or notification of another 
government agency. These types of events could be maintained at 4 hours 
so that information is available on a more timely basis to respond to 
heightened public concern about such events. In another example, events 
related to environmental protection are sometimes reportable to another 
agency, which is the lead agency for the matter, with a different time 
limit, such as 12 hours. These types of events could be reported to the 
NRC at approximately the same time as they are reported to the other 
agency.
    Most comments on the proposed rule supported the proposal to use 
just three basic levels of required reporting times in 10 CFR 50.72 and 
10 CFR 50.73 (1 hour, 8 hours, and 60 days), as indicated in the 
proposed rule, in the interest of simplicity. They indicated that 
additional levels of reporting are not needed. They also agreed with 
the revised reporting times based on importance to risk and extending 
the required reporting times consistent with the need for prompt NRC 
action. Additionally, they noted that the increased time for submittal 
of LERs will allow for completion of required engineering evaluations 
after event discovery, provide for more complete and accurate LERs, and 
result in fewer LER revisions and supplemental reports.
    One comment letter, from the State of North Carolina, recommended 
maintaining the required reporting time at 4 hours for:
    (1) Any airborne radioactive release that, when averaged over a 
time period of 1 hour, results in concentrations in an unrestricted 
area that exceed 20 times the applicable concentration specified in 
Appendix B to Part 20, Table 2, Column 1;
    (2) Any liquid effluent release that, when averaged over a time of 
1 hour, exceeds 20 times the applicable concentration specified in 
Appendix B to Part 20, Table 2, Column 2, at the point of entry into 
the receiving waters (i.e., unrestricted area) for all radionuclides 
except tritium and dissolved noble gases;
    (3) Any event requiring the transport of a radioactively 
contaminated person to an offsite medical facility for treatment; and
    (4) Any event or situation, related to the health and safety of the 
public or onsite personnel, or protection of the environment, for which 
a news release is planned or notification to other government agencies 
has been or will be made. Such an event may include an onsite fatality 
or inadvertent release of radioactively contaminated materials.
    The letter indicated that the information from such events are of 
interest to the public and public officials. Furthermore, the State's 
Division of Radiation Protection (DRP) provides independent advice to 
State decision-makers as part of its emergency preparedness function. 
Any delay in providing the information to the DRP may prevent or delay 
decisions on public health or public announcements. State agencies may 
be able to get the information from licensees, even under the proposed 
rule. However, this can be difficult to do when an incident is actually 
occurring unless the NRC's rules mandate the reporting within a prompt 
and well-defined period of time.
    Similar comments were received from the State of Illinois regarding 
the ANPR.
    Response: After consideration of the comments, and the potential 
need for NRC action, the final rule employs four basic levels of 
required reporting times in 10 CFR 50.72 and 10 CFR 50.73 (1 hour, 4 
hours, 8 hours, and 60 days). Although this is not as simple as using 
just three levels, as was indicated in the proposed rule, it allows 
more flexibility in matching the required reporting time to the 
potential need for NRC action.
    The final rule requires 4-hour reporting, if the event was not 
reported in 1 hour, for an event or situation, related to the health 
and safety of the public or onsite personnel, or protection of the 
environment, for which a news release is planned or notification to 
other government agencies has been or will be made. Such an event may 
include an onsite fatality or inadvertent release of radioactively 
contaminated materials. This is the same as previously required. These 
reports are needed promptly because they involve events where there may 
be a need for the NRC to respond to heightened public concern.
    The final rule also requires 4-hour reporting, if the event was not 
reported in 1 hour, for unplanned transients. These are events where 
there may be a need for the NRC to take a reasonably prompt action, 
such as partially activating its response plan to monitor the course of 
the event. In summary, they are:
    (a) An event that resulted or should have resulted in ECCS 
discharge into the reactor coolant system (RCS) as a result of a valid 
signal, except when it results from and is part of a pre-planned 
sequence during testing or operation. Previously this was a 1-hour 
report.
    (b) Initiation of a shutdown required by the plant's Technical 
Specifications. Previously this was a 1-hour report.
    (c) A reactor scram or reactor trip when the reactor is critical, 
except when it results from and is part of a pre-planned sequence 
during testing or operation. Previously, actuation of any engineered 
safety feature (ESF), including the reactor protection system (RPS), 
was a 4-hour report.
    Three criteria are deleted from Sec. 50.72 because they are not 
needed in order to obtain prompt notification of events. They are 
retained in Sec. 50.73, however, because they are needed in order to 
obtain written LERs. In summary, they are:
    (a) A natural phenomenon or other external event that poses an 
actual threat to plant safety, or significantly hampers site personnel 
in the performance of duties necessary for safe operation. Events of 
this type are captured by declaration of an Emergency Class, which is 
reportable within 1 hour.
    (b) An internal event that poses an actual threat to plant safety, 
or significantly hampers site personnel in the performance of duties 
necessary for safe operation, including fires, toxic gas releases, or 
radioactive releases. Events of this type are captured by declaration 
of an Emergency Class, which is reportable within 1 hour.
    (c) An airborne radioactive release, or liquid effluent release, 
that exceeds specific limits. Releases that are large enough to warrant 
prompt notification are captured by declaration of an

[[Page 63774]]

Emergency Class, which is reportable within 1 hour after the 
declaration. Releases that involve a public announcement or 
notification to another agency are reportable within 1 hour after the 
announcement or notification.
    For the remaining events reportable under Sec. 50.72, the final 
rule requires 8-hour reporting, if not reported in 1 hour or 4 hours. 
These are events where there may be a need for the NRC to take an 
action within about a day, such as initiating a special inspection or 
investigation. In summary, they are:
    (a) The plant including its principal safety barriers being in a 
seriously degraded condition, or the plant being in an unanalyzed 
condition that significantly degrades plant safety.
    (b) A valid actuation of any system listed in paragraph 
(b)(3)(iv)(B), except when the actuation results from and is part of a 
pre-planned sequence during testing or reactor operation.
    (c) An event or condition that at the time of discovery could have 
prevented fulfillment of the safety function of structures or systems 
needed to shut down the reactor, remove residual heat, control the 
release of radioactive material, or mitigate an accident.
    (d) Transport of a radioactively contaminated person to an offsite 
medical facility.
    (e) A major loss of emergency assessment capability, offsite 
response capability, or offsite communications capability.
    Comment E (Eliminate all reporting of invalid ESF actuations): The 
proposed rule would have eliminated the requirement for a telephone 
notification, under 10 CFR 50.72, for an ESF actuation if it is an 
invalid automatic actuation or an unintentional manual actuation. It 
was stated that invalid actuations are generally less significant than 
valid actuations because they do not involve plant conditions (e.g., 
low reactor coolant system pressure) that would warrant system 
actuation. Instead, they result from other causes (such as a dropped 
electrical lead during testing).
    The proposed rule would not have eliminated the requirement for a 
written LER for such events. It was stated that there is still a need 
for reporting, because the reports are used in making estimates of 
equipment reliability parameters, which in turn are needed to support 
the Commission's move towards risk-informed regulation.
    Most commenters indicated that invalid ESF actuations should not be 
reported under 10 CFR 50.73 unless the actuation impacts the plant such 
that other reporting criteria are independently met. They stated that 
contrary to the NRC's expectations, reporting of invalid actuations 
will not provide the information needed to estimate equipment 
reliability parameters. This information should be collected by other 
less burdensome mechanisms, such as the Equipment Performance and 
Information Exchange (EPIX) system and Maintenance Rule reports.
    Response: The NRC disagrees with many of the comments. Invalid 
actuations do provide information needed in estimating equipment 
reliability because they constitute unplanned demands. The response to 
unplanned demands may or may not differ significantly from those of 
planned test demands. Thus, in making reliability estimates, the 
results from unplanned demands are compared against those from planned 
test demands to determine whether or not it is appropriate to combine 
them. As indicated in the Commission Paper SECY-97-101, May 7, 1997, 
``Proposed Rule, 10 CFR 50.76, Reporting Reliability and Availability 
Information for Risk-significant Systems and Equipment,'' Attachment 3, 
this is one of the categories of information that the NRC relies upon 
in order to make equipment reliability estimates.
    As also discussed in SECY-97-101, EPIX is a voluntary program which 
does not provide a break out of invalid actuations and their results. 
The fact that ESF actuations are reported in written LERs was one of 
the key factors in making the determination that the NRC could work 
around weaknesses in the EPIX data in order to develop reliability 
estimates.
    Reports developed under the maintenance rule, 10 CFR 50.65, are not 
submitted to the NRC.
    Regardless, the Commission agrees that a reduction in unnecessary 
burden is warranted. Accordingly, the final rule takes the following 
approach:
    (a) The requirement to provide a telephone notification under 
Sec. 50.72 for an invalid ESF actuation is eliminated, as was indicated 
in the proposed rule.
    (b) The requirement to report these events under Sec. 50.73 is 
retained. However, the licensee is given the option of providing a 
telephone report rather than a written LER. This is far less 
burdensome. In this case, the telephone notification has the same due 
date as the LER would have (60 days) because the information is not 
needed immediately.
    Comment F (Eliminate reporting of high pressure coolant injection 
(HPCI) inoperability): As indicated in the 1983 Statements of 
Considerations for 10 CFR 50.72 and 50.73, failure or inoperability of 
a single train system, such as the HPCI system in BWRs, is considered 
to constitute an ``event or condition that alone could have prevented 
the fulfillment of the safety function of structures or systems that 
are needed to: (A) Shut down the reactor * * *; (B) Remove residual 
heat; (C) Control the release of radioactive material; or (D) Mitigate 
the consequences of an accident.''
    Most commenters indicated that inoperability of HPCI does not of 
itself constitute a condition that would prevent the fulfillment of a 
safety function. Therefore, there is no benefit in reporting of HPCI 
inoperability if it has no affect on the ability to fulfill a safety 
function. BWR design considers HPCI inoperability and provides 
supporting systems such as reactor core isolation cooling (RCIC), Core 
Spray, and automatic depressurization system (ADS). This is supported 
by the relatively long Allowed Outage Time for HPCI in the Standard 
Technical Specifications (i.e., 14 days). If, in the event of HPCI 
inoperability, it can be shown that these systems are available and 
capable of fulfilling the safety function without HPCI, the event 
should not be reportable. Reporting HPCI inoperability in these cases 
has no meaning for event reporting and appears to be solely a data 
gathering exercise.
    Additionally, the reactor oversight process uses a performance 
indicator for Safety System Functional Failures based on 10 CFR 50.73 
reports. These indicators count failures of single train systems (such 
as HPCI), assuming that the event report documents a safety system 
failure. Reporting HPCI inoperability when there is no impact on the 
overall capability to fulfill the safety function (e.g., remove 
residual heat) will result in an overly conservative and detrimental 
assessment of this indicator.
    Response: As indicated in the 1983 Statements of Considerations for 
10 CFR 50.72 and 50.73, the purpose of this reporting criterion is to 
capture failure, inoperability, etc. on the basis of a structure or 
system. Thus, if an event or condition could have prevented fulfillment 
of the safety function of a system (i.e., by that system), it is 
reportable even if other system(s) could have performed the same safety 
function(s).
    Also, in its assessment of plant performance, the NRC uses a 
performance indicator that includes failure or inoperability of single 
train systems such as HPCI. Thus, elimination of the requirement to 
report such events would be contrary to one of the objectives of the 
rulemaking--to

[[Page 63775]]

maintain consistency with the NRC's actions to improve integrated plant 
performance.
    Comment G (Allow 8 hours after discovery for telephone reporting): 
Section 50.72(b)(3) states ``* * * the licensee shall notify the NRC as 
soon as practical and in all cases, within eight hours of the 
occurrence of any of the following: * * *'' The comment letter states 
that this should be revised to say ``* * * the licensee shall notify 
the NRC as soon as practical and in all cases, within eight hours of 
the occurrence or discovery of any of the following: * * *.'' The 
addition of the term ``or discovery'' provides for those events that 
are discovered to have occurred in the past, remained undetected for 
sometime, and presently exist.
    Response: The NRC disagrees. Addition of the term ``or discovery,'' 
as suggested by the comment, is not necessary. As they have in the 
past, the reporting guidelines address those limited cases, such as 
discovery of an existing but previously unrecognized condition, where 
it may be necessary to undertake an evaluation in order to determine if 
an event or condition is reportable. In other cases, where telephone 
reporting is required, the event should be reported as soon as 
practical and not later than the specified time limit.
    Comment H (Eliminate telephone reporting for non-critical scrams): 
Most commenters recommended that telephone reporting of RPS actuation 
(reactor scrams) be limited to those occurring from a critical 
condition.
    Response: The NRC partially disagrees. A valid scram, even from a 
subcritical condition, is indicative of an event with enough 
significance that the NRC should screen and/or review it on the day it 
occurs, rather than waiting for submittal of a written LER. However, 
telephone reporting under section 50.72 has been eliminated for invalid 
scrams from a subcritical condition.
    Comment I (Limit reporting to conditions that do prevent 
fulfillment of a required function): Regarding section 50.72(b)(2)(v), 
which indicates that licensees shall report: ``Any event or condition 
that at the time of discovery could have prevented the fulfillment of 
the safety function of structures or systems that are needed to: * * 
*,'' this should be revised to read as follows: ``Any event or 
condition that at the time of discovery is preventing the ability to 
fulfill the safety function of structures or systems that are needed 
to: * * *''
    This change is required to reflect the correct tense of the 
existence of an event or condition, rather than past speculation. 
Because of past confusion pertaining to the interpretation of this 
area, it is suggested that further discussion be included in the 
statements of consideration explaining that ``is preventing'' 
represents actual conditions and does not imply that further failures 
should be speculated.
    Response: The NRC does not agree. The term ``could have prevented'' 
reflects the meaning of the rule. It means that, at the time of 
discovery, the condition could have prevented fulfillment of the 
function (for example, had there been a demand for the function). This 
includes but is not limited to the case where, at the time of 
discovery, the condition is actually preventing fulfillment of the 
function.
    This Statement of Considerations and the reporting guidelines 
indicate that, in evaluating reportability under this criterion, it is 
not necessary to postulate an additional random single failure.
    Comment J (Human performance data in LERs): Section 
50.73(b)(2)(ix)(J) previously required that the narrative section of an 
LER include the following specific information as appropriate for the 
particular event:
    ``(1) Operator actions that affected the course of the event, 
including operator errors, procedural deficiencies, or both, that 
contributed to the event.
    (2) For each personnel error, the licensee shall discuss:
    (i) Whether the error was a cognitive error (e.g., failure to 
recognize the actual plant condition, failure to realize which systems 
should be functioning, failure to recognize the true nature of the 
event) or a procedural error;
    (ii) Whether the error was contrary to an approved procedure, was a 
direct result of an error in an approved procedure, or was associated 
with an activity or task that was not covered by an approved procedure;
    (iii) Any unusual characteristics of the work location (e.g., heat, 
noise) that directly contributed to the error; and
    (iv) The type of personnel involved (i.e., contractor personnel, 
utility-licensed operator, utility non-licensed operator, other utility 
personnel).''
    The proposed amendment would have changed section 
50.73(b)(2)(ii)(J) to simply state: ``For each human performance 
related problem that contributed to the event, the licensee shall 
discuss the cause(s) and circumstances.'' It was stated that the 
current rule is more detailed than necessary. Details would continue to 
be provided in the reporting guidelines, as indicated in section 5.2.1 
of the draft of Revision 2 to NUREG-1022.
    Most commenters recommended that, instead of adopting the wording 
in the proposed rule, section 50.73(b)(2)(ii)(J) be revised to state: 
``For each root cause personnel error, the licensee shall discuss the 
cause(s) and circumstances.'' They stated that the shift from 
``personnel error'' and the implied ``root cause'' to ``human 
performance related problem'' and ``contributing factors'' would 
greatly increase the scope of investigation and burden to the licensee. 
They also stated that it is only appropriate to require discussion of 
personnel error root causes.
    Response: The intent of the proposed change was to clarify the 
requirements, not to expand them. Accordingly, the final rule states 
``For each human performance related root cause, the licensee shall 
discuss the cause(s) and circumstances.'' This limits the requirement 
to discussion of root causes of the event. It would not be appropriate, 
or consistent with the previous requirement discussed above, to limit 
the requirement to discussion of personnel error root causes, as 
opposed to procedural deficiency root causes, for example.
    Comment K (Do not require additional availability data in LERs): 
Section 50.73(b)(3) requires that the assessment of safety consequences 
in an LER include the availability of systems or components that could 
have performed the same functions as systems or components that failed 
during the event. Proposed section 50.73(b)(3)(ii) would add a 
requirement that the assessment also include the availability of 
systems or components that: ``Are included in emergency or operating 
procedures and could have been used to recover from the event in case 
of an additional failure in the systems actually used for recovery.''
    Most commenters objected to this new provision, on the grounds that 
it adds significant burden without adding value. They stated that 
reporting should be based on existing plant conditions. Emergency 
operating procedures provide direction for use of many plant systems. 
If additional failures must be postulated, multiple systems would be 
required to be included in the LER for each safety function. There 
exists an infinite combination of failures that could be postulated. 
This unbounded requirement would result in a large amount of additional 
information that would be of minimal use. The assessment of the safety 
consequences and implications of the event would become cluttered with 
hypothetical additional failures and possible plant responses. Some 
commenters stated that the proposed requirement would require licensees 
to speculate on actions that could have been taken, and it would

[[Page 63776]]

add significant burden with no added value.
    Response: The purpose of the proposed change was to ensure that 
LERs contain sufficient information to support a risk assessment of the 
event. Usually there is enough information, or there is nearly enough 
information and the NRC staff can telephone the licensee to obtain any 
additional information needed. Section 50.73(b)(2)(6) requires that 
LERs include ``The name and telephone number of a person within the 
licensee's organization who is knowledgeable about the event and can 
provide additional information concerning the event and the plant's 
characteristics.'' Further, Section 50.73(c) provides that the NRC may 
require submittal of additional information if necessary for complete 
understanding of an unusually complex or significant event.
    However, for those events that occur when the plant is shutdown, it 
has been difficult to obtain enough information because it cannot be 
assumed that equipment that is normally operable and available during 
operation is available during plant shutdown. Accordingly, in the final 
rule there is a requirement for additional availability information. To 
eliminate unnecessary burden, the requirement for additional 
availability data is limited to shutdown events. Also, it is revised to 
simply require providing the availability of systems needed to shut 
down the reactor and maintain safe shutdown conditions, remove residual 
heat, control the release of radioactive material, or mitigate an 
accident. This will eliminate potential difficulties in deciding what 
combinations of failures should be postulated for the purpose of 
deciding which systems to address.
    Comment L (The rule should stand alone): Licensees must use both 
the rule and NUREG-1022, Rev. 2, to determine reportability of 
conditions. The rule should be a stand-alone document written simply 
enough to be understood without the need for a 100+ page guidance 
document.
    Response: The NRC does not agree that it is necessary to eliminate 
the detailed event reporting guidelines and/or include a similar level 
of detail in the rule. Generally speaking, the rule language cannot be 
precise enough to cover all the situations that might be governed by 
the rule and require clarification. Furthermore, in response to the 
ANPR, most commenters expressed the need for timely guidance on the 
final rule. Finally, the NRC has reviewed the guidelines and modified 
them where necessary to ensure they are consistent with the final rule.
    Comment M (The terms ``significant'' and ``serious'' are not 
defined in the rule): One commenter stated that the terms 
``significantly affects'' and ``seriously degraded'' are not defined 
anywhere in the proposed rule.
    Response: The NRC does not agree that it is necessary to define 
these terms in the rule. The term ``unanalyzed condition that 
significantly affects plant safety,'' which was used in the proposed 
rule, is changed to ``unanalyzed condition that significantly degrades 
plant safety'' in the final rule, to make it clear that only matters 
with a negative effect on safety are reportable. Its meaning is defined 
by the same examples that have served since 1983 to define the term 
``unanalyzed condition that significantly compromises plant safety.'' 
These are: (1) Multiple functionally related safety grade components 
out of service; (2) accumulation of voids that could inhibit the 
ability to adequately remove heat from the reactor core, particularly 
under natural circulation conditions; and (3) voiding in instrument 
lines that results in erroneous indication causing the operator to 
misunderstand the true condition of the plant. Also, two new examples 
have been added. They are: (1) Discovery that a system required to meet 
the single failure criterion does not do so; and (2) discovery that the 
fire protection system does not protect at least one safe shutdown 
train in the event of fire in a given area. All of these examples are 
discussed in the Statement of Considerations for the final rule as well 
as the reporting guidelines.
    The term ``condition of the nuclear power plant, including its 
principal safety barriers, being seriously degraded'' is defined by 
guidance that is very similar to the guidance which has defined it 
since 1983. Specifically, the guidance states that this criterion 
applies to material (e.g., metallurgical or chemical) problems that 
cause abnormal degradation of or stress upon the principal safety 
barriers (i.e., the fuel cladding, reactor coolant system pressure 
boundary, or the containment) such as:
    (1) Fuel cladding failures in the reactor, or in the storage pool, 
that exceed expected values, or that are unique or widespread, or that 
are caused by unexpected factors.
    (2) Welding or material defects in the primary coolant system which 
cannot be found acceptable under ASME Section XI, IWB-3600, 
``Analytical Evaluation of Flaws'' or ASME Section XI, Table IWB-3410-
1, ``Acceptance Standards.''
    (3) Serious steam generator tube degradation.
    (4) Low temperature over pressure transients where the pressure-
temperature relationship violates pressure-temperature limits derived 
from Appendix G to 10 CFR Part 50 (e.g., TS pressure-temperature 
curves).
    (5) Loss of containment function or integrity, including 
containment leak rate tests where the total containment as-found, 
minimum-pathway leak rate exceeds the limiting condition for operation 
(LCO) in the facility's TS.
    This guidance is discussed in further detail below under the 
heading ``Principal safety barrier seriously degraded.''
    Comment N (False elevated sense of problems): In addition to the 
points discussed above under the heading ``Comment E,'' some commenters 
stated that reporting of invalid actuations will convey a false 
elevated sense of problems to the general public, causing undue alarm 
for situations that actually represent little or no safety or risk 
significance. Therefore, the new rule should not require invalid 
actuations to be reported.
    Response: The NRC does not agree that it is necessary to eliminate 
reporting for invalid actuations in order to avoid conveying a false 
elevated sense of problems to the general public. As discussed in the 
response to Comment E, there is a need for reporting of these events 
because they are used in making estimates of equipment reliability 
parameters, which in turn are needed to support the NRC's move towards 
risk-informed regulation. Invalid actuations have been reportable since 
1983 under the previous rules, pursuant to both sections 50.72 and 
50.73. No undue public alarm about such invalid actuations has been 
apparent to the NRC. The commenters did not identify any specific 
situation or provide any anecdotal evidence that reporting such invalid 
actuations has caused undue public alarm.
    Comment O (Eliminate reporting of missing fire barriers): One 
commenter stated that the proposed rule notice at Page 36299, first 
column, the example pertaining to missing or degraded fire barriers 
basically equates such conditions with degraded principal safety 
barriers (i.e., fuel cladding, reactor coolant pressure boundary, and 
containment). This is inappropriate and should be deleted.
    Response: The NRC does not agree. The example indicates that a 
condition is reportable, as an unanalyzed condition that significantly 
affects plant safety, ``if fire barriers are found to be missing such 
that the required degree of separation for redundant safe shutdown 
trains is lacking.'' This would mean

[[Page 63777]]

that, if a fire occurs in the given area, no safe shutdown trains would 
be protected to an acceptable degree. Because Probabilistic Risk 
Assessment (PRA) studies continue to indicate that fire is a dominant 
contributor to risk, the inability to guarantee one train of safe 
shutdown capability, as required, is considered to be a condition that 
significantly degrades safety.
    Comment P (Applicability of the backfit rule--no basis was stated): 
One commenter stated that in the proposed rule at Page 36303, Section 
VI., Backfit Analysis, the NRC stated that 10 CFR 50.109 does not apply 
without giving any basis for the claim.
    Response: The discussion below, entitled Backfit Analysis, has been 
modified to provide the basis for the conclusion that 10 CFR 50.109 
does not apply.
    Comment Q (Modify ``unanalyzed condition that significantly affects 
safety''): Most commenters stated that in section 50.72(b)(2)(ii)(B), 
the phrase ``significantly affects plant safety'' has no positive or 
negative connotation. Reword the section to read, ``The nuclear power 
plant being in an unanalyzed condition that significantly degrades 
plant safety.''
    Response: The NRC agrees. The phrase is revised as recommended for 
the reason stated.
    Comment R (Recognize risk-significance factors): One commenter 
stated that Section 50.73(a)(1) fails to recognize any risk 
significance factors.
    Response: The NRC does not agree. Section 50.73(a)(1) is general in 
nature and indicates that, unless otherwise specified in section 50.73, 
the licensee shall report an event if it occurred within the last three 
years regardless of the plant mode or power level, and regardless of 
the significance of the structure, system, or component that initiated 
the event. Risk factors are recognized elsewhere in section 50.73. For 
example, the requirement to report an event or condition that could 
have prevented fulfillment of the safety function of structures or 
systems is limited to those structures or systems that are needed to 
perform specific safety functions. The list of systems for which 
actuation must be reported is based on risk-significance. Lack of 
significance is the reason for the elimination of reporting for late 
surveillance tests where the equipment, when tested, is functional. It 
is also the basis for eliminating several other requirements, such as 
immediate notification under section 50.72 for many invalid actuations.
    Comment S (Modify ``operation or condition prohibited by TS''): 
Section 50.73(a)(2)(i)(B) should be revised to read, ``Any operation or 
condition occurring within three years of the date of discovery which 
was prohibited by the plant's CURRENT Technical Specifications.'' This 
rewrite would direct plants that recently converted to Improved 
Standard Technical Specifications to apply the current requirements to 
the identified condition, rather than having to consider the previous 
requirements under old Technical Specifications which are no longer 
applicable.
    Response: The NRC agrees. The issue involves the following 
scenario. A licensee discovers a historical operation or condition that 
was prohibited by the TS in effect at the time the operation or 
condition occurred. However, the prohibition has subsequently been 
removed from the TS. The event is not considered significant because 
subsequently the operation or condition was found to be acceptable and 
the Technical Specifications have been revised to permit it. 
Accordingly, the final rule eliminates the requirement to report such 
events.
    Comment T (Reporting burden would not be decreased): In addition to 
the points discussed above under the heading ``Comment A,'' one 
commenter disagreed with the NRC's assessment that the proposed rule 
would represent an overall decrease in burden. This disagreement was 
based on the following points:
    (a) (Telephone notifications are less burdensome than written 
LERs): Although the proposed rule would have decreased the number of 
phone-in reports pursuant to 10 CFR 50.72, the commenter believes this 
burden is very small when compared with the burden of processing and 
submitting Licensee Event Reports (LERs) pursuant to 10 CFR 50.73.
    (b) (Actuation of systems that are currently excluded systems would 
become reportable): In the proposed rule, systems that were excluded 
from reporting requirements via previous rulemaking because they 
represented little or no safety significance have been reinstated 
(e.g., Reactor Water Cleanup System). Such action will now lead to 
reporting all isolations, even those with no safety significance.
    (c) (Systems not classified as ESF would be treated as ESF): 
Systems that are not classified as Engineered Safety Features (ESF) 
will now be treated as ESF (e.g., Reactor Core Isolation Cooling 
System).
    (d) (Invalid actuations would be added to the reporting 
requirements): Invalid actuations are now included in the reporting 
requirements. The impact of this change is that the clarifications for 
what used to be reportable have been deleted. Therefore, the proposed 
rule would treat all isolations or movements of a component as 
reportable regardless of safety significance.
    (e) (The requirements for human performance data would be 
increased): The scope of information requested for human performance 
events has substantially increased, going well beyond previous direct 
root cause to now include associated contributing factors.
    Response: The NRC believes that reporting burden will be decreased 
for the reasons described in the regulatory analysis. With regard to 
the specific bases cited for this comment:
    (a) The NRC agrees that a telephone notification is less burdensome 
than a written LER. However, this does not mean that the reporting 
burden would be increased, or maintained, unless there is some increase 
in the number of LERs required under the final rule. This is not the 
case.
    (b) The NRC does not agree that the proposed rule would have made 
actuation of previously excluded systems reportable. The previously 
excluded systems are: (i) Reactor water clean-up system; (ii) control 
room emergency ventilation system; (iii) reactor building ventilation 
system; (iv) fuel building ventilation system; or (v) auxiliary 
building ventilation system. None of these appeared on the proposed 
list of systems for which actuation would be reportable.
    (c) The NRC believes that system actuations added by adoption of 
the proposed list of systems are outweighed by system actuations 
eliminated.
    (d) The NRC does not agree that invalid actuations are being added 
to the reporting requirements, because they were already in the 
reporting requirements.
    (e) See the response to Comment J.
    Comment U (Incentive to disable safety systems): In addition to the 
points discussed above under the heading ``Comment E,'' one commenter 
indicated that reporting of invalid system actuations provided an 
incentive to disable safety systems.
    Response: The NRC does not agree that it is necessary to eliminate 
reporting for invalid actuations to avoid creating an incentive to 
disable safety systems during maintenance activities to avoid the 
possibility of reporting an inadvertent actuation.
    As discussed in the response to Comment E, there is a need for 
reporting of these events because they are used in making estimates of 
equipment reliability parameters, which in turn are

[[Page 63778]]

needed to support the NRC's move towards risk-informed regulation. 
Also, in the final rule, licensees are not required to provide an 
immediate notification under Section 50.72 for an invalid system 
actuation. Furthermore, in the final rule licensees have the option of 
providing a telephone notification within 60 days, rather than 
submitting a written LER, for an invalid system actuation. These 
changes provide a drastic reduction in the burden of reporting for 
invalid system actuations. This burden reduction mitigates against any 
incentive to disable safety systems during maintenance in order to 
avoid the possibility of reporting an invalid actuation.
    Comment V (Amend 10 CFR 76.120(d)(2) to allow 60 days): One 
commenter noted that the NRC plans to consider the idea of expanding 
the 60-day deadline for written reports to other regulations. The 
commenter recommended amending 10 CFR 76.120(d)(2) to allow 60 days for 
written reports required under that regulation.
    Response: The NRC continues to plan to evaluate the need for 
rulemaking to modify 10 CFR Parts 72 and 73, including the suggestion 
that 60 days be allowed for written reports required under 10 CFR 72.75 
and 73.71. As part of that effort, the NRC will also consider the 
suggestion that 60 days be allowed for written reports required under 
10 CFR 76.120(d)(2).
    Comment W (Enforcement levels): Some commenters indicated that the 
proposed characterization of Enforcement Level III for failure to 
provide a required 1-hour or 8-hour non-emergency telephone 
notification is too harsh in most cases. They indicated that in most 
cases the information provided in these non-emergency notifications has 
low safety significance.
    Response: As discussed further below under the heading 
``Enforcement,'' the philosophy of the Enforcement Policy changes is to 
base the significance of the reporting violation on its impact on the 
NRC's ability to provide proper oversight of licensee activities. 
Accordingly, in some cases, Severity Level III is appropriate for 
failure to make a required telephone notification and in other cases it 
is not.
    Comment X (LER format and content): One commenter recommended that 
the NRC reconsider a ``check the box'' approach. The commenter 
indicated that such an approach could be crafted to make LER data entry 
easier, more consistent, and less ambiguous, without making LERs more 
difficult for the general public to understand.
    Response: The NRC does not believe it is feasible to adopt a 
``check the box'' in the final rule because the proposed rule did not 
include a proposal along those lines and development of a sound system 
would take considerable time, delaying issuance of the final rule.
    Comment Y (Coordinate with performance indicator efforts): One 
commenter suggested careful coordinated consideration among the NRC 
staff responsible for this rulemaking and those responsible for 
performance indicator efforts to ensure that reports submitted under 10 
CFR 50.73(a)(2)(v) are not being misapplied.
    Response: The NRC agrees and the suggested coordination has taken 
place, and will continue in the future as well. As a result, it is not 
expected that the NRC will misapply reports submitted under 10 CFR 
50.73(a)(2)(v).
    Comment Z: One commenter recommended that telephone notifications 
due within 8 hours should only be required when activation of the NRC 
emergency response organization is actually required.
    Response: The NRC does not agree that this is a feasible approach 
because activation of the NRC's emergency response organization is not 
a simple function of the reporting criterion under which an event is 
considered to be reportable. For example, the emergency response 
organization is sometimes activated for events which, at the time of 
reporting, are considered to correspond to lower levels of Emergency 
Classes or non-emergency reporting criteria.
    Comment AA (Do not include criteria for reporting degraded steam 
generator tubes): The Statement of Considerations for the proposed rule 
and the Draft Revision 2 to NUREG-1022 would indicate that steam 
generator tube degradation is considered serious, and thus reportable 
as a seriously degraded reactor coolant system boundary, if the tubing 
fails to meet specific performance criteria involving margin against 
burst and accident induced leakage rate. Most commenters proposed that 
this guidance be deleted. They stated that the position was based on a 
Draft Regulatory Guide (DG-1074, Steam Generator Tube Integrity) that 
has not been approved. Discussions between the industry and the NRC are 
being held to define the steam generator program and Technical 
Specification requirements. Some of the examples provided in the 
proposed section are contrary to agreements that have been made between 
the industry and the NRC staff. Recognizing that these agreements are 
still evolving, the proposed revisions to the rule(s) and NUREG-1022 
must agree with the final positions on steam generator issues.
    Response: The details have been removed from the Statement of 
Considerations. The details in the final Revision 2 to NUREG-1022 have 
been modified to reflect the NRC staff's current thinking. The guidance 
is consistent with the steam generator tube integrity performance 
criteria and reporting guidelines currently under discussion. This 
reporting is needed to permit the staff to determine if further inquiry 
or action might be needed before the plant is restarted.
    The NRC does not agree that it is necessary to delay issuance of 
this reporting guidance pending staff endorsement of the NEI 97-06 
initiative. The NUREG-1022 guidance merely provides reasonable examples 
of degraded steam generator tube conditions which the NRC needs to 
evaluate. If it is determined in the future that different detailed 
guidance is needed, it can be issued at that time.

III. Discussion

1. Objectives

    The purposes of sections 50.72 and 50.73 remain the same because 
the basic needs remain the same. The essential purpose of section 50.72 
is `` * * * to provide the Commission with immediate reporting of * * * 
significant events where immediate Commission action to protect the 
public health and safety may be required or where the Commission needs 
timely and accurate information to respond to heightened public 
concern.'' (48 FR 39039; August 29, 1983). Section 50.73 ``* * * 
identifies the types of reactor events and problems that are believed 
to be significant and useful to the NRC in its effort to identify and 
resolve threats to public safety. It is designed to provide the 
information necessary for engineering studies of operational anomalies 
and trends and patterns analysis of operational occurrences. The same 
information can be used for other analytic procedures that will aid in 
identifying accident precursors.'' (48 FR 33851; July 26, 1983).
    The objectives of these final amendments are as follows:
    (1) To better align the reporting requirements with the NRC's needs 
for information to carry out its safety mission. An example is 
extending the required initial reporting times for some events, 
consistent with the time at which the reports are needed for NRC 
action.
    (2) To reduce unnecessary reporting burden, consistent with the 
NRC's needs. An example is eliminating the

[[Page 63779]]

reporting of design and analysis defects and deviations with little or 
no risk-or safety-significance.
    (3) To clarify the reporting requirements where needed. An example 
is clarifying the criteria for reporting design or analysis defects or 
deviations.
    (4) Any changes should be consistent with NRC actions to improve 
integrated plant assessments. For example, reports that are needed in 
the assessment process should not be eliminated.

2. Section by Section Discussion of Final Amendments

    General requirements and reportable events [section 50.72(a)(1) and 
section 50.73(a)(1)]. The term ``if it occurred within 3 years of the 
date of discovery'' is added to eliminate reporting for conditions that 
have not existed during the three years before discovery. Such a 
historical event has less significance, and assessing reportability for 
earlier times can consume considerable resources. For example, assume 
that a procedure is found to be unclear and, as a result, a question is 
raised as to whether the plant was ever operated in a prohibited 
condition. If operation in the prohibited condition is likely, the 
answer would be reasonably apparent based on the knowledge and 
experience of the plant's operators and/or a review of operating 
records for the past three years. The effort required to review all 
records older than three years in order to rule out the possibility is 
not warranted.
    A sentence is added to indicate that for an invalid actuation 
reported under section 50.73(a)(2)(iv) the licensee may, at its option, 
provide a telephone notification to the NRC Operations Center within 60 
days after discovery of the event in lieu of submitting a written LER. 
For this type of event, a telephone notification will provide the 
information needed and impose less burden than an LER.
    General requirements [section 50.72(a)(5)]. The requirement to 
inform the NRC of the type of report being made (i.e., Emergency Class 
declared, non-emergency 1-hour report, or non-emergency 8-hour report) 
is revised to refer to paragraph (a)(1) instead of referring to 
paragraph (a)(3) to correct a typographical error.
    Required initial reporting times [sections 50.72(a)(5), (b)(1), 
(b)(2), and new section 50.72(b)(3); and sections 50.73(a)(1) and (d)]. 
In the final amendments, declaration of an Emergency Class continues to 
be reported immediately after notification of appropriate State or 
local agencies and not later than 1-hour after declaration. This 
includes declaration of an Unusual Event, the lowest Emergency Class.
    Deviations from Technical Specifications authorized pursuant to 10 
CFR 50.54(x) continue to be reported as soon as practical and in all 
cases within 1 hour of occurrence. These two criteria capture those 
events where there may be a need for immediate action by the NRC to 
protect public health and safety.
    The requirement to report an event or situation, related to the 
health and safety of the public or onsite personnel, or protection of 
the environment, for which a news release is planned or notification to 
other government agencies has been renumbered from section 
50.72(b)(2)(vi) to section 50.72(b)(2)(xi). In other respects this 
reporting criterion is unchanged, and the event is reportable within 4 
hours, if not reported within 1 hour. This provides the information at 
the time it may be needed to respond to heightened public concern.
    The requirement to report a natural phenomenon or other external 
event that poses an actual threat to plant safety or significantly 
hampers site personnel in the performance of duties necessary for safe 
operation in section 50.72(b)(1)(iii) is deleted. Events of this type 
are captured by declaration of an Emergency Class, which is reportable 
within 1 hour.
    The requirement to report an internal event that poses an actual 
threat to plant safety, or significantly hampers site personnel in the 
performance of duties necessary for safe operation, including fires, 
toxic gas releases, or radioactive releases in section 50.72(b)(1)(vi) 
is deleted. Events of this type are captured by declaration of an 
Emergency Class, which is reportable within 1 hour.
    The requirement to report an airborne radioactive release or liquid 
effluent release that exceeds specific limits in section 50.72 
(b)(2)(iv) is deleted. Releases that are large enough to warrant prompt 
notification are captured by declaration of an Emergency Class, which 
is reportable within 1 hour after the declaration. Releases that 
involve a news release or notification to other government agencies are 
reportable within 4 hours of the occurrence.
    The remaining non-emergency events that are reportable by telephone 
under 10 CFR 50.72 are reportable as soon as practical and in all cases 
within 4 hours or 8 hours (instead of within 1 hour or 4 hours as was 
previously required). This reduces the unnecessary burden of rapid 
reporting, while:
    (1) Capturing, within 4 hours, those events where there may be a 
need for the NRC to take a reasonably prompt action, such as partially 
activating its response plan to monitor the course of the event.
    (2) Capturing, within 8 hours, those events where there may be a 
need for the NRC to take an action within about a day, such as 
initiating a special inspection or investigation.
    See the response to Comment D, above, for further discussion.
    Written LERs are due within 60 days after discovery of a reportable 
event or condition (instead of within 30 days as was previously 
required). Changing the time limit from 30 days to 60 days does not 
imply that licensees should take longer than they previously did to 
develop and implement corrective actions. They should continue to do so 
on a time scale commensurate with the safety significance of the issue. 
However, for those cases where it does take longer than thirty days to 
complete a root cause analysis, this change will result in fewer LERs 
that require amendment (by submittal of an amended report).
    The term ``within 30 days of the discovery of a reportable event or 
situation'' is deleted from section 50.73(d). This provision is 
redundant to the provisions of section 50.73(a)(1), which requires that 
a licensee submit an LER within 60 days after discovery of an event 
described in section 50.73(a). Retaining the time limit, which is now 
60 days, in section 50.73(d) would create a conflict with sections 
20.2201 and 20.2203 which require licensees to submit LERs for the 
events described in those sections within 30 days and in accordance 
with section 50.73(d).
    Operation or condition prohibited by technical specifications 
[section 50.73(a)(2)(i)(B)]. This criterion is modified to eliminate 
reporting if the Technical Specification is administrative in nature. 
Violations of administrative Technical Specifications have generally 
not been considered to warrant submittal of an LER, and since 1983 when 
the LER rule was issued the NRC's event reporting guidelines have 
excluded almost all cases of such reporting. This change makes the 
plain wording of the rule consistent with that guidance.
    Also, this criterion is modified to eliminate reporting if the 
event consisted solely of a case of a late surveillance test where the 
oversight is corrected, the test is performed, and the equipment is 
found to be functional. This type of event has not proven to be 
significant because the equipment remained functional.
    Finally, this criterion is modified to eliminate reporting of an 
operation or

[[Page 63780]]

condition that occurred in the past and was prohibited at that time if, 
prior to discovery of the event, the Technical Specifications were 
revised such that the operation or condition is no longer prohibited. 
Such an event would have little or no significance because, by the time 
of discovery, the operation or condition would have been determined to 
be acceptable and thus permissible under current Technical 
Specifications.
    The NRC expects licensees to include violations of the Technical 
Specifications in their corrective action programs, which are subject 
to NRC audit.
    Condition of the nuclear power plant, including its principal 
safety barriers, being seriously degraded [former sections 
50.72(b)(1)(ii) and (b)(2)(i), replaced by new section 
50.72(b)(3)(ii)(A); and section 50.73(a)(2)(ii), renumbered to 
50.73(a)(2)(ii)(A)]. Previously, 10 CFR 50.72(b)(1)(ii) and (b)(2)(i) 
provided the following distinction. During operation, a seriously 
degraded plant, including its principal safety barriers, was reportable 
within one hour. An event discovered while shutdown that had it been 
discovered during operation would have resulted in a seriously degraded 
plant, including its principal safety barriers, was reportable within 4 
hours. The new 10 CFR 50.72(b)(3)(ii)(A) eliminates the distinction 
because there are no longer separate 1-hour and 4-hour categories of 
non-emergency reports for this criterion. There are only 8-hour non-
emergency reports for this criterion.
    Unanalyzed condition that significantly degrades plant safety 
[sections 50.72(b)(1)(ii)(A) and (b)(2)(i), replaced by new section 
50.72(b)(3)(ii)(B); and section 50.73(a)(2)(ii)(A), renumbered to 
50.73(a)(2)(ii)(B)]. Previously, 10 CFR 50.72(b)(1)(ii)(A) and 
(b)(2)(i) provided the following distinction. During operation, an 
unanalyzed condition that significantly compromised plant safety was 
reportable within 1 hour. An event discovered while shut down that had 
it been discovered during operation would have resulted in an 
unanalyzed condition that significantly compromised plant safety was 
reportable within 4 hours. The new 10 CFR 50.72(b)(2)(ii)(B) eliminates 
this distinction because there are no longer separate 1-hour and 4-hour 
categories of non-emergency reports for this reporting criterion. There 
are only 8-hour non-emergency reports for this criterion.
    In addition, the new 10 CFR 50.72(b)(2)(ii)(B) and 
50.73(a)(2)(ii)(B) refer to a condition that significantly degrades 
plant safety rather than a condition that significantly compromises 
plant safety. This is an editorial change intended to better reflect 
the nature of the criterion.
    Condition that is outside the design basis of the plant [old 
section 50.72(b)(2)(ii)(B); and old section 50.73(a)(2)(ii)(B)]. This 
criterion is deleted. A condition outside the design basis of the plant 
is still required to be reported if it is significant enough to qualify 
under one or more of the following criteria.
    Plant safety significantly degraded. If a condition outside the 
design basis of the plant (or any other unanalyzed condition) is 
significant enough that, as a result, plant safety is significantly 
degraded, the condition is reportable under sections 50.72(b)(2)(ii)(B) 
and 50.73(a)(2)(ii)(B) [i.e., an unanalyzed condition that 
significantly degrades plant safety].
    As was previously indicated in the 1983 Statements of 
Considerations for 10 CFR 50.72 and 50.73, with regard to an unanalyzed 
condition that significantly compromises plant safety, ``The Commission 
recognizes that the licensee may use engineering judgment and 
experience to determine whether an unanalyzed condition existed. It is 
not intended that this paragraph apply to minor variations in 
individual parameters, or to problems concerning single pieces of 
equipment. For example, at any time, one or more safety-related 
components may be out of service due to testing, maintenance, or a 
fault that has not yet been repaired. Any trivial single failure or 
minor error in performing surveillance tests could produce a situation 
in which two or more often unrelated, safety-grade components are out-
of-service. Technically, this is an unanalyzed condition. However, 
these events should be reported only if they involve functionally 
related components or if they significantly compromise plant safety,'' 
(48 FR 39042; August 29, 1983 and 48 FR 33856, July 26, 1983).
    ``When applying engineering judgment, and there is a doubt 
regarding whether to report or not, the Commission's policy is that 
licensees should make the report,'' (48 FR 39042; August 29, 1983).
    ``For example, small voids in systems designed to remove heat from 
the reactor core which have been previously shown through analysis not 
to be safety significant need not be reported. However, the 
accumulation of voids that could inhibit the ability to adequately 
remove heat from the reactor core, particularly under natural 
circulation conditions, would constitute an unanalyzed condition and 
would be reportable,'' (48 FR 39042; August 29, 1983 and 48 FR 33856, 
July 26, 1983).
    ``In addition, voiding in instrument lines that results in an 
erroneous indication causing the operator to misunderstand the true 
condition of the plant is also an unanalyzed condition and should be 
reported,'' (48 FR 39042; August 29, 1983 and 48 FR 33856, July 26, 
1983).
    Furthermore, beyond the examples given in 1983, examples of 
reportable events include discovery that a system required to meet the 
single failure criterion does not do so.
    In another example, if fire barriers are found to be missing, such 
that the required degree of separation for redundant safe shutdown 
trains is lacking, the event is reportable. On the other hand, if a 
fire wrap, to which the licensee has committed, is missing from a safe 
shutdown train but another safe shutdown train is available in a 
different fire area, protected such that the required separation for 
safe shutdown trains is still provided, the event is not reportable.
    Structure or system not capable of performing its specified safety 
function. If a design or analysis defect or deviation (or any other 
event or condition) is significant enough that, as a result, a 
structure or system is not capable of performing its specified safety 
functions, the condition is reportable under sections 50.72(b)(3)(v) 
and 50.73(a)(2)(v) [i.e., an event or condition that could have 
prevented the fulfillment of the safety function of structures or 
systems that are needed to: shut down the reactor * * *; remove 
residual heat; control the release of radioactive material; or mitigate 
the consequences of an accident].
    For example, in one case an annual inspection indicated that some 
bearings were wiped or cracked on both emergency diesel generators 
(EDGs). Although the EDGs were running prior to the inspection, the 
event was reportable because there was reasonable doubt about the 
ability of the EDGs to operate for an extended period of time, as 
required.
    Train inoperable longer than allowed. If a design or analysis 
defect or deviation (or any other event or condition) is significant 
enough that, as a result, one train of a multiple train system 
controlled by the plant's TS is not capable of performing its specified 
safety functions for a period of time longer than allowed by the TS, 
the condition is reportable under section 50.73(a)(2)(i)(B) [i.e., an 
operation or condition prohibited by TS].
    For example, if it is found that an exciter panel for one EDG lacks

[[Page 63781]]

appropriate seismic restraints because of a design, analysis, or 
construction inadequacy and, as a result, there is reasonable doubt 
about the EDG's ability to perform its specified safety functions 
during and after a Safe Shutdown Earthquake (SSE), the event would be 
reportable.
    Or, for example, if it is found that a loss of offsite power could 
cause a loss of instrument air and, as a result, there is reasonable 
doubt about the ability of one train of the auxiliary feedwater system 
to perform its specified safety functions for certain postulated steam 
line breaks, the event would be reportable.
    Principal safety barrier seriously degraded. If a condition outside 
the design basis of the plant (or any other event or condition) is 
significant enough that, as a result, a principal safety barrier is 
seriously degraded, it is reportable under sections 50.72(b)(3)(ii)(A) 
and 50.73(a)(2)(ii)(A) [i.e., any event or condition that results in 
the condition of the nuclear power plant, including its principal 
safety barriers, being seriously degraded]. This reporting criterion 
applies to material (e.g., metallurgical or chemical) problems that 
cause abnormal degradation of or stress upon the principal safety 
barriers (i.e., the fuel cladding, reactor coolant system pressure 
boundary, or the containment) such as:
    (i) Fuel cladding failures in the reactor, or in the storage pool, 
that exceed expected values, or that are unique or widespread, or that 
are caused by unexpected factors.
    (ii) Welding or material defects in the primary coolant system 
which cannot be found acceptable under ASME Section XI, IWB-3600, 
``Analytical Evaluation of Flaws'' or ASME Section XI, Table IWB-3410-
1, ``Acceptance Standards.''
    (iii) Serious steam generator tube degradation.
    (iv) Low temperature over pressure transients where the pressure-
temperature relationship violates pressure-temperature limits derived 
from Appendix G to 10 CFR Part 50 (e.g., TS pressure-temperature 
curves).
    (v) Loss of containment function or integrity, including 
containment leak rate tests where the total containment as-found, 
minimum-pathway leak rate exceeds the limiting condition for operation 
(LCO) in the facility's TS.\1\
---------------------------------------------------------------------------

    \1\ The LCO typically employs La, which is defined in Appendix J 
to 10 CFR Part 50 as the maximum allowable containment leak rate at 
pressure Pa, the calculated peak containment internal pressure 
related to the design basis accident. Minimum-pathway leak rate 
means the minimum leak rate that can be attributed to a penetration 
leakage path; for example, the smaller of either the inboard or 
outboard valve's individual leak rates.
---------------------------------------------------------------------------

    Common cause inoperability of independent trains or channels. If a 
condition outside the design basis of the plant (or any other event or 
condition) is significant enough that, as a result, independent trains 
or channels become inoperable, it would be reportable under section 
50.73(a)(2)(vii) [i.e., an event where a single cause or condition 
caused independent trains or channels to become inoperable]. For 
example, in one case it was found that independent circuit breakers, 
required to operate after a Loss-of-Coolant Accident (LOCA), were not 
qualified for the expected radiation levels (and were thus considered 
inoperable). In another example, a wiring error caused independent 
containment isolation valves to be incapable of properly closing (i.e., 
they would not close tightly because they would stop closing based on 
limit switch operation rather than torque).
    Single Cause that Could Have Prevented Fulfillment of the Safety 
Functions of Trains or Channels in Different Systems. Finally, a 
condition outside the design basis of the plant (or any other event or 
condition) would be reportable if it is significant enough that, as a 
result of a single cause, it could have prevented the fulfillment of a 
safety function for two or more trains or channels in different systems 
that are needed to:
    (1) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (2) Remove residual heat;
    (3) Control the release of radioactive material; or
    (4) Mitigate the consequences of an accident.
    This new criterion is contained in sections 50.73(a)(2)(ix)(A) and 
(B), as discussed below.
    Single Cause that Could Have Prevented Fulfillment of the Safety 
Functions of Trains or Channels in Different Systems. [new sections 
50.73(a)(2)(ix)(a) and (B)]. This new criterion requires reporting any 
event or condition that as a result of a single cause could have 
prevented the fulfillment of a safety function for two or more trains 
or channels in different systems that are needed to:
    (1) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (2) Remove residual heat;
    (3) Control the release of radioactive material; or
    (4) Mitigate the consequences of an accident.
    Events covered by this new criterion may include cases of 
procedural error, equipment failure, and/or discovery of a design, 
analysis, fabrication, construction, and/or procedural inadequacy. 
However, licensees are not required to report an event pursuant to this 
criterion if the event results from:
    (1) A shared dependency among trains or channels that is a natural 
or expected consequence of the approved plant design; or
    (2) Normal and expected wear or degradation.
    Subject to the two exclusions stated above, this criterion captures 
those events where a single cause could have prevented the fulfillment 
of the safety function of multiple trains or channels, but the event:
    (1) Would not be captured by Secs. 50.73(a)(2)(v) and 
50.72(b)(3)(v) [event or condition that could have prevented 
fulfillment of the safety function of structures and systems needed to 
. . .] because the affected trains or channels are in different 
systems; and
    (2) Would not be captured by Sec. 50.73(a)(2)(vii) [common cause 
inoperability of independent trains or channels] because the affected 
trains or channels are either:
    (i) Not assumed to be independent in the plant's safety analysis; 
or
    (ii) Not both considered to be inoperable.
    This new criterion is closely related to Secs. 50.73(a)(2)(v) and 
50.72(b)(3)(v) [event or condition that could have prevented 
fulfillment of the safety function of structures and systems needed to: 
shut down the reactor and maintain it in a safe shutdown condition; 
remove residual heat; control the release of radioactive material; or 
mitigate the consequences of an accident]. Specifically:
    The meaning of the term ``could have prevented the fulfillment of 
the safety function'' is essentially the same for this new criterion as 
it is for Secs. 50.73(a)(2)(v) and 50.72(b)(3)(v) [i.e., there was a 
reasonable expectation of preventing the fulfillment of the safety 
function(s) involved]. However, in contrast to Secs. 50.73(a)(2)(v) and 
50.72(b)(3)(v), reporting under this new criterion applies to trains or 
channels in different systems. Thus, for this new criterion, the safety 
function that is affected may be different in different trains or 
channels.
    In contrast to Secs. 50.73(a)(2)(v) and 50.72(b)(3)(v), reporting 
under this new criterion applies only to a single cause. Also, in 
contrast to Secs. 50.73(a)(2)(v) and 50.72(b)(3)(v), this new criterion 
does not apply to an event that results from

[[Page 63782]]

a shared dependency among trains or channels that is a natural or 
expected consequence of the approved plant design. For example, this 
new criterion does not capture failure of a common electrical power 
supply that disables Train A of AFW and Train A of High Pressure Safety 
Injection (HPSI), because their shared dependency on the single power 
supply is a natural or expected consequence of the approved plant 
design.
    Similar to Secs. 50.73(a)(2)(v) and 50.72(b)(3)(v), this new 
criterion does not capture events or conditions that result from normal 
and expected wear or degradation. For example, consider pump bearing 
wear that is within the normal and expected range. In the case of two 
pumps in different systems, this new criterion categorically excludes 
normal and expected wear. In the case of two pumps in the same system, 
normal and expected wear should be adequately addressed by normal plant 
operating and maintenance practices and thus should not indicate a 
reasonable expectation of preventing fulfillment of the safety function 
of the system.
    This criterion pertains only to written LERs required by 10 CFR 
50.73. Telephone notifications are not required under this criterion.
    It is estimated that the combination of removing the previous 
requirement to report a condition outside the design basis of the plant 
and adding this criterion will, on balance, result in fewer reports. In 
addition, the events reportable under this criterion are events that 
would likely have been considered reportable under the previous 
requirement to report a condition outside the design basis of the 
plant.
    An example of an event that would be reportable under this 
criterion is as follows. During testing, two containment isolation 
valves failed to function as a result of improper air gaps in the 
solenoid operated valves that controlled the supply of instrument air 
to the containment isolation valves. The valves were powered from the 
same electrical division. Thus, Sec. 50.73(a)(2)(vii) [common cause 
inoperability of independent trains or channels] would not apply. The 
two valves isolated fluid process lines in two different systems. Thus 
Sec. 50.73(a)(2)(v) [condition that could have prevented fulfillment of 
the safety function of a structure or system] would apply only if 
engineering judgment indicates there was a reasonable expectation of 
preventing fulfillment of the safety function for redundant valves 
within the same system.\2\ However, this new criterion would certainly 
apply if a single cause (such as a design inadequacy) induced the 
improper air gaps, thus preventing fulfillment of the safety function 
of two trains or channels in different systems.
---------------------------------------------------------------------------

    \2\ Or, alterantively, there was reasonable doubt that the 
safety function would have been fulfilled if the affected trains has 
been called upon to perform them.
---------------------------------------------------------------------------

    Another example of an event reportable under this criterion is as 
follows. A motor operated valve in one train of a system was found with 
a crack 75 percent through the stem. Although the valve stem did not 
fail, engineering evaluation indicated that further cracking would 
occur which could have prevented fulfillment of its safety function. As 
a result, the train was not considered capable of performing its 
specified safety function and the valve stem was replaced with a new 
one.
    The root cause was determined to be environmentally assisted stress 
corrosion cracking which resulted from installation of an inadequate 
material some years earlier. The same inadequate material had been 
installed in a similar valve in a different system at the same time. 
The similar valve was exposed to similar environmental conditions as 
the first valve.
    The condition is reportable under this new criterion if engineering 
judgment indicates that there was a reasonable expectation of 
preventing fulfillment of the safety function of both affected trains. 
This depends on details such as whether the second valve stem was also 
significantly degraded and, if not, whether any future degradation of 
the second valve stem would have been discovered and corrected, as a 
result of routine maintenance programs, before it could become 
problematic.
    Additional examples may be found in event reporting guidelines in 
NUREG-1022, Revision 2.
    Condition not covered by the plant's operating and emergency 
procedures [former section 50.72(b)(1)(ii)(C); and former section 
50.73(a)(2)(ii)(C)]. This criterion is deleted because it does not 
result in worthwhile reports aside from those that are captured by 
other reporting criteria such as:
    (1) An unanalyzed condition that significantly degrades plant 
safety;
    (2) An event or condition that could have prevented the fulfillment 
of the safety function of structures or systems that are needed to: 
shutdown the reactor and maintain it in a safe shutdown condition; 
remove residual heat; control the release of radioactive material; or 
mitigate the consequences of an accident;
    (3) An event or condition that results in the condition of the 
nuclear power plant, including its principal safety barriers, being 
seriously degraded;
    (4) An operation or condition prohibited by the plant's TS;
    (5) An event or condition that results in actuation of an ESF;
    (6) An event that poses an actual threat to the safety of the 
nuclear power plant or significantly hampers site personnel in the 
performance of duties necessary for the safe operation of the nuclear 
power plant.
    External event that poses an actual threat or significantly hampers 
personnel [former section 50.72(b)(1)(iii), deleted; and section 
50.73(a)(2)(iii)]. This criterion requires reporting a natural 
phenomenon or other external event that poses an actual threat to plant 
safety, or significantly hampers site personnel in the performance of 
duties necessary for safe operation. Section 50.72(b)(1)(iii) is 
deleted because it is redundant to section 50.72(a)(1)(i). That is, 
events of this type are captured by declaration of an Emergency Class, 
which is reportable within 1 hour. Section 50.73(a)(2)(iii) is retained 
in order to ensure submittal of an LER. This provision is not redundant 
because there is no criterion in section 50.73 that generally requires 
an LER for declaration of an Emergency Class.
    System actuation [old sections 50.72(b)(1)(iv) and (b)(2)(ii), 
replaced by new sections 50.72(b)(2)(iv)(A), (b)(2)(iv)(B), and 
(b)(3)(iv); and section 50.73(a)(2)(iv)]. Previously, sections 
50.72(b)(1)(iv) and (b)(2)(ii) provided the following distinction: an 
event that results or should have resulted in ECCS discharge into the 
reactor coolant system as a result of a valid signal was reportable 
within 1 hour; any other engineered safety feature (ESF) actuation, 
including reactor protection system (RPS) actuation, was reportable 
within 4 hours. The new 10 CFR 50.72(b)(2)(iv)(A) requires reporting an 
event that results or should have resulted in ECCS discharge into the 
reactor coolant system as a result of a valid signal within 4 hours. 
The new section 50.72(b)(2)(iv)(B) requires reporting a reactor scram 
during critical operation within 4 hours. The new section 
50.72(b)(3)(iv) requires reporting other ESF actuations within 8 hours. 
See the response to Comment D, above, for further discussion.
    The new section 50.72(b)(2)(iv) eliminates telephone reporting for 
invalid actuations, except for actuation of the RPS when the reactor is 
critical. These events are not significant and thus telephone reporting 
is not needed. The final amendments do not eliminate

[[Page 63783]]

the requirement for reporting of an invalid actuation under 10 CFR 
50.73. There is still a need for reporting of these events because they 
are used in making estimates of equipment reliability parameters, which 
in turn are needed to support the Commission's move towards risk-
informed regulation. However, for an invalid actuation reported under 
section 50.73(a)(2)(iv), other than actuation of the RPS when the 
reactor is critical, section 50.73(a)(1) provides the option of making 
a telephone report to the NRC Operations Center within 60 days instead 
of submitting a written LER. The telephone report is far less 
burdensome. Sixty days is an appropriate time because the information 
is not needed immediately. (See the response to Comment E above for 
further discussion of this need.)
    Previously, the rules generally required reporting the actuation of 
any ESF including the RPS. The final rule, instead, generally requires 
reporting for actuation of specific listed systems. These systems are:
    (1) Reactor protection system (RPS) including: reactor scram or 
reactor trip.
    (2) General containment isolation signals affecting containment 
isolation valves in more than one system or multiple main steam 
isolation valves (MSIVs).
    (3) Emergency core cooling systems (ECCS) for pressurized water 
reactors (PWRs) including: high-head, intermediate-head, and low-head 
injection systems and the low pressure injection function of residual 
(decay) heat removal systems.
    (4) ECCS for boiling water reactors (BWRs) including: high-pressure 
and low-pressure core spray systems; high-pressure coolant injection 
system; low pressure injection function of the residual heat removal 
system.
    (5) BWR reactor core isolation cooling system; isolation condenser 
system; and feedwater coolant injection system.
    (6) PWR auxiliary or emergency feedwater system.
    (7) Containment heat removal and depressurization systems, 
including containment spray and fan cooler systems.
    (8) Emergency ac electrical power systems, including: emergency 
diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs 
at the Oconee Station; and BWR dedicated Division 3 EDGs.
    (9) Emergency service water (ESW) systems that do not normally run 
and that serve as ultimate heat sinks. ESW system actuations are 
reportable under section 50.73 only.
    This approach provides for consistent reporting of actuations for 
these highly risk-significant systems. At the same time, it eliminates 
reporting of actuations for systems of lesser risk-significance, such 
as actuation of ventilation systems that are considered to be ESFs.
    Section 50.72 excludes reporting for an actuation that resulted 
from and was part of a pre-planned sequence during testing or reactor 
operation. It further excludes reporting of an invalid actuation, 
except for a reactor scram or reactor trip when the reactor is 
critical.
    A valid actuation is one that results from either a ``valid 
signal'' or an intentional manual initiation. A ``valid signal'' is one 
that results from actual plant conditions or parameters satisfying the 
requirements for system actuation. An invalid actuation is one that 
does not meet the criteria for being valid.
    Section 50.73 also excludes reporting for an actuation that 
resulted from and was part of a pre-planned sequence during testing or 
reactor operation. It further excludes reporting of an invalid 
actuation that occurred when the system was properly removed from 
service or an invalid actuation that occurred after the safety function 
had been already completed.
    For those invalid actuations which are reportable under section 
50.73, a licensee may provide a telephone notification within 60 days, 
rather than submitting an LER. This option to provide a telephone 
notification rather than an LER does not apply, however, to a reactor 
scram or reactor trip that occurs while the reactor is critical.
    Event or condition that could have prevented the fulfillment of the 
safety function of structures or systems that are needed to: shut down 
the reactor and maintain it in a safe shutdown condition; remove 
residual heat; control the release of radioactive material; or mitigate 
the consequences of an accident [former section 50.72(b)(2)(iii), 
replaced by new sections 50.72(b)(3)(v) and (b)(3)(vi); and sections 
50.73(a)(2)(v) and (a)(2)(vi)]. The phrase ``event or condition that 
alone could have prevented the fulfillment of the safety function of 
structures or systems * * *'' is clarified by deleting the word 
``alone''. This clarifies the requirements by more clearly reflecting 
the principle that it is necessary to consider other existing plant 
conditions in determining the reportability of an event or condition 
under this criterion. For example, if one train of a two train system 
is incapable of performing its safety function for one reason, and the 
other train is incapable of performing its safety function for a 
different reason, the event is reportable.
    The term ``at the time of discovery'' is added to section 
50.72(b)(3)(v) to eliminate telephone notification for a condition that 
no longer exists or no longer has an effect on required safety 
functions. For example, it might be discovered that at some time in the 
past both trains of a two train system were incapable of performing 
their safety function, but the condition was subsequently corrected and 
no longer exists. In another example, while the plant is shutdown, it 
might be discovered that during a previous period of operation a system 
was incapable of performing its safety function, but the system is not 
currently required to be operable. These events are considered 
significant, and an LER is required, but there is no need for telephone 
notification.
    A new paragraph, section 50.72(b)(3)(vi) is added to clarify 
section 50.72. The new paragraph explicitly states that telephone 
reporting is not required under section 50.72(b)(2)(v) for single 
failures if redundant equipment in the same system was operable and 
available to perform the required safety function. That is, although 
one train of a system may be incapable of performing its safety 
function, reporting is not required under this criterion if that system 
is still capable of performing the safety function. This is the same 
principle that was and continues to be stated explicitly in section 
50.73(a)(2)(vi) with regard to written LERs.
    Airborne radioactive release or liquid effluent release [former 
section 50.72(b)(2)(iv), deleted; and section 50.73(a)(2)(viii), 
retained; and former section 50.73(a)(2)(ix), deleted]. These criteria 
require reporting releases of radioactive material at a very low level 
because, for a power reactor, such a release would indicate a breakdown 
in the licensee's programs to control releases--not because of the 
impact of such a release.
    Section 50.72(b)(2)(iv) is deleted because immediate notification 
is not needed for releases at such a low level. Declaration of an 
Emergency Class, which occurs at a somewhat higher (but still low) 
level, captures releases that are large enough to warrant immediate 
notification. Declaration of an Emergency Class is reportable within 1 
hour under section 50.72(a)(1)(i).
    Section 50.73(a)(2)(viii) is retained in order to ensure submittal 
of an LER. Even if the release is very small, the NRC needs to review 
the event and consider whether action is needed to ensure the cause is 
addressed at other plants as appropriate. There is no criterion in 
section 50.73 that generally

[[Page 63784]]

requires an LER for declaration of an Emergency Class.
    Section 50.73(a)(2)(vix) is deleted because it is not correct. It 
indicated that reporting under section 50.73(a)(2)(viii) satisfied the 
requirements of section 20.2203(a)(3). However, some events captured by 
section 20.2203(a)(3) are not captured by section 50.73(a)(2)(viii).
    Internal event that poses an actual threat or significantly hampers 
personnel [former section 50.72(b)(1)(vi), deleted; and section 
50.73(a)(2)(x)]. This criterion requires reporting an internal event 
that poses an actual threat to plant safety, or significantly hampers 
site personnel in the performance of duties necessary for safe 
operation, including fires, toxic gas releases, or radioactive 
releases. Section 50.72(b)(1)(vi) is deleted because it is redundant to 
section 50.72(a)(1)(i). That is, events of this type are captured by 
declaration of an Emergency Class, which is reportable within 1 hour. 
Section 50.73(a)(2)(x) is retained in order to ensure submittal of an 
LER. This provision is not redundant because there is no criterion in 
section 50.73 that generally requires an LER for declaration of an 
Emergency Class.
    Major loss of emergency assessment capability, offsite response 
capability, or communication capability [former section 50.72(b)(2)(v), 
replaced by new section 50.72(b)(3)(xii)]. The new section is modified 
by adding the word ``offsite'' in front of the term ``communications 
capability'' to make it clear that the requirement does not apply to 
internal plant communication systems.
    Contents of LERs [section 50.73(b)(2)(ii)(F)]. Paragraph (F) is 
revised to correct the address of the NRC Library.
    Spent fuel storage cask problems [former sections 50.72(b)(2)(vii) 
and 72.216(a)(1), (a)(2), and (b)]. The provisions of section 
50.72(b)(2)(vii) are deleted because these reporting criteria are 
redundant to the reporting criteria contained in sections 72.216(a)(1) 
and (a)(2). Repetition of the same reporting criteria in different 
sections of the rules added unnecessary complexity and was inconsistent 
with the current practice in other areas, such as reporting of 
safeguards events as required by section 73.71.
    Sections 72.216(a)(1) and (a)(2) place upon general licensees the 
same reporting criteria as are placed on specific licensees under 
sections 72.75(b)(2) and (b)(3). To avoid duplication in Part 72, 
sections 72.216(a)(1) and (a)(2) are deleted and section 72.216(c) is 
abridged to simply require that the general licensee shall make initial 
and written reports in accordance with sections 72.74 and 72.75. These 
changes eliminate a reference in section 72.216(a) to section 
50.72(b)(2)(vii), now deleted, which had established the time limit for 
initial notification by general licensees. The same time limit is 
placed on general licensees by including them within the scope of 
section 72.75(b). Section 72.216(b) is also deleted because its 
requirements for a written report are encompassed by section 
72.75(d)(2).
    Exemptions [section 50.73(f)]. The provisions of this section are 
deleted because the exemption provisions in section 50.12 provide for 
granting of exemptions when they are warranted. Including another, 
section-specific exemption provision in section 50.73 adds unnecessary 
complexity to the rules.

3. Revisions to Event Reporting Guidelines in NUREG-1022

    A report, NUREG-1022, Revision 2, ``Event Reporting Guidelines, 10 
CFR 50.72 and 50.73,'' is being made available concurrently with the 
final amendments to 10 CFR 50.72 and 50.73. The report is available for 
inspection in the NRC Public Document Room or it may be viewed and 
downloaded electronically via the interactive rulemaking web site 
established by the NRC for this rulemaking, as discussed above under 
the heading ADDRESSES. Single copies may be obtained from the contact 
listed above under the heading For Further Information Contact. In the 
report, guidance that is considered to be new or different in a 
meaningful way, relative to that provided in NUREG-1022, Revision 1, is 
indicated by underlining the appropriate text.

4. Reactor Oversight

    The NRC is implementing revisions to the process for oversight of 
operating reactors, including inspection, assessment, and enforcement 
processes. In connection with this effort, the NRC has considered the 
kinds of event reports that would be eliminated by the final rules and 
concluded that the changes are consistent with the oversight process.
    In connection with the proposed rule, public comment was invited on 
whether or not this is the case. In particular, it was requested that 
if any examples to the contrary are known they be identified. None were 
identified.

5. Enforcement

    The NRC intends to modify its existing enforcement policy in 
connection with the final amendments to sections 50.72 and 50.73. The 
philosophy of the changes is to base the significance of the reporting 
violation on the impact on the NRC's ability to provide proper 
oversight of licensee activities. For example, a late report may impact 
the ability of the NRC to fulfill its obligations of fully 
understanding issues that are required to be reported in order to 
accomplish its public health and safety mission, which in many cases 
involves reacting to reportable issues or events. As such, the NRC 
intends to revise the Enforcement Policy, NUREG-1600 \3\ as follows:
---------------------------------------------------------------------------

    \3\ The examples refer to those published in NUREG-1600, 
``General Statement of Policy and Procedure for NRC Enforcement 
Actions,'' dated May 1, 2000.
---------------------------------------------------------------------------

    (1) Supplement I.C--Examples of Severity Level III violations.
    (a) Example 11 will be revised to read as follows--A failure to 
provide the required 1-hour telephone notification of an emergency 
action taken pursuant to 10 CFR 50.54(x).
    (b) An additional example will be added that will read as follows--
A failure to provide a required 1-hour, 4-hour or 8-hour non-emergency 
telephone notification pursuant to 10 CFR 50.72, that substantially 
impacts agency response.
    (c) An additional example will be added that will read as follows--
A late 4-hour or 8-hour notification that substantially impacts agency 
response.
    (2) Supplement I.D--Examples of Severity Level IV violations.
    (a) Example 4, will be revised to read as follows--A failure to 
provide a required 60-day written LER pursuant to 10 CFR 50.73.
    These changes in the Enforcement Policy will be consistent with the 
overall objective of the rule change of better aligning the reporting 
requirements with the NRC's reporting needs. The Enforcement Policy 
changes will correlate the Severity Level of the infractions with the 
relative importance of the information needed by the NRC.
    Section IV.A.3 of the Enforcement Policy provides that the Severity 
Level of an untimely report may be reduced depending on the individual 
circumstances. In deciding whether the Severity Level should be reduced 
for an untimely 1-hour, 4-hour, or 8-hour non-emergency report, the 
impact that the failure to report had on any agency response will be 
considered. For example, if a delayed 8-hour reportable event impacted 
the timing of a followup inspection that was deemed necessary, then the 
Severity Level will not normally be reduced. Similarly, a late 
notification that delayed the NRC's

[[Page 63785]]

ability to perform an engineering analysis of a condition to determine 
if additional regulatory action was necessary will generally not be 
considered for disposition at a reduced Severity Level.

6. Electronic Reporting

    The NRC is currently in the process of implementing an electronic 
document management and reporting program, known as the Agency Wide 
Document Access and Management System (ADAMS) that will provide for 
electronic submittal of many types of reports, including LERs. 
Accordingly, no separate rulemaking effort to provide for electronic 
submittal of LERs is necessary.

7. State Input

    Many States (Agreement States and Non-Agreement States) have 
agreements with power reactors to inform the States of plant issues. 
State reporting requirements are frequently triggered by NRC reporting 
requirements. Accordingly, the NRC sought State comment on issues 
related to the proposed amendments via letters to State Liaison 
Officers as well as by a specific request in the Federal Register 
notice on the proposed rule. Comments on the proposed rule were 
received from one State agency, as discussed above under the heading 
``Comment D.''

8. Plain Language

    The President's Memorandum dated June 1, 1998, entitled, ``Plain 
Language in Government Writing,'' directed that the Federal 
Government's writing be in plain language. The NRC requested comments 
on the proposed rule specifically with respect to the clarity and 
effectiveness of the language used. A number of suggestions aimed at 
improving the clarity and effectiveness of the language used were 
received and incorporated into the final rule.

IV. Environmental Impact: Categorical Exclusion

    The NRC has determined that this proposed regulation is the type of 
action described in categorical exclusion 10 CFR 51.22(c)(3)(iii). 
Therefore, neither an environmental impact statement nor an 
environmental assessment has been prepared for this proposed 
regulation.

V. Backfit Analysis

    The NRC has determined that the backfit rule, 10 CFR 50.109, does 
not apply to information collection and reporting requirements such as 
those contained in the final rule because they do not impose backfits 
as defined in 10 CFR 50.109(a)(1). Therefore, a backfit analysis has 
not been prepared. However, as discussed below, the NRC has prepared a 
regulatory analysis for the proposed rule, which examines the costs and 
benefits of the proposed requirements in this rule. The Commission 
regards the regulatory analysis as a disciplined process for assessing 
information collection and reporting requirements to determine that the 
burden imposed is justified in light of the potential safety 
significance of the information to be collected.

VI. Regulatory Analysis

    The NRC prepared a draft regulatory analysis for the proposed rule 
to examine the costs and benefits of the alternatives considered by the 
NRC, and public comments on this analysis were requested in connection 
with the proposed rule. As discussed above under the heading ``Comment 
T,'' some commenters disagreed with the proposition that the rule would 
reduce reporting burden. These comments were addressed by incorporating 
changes into the final rule, such that the assumptions in the draft 
regulatory analysis are sustained, and no changes have been made to the 
regulatory analysis. The regulatory analysis is available for 
inspection in the NRC Public Document Room or it may be viewed and 
downloaded electronically via the interactive rulemaking web site 
established by NRC for this rulemaking, as discussed above under the 
heading ADDRESSES. Single copies may be obtained from the contact 
listed above under the heading FOR FURTHER INFORMATION CONTACT.

VII. Paperwork Reduction Act Statement

    This final rule amends information collection requirements that are 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). This rule has been reviewed and approved by the Office of 
Management and Budget, approval numbers 3150-0011 and 3150-0104.
    The annual public reporting burden for the currently existing 
reporting requirements in 10 CFR 50.72 and 50.73 is estimated to 
average about 700 hours per nuclear power reactor, including the time 
for reviewing instructions, searching existing data sources, gathering 
and maintaining the data needed, and completing and reviewing the 
information collection. It is estimated that the proposed amendments 
would impose a one-time implementation burden of about 200 hours per 
reactor. The recurring annual information collection burden is 
estimated to be reduced by 132 hours per reactor.
    Send comments on any aspect of this information collection, 
including suggestions for reducing this burden, to the Records 
Management Branch (T-6E6), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001 or by Internet electronic mail to 
<a href="/cdn-cgi/l/email-protection#d5979f86e4959b8796fb929a83"><span class="__cf_email__" data-cfemail="094b435a3849475b4a274e465f">[email protected]</span></a>; and to the Desk Officer, Office of Information and 
Regulatory Affairs, NEOB-10202, (3150-0011 AND 3150-0104); Office of 
Management and Budget, Washington, DC 20503.

Public Protection Notification

    If a means used to impose an information collection does not 
display a currently valid OMB control number, the NRC may not conduct 
or sponsor, and a person is not required to respond to, an information 
collection.

VIII. Regulatory Flexibility Act 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 proposed 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).

IX. 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.

X. National Technology Transfer and Advancement Act

    The National Technology Transfer and Advancement Act of 1995, Pub. 
L. 104-113, requires that Federal agencies use technical standards 
developed or adopted by voluntary consensus standards bodies unless the 
use of such a standard is inconsistent with applicable law or otherwise 
impractical. There are no consensus standards regarding the reporting 
of safety information by nuclear power plant licensees to regulatory 
authorities that would apply to the requirements imposed by this rule. 
Thus, the provisions of the Act do not apply to this rule.

[[Page 63786]]

XI. Final Amendments

List of Subjects

10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
prevention, Intergovernmental relations, Nuclear power plants and 
reactors, Radiation protection, Reactor siting criteria, Reporting and 
recordkeeping requirements.

10 CFR Part 72

    Criminal penalties, Manpower training programs, Nuclear materials, 
Occupational safety and health, Reporting and recordkeeping 
requirements, Security measures, Spent fuel.

    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. 553, the NRC is adopting the 
following amendments to 10 CFR Part 50 and 10 CFR Part 72.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    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, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 
68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. L. 91-
190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(D.D.), and 
50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 
2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 
185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and 
Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 
U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 
Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also 
issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 
50.78 also issued under sec. 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).

    2. Section 50.72 is amended by revising paragraphs (a) and (b) to 
read as follows:


Sec. 50.72  Immediate notification requirements for operating nuclear 
power reactors.

    (a) General requirements.\1\ (1) Each nuclear power reactor 
licensee licensed under Sec. 50.21(b) or Sec. 50.22 of this part shall 
notify the NRC Operations Center via the Emergency Notification System 
of:
---------------------------------------------------------------------------

    \1\ Other requirements for immediate notification of the NRC by 
licensed operating nuclear power rectors are contained elsewhere in 
this chapter, in particular Secs. 20.1906, 20.2202, 50.36, 72.216, 
and 73.71.
---------------------------------------------------------------------------

    (i) The declaration of any of the Emergency Classes specified in 
the licensee's approved Emergency Plan; \2\ or
---------------------------------------------------------------------------

    \2\ These Emergency Classes are addressed in Appendix E of this 
part.
---------------------------------------------------------------------------

    (ii) Those non-emergency events specified in paragraph (b) of this 
section that occurred within three years of the date of discovery.
    (2) If the Emergency Notification System is inoperative, the 
licensee shall make the required notifications via commercial telephone 
service, other dedicated telephone system, or any other method which 
will ensure that a report is made as soon as practical to the NRC 
Operations Center.\3\
---------------------------------------------------------------------------

    \3\ Commercial telephone number of the NRC Operations Center is 
(301) 816-5100.
---------------------------------------------------------------------------

    (3) The licensee shall notify the NRC immediately after 
notification of the appropriate State or local agencies and not later 
than one hour after the time the licensee declares one of the Emergency 
Classes.
    (4) The licensee shall activate the Emergency Response Data System 
(ERDS) \4\ as soon as possible but not later than one hour after 
declaring an Emergency Class of alert, site area emergency, or general 
emergency. The ERDS may also be activated by the licensee during 
emergency drills or exercises if the licensee's computer system has the 
capability to transmit the exercise data.
---------------------------------------------------------------------------

    \4\ Requirements for ERDS are addressed in Appendix E, Section 
VI.
---------------------------------------------------------------------------

    (5) When making a report under paragraph (a)(1) of this section, 
the licensee shall identify:
    (i) The Emergency Class declared; or
    (ii) Paragraph (b)(1), ``One-hour reports,'' paragraph (b)(2), 
``Four-hour reports,'' or paragraph (b)(3), ``Eight-hour reports,'' as 
the paragraph of this section requiring notification of the non-
emergency event.
    (b) Non-emergency events--(1) One-hour reports. If not reported as 
a declaration of an Emergency Class under paragraph (a) of this 
section, the licensee shall notify the NRC as soon as practical and in 
all cases within one hour of the occurrence of any deviation from the 
plant's Technical Specifications authorized pursuant to Sec. 50.54(x) 
of this part.
    (2) Four-hour reports. If not reported under paragraphs (a) or 
(b)(1) of this section, the licensee shall notify the NRC as soon as 
practical and in all cases, within four hours of the occurrence of any 
of the following:
    (i) The initiation of any nuclear plant shutdown required by the 
plant's Technical Specifications.
    (ii)-(iii) [Reserved]
    (iv)(A) Any event that results or should have resulted in emergency 
core cooling system (ECCS) discharge into the reactor coolant system as 
a result of a valid signal except when the actuation results from and 
is part of a pre-planned sequence during testing or reactor operation.
    (B) Any event or condition that results in actuation of the reactor 
protection system (RPS) when the reactor is critical except when the 
actuation results from and is part of a pre-planned sequence during 
testing or reactor operation.
    (v)-(x) [Reserved]
    (xi) Any event or situation, related to the health and safety of 
the public or onsite personnel, or protection of the environment, for 
which a news release is planned or notification to other government 
agencies has been or will be made. Such an event may include an onsite 
fatality or inadvertent release of radioactively contaminated 
materials.
    (3) Eight-hour reports. If not reported under paragraphs (a), 
(b)(1) or (b)(2) of this section, the licensee shall notify the NRC as 
soon as practical and in all cases within eight hours of the occurrence 
of any of the following:
    (i) [Reserved]
    (ii) Any event or condition that results in:
    (A) The condition of the nuclear power plant, including its 
principal safety barriers, being seriously degraded; or
    (B) The nuclear power plant being in an unanalyzed condition that 
significantly degrades plant safety.
    (iii) [Reserved]
    (iv)(A) Any event or condition that results in valid actuation of 
any of the systems listed in paragraph (b)(3)(iv)(B) of this section, 
except when the actuation results from and is part of a pre-planned 
sequence during testing or reactor operation.
    (B) The systems to which the requirements of paragraph 
(b)(3)(iv)(A) of this section apply are:

[[Page 63787]]

    (1) Reactor protection system (RPS) including: Reactor scram and 
reactor trip. \5\
---------------------------------------------------------------------------

    \5\ Actuation of the RPS when the reactor is critical is 
reportable under paragraph (b)(2)(iv)(B) of this section.
---------------------------------------------------------------------------

    (2) General containment isolation signals affecting containment 
isolation valves in more than one system or multiple main steam 
isolation valves (MSIVs).
    (3) Emergency core cooling systems (ECCS) for pressurized water 
reactors (PWRs) including: High-head, intermediate-head, and low-head 
injection systems and the low pressure injection function of residual 
(decay) heat removal systems.
    (4) ECCS for boiling water reactors (BWRs) including: High-pressure 
and low-pressure core spray systems; high-pressure coolant injection 
system; low pressure injection function of the residual heat removal 
system.
    (5) BWR reactor core isolation cooling system; isolation condenser 
system; and feedwater coolant injection system.
    (6) PWR auxiliary or emergency feedwater system.
    (7) Containment heat removal and depressurization systems, 
including containment spray and fan cooler systems.
    (8) Emergency ac electrical power systems, including: Emergency 
diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs 
at the Oconee Station; and BWR dedicated Division 3 EDGs.
    (v) Any event or condition that at the time of discovery could have 
prevented the fulfillment of the safety function of structures or 
systems that are needed to:
    (A) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (B) Remove residual heat;
    (C) Control the release of radioactive material; or
    (D) Mitigate the consequences of an accident.
    (vi) Events covered in paragraph (b)(3)(v) of this section may 
include one or more procedural errors, equipment failures, and/or 
discovery of design, analysis, fabrication, construction, and/or 
procedural inadequacies. However, individual component failures need 
not be reported pursuant to paragraph (b)(3)(v) of this section if 
redundant equipment in the same system was operable and available to 
perform the required safety function.
    (vii)-(xi) [Reserved]
    (xii) Any event requiring the transport of a radioactively 
contaminated person to an offsite medical facility for treatment.
    (xiii) Any event that results in a major loss of emergency 
assessment capability, offsite response capability, or offsite 
communications capability (e.g., significant portion of control room 
indication, Emergency Notification System, or offsite notification 
system).
* * * * *

    3. Section 50.73 is amended by revising sections (a), 
(b)(2)(ii)(F), (b)(2)(ii)(J), (b)(3), (d), and (e) and by removing and 
reserving paragraph (f) to read as follows:


Sec. 50.73  Licensee event report system.

    (a) Reportable events. (1) The holder of an operating license for a 
nuclear power plant (licensee) shall submit a Licensee Event Report 
(LER) for any event of the type described in this paragraph within 60 
days after the discovery of the event. In the case of an invalid 
actuation reported under Sec. 50.73(a)(2)(iv), other than actuation of 
the reactor protection system (RPS) when the reactor is critical, the 
licensee may, at its option, provide a telephone notification to the 
NRC Operations Center within 60 days after discovery of the event 
instead of submitting a written LER. Unless otherwise specified in this 
section, the licensee shall report an event if it occurred within three 
years of the date of discovery regardless of the plant mode or power 
level, and regardless of the significance of the structure, system, or 
component that initiated the event.
    (2) The licensee shall report:
    (i)(A) The completion of any nuclear plant shutdown required by the 
plant's Technical Specifications.
    (B) Any operation or condition which was prohibited by the plant's 
Technical Specifications except when:
    (1) The Technical Specification is administrative in nature;
    (2) The event consisted solely of a case of a late surveillance 
test where the oversight was corrected, the test was performed, and the 
equipment was found to be capable of performing its specified safety 
functions; or
    (3) The Technical Specification was revised prior to discovery of 
the event such that the operation or condition was no longer prohibited 
at the time of discovery of the event.
    (C) Any deviation from the plant's Technical Specifications 
authorized pursuant to Sec. 50.54(x) of this part.
    (ii) Any event or condition that resulted in:
    (A) The condition of the nuclear power plant, including its 
principal safety barriers, being seriously degraded; or
    (B) The nuclear power plant being in an unanalyzed condition that 
significantly degraded plant safety.
    (iii) Any natural phenomenon or other external condition that posed 
an actual threat to the safety of the nuclear power plant or 
significantly hampered site personnel in the performance of duties 
necessary for the safe operation of the nuclear power plant.
    (iv)(A) Any event or condition that resulted in manual or automatic 
actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of 
this section, except when:
    (1) The actuation resulted from and was part of a pre-planned 
sequence during testing or reactor operation; or
    (2) The actuation was invalid and;
    (i) Occurred while the system was properly removed from service; or
    (ii) Occurred after the safety function had been already completed.
    (B) The systems to which the requirements of paragraph 
(a)(2)(iv)(A) of this section apply are:
    (1) Reactor protection system (RPS) including: reactor scram or 
reactor trip.
    (2) General containment isolation signals affecting containment 
isolation valves in more than one system or multiple main steam 
isolation valves (MSIVs).
    (3) Emergency core cooling systems (ECCS) for pressurized water 
reactors (PWRs) including: high-head, intermediate-head, and low-head 
injection systems and the low pressure injection function of residual 
(decay) heat removal systems.
    (4) ECCS for boiling water reactors (BWRs) including: high-pressure 
and low-pressure core spray systems; high-pressure coolant injection 
system; low pressure injection function of the residual heat removal 
system.
    (5) BWR reactor core isolation cooling system; isolation condenser 
system; and feedwater coolant injection system.
    (6) PWR auxiliary or emergency feedwater system.
    (7) Containment heat removal and depressurization systems, 
including containment spray and fan cooler systems.
    (8) Emergency ac electrical power systems, including: emergency 
diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs 
at the Oconee Station; and BWR dedicated Division 3 EDGs.
    (9) Emergency service water systems that do not normally run and 
that serve as ultimate heat sinks.
    (v) Any event or condition that could have prevented the 
fulfillment of the safety function of structures or systems that are 
needed to:

[[Page 63788]]

    (A) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (B) Remove residual heat;
    (C) Control the release of radioactive material; or
    (D) Mitigate the consequences of an accident.
    (vi) Events covered in paragraph (a)(2)(v) of this section may 
include one or more procedural errors, equipment failures, and/or 
discovery of design, analysis, fabrication, construction, and/or 
procedural inadequacies. However, individual component failures need 
not be reported pursuant to paragraph (a)(2)(v) of this section if 
redundant equipment in the same system was operable and available to 
perform the required safety function.
    (vii) Any event where a single cause or condition caused at least 
one independent train or channel to become inoperable in multiple 
systems or two independent trains or channels to become inoperable in a 
single system designed to:
    (A) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (B) Remove residual heat;
    (C) Control the release of radioactive material; or
    (D) Mitigate the consequences of an accident.
    (viii)(A) Any airborne radioactive release that, when averaged over 
a time period of 1 hour, resulted in airborne radionuclide 
concentrations in an unrestricted area that exceeded 20 times the 
applicable concentration limits specified in appendix B to part 20, 
table 2, column 1.
    (B) Any liquid effluent release that, when averaged over a time 
period of 1 hour, exceeds 20 times the applicable concentrations 
specified in appendix B to part 20, table 2, column 2, at the point of 
entry into the receiving waters (i.e., unrestricted area) for all 
radionuclides except tritium and dissolved noble gases.
    (ix)(A) Any event or condition that as a result of a single cause 
could have prevented the fulfillment of a safety function for two or 
more trains or channels in different systems that are needed to:
    (1) Shut down the reactor and maintain it in a safe shutdown 
condition;
    (2) Remove residual heat;
    (3) Control the release of radioactive material; or
    (4) Mitigate the consequences of an accident.
    (B) Events covered in paragraph (a)(2)(ix)(A) of this section may 
include cases of procedural error, equipment failure, and/or discovery 
of a design, analysis, fabrication, construction, and/or procedural 
inadequacy. However, licensees are not required to report an event 
pursuant to paragraph (a)(2)(ix)(A) of this section if the event 
results from:
    (1) A shared dependency among trains or channels that is a natural 
or expected consequence of the approved plant design; or
    (2) Normal and expected wear or degradation.
    (x) Any event that posed an actual threat to the safety of the 
nuclear power plant or significantly hampered site personnel in the 
performance of duties necessary for the safe operation of the nuclear 
power plant including fires, toxic gas releases, or radioactive 
releases.
    (b) * * * 
    (2) * * * 
    (ii) * * * 
    (F) The Energy Industry Identification System component function 
identifier and system name of each component or system referred to in 
the LER.
    (1) The Energy Industry Identification System is defined in: IEEE 
Std 803-1983 (May 16, 1983) Recommended Practice for Unique 
Identification in Power Plants and Related Facilities--Principles and 
Definitions.
    (2) IEEE Std 803-1983 has been approved for incorporation by 
reference by the Director of the Federal Register in accordance with 5 
U.S.C. 552(a) and 1 CFR part 51.
    (3) A notice of any changes made to the material incorporated by 
reference will be published in the Federal Register. Copies may be 
obtained from the Institute of Electrical and Electronics Engineers, 
445 Hoes Lane, P.O. Box 1331, Piscataway, NJ 08855-1331. IEEE Std 803-
1983 is available for inspection at the NRC's Technical Library, which 
is located in the Two White Flint North Building, 11545 Rockville Pike, 
Rockville, Maryland 20852-2738; and at the Office of the Federal 
Register, 800 North Capitol Street, Suite 700, NW, Washington, DC 2001.
* * * * *
    (J) For each human performance related root cause, the licensee 
shall discuss the cause(s) and circumstances.
* * * * *
    (3) An assessment of the safety consequences and implications of 
the event. This assessment must include:
    (i) The availability of systems or components that could have 
performed the same function as the components and systems that failed 
during the event, and
    (ii) For events that occurred when the reactor was shutdown, the 
availability of systems or components that are needed to shutdown the 
reactor and maintain safe shutdown conditions, remove residual heat, 
control the release of radioactive material, or mitigate the 
consequences of an accident.
* * * * *
    (d) Submission of reports. Licensee Event Reports must be prepared 
on Form NRC 366 and submitted to the U.S. Nuclear Regulatory 
Commission, as specified in Sec. 50.4.
    (e) Report legibility. The reports and copies that licensees are 
required to submit to the Commission under the provisions of this 
section must be of sufficient quality to permit legible reproduction 
and micrographic processing.
    (f) [Reserved]
* * * * *

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

    4. The authority citation for part 72 continues to read as follows:

    Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 
184, 186, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 954, 955, as 
amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2071, 2073, 
2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 2234, 2236, 
2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 688, as 
amended (42 U.S.C. 5841, 5842, 5846); Pub. L. 95-601, sec. 10, 92 
Stat. 2951 as amended by Pub. L. 102-486, sec. 7902, 106 Stat. 3123 
(42 U.S.C. 5851); sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 
4332); secs. 131, 132, 133, 135, 137, 141, Pub. L. 97-425, 96 Stat. 
2229, 2230, 2232, 2241, sec. 148, Pub. L. 100-203, 101 Stat. 1330-
235 (42 U.S.C. 10151, 10152, 10153, 10155, 10157, 10161, 10168).

    Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), 
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 
10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat. 955 
(42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 U.S.C. 
10154). Section 72.96(d) also issued under sec. 145(g), Pub. L. 100-
203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also issued 
under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-425, 96 
Stat. 2202, 2203, 2204, 2222, 2224, (42 U.S.C. 10101, 10137(a), 
10161(h)). Subparts K and L are also issued under sec. 133, 98 Stat. 
2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 (42 U.S.C. 
10198).

    5. Section 72.216 is revised to read as follows:


Sec. 72.216  Reports.

    (a) [Reserved]

[[Page 63789]]

    (b) [Reserved]
    (c) The general licensee shall make initial and written reports in 
accordance with Secs. 72.74 and 72.75.

    Dated at Rockville, Maryland, this 18th day of October, 2000.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook
Secretary of the Commission.
[FR Doc. 00-27283 Filed 10-24-00; 8:45 am]
BILLING CODE 7590-01-P


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