[Federal Register Volume 64, Number 128 (Tuesday, July 6, 1999)]
[Proposed Rules]
[Pages 36291-36307]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-16934]
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Proposed Rules
Federal Register
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This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 64, No. 128 / Tuesday, July 6, 1999 /
Proposed Rules
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 50 and 72
RIN 3150-AF98
Reporting Requirements for Nuclear Power Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission is proposing to amend the
event reporting requirements for nuclear power reactors: to update the
current rules, including reducing or eliminating the reporting burden
associated with events of little or no safety significance; and to
better align the rules with the NRC's needs for information 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 it is needed for prompt NRC action. Also, a
draft report, NUREG-1022, Revision 2, is being made available for
public comment concurrently with the proposed amendments.
DATES: Submit comments on or before September 20, 1999. Comments
received after this date will be considered if it is practical to do
so, but the Commission is able to ensure consideration only for
comments received on or before this date.
ADDRESSES: Mail comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. ATTN: Rulemakings and
Adjudications Staff.
Deliver comments to: 11555 Rockville Pike, Rockville, Maryland,
between 7:30 a.m. and 4:15 p.m. Federal workdays.
Electronic comments may be provided via the NRC's interactive
rulemaking website through the NRC home page (http://www.nrc.gov). From
the home page, select ``Rulemaking'' from the tool bar at the bottom of
the page. The interactive rulemaking website can then be accessed by
selecting ``Rulemaking Forum.'' This site provides the ability to
upload comments as files (any format), if your web browser supports
that function. For information about the interactive rulemaking
website, contact Ms. Carol Gallagher, (301) 415-5905; e-mail
[email protected]
Certain documents related to this rulemaking, including comments
received, the transcripts of public meetings held, the draft regulatory
analysis and the draft report NUREG-1022, Revision 2 may be examined at
the NRC Public Document Room, 2120 L Street, NW, (Lower Level),
Washington, DC. These same documents also may be viewed and downloaded
electronically via the interactive rulemaking web site established by
NRC for this rulemaking.
FOR FURTHER INFORMATION CONTACT: Dennis P. Allison, Office of Nuclear
Reactor Regulation, Washington, DC 20555-0001, telephone (301) 415-
1178, e-mail dpa@nrc.gov.
SUPPLEMENTARY INFORMATION:
Contents
I. Background
II. Rulemaking Initiation
III. Analysis of Comments
IV. Discussion
1. Objectives of Proposed Amendments
2. Discussion of Proposed Amendments
3. Revisions to Reporting Guidelines in NUREG-1022
4. Reactor Oversight
5. Reporting of Historical Problems
6. Reporting of Component Problems
7. Enforcement
8. Electronic Reporting
9. Schedule
10. State Input
V. Environmental Impact: Categorical Exclusion
VI. Backfit Analysis
VII. Regulatory Analysis
VIII. Paperwork Reduction Act Statement
IX. Regulatory Flexibility Certification
X. Proposed Amendments
I. Background
Section 50.72 has been in effect, with minor modifications, since
1983. Its essential purpose 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 has also been in effect, with minor modification,
since 1983. Its essential purpose is to identify ``* * * 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).
II. Rulemaking Initiation
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
whether the NRC should proceed with rulemaking to modify the event
reporting requirements in 10 CFR 50.72, ``Immediate notification
requirements for operating nuclear power reactors,'' and 50.73,
``Licensee event report system,'' and several concrete proposals were
provided for comment.
A public meeting was held to discuss the ANPR at NRC Headquarters
on August 21, 1998. The ANPR was also discussed, along with other
topics, at a public meeting on the role of industry in nuclear
regulation in Rosemont, Illinois on September 1, 1998. The public
comment period on the ANPR closed on September 21, 1998. A comment from
the Nuclear Energy Institute (NEI) proposed conducting ``table top
exercises'' early in the development and review process to test key
parts of the requirements and guidance for clarity and consistency.
That comment was accepted and a third public meeting was held on
November 13, 1998 to discuss issues of clarity and consistency in the
contemplated approach. Transcripts of these meetings are available for
inspection in the NRC Public Document Room or they may be viewed and
downloaded electronically via the interactive rulemaking web site
established by NRC for this rulemaking, as discussed above under the
heading
[[Page 36292]]
ADDRESSES. Single copies may be obtained from the contact listed above
under the heading For Further Information Contact.
III. Analysis of Comments
The comment period for the ANPR expired September 21, 1998. Twenty-
one comment letters were received, representing comments from sixteen
nuclear power plant licensees (utilities), two organizations of
utilities, two States and one public interest group. A list of comment
letters is provided below. The comment letters expressed support for
amending the rules along the general lines of the objectives discussed
in the ANPR. Most of the letters also provided specific recommendations
for changes to the contemplated amendments discussed in the ANPR. In
addition to the written comments received, the ANPR has been the
subject of three public meetings as discussed above under the heading
BACKGROUND, and comments made at those meetings have also been
considered.
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 ANPR; or applicable to another area or activity outside the
scope of sections 50.72 and 50.73).
Comment 1: Several comments recommended amending 10 CFR 50.73 to
allow 60 days (instead of the current 30 days) for submittal of
Licensee Event Reports (LERs). They indicated that this would allow a
more reasonable time to determine the root causes of events and lead to
fewer amended reports.
Response: The comments are accepted for the reason stated above.
The proposed rule would change the time limit to 60 days.
Comment 2: Two comments suggested a need to establish starting
points for reporting time clocks that are clear and not subject to
varied interpretations.
Response: The reporting guidelines in this area have been reviewed
for clarity. Some editorial clarifications are proposed in section 2.5
of the draft of Revision 2 to NUREG-1022, which is being made available
for public comment concurrently with the proposed rule, as discussed
below under the heading ``Revisions to Reporting Guidelines in NUREG-
1022.''
Comment 3: Many comments opposed adopting a check the box approach
for human performance and other information in LERs (as was proposed in
the ANPR, with the objective of reducing reporting burden). They
indicated that adopting a check the box approach would result in
substantial implementation problems, and recommended continuing to rely
on the narrative description which provides adequate information. One
comment opposed the idea of a check the box approach on the grounds
that it would make LERs more difficult for the general public to
understand. A few comments supported the check the box approach.
Response: The intent of the check the box approach was to reduce
the effort required in reporting; however, the majority of comments
indicate this would not be the case. Accordingly, the proposed rule
does not reflect adoption of a check the box approach.
Comment 4: Several comments opposed codifying the current
guidelines for reporting human performance information in LERs (i.e.,
adding the detailed guidelines to the rule, as was proposed in the
ANPR). They recommended leaving the rule unchanged in this regard,
indicating that sufficient information is being provided under the
current rule and guidelines.
Response: The comments are partially accepted. The proposed rule
would not codify the reporting guidelines (as proposed in the ANPR) for
the reasons stated above.
However, the proposed rule would simplify the requirement. It is
not necessary to specify the level of detail provided in the current
rule. Accordingly, the amended paragraph would simply require a
discussion of the causes and circumstances for any human performance
related problems that contributed to the event. 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. This draft report is
being made available for public comment concurrently with the proposed
rule, as discussed below under the heading ``Revisions to Reporting
Guidelines in NUREG-1022.''
Comment 5: Several comments opposed codifying a list of specific
systems for which actuation must be reported (by naming the systems in
10 CFR 50.72 and 50.73, as was proposed in the ANPR). They indicated
that a system's contribution to risk can vary widely from plant to
plant, which precludes construction of a valid universal list. They
recommended that, instead, actuation be reported only for those systems
that are specified to be engineered safety features (ESFs) in the final
safety analysis report (FSAR).
Response: The proposed rule would include a list of systems for
which actuation would be reported. However, the concern is recognized
and public comment will be specifically invited on several alternatives
to the proposed rule.
Comment 6: Several comments opposed changing the criteria in 10 CFR
50.72 and 50.73 which require reporting any event or condition that
alone could have prevented the fulfillment of the safety function of
structures or systems * * *. The change proposed in the ANPR would have
substituted the phrase ``alone or in combination with other existing
conditions'' for the word ``alone'' in this criterion. The comments
indicated that this would add confusion, the rule as currently worded
is sufficiently clear, and the need to consider other existing plant
conditions in evaluating reportability is understood and uniformly
implemented. They recommended leaving the rule unchanged in this
regard.
Response: The comments are partially accepted. The requirement
would not be changed by substituting the phrase ``alone or in
combination with other existing conditions'' for the word ``alone'' in
this criterion (as proposed in the ANPR).
However, the proposed amendments would change the rules by deleting
the word ``alone,'' so that they would require reporting ``any event or
condition that could have prevented fulfillment of the safety function
of structures or systems * * *.'' This would simplify the wording,
rather than making it more complicated. It is not intended to change
the meaning of the requirement, but to make the meaning more apparent
in the wording of the rule. The following points, which are relevant to
this question, would continue to be made clear in the reporting
guidelines. See section 3.2.7 of the draft of Revision 2 to NUREG-1022,
which is being made available for public comment concurrently with the
proposed rule, as discussed below under the heading ``Revisions to
Reporting Guidelines in NUREG-1022.''
(1) It is not necessary to assume an additional random single
failure in evaluating reportability. (If such an assumption were
necessary, inoperability of a single train would generally be
reportable under this criterion.)
(2) It is necessary to consider other existing conditions in
determining reportability. (For example, if Train A fails at a time
when Train B is out of service for maintenance, the event is
reportable.)
(3) The event is reportable regardless of whether or not a system
was called upon to perform its safety function. (For example, if an
emergency core cooling system [ECCS] was incapable of performing its
specified safety
[[Page 36293]]
functions, the event is reportable even if there was no call for the
ECCS function.)
(4) The event is reportable regardless of whether or not a
different system was capable of performing the safety function. (For
example, if the onsite power system failed, the event is reportable
even if the offsite power system was available and capable of
performing its safety functions.)
Comment 7: Several comments recommended changing 10 CFR 50.72 and
50.73 to exclude reporting an invalid actuation of an ESF. (An invalid
actuation is one that does not result from a plant condition that
warrants ESF initiation.)
Response: The comments are partially accepted. The proposed
amendments would eliminate the requirement for telephone notification
of an invalid actuation under 10 CFR 50.72. Invalid actuations are
generally less significant than valid actuations because they do not
involve plant conditions (e.g., low reactor coolant system pressure)
conditions that would warrant system actuation. Instead, they result
from other causes such as a dropped electrical lead during testing).
However, the proposed amendments would not eliminate the
requirement for a written report of an invalid actuation under 10 CFR
50.73. There is still a need for reporting of invalid actuations
because they are needed to make estimates of equipment reliability
parameters, which in turn are needed to support the Commission's move
towards risk-informed regulation. This is discussed further in a May 7,
1997 Commission paper, SECY-97-101, ``Proposed Rule, 10 CFR 50.76,
Reporting Reliability and Availability Information for Risk-significant
Systems and Equipment,'' Attachment 3.
Comment 8: Several comments recommended changing 10 CFR 50.72 and
50.73 to limit certain reports to current events and conditions. That
is, they recommended that an event or condition that could have
prevented the fulfillment of the safety function of structures or
systems * * * be reported:
(1) By telephone under 10 CFR 50.72(b)(2)(iii) only if it currently
exists, and
(2) By written LER under 10 CFR 50.73(a)(2)(v) only if it existed
within the previous two years.
For a ``historical'' event or condition of this type (i.e., one
which might have been significant at one time but has since been
corrected) there is less significance than there is for a current event
and, thus, immediate notification under 50.72(b)(2)(iii) is not
warranted. With regard to 50.73(a)(2)(v), two years encompasses at
least one operating cycle. Considerable resources are expended when it
is necessary to search historical records older than this to make past
operability determinations, and this is not warranted by the lesser
significance of historical events older than two years.
Response: The comments are partially accepted, for the reasons
stated above. That is, under the proposed rules, an event or condition
that could have prevented the fulfillment of the safety function of
structures or systems * * * would be reported by telephone under 10 CFR
50.72(b)(2)(iii) only if it exists at the time of discovery. An event
or condition that could have prevented the fulfillment of the safety
function of structures or systems * * * would be reported by written
LER under 10 CFR 50.73(a)(2)(v) only if it existed within the previous
three years.
In addition, although not recommended in the comments, under the
proposed rule an operation or condition prohibited by the plant's
Technical Specifications would be reported under 50.73(a)(2)(i)(B) only
if it existed within the previous three years. For this criterion as
well, considerable resources are expended when it is necessary to
search historical records older than three years to make past
operability determinations, and this is not warranted by the lesser
significance of historical events older than three years.
Three years is proposed, rather than two years as suggested in the
comments, because the NRC staff trends plant performance indicators
over a period of three years to ensure inclusion of periods of both
shut down and operation.
Comment 9: Several comments opposed using the term risk-significant
(or significant) in the absence of a clear definition.
Response: The term ``significant'' would be used in two criteria in
the proposed rules. In the first criterion, sections 50.72 and 50.73
would require reporting an unanalyzed condition that significantly
affects plant safety. In this context the term ``significant'' would be
defined by examples, five of which are discussed below under the
heading ``Condition that is outside the design basis of the plant.'' In
the second criterion, section 50.73 would require reporting when a
component's ability to perform its safety function is significantly
degraded and the condition could reasonably be expected to affect other
similar components in the plant. Again, the term ``significant'' would
be defined by examples, six of which are discussed below under the
heading ``Significantly degraded components.''
Comment 10: Several comments recommended changing 10 CFR 50.72 and
50.73 to exclude reporting of an unanalyzed condition that
significantly compromised plant safety on the basis that it is
redundant to other reporting criteria.
Response: The comment is not accepted. Several types of worthwhile
reports have been identified that could not readily be captured by
other criteria as discussed further below under the heading ``Condition
that is outside the design basis of the plant.''
Comment 11: Several comments recommended amending 10 CFR 50.72 and
50.73 to exclude reporting of a seriously degraded principal safety
barrier on the basis that it is redundant to other reporting criteria.
Response: The comments are not accepted. This criterion captures
some worthwhile reports that would not be captured by other criteria,
such as significant welding or material defects in the primary coolant
system. However, some clarifications are proposed in Section 3.2.4 of
the draft reporting guidelines, to better indicate which events are
serious enough to qualify for reporting under this criterion.
Comment 12: One comment recommended that, with regard to a
condition or operation prohibited by the plant's Technical
Specifications, reporting should be eliminated for violation of all
administrative Technical Specifications.
Response: The comment is partially accepted. The proposed rule
would eliminate reporting for Technical Specifications that are
administrative in nature. The reporting guidelines would not change. As
stated in the current reporting guidelines in NUREG-1022, Revision 1,
failure to meet administrative Technical Specifications requirements is
reportable only if it results in violations of equipment operability
requirements, or had a similar detrimental effect on a licensee's
ability to safely operate the plant. For example, operation with less
than the required number of people on shift would constitute operation
prohibited by the Technical Specifications. However, a change in the
plant's organizational structure that has not yet been approved as a
Technical Specification change would not. An administrative procedure
violation or failure to implement a procedure, such as failure to lock
a high radiation area door, that does not have a direct impact on the
safe operation of the plant, is generally not reportable under this
criterion.
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Comment 13: One comment recommended changing 10 CFR 50.73 to
require that LERs identify: (1) How many opportunities to detect the
problem were missed and (2) corrective actions to prevent future
misses.
Response: No changes are proposed. If missed opportunities are
identified and are significant to the event, they should be captured by
the current requirements to provide a comprehensive description of the
event and to describe corrective actions if they are significant to the
event.
Comment 14: With regard to design issues, one comment recommended
including language in the rules or their statements of considerations
encouraging a voluntary report under 10 CFR 50.9 for a newly discovered
design issue which is not otherwise reportable at the plant where first
discovered (because the affected systems can still perform their
specified safety functions) but which might have a significant impact
on generic design issues at other plants.
Response: A statement encouraging submittal of voluntary LERs is
included in the reporting guidelines. In addition, the guidelines would
indicate that any significant degradation that could reasonably be
expected to affect multiple similar components in the plant should be
reported.
Comment 15: Several comments opposed placing a condition, related
to systematic non-compliance, on the elimination of reporting of late
surveillance tests (as proposed in the ANPR) under 10 CFR 50.73. The
condition would be burdensome because licensees would need to track
instances of missed surveillance tests in given time periods.
Response: The proposed rule does not contain this condition.
Reporting for the purpose of identifying systematic non-compliance is
not needed because NRC resident inspectors routinely review plant
problem lists, and thus would be aware of any systematic non-compliance
in this area if it occurs.
Comment 16: One comment recommended changing the rules to allow
licensees to rely on notifications made to resident inspectors, which
could eliminate the need to make a telephone notification via the
emergency notification system (ENS) and/or submit a written LER, at
least for some events or conditions. They indicated, for example, this
should be adequate where the event is a decision to issue a news
release.
Response: No changes are proposed. Telephone notifications to the
NRC Operations Center, when required, are needed to ensure that the
event can be promptly reviewed. This includes notification of the NRC
Headquarters Emergency Officers and the Regional Duty Officer and
consideration of whether to activate NRC incident response procedures.
Written LERs, when required, are needed to ensure that events can be
systematically reviewed for safety significance.
Comment 17: Some comments opposed amending 10 CFR 50.73 to require
additional information regarding equipment availability for shutdown
events (as proposed in the ANPR) to support staff probabilistic risk
assessments (PRAs). They indicated that it is rare that sufficient
information is not available in an LER.
Response: The proposed rule would require such information.
Frequently, when shutdown events are subjected to a probabilistic risk
analysis, it is necessary to call the plant to determine the status of
systems and equipment. The proposed rule would eliminate much of that
need.
Comment 18: Several comments recommended deleting 10 CFR
50.72(b)(2)(i), ``Any event found while the reactor is shut down, 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.'' The comments indicated
that because the plant would be shutdown, there is no need for
immediate NRC action.
Response: The requirement for telephone reporting would not be
entirely eliminated because, if a principal safety barrier is
significantly degraded or a condition that significantly affects plant
safety exists; the event may be significant enough that the NRC would
need to initiate actions [such as contacting the plant to better
understand the event and/or initiating a special inspection or
investigation] within about a day even if the plant is shutdown.
However, in the proposed rule this specific criterion would be
combined with 10 CFR 50.72(b)(1)(ii), ``Any event or condition during
plant operation that results in the condition of the nuclear power
plant, including its principal safety barriers, being seriously
degraded or * * * '' Also, the term ``unanalyzed condition that
significantly compromises plant safety'' would be deleted. In
combination with other changes, this would result in the following
criterion for telephone notification ``Any event or condition that
results in the condition of the nuclear power plant, including its
principal safety barriers, being seriously degraded.''
Comment 19: Some comments recommended that the NRC use enforcement
discretion during the rulemaking process to provide early relief with
regard to reporting a condition outside the design basis of the plant
and/or a late surveillance test (condition or operation prohibited by
Technical Specifications).
Response: The current rules will continue to apply until final
revised rules are issued and become effective. However in
dispositioning any violation, the risk-and safety-significance of the
violation will be an important consideration. Establishing an interim
enforcement discretion policy would involve the same critical elements
as developing the revised rule and guidance including a provision for
public comment. This would complicate the rulemaking process, and
essentially constitute a prediction of its final outcome, which may or
may not turn out to be correct.
Comment 20: Several comment letters opposed the idea of tying
enforcement criteria (i.e., violation severity levels) to reporting
criteria. They indicated this could have an unintended adverse effect
on reporting and the resources consumed because in matching an event
with a reporting criterion, a licensee would essentially be forced to
make a preliminary determination of severity level.
Response: The comments are not accepted. The proposed changes to
the enforcement criteria, are discussed below under the heading
``Enforcement.''
Comment 21: As requested by the ANPR, a number of comments
identified reactor reporting requirements other than sections 50.72 and
50.73 where changes are warranted.
Response: Comments regarding changes to reactor reporting
requirements other than sections 50.72 and 50.73 will be addressed in a
separate action. A Commission paper on that subject was submitted on
January 20, 1999, SECY-99-022, ``Rulemaking to Modify Reporting
Requirements for Power Reactors'' and the Commission issued a Staff
Requirements Memorandum on March 19, 1999 directing the staff to
proceed with planning and scheduling.
Comment 22: One comment recommended changing the required initial
reporting time for some events to `` * * * within 8 hours or by the
beginning of the next business day,'' instead of simply specifying `` *
* * within 8 hours.'' The comment indicated it does not appear that
the
[[Page 36295]]
NRC takes action on these events during non-business hours.
Response: The comment is not accepted. The NRC needs these reports
in time to call the plant to find out more about the event and/or
initiate a special inspection or an investigation, if warranted, within
a day. Sometimes these actions are taken during non-business hours.
Comment 23: One comment recommended that an event or condition that
could have prevented fulfillment of the safety function of structures
or systems. * * * should be reportable only when the time limits of the
TS are exceeded. It indicated that if the time limits are not exceeded
the event is not significant enough to warrant reporting.
Response: The comment is not accepted. Generally, standard TS
require commencement of shutdown within one hour if an important
system, such as emergency ac power, is inoperable. However, the stated
reason for allowing one hour before commencing the shutdown is to
provide time to prepare for an orderly shutdown. Also, the condition
might have lasted much longer than one hour before it was discovered.
Finally, an event that results in a safety system failure (or inability
to perform its function) is generally significant enough to warrant NRC
review.
Comment 24: One comment from the State of Ohio recommended that,
although rule changes are not necessary, emphasis should be placed on
positive notification of State and local agencies of emergency
conditions before calling the NRC.
Response: The comment is accepted. It arose from a weakness in the
NRC's response to an event at the Davis-Besse plant. Because there were
considerable difficulties in establishing telephone communications with
the plant at the time of the event, NRC Operations Center personnel
requested that the licensee remain on the line and said that the NRC
would notify the State. However, the NRC did not do so in a timely
manner. Training and procedure changes have been implemented to ensure
this type of problem will not reoccur.
Comment 25: One comment letter, from the State of Illinois, stated
the following: ``In section 50.72 of the advance notice of proposed
rulemaking, seven non-emergency events listed as (f), are proposed to
be reported in eight hours instead of one hour. Of those seven events,
six (specifically, (ii), (iii), (iv), (v), (vi), and (vii)) would
probably be classified as emergency events under existing emergency
plans at an Illinois site * * *. This will cause reporting confusion
during an event at a time when clarity is necessary. These six events
should all be reported as emergency events, not non-emergency events.
EAL thresholds in licensee emergency plans should be required to
reflect them clearly. All of these events would affect the State of
Illinois' response and our emergency plans. NRC must reconsider the
categories of non-emergency events in the context of the current
guidance to licensees for classifying EALs to ensure there is a clear
distinction between emergency and non-emergency reportable events.''
Response: Section 50.72 has been reviewed, and appears to be clear
in this regard. It indicates the following:
(1) Any declaration of an Emergency Class is reportable pursuant to
10 CFR 50.72(a)(1)(i) and (a)(3),
(2) The conditions listed in paragraph (b)(1), ``One-hour
reports,'' are reportable pursuant to paragraph (b)(1) if not reported
as a declaration of an Emergency Class under paragraph (a), and
(3) The conditions listed in paragraph (b)(2), ``Eight-hour
reports, are reportable pursuant to paragraph (b)(2), if not reported
under paragraphs (a) or (b)(1).
Comment 26: One comment letter, from the State of Illinois, opposed
relaxing the required initial reporting time from 4 hours to 8 hours
for the following types of events:
(i) Airborne radioactive release that results in concentrations
over 20 times allowable levels in an unrestricted area;
(ii) Liquid effluent in excess of 20 times allowable concentrations
released to an unrestricted area;
(iii) Radioactively contaminated person transported to an offsite
medical facility for treatment;
(iv) News release or other government agency notification related
to the health and safety of the public or onsite personnel, or
protection of the environment.
The comment further indicated: ``It is of paramount importance that
those charged with regulating and monitoring the public impact of
radiological releases are being kept informed of unplanned releases in
a timely manner. Illinois law requires that we perform independent
assessments, decide what actions may be necessary to protect the
public, and assist in informing the public regarding any radiological
risk. Should follow-up action to a release be necessary, then the less
time that has elapsed, the better the state is able to respond in a
timely and appropriate manner. We oppose any reduction in notification
requirements for unplanned radiation releases from a site regardless of
the source or quantity.
Timeliness is also important for items of obvious public interest.
News of seemingly small events spreads quickly, particularly in local
communities around the power plants. Delayed reporting of such events
means that we will be unprepared to respond to queries from local
officials, or the media, with a resultant loss of public confidence.
Therefore, we also oppose any reduction in notification requirements
for newsworthy events.''
Response: In the interest of simplicity, the proposed amendments
would maintain just three basic levels of required reporting times in
10 CFR 50.72 and 50.73 (1 hour, 8 hours, and 60 days). However, the
concern is recognized and public comment is specifically invited on the
question of whether additional levels should be introduced to better
correspond to particular types of events, as discussed below under the
heading ``Required Initial Reporting Times.'' Also, if in a final rule
the NRC should relax the time limit to 8 hours, a State would not be
precluded from obtaining reports earlier than 8 hours.
Comment 27: Two comment letters addressed coordination with States.
The comment letter from Florida Power & Light Company stated ``The
NRC's Public workshop on August 21, 1998, touched on a number of
examples where opportunities exist to reduce reporting burdens. An
industry representative commented that licensees sometimes have to
report the same event to state agencies and the NRC provided one such
example. FPL concurs with the recommendation that the time requirement
for reporting an event to the NRC and to the state should be consistent
wherever practical and possibly in some cases eliminated.''
The comment letter from Northeast Nuclear Energy Company stated
``Northeast Nuclear Energy Company agrees with extending the non-
emergency prompt notifications to eight hours. This would help to
eliminate unnecessary reports and retractions. However, it is necessary
to have the individual states closely involved with the rule change
since they may have requirements that are more restrictive or conflict
with the proposed rulemaking. For example, in Connecticut all 10 CFR
50.72 reports require notification of the state within one hour.''
Response: The ANPR specifically requested State input. In addition,
a letter requesting input was sent to each State. Written comments were
received from the State of Ohio and the State of Illinois. In addition,
representatives
[[Page 36296]]
from several States attended one of the public meetings on the ANPR.
The NRC will continue to solicit State input as the rulemaking process
proceeds.
Comment 28: One comment recommended eliminating two of the
requirements for immediate followup notification during the course of
an event, section 50.72(c)(2)(i), the results of ensuing evaluations or
assessments of plant conditions, and section 50.72(c)(2)(ii), the
effectiveness of response or protective measures taken. The comment
indicated that the requirements continue to apply after the event and
that they require reporting even if, for example, the result of a
further analysis does not change the initial report.
Response: The comment is not accepted. The requirements for
followup reporting apply only during the course of the event. Followup
reports are needed while the event is ongoing. For example, if an
analysis is completed during an ongoing event, and it confirms an
earlier estimate of how long it will take to uncover the reactor core
if electric power is not restored, that information may very well be
useful for the purpose of evaluating the need for protective measures
(evacuation).
Comment 29: One comment recommended clarifying the reporting
requirements for problems identified by NRC inspectors.
Response: No changes are proposed. The current reporting guidelines
include a paragraph making it clear that an event must be reported via
telephone notification and/or written LER, as required, regardless of
whether it had been discussed with NRC staff personnel or was
identified by NRC personnel.
Comment 30: Several comments recommended changing the requirements
in 50.46(a)(iii)(2) for reporting errors in or corrections to ECCS
analyses.
Response: These comments will be addressed in a separate action
(along with other comments on reporting requirements other than
sections 50.72 and 50.73).
Comment 31: Some comments raised issues regarding plant-specific
reporting requirements contained in Technical Specifications (or other
parts of the operating license). One suggestion was that 10 CFR 50.72
and 50.73 should be changed to address these issues. Another suggestion
was that a Generic Letter be issued indicating that the NRC would be
receptive to requests for license amendments to eliminate specific
reporting requirements.
Response: No changes are proposed for sections 50.72 and 50.73,
which identify generic reporting requirements. It is not feasible or
appropriate to address the specific reporting requirements contained in
individual operating licenses in this format.
The idea of issuing a generic communication to specific requests
for license amendments will be addressed (along with other comments on
reporting requirements beyond the scope of sections 50.72 and 50.73) in
a separate action.
Comment 32: One comment recommended that in section 50.72(b)(1)(v),
the word ``offsite'' be added before ``communications capability'' to
make it clear that what must be reported is a loss of communications
with outside agencies, not internal plant communications systems.
Response: The comment is accepted. In the proposed rule the word
``offsite'' would be added.
Comment 33: Several comments suggested that the NRC should define
its needs relative to the information provided in LERs.
Response: The essential purpose of the LER rule is to identify 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. The rule is designed to provide the
information necessary for engineering studies of operational anomalies
and trends, and patterns analysis of operational occurrences. To this
end, the information required in LERs is generally needed to understand
the event, its significance, and its causes in order to determine
whether generic or plant specific action is needed to preclude
recurrence. Some further specific functions are discussed below.
It is necessary to identify and analyze events and conditions that
are precursors to potential severe core damage, to discover emerging
trends or patterns of potential safety significance, to identify events
that are important to safety and their associated safety concerns and
root causes, to determine the adequacy of corrective actions taken to
address the safety concerns, and to assess the generic applicability of
events.
The NRC staff reviews each LER to identify those individual events
or generic situations that warrant additional analysis and evaluation.
The staff identifies repetitive events and failures and situations
where the frequency or the combined significance of reported events may
be cause for concern. The NRC staff reviews past operating history for
similar events and initiates a generic study, as appropriate, to focus
upon the nature, cause, consequences and possible corrective actions
for the particular situation or concern.
The NRC staff uses the information reported in LERs in confirming
licensing bases, studying potentially generic safety problems,
assessing trends and patterns of operational experience, monitoring
performance, identifying precursors of more significant events, and
providing operational experience to the industry.
The NRC determines whether events meet the criteria for reporting
as an Abnormal Occurrence Report to Congress or for reporting to the
European Nuclear Energy Agency (NEA).
The information from LERs is widely used within the nuclear
industry, both nationally and internationally. The industry's Institute
of Nuclear Power Operation (INPO) uses LERs as a basis for providing
operational safety experience feedback data to individual utilities
through such documents as significant operating experience reports,
significant event reports, significant events notifications, and
operations and maintenance reminders. U.S. vendors and nuclear steam
system suppliers, as well as other countries and international
organizations, use LER data as a source of operational experience data.
Comment 34: Some comments indicated that the licensing basis should
be defined.
Response: No changes are proposed. The term ``licensing basis'' is
not explicitly used in the event reporting rules or the draft reporting
guidelines. It can come into play, via Generic Letter (GL) 91-18,
``Information to Licensees Regarding two NRC Inspection Manual Sections
on Resolution of Degraded and Nonconforming Conditions and on
Operability,'' in determining what the ``specified safety function'' of
a system is. This relates to whether an event is reportable as an event
or condition that could have prevented the fulfillment of the safety
function of structures or systems * * * and/or an operation or
condition prohibited by the plant's technical specification (TS).
However, any unsettled details regarding exactly which commitments are
included in the licensing basis (for example because of differences
between the definitions in GL 91-18 and 10 CFR 54.3) are not of a
nature that would change the determination of whether or not a system
is capable of performing its specified safety functions (i.e.,
operable).
[[Page 36297]]
Comment 35: Several comments recommended conducting tabletop
exercises (public meetings) early in the drafting process, involving
licensees, inspectors, and headquarters personnel to discuss the draft
amendments and associated and guidance.
Response: The Commission agrees. The recommended public meeting was
held on November 13, 1998.
Comment 36: Several comments recommended conducting a workshop
(public meeting) early during the public comment period to discuss the
proposed rule and draft guidance.
Response: The Commission agrees. The recommended workshop has been
added to the schedule.
Comment 37: Several comments recommended that the reporting
guidelines be revised concurrently with the rules.
Response: The Commission agrees. Draft guidelines are being made
available for comment concurrent with the proposed rules.
Comment 38: Several comment letters recommended reviewing
enforcement criteria at the same time the rule is being developed to
ensure consistent application of enforcement to reporting.
Response: The comment is accepted. The Enforcement Policy is being
reviewed concurrently with development of the rule.
IV. Discussion
1. Objectives of Proposed Amendments
The purpose of sections 50.72 and 50.73 would remain the same
because the basic needs remain the same. The objectives of the proposed
amendments would be as follows:
(1) To better align the reporting requirements with the NRC's
current reporting needs. An example is extending the required initial
reporting times for some events, consistent with the need for timely
NRC action. Another example is changing the criteria for reporting
system actuations, to obtain reporting that is more consistent with the
risk-significance of the systems involved.
(2) To reduce the reporting burden, consistent with the NRC's
reporting needs. An example is eliminating the reporting of design and
analysis defects and deviations of 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) To maintain consistency 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 Proposed Amendments
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)
would be 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 sections 50.73(a)(1) and (d)]. In the proposed amendments,
declaration of an emergency class would continue to be reported
immediately after notification of appropriate State or local agencies
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) would 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.
Non-emergency events that are reportable by telephone under 10 CFR
50.72 would be reportable as soon as practical and in all cases within
8 hours (instead of within 1 hour or 4 hours as is currently required).
This would reduce the burden of rapid reporting, while still capturing
those events where there may be a need for the NRC to contact the plant
to find out more about the event and/or initiate a special inspection
or investigation within about a day.
Written LERs would be due within 60 days after discovery of a
reportable event or condition (instead of within 30 days as is
currently 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 would result
in fewer LERs that require amendment (by submittal of an additional
report).
The Performance Indicator (PI) program and the future risk-based
performance indicator program provide valued input to regulatory
decisions (e.g. Senior Management Meetings). Adding 30 days to the
delivery of data supplying these programs would result in the reduction
in the currency and value of these indicators to senior managers. With
respect to the Accident Sequence Precursor program, the additional 30
days will add a commensurate amount of time to each individual event
assessment since Licensee Event Reports (LERs) are the main source of
data for these analyses. The delivery date for the annual Accident
Sequence Precursor report would also slip accordingly. The NRC staff
would have to make more extensive use of Immediate Notifications (10
CFR 50.72) and event followup to compensate in part for the Licensee
Event Report (LER) reporting extension.
In the interest of simplicity, the proposed amendments would
maintain just three basic levels of required reporting times in 10 CFR
50.72 and 50.73 (1 hour, 8 hours, and 60 days). However public comment
is specifically invited on the question of whether additional levels
should be introduced to better correspond or particular types of
events. For example, 10 CFR 50.72 currently requires 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.
Operation or condition prohibited by TS [section
50.73(a)(2)(i)(B)]. The term ``during the previous three years'' would
be added to eliminate written LERs for conditions that have not existed
during the previous three years. Such a historical event would now have
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 should 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 very considerable
[[Page 36298]]
effort required to review all records older than three years, in order
to rule out the possibility, would not be warranted.
In addition, this criterion would be modified to eliminate
reporting if the technical specification is administrative in nature.
Violation of administrative technical specifications have generally not
been considered to warrant submittal of an LER, and since 1983 when the
rule was issued the staff's reporting guidance has excluded almost all
cases of such reporting. This change would make the plain wording of
the rule consistent with that guidance.
Finally, this criterion would be 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.
Condition of the nuclear power plant, including its principal
safety barriers, being seriously degraded [current sections
50.72(b)(1)(ii) and (b)(2)(i), replaced by new section 50.72(b)(2)(ii),
and section 50.73(a)(2)(ii)]. Currently, 10 CFR 50.72(b)(1)(ii) and
(b)(2)(i) provide the following distinction: a qualifying event or
condition during operation is initially reportable in one hour; a
condition discovered while shutdown that would have qualified if it had
it been discovered during operation is initially reportable in four
hours. The new 10 CFR 50.72(b)(2)(ii) would eliminate the distinction
because there would no longer be separate 1-hour and 4-hour categories
of non-emergency reports for this criterion. There would only be 8-hour
non-emergency reports for this criterion.
Unanalyzed condition that significantly compromises plant safety
[sections 50.72(b)(1)(ii)(A) and (b)(2)(i), and section
50.73(a)(2)(ii)(A); replaced by new section 50.72(b)(2)(ii)(B), and
section 50.73(a)(2)(ii)(B)]. Currently, 10 CFR 50.72(b)(1)(ii)(A) and
(b)(2)(i) provide the following distinction: a qualifying event or
condition during operation is initially reportable in one hour; a
condition discovered while shutdown that would have qualified if it had
it been discovered during operation is initially reportable in four
hours. The new 10 CFR 50.72(b)(2)(ii)(B) would eliminate the
distinction because there would no longer be separate 1-hour and 4-hour
categories of non-emergency reports for this reporting criterion. There
would only be 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) would refer to a condition that significantly
affects 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 [current
Section 50.72(b)(2)(ii)(B) and section 50.73(a)(2)(ii)(B)]. This
criterion would be deleted. However, a condition outside the design
basis of the plant would still be reported if it is significant enough
to qualify under one or more of the following criteria.
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 would not be capable of performing its specified safety
functions, the condition would be reportable under sections
50.72(b)(2)(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: (A) Shut down * * *].
For example, during testing of 480 volt safety-related breakers,
one breaker would not trip electrically. The cause was a loose
connection, due to a lug that was too large for a connecting wire.
Other safety related breakers did not malfunction, but they had the
same mismatch. The event would be reportable because the incompatible
lugs and wires could have caused one or more safety systems to fail to
perform their specified safety function(s).
Another example is as follows. 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 would be reportable because there was reasonable doubt about the
ability of the EDGs to operate for an extended period of time, as
required.
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, and thus the train is
inoperable longer than allowed by the TS, the condition would be
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
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 a certain postulated
steam line breaks, the event would be reportable.
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 affected, the condition would be reportable
under sections 50.72(b)(2)(ii)(B) and 50.73(a)(2)(ii)(B) [i.e., an
unanalyzed condition that significantly affects 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.''
\1\
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\1\ 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.'' \2\
---------------------------------------------------------------------------
\2\ 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
[[Page 36299]]
unanalyzed condition and would be reportable.'' \3\
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\3\ 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.'' \4\
---------------------------------------------------------------------------
\4\ 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.
---------------------------------------------------------------------------
Furthermore, beyond the examples given in 1983, examples of
reportable events would 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 would be 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 would not be
reportable.
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 would be reportable
under sections 50.72(b)(2)(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) Steam generator tube degradation in the following
circumstances:
(1) The severity of degradation corresponds to failure to maintain
structural safety factors. The structural safety factors implicit in
the licensing basis are those described in Regulatory Guide 1.121.
These safety factors include a margin of 3.0 against gross failure or
burst under normal plant operating conditions, including startup,
operation in the power range, hot standby, and cooldown, and all
anticipated transients that are included in the plant design
specification.
(2) The calculated potential primary-to-secondary leak rate is not
consistent with the plant licensing basis. The licensing basis accident
analyses typically assume [for accidents other than a steam generator
tube rupture (SGTR)] a 1 gpm primary-to-secondary leak rate concurrent
with the accident to demonstrate that the radiological consequences
satisfy 10 CFR Part 100 and GDC-19. In these instances, degradation
which may lead to leakage above 1 gpm under accident conditions, other
than a SGTR, would exceed the threshold. For some units, the staff has
approved accident leakages above 1 gpm subject to updating the
licensing basis accident analyses to reflect this amount of leakage and
subject to risk implications being acceptable.\5\
---------------------------------------------------------------------------
\5\ In addition, if the extent of degradation is great (i.e., if
many tubes are degraded or defective), a telephone notification and
a written LER should be provided. The plant's TS typically provide
specific requirements indicating when reporting is required (based
on the number of tubes degraded or defective in terms of ``percent
inspected'') and those requirements should be used to determine
reportability.
---------------------------------------------------------------------------
(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.\6\
---------------------------------------------------------------------------
\6\ 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.
---------------------------------------------------------------------------
Finally, a condition outside the design basis of the plant (or any
other event or condition) would be reportable if a component is in a
degraded or non-conforming condition such that the ability of a
component to perform its specified safety function is significantly
degraded and the condition could reasonably be expected to apply to
other similar components in the plant. This new criterion is contained
in section 50.73(a)(2)(ii)(C) as discussed below.
As a result, these proposed amendments would focus the reporting of
conditions outside the design basis of the plant to the safety
significant issues while reducing the number of reports under the
current rules in order to minimize the reporting of less significant
issues. In particular, the proposed amendments will help ensure that
significant safety problems that could reasonably be expected to be
applicable to similar components at the specific plant or at other
plants will be identified and addressed although the specific licensee
might determine that the system or structure remained operable, or that
technical specification requirements were met. The proposed rules will
provide that, consistent with the NRC's effort to obtain information
for engineering studies of operational anomalies and trends and
patterns analysis of operational occurrences, the NRC would be able to
monitor the capability of safety-related components to perform their
design-basis functions.
Significantly degraded component(s) [section 50.73(a)(2)(ii)(C)].
This new reporting criterion would require reporting if a component is
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 apply to
other similar components in the plant. It would be added 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. On the other hand,
if the degradations are not significant or the condition does not have
generic implications, reporting would not be required under this
criterion.
For example, at one plant several normally open valves in the low
pressure safety injection system were routinely closed to support
quarterly surveillance testing of the system. In reviewing the design
basis and associated calculations, it was determined that the
capability of the valves to open in the event of a large break loss-of-
coolant accident (LOCA) combined with degraded grid voltage during a
surveillance test was degraded. The licensee concluded that the valves
would still be able to reopen under the postulated conditions and
considered them operable. However, that conclusion could not be
supported using the conservative standards established by Generic
Letter 89-10. Pending determination of final corrective action,
administrative procedures were implemented to preclude closing the
valves. The event would be reportable because the
[[Page 36300]]
capability of a component to perform its specified safety functions was
significantly degraded and the same condition could reasonably be
expected to apply to other similar components.
In another example, during a routine periodic inspection, jumper
wires in the valve operators for three valves were found contaminated
with grease which was leaking from the limit switch gear box. The cause
was overfilling of the grease box, as a result of following a generic
maintenance procedure. The leakage resulted in contamination and
degradation of the electrical components which were not qualified for
exposure to grease. This could result in valve malfunction(s). The
conditions were corrected and the maintenance procedures were changed.
The event would be reportable because the capability of several similar
components to perform their specified safety functions could be
significantly degraded.
In a further example, while processing calculations it was
determined that four motor operated valves within the reactor building
were located below the accident flood level and were not qualified for
that condition. Pending replacement with qualified equipment, the
licensee determined that three of the valves had sufficiently short
opening time that their safety function would be completed before they
were submerged. The fourth valve was normally open and could remain
open. After flooding, valve position indication could be lost, but
valve position could be established indirectly using process parameter
indications. The event would be reportable because the capability of
several similar components to perform their specified safety functions
could be significantly degraded.
An example of an event that would not be reportable is as follows.
The motor on a motor-operated valve (MOV) burned out after repeated
cycling for testing. This event would not be reportable because it is a
single component failure, and while there might be similar MOVs in the
plant, there is not a reasonable basis to think that other MOVs would
be affected by this same condition. On the other hand, if several MOVs
had been repeatedly cycled and then after some extended period of time
one of the MOVs was found inoperable or significantly degraded because
of that cycling, then the condition would be reportable.
Minor switch adjustments on MOVs would not be reported where they
do not significantly affect the ability of the MOV to carry out its
design-basis function and the cause of the adjustments is not a generic
concern.
At one plant the switch on the radio transmitter for the auxiliary
building crane was used to handle a spent fuel cask while two
protective features had been defeated by wiring errors. A new radio
control transmitter had been procured and placed in service. Because
the new controller was wired differently than the old one, the drum
overspeed protection and spent fuel pool roof slot limit switch were
inadvertently defeated. While the crane was found to be outside its
design basis, this condition would not be reportable because the switch
wiring deficiency could not reasonably be expected to affect any other
components at the plant.
Condition not covered by the plant's operating and emergency
procedures [section 50.72(b)(2)(ii)(C), and section
50.73(a)(2)(ii)(C)]. This criterion would be deleted because it does
not result in worthwhile reports aside from those that would be
captured by other reporting criteria such as:
(1) An unanalyzed condition that significantly affects 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:
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;
(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 any of the
systems listed in the rules, as amended;
(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.
Manual or automatic actuation of any engineered safety feature ESF
[current sections 50.72(b)(1)(iv) and (b)(2)(ii), replaced by new
sections 50.72(b)(2)(iv), and section 50.73(a)(2)(iv)]. Currently,
sections 50.72(b)(1)(iv) and (b)(2)(ii) provide the following
distinction: an event that results or should have resulted in ECCS
discharge into the reactor coolant system is initially reportable
within 1 hour; other ESF actuations are initially reportable within 4
hours. The new 10 CFR 50.72(b)(2)(iv) would eliminate this distinction
because there would no longer be separate 1-hour and 4-hour categories
of non-emergency reports for this criterion. There would only be 8-hour
non-emergency reports for this criterion.
The new section 50.72(b)(2)(iv) would eliminate telephone reporting
for invalid automatic actuation or unintentional manual actuation.
These events are not significant and thus telephone reporting is not
needed. However, the proposed amendments would not eliminate the
requirement for a written report 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. (See 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).
The term ``any engineered safety feature (ESF), including the
reactor protection system (RPS),'' which currently defines the systems
for which actuation must be reported in section 50.72(b)(2)(iv) and
section 50.73(a)(2)(iv), would be replaced by a specific list of
systems. The current definition has led to confusion and variability in
reporting because there are varying definitions of what constitutes an
ESF. For example, at some plants systems that are known to have high
risk significance, such as emergency ac power, auxiliary feedwater, and
reactor core isolation cooling are not considered ESFs. Furthermore, in
many cases systems with much lower levels of risk significance, such as
control room ventilation systems, are considered to be ESFs.
In the proposed amendments actuation would be reportable for the
specific systems named in sections 50.72(b)(2)(iv) and 50.73(a)(2)(iv).
This would result in consistent reporting of events that result in
actuation of these highly risk-significant systems. Reasonable
consistency in reporting actuation of highly risk-significant systems
is needed to support estimating equipment reliability parameters, which
is important to several aspects of the move towards more risk-informed
regulation, including more risk-informed monitoring of plant
performance.
The specific list of systems in the proposed rule would also
eliminate reporting for events of lesser significance, such as
actuation of control room ventilation systems.
The specific list of systems in the proposed rule is similar to the
list of systems currently provided in the reporting guidelines in
NUREG-1022,
[[Page 36301]]
Revision 1, with some minor revisions. It is based on systems for which
actuation is frequently reported, and systems with relatively high
risk-significance based on a sampling of plant-specific PRAs (see Draft
Regulatory Guide DG-1046, ``Guidelines for Reporting Reliability and
Availability Information for Risk-Significant Systems and Equipment in
Nuclear Power Plants,'' particularly Tables C-1 through C-5).
This proposal to list the systems in the rule is controversial and
public comment is specifically invited in this area. In particular,
three principal alternatives to the proposed rule have been identified
for comment:
(1) Maintain the status quo. Under this alternative, the rule would
continue to require reporting for actuation of ``any ESF.'' The
guidance would continue to indicate that reporting should include as a
minimum the system on the list.
(2) Require use of a plant-specific, risk-informed list. Under this
alternative, the list of systems would be risk-informed, and plant-
specific. Licensees would develop the list based on existing PRA
analyses, judgment, and specific plant design. No list would be
provided in the rule.
(3) Return to the pre-1998 situation (i.e., before publication of
the reporting guidance in NUREG-1022, Revision 1). Under this
alternative, the rule would continue to require reporting for actuation
of ``any ESF.'' The guidance would indicate that reporting should
include those systems identified as ESF's for each particular plant
(e.g., in the FSAR).
With regard to this third alternative, it may be noted that this
approach has the advantage of clarity and simplicity. There would be no
need to develop a new list, and this is the practice that was followed
from 1984-1997 without creating major problems. However, the lists of
ESFs are not based on risk-significance. For example, emergency diesel
generators (EDGs) are known to be highly risk-significant; however, at
six plants, the EDGs are not considered to be ESFs. Similarly,
auxiliary feedwater (AFW), systems at pressurized water reactors (PWRs)
are known to be highly risk-significant; however, at a number of plants
these systems are not considered to be ESFs. Also, reactor core
isolation cooling (RCIC) systems at boiling water reactors (BWRs) are
known to be highly risk significant; however, at a number of plants
these systems are not considered to be ESFs. In contrast, at many
plants, systems with much lower levels of risk significance, such as
control room ventilation systems, are considered to be ESFs.
Event or condition that could have prevented fulfillment of the
safety function of structures or systems that * * * [current sections
50.72(b)(1)(ii) and (b)(2)(i), replaced by new sections 50.72(b)(2)(v)
and (vi), and sections 50.73(a)(2)(v) and (vi)] The phrase ``event or
condition that alone could have prevented the fulfillment of the safety
function of structures or systems.* * * '' would be 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'' would be added to section
50.72(b)(2)(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 some time ago 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 would be required, but there would be no need
for telephone notification.
The phrase ``occurring within three years of the date of
discovery'' would be added to section 50.73(a)(2)(v) to eliminate
written LERs for conditions that have not existed during the previous
three years. Such a historical event would now have less significance,
and assessing reportability for earlier times can consume considerable
resources. For example, assume that during a design review a
discrepancy is found that affects the ability of a system to perform
its safety function in a given specific configuration. If it is likely
that the safety function could have been prevented, the answer should
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 very considerable effort required to review all
records older than three years, in order to rule out the possibility,
would not be warranted.
A new paragraph, section 50.72(b)(2)(vi) would be added to clarify
section 50.72. The new paragraph would explicitly state 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 is currently stated explicitly in section
50.73(a)(2)(vi) with regard to written LERs.
Major loss of emergency assessment capability, offsite response
capability, or communication capability [current section
50.72(b)(2)(v), new section 50.72(b)(2)(xiii)]. The new section would
be 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.
Airborne radioactive release * * * and liquid effluent release * *
* [section 50.72(b)(2)(viii) and sections 50.73(a)(2)(viii) and
50.73(a)(2)(ix)]. The statement indicating reporting under section
50.72(b)(2)(viii) satisfies the requirements of section 20.2202 would
be removed because it would not be correct. For example, some events
captured by section 20.2202 would not be captured by section
50.72(b)(2)(viii). Also, the statement indicating that reporting under
section 50.73(a)(2)(viii) satisfies the requirements of section
20.2203(a)(3) would be deleted because it would not be correct. Some
events captured by section 20.2203(a)(3) would not be captured by
section 50.73(a)(2)(viii).
The proposed extension of reporting deadlines to 8 hours in section
50.72 and 60 days in section 50.73 raises questions about whether
similar changes should be made to Parts 20, 30, 40, 70, 72 and 76. The
merits of such changes, which may vary for different types of
licensees, will be addressed in separate actions.
Contents of LERs [sections 50.73(b)(2)(ii)(F) and
50.73(b)(2)(ii)(J)]. Paragraph (F) would be revised to correct the
address of the NRC Library.
Paragraph (J) currently requires that the narrative section include
the following specific information as appropriate for the particular
event:
``(1) Operator actions that affected the course of the event,
including operator
[[Page 36302]]
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 change section 50.73(b)(2)(ii)(J) to
simply require that the licensee discuss the causes and circumstances
for each human performance related problem that contributed to the
event. It is not necessary to specify the level of detail provided in
the current rule, which is more appropriate for guidance. 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. This draft
report is being made available for public comment concurrently with the
proposed rule, as discussed below under the heading ``Revisions to
Reporting Guidelines in NUREG-1022.''
Spent fuel storage cask problems [current sections 50.72(b)(2)(vii)
and 72.16(a)(1), (a)(2), (b) and (c)]. Section 50.72(b)(2)(vii) would
be deleted because these reporting criteria are redundant to the
reporting criteria contained in sections 72.216(a)(1), (a)(2), and (b).
Repetition of the same reporting criteria in different sections of the
rules adds unnecessary complexity and is inconsistent with the current
practice in other areas, such as reporting of safeguards events as
required by section 73.71.
Also, a conforming amendment would be made to section 72.216. This
is necessary because section 72.216(a) currently relies on section
50.72(b)(2)(vii), which would be deleted, to establish the time limit
for initial notification. The amended section 72.216 would refer to
sections 72.74 and 72.75 for initial notification and followup
reporting requirements.
Assessment of Safety Consequences [section 50.73(b)(3)]. This
section currently requires that an LER include an assessment of the
safety consequences and implications of the event. This assessment must
include the availability of other systems or components that could have
performed the same function as the components and systems that failed
during the event. It would be modified by adding a requirement to also
include the status of components and systems 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.'' This information is needed to better support the
NRC's assessment of the risk-significance of reported events.
Exemptions [section 50.73(f)]. This provision would be deleted
because the exemption provisions in section 50.12 provide for granting
of exemptions as warranted. Thus, including another, section-specific
exemption provision in section 50.73 adds unnecessary complexity to the
rules.
3. Revisions to Reporting Guidelines in NUREG-1022
A draft report, NUREG-1022, Revision 2, ``Event Reporting
Guidelines, 10 CFR 50.72 and 50.73,'' is being made available for
public comment concurrently with the proposed amendments to 10 CFR
50.72 and 50.73. The draft 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
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 draft
report, guidance that is considered to be new or different is a
meaningful way, relative to that provided in NUREG-1022, Revision 1, is
indicated by redlining the appropriate text.
4. Reactor Oversight
The NRC is developing revisions to 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 proposed rules
and believes that the changes would not have a deleterious effect on
the oversight process. Public comment is invited on whether or not this
is the case. In particular, it is requested that if any examples to the
contrary are known they be identified.
5. Reporting of Historical Problems
As discussed above, provisions would be added to sections
50.73(a)(2)(i)(B) and 50.73(a)(2)(v) to eliminate reporting of a
condition or event that did not occur within three years of the date of
discovery. (See the response to Comment 8, the discussion under the
heading ``Operation or condition prohibited by TS,'' and the discussion
under the heading ``Event or condition that could have prevented
fulfillment of the safety function of structures or systems that * * *
'') Public comment is invited on whether such historical events and
conditions should be reported (rather than being excluded from
reporting, as proposed). Public comment is 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).
6. Reporting of Component Problems
As discussed above, a new reporting criterion would be added to
require reporting if a component is 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 apply to other similar components in
the plant. (See the response to Comment 14 and the discussion under the
heading ``Significantly degraded component(s) [section
50.73(a)(2)(ii)(C)].'') Public comment is invited on whether this
proposed new criterion would accomplish its stated purpose--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.
Public comment is also invited on whether the proposed new criterion
would be subject to varying interpretations by licensees and
inspectors.
7. Enforcement
The NRC intends to modify its existing enforcement policy in
connection with the proposed amendments to sections 50.72 and 50.73.
The philosophy of the proposed changes is to base the significance of
the reporting violation on: (1) The reporting requirement, which will
require reporting within time frames more commensurate with the
significance of the underlying issues than the current rule; and (2)
the impact that a late report may have on the ability of the NRC to
[[Page 36303]]
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, Rev. 1 as follows:
(1) Appendix B, Supplement I.C--Examples of Severity Level III
violations.
(a) Example 14 would be revised to read as follows--A failure to
provide the required one hour telephone notification of an emergency
action taken pursuant to 10 CFR 50.54(x).
(b) An additional example would be added that would read as
follows--A failure to provide a required 1-hour or 8-hour non-emergency
telephone notification pursuant to 10 CFR 50.72.
(c) An additional example would be added that would read as
follows--A late 8-hour notification that substantially impacts agency
response.
(2) Appendix B, Supplement I.D--Examples of Severity Level IV
violations.
(a) Example 4, would 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 would 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 would correlate the Severity Level of the infractions
with the relative importance of the information needed by the NRC.
Section IV.D 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 or 8-hour non-emergency report the impact that
the failure to report had on any agency response would 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 would not normally be reduced. Similarly, a late notification
that delayed the NRC's ability to perform an engineering analysis of a
condition to determine if additional regulatory action was necessary
would generally not be considered for disposition at a reduced Severity
Level. Additionally, late reports filed in cases where the NRC had to
prompt the licensee to report would generally not be subject to
disposition at reduced Severity Level and the Severity Level for
failure to submit a timely Licensee Event Report (LER) would not be
reduced to a minor violation.
In accordance with Appendix C of the Enforcement Policy, `` Interim
Enforcement Policy for Severity Level IV Violations Involving
Activities of Power Reactor Licensees,'' the failure to file a 60-day
LER would normally be dispositioned as a Non-Cited Violation (NCV).
Repetitive failures to make LER reports indicative of a licensee's
inability to recognize reportable conditions, such that it is not
likely that the NRC will be made aware of operational, design and
configuration issues deemed reportable pursuant to 10 CFR 50.73, will
be considered for categorization at Severity Level III. This
disposition may be warranted since such licensee performance impacts
the ability of the NRC to fulfill its regulatory obligations.
8. Electronic Reporting
The NRC is currently planning to implement an electronic document
management and reporting program, known as the Agency-wide Document
Access and Management System (ADAMS), that will in general provide for
electronic submittal of many types of reports, including LERs.
Accordingly, no separate rulemaking effort to provide for electronic
submittal of LERs is contemplated.
9. Schedule
The current schedule is as follows:
08/99--Conduct public workshop to discuss proposed rule and draft
reporting guidelines (separate notice with workshop details will be
published later this month).
August 5, 1999--Public comments due to OMB
September 7, 1999--Receive OMB approval
September 20, 1999--Public comments due to NRC
10/01/99--Provide final rule and guidelines to NRC staff rulemaking
group
11/05/99--Provide final rule and guidelines to the formal concurrence
chain
01/14/00--Provide final rule and guidelines to CRGR and ACRS
02/11/00--Complete briefings of CRGR and ACRS
03/10/00--Provide final rule and guidelines to Commission
04/07/00--Publish final rule and guidelines
10. 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 seeks State comment on issues
related to the proposed amendments to power reactor reporting
requirements.
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 requests comments on
this proposed rule specifically with respect to the clarity and
effectiveness of the language used. Comments should be sent to the
address listed above.
V. 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.
VI. 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 proposed rule. 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.
VII. Regulatory Analysis
The Commission has prepared a draft regulatory analysis on this
proposed rule. The analysis examines the costs and benefits of the
alternatives considered by the Commission. The draft 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.''
The Commission requests public comment on this draft analysis.
Comments on the draft analysis may be
[[Page 36304]]
submitted to the NRC as discussed above under the heading ADDRESSES.
VIII. Paperwork Reduction Act Statement
This proposed rule would amend information collection requirements
that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501
et seq.). This rule has been submitted to the Office of Management and
Budget for review and approval of the information collection
requirements.
The public reporting burden for the currently existing reporting
requirements in 10 CFR 50.72 and 50.73 is estimated to average about
790 hours per response (i.e., per commercial nuclear power reactor per
year) 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, after which there would be a recurring
annual burden reduction of about 200 hours per reactor per year. The
U.S. Nuclear Regulatory Commission is seeking public comment on the
potential impact of the information collection contained in the
proposed rule and on the following issues:
Is the proposed information collection necessary for the proper
performance of the NRC, including whether the information will have
practical utility?
Is the estimate of burden accurate?
Is there a way to enhance the quality, utility, and clarity of the
information to be collected?
How can the burden of the information collection be minimized,
including the use of automated collection techniques?
Send comments on any aspect of this proposed information
collection, including suggestions for reducing this burden, to the
Information and Records Management Branch (T-5 F33), U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001 or by Internet
electronic mail to [email protected]; and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150AF98), Office of
Management and Budget, Washington, DC 20503.
Comments to OMB on the information collections or on the above
issues should be submitted by August 5, 1999. Comments received after
this date will be considered if it is practical to do so, but
consideration cannot be ensured for comments received after this date.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, an information collection unless it displays a currently
valid OMB control number.
IX. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act (5 U.S.C.
605(b)), the Commission certifies that this rule will not, if
promulgated, 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).
X. Proposed 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, and 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 proposing to
adopt 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.7 (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:
---------------------------------------------------------------------------
\7\ Other requirements for immediate notification of the NRC by
licensed operating nuclear power reactors are contained elsewhere in
this chapter, in particular Secs. 20.1906, 20.2202, 50.36, 72.74,
72.75, and 73.71.
---------------------------------------------------------------------------
(i) The declaration of any of the Emergency Classes specified in
the licensee's approved Emergency Plan; 8 or
---------------------------------------------------------------------------
\8\ These Emergency Classes are addressed in Appendix E of this
part.
---------------------------------------------------------------------------
(ii) Of those non-Emergency events specified in paragraph (b) of
this section.
(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.9, 10
---------------------------------------------------------------------------
\9\ Commercial telephone number of the NRC Operations Center is
(301) 816-5100.
\10\ [Reserved]
---------------------------------------------------------------------------
(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) 11 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
[[Page 36305]]
system has the capability to transmit the exercise data.
---------------------------------------------------------------------------
\11\ 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) Either paragraph (b)(1), ``One-Hour Report,'' or paragraph
(b)(2) ``Eight-Hour Report,'' 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 the 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) Eight-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 eight 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) 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 affects plant safety.
(iii) Any natural phenomenon or other external condition 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 plant.
(iv)(A) Any event or condition that results in intentional manual
actuation or valid automatic actuation of any of the systems listed in
paragraph (b)(2)(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)(2)(iv)(A) of this section apply are:
(1) Reactor protection system (reactor scram, reactor trip).
(2) 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.
(3) ECCS for boiling water reactors (BWRs) including: high-pressure
and low-pressure core spray systems; high-pressure coolant injection
system; feedwater coolant injection system; low pressure injection
function of the residual heat removal system; and automatic
depressurization system.
(4) BWR isolation condenser system and reactor core isolation
cooling system.
(5) PWR auxiliary feedwater system.
(6) Containment systems including: containment and reactor vessel
isolation systems (general containment isolation signals affecting
numerous valves and main steam isolation valve [MSIV] closure signals
in BWRs) and containment heat removal and depressurization systems,
including containment spray and fan cooler systems.
(7) Emergency ac electrical power systems, including: emergency
diesel generators (EDGs) and their associated support systems;
hydroelectric facilities used in lieu of EDGs at the Oconee Station;
safety related gas turbine generators; BWR dedicated Division 3 EDGs
and their associated support systems; and station blackout diesel
generators (and black-start gas turbines that serve a similar purpose)
which are started from the control room and included in the plant's
operating and emergency procedures.
(8) Anticipated transient without scram (ATWS) mitigating systems.
(9) Service water (standby emergency service water systems that do
not normally run).
(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)(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 this paragraph if redundant equipment in
the same system was operable and available to perform the required
safety function.
(vii) [Reserved]
(viii)(A) 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.
(B) 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.
(ix) Any 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 including fires, toxic gas releases, or radioactive
releases.
(x) Any event requiring the transport of a radioactively
contaminated person to an offsite medical facility for treatment.
(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.
(xii) 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. Unless otherwise specified in
this section, the licensee shall report an event 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 occurring within three years of the
date of discovery which was prohibited by the plant's Technical
Specifications, except when:
(1) The technical specification is administrative in nature; or
[[Page 36306]]
(2) The event consists 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 capable of performing its specified safety
functions.
(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;
(B) The nuclear power plant being in an unanalyzed condition that
significantly affects plant safety; or
(C) 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.
(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 (reactor scram, reactor trip).
(2) 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.
(3) ECCS for boiling water reactors (BWRs) including: high-pressure
and low-pressure core spray systems; high-pressure coolant injection
system; feedwater coolant injection system; low pressure injection
function of the residual heat removal system; and automatic
depressurization system.
(4) BWR isolation condenser system and reactor core isolation
cooling system.
(5) PWR auxiliary feedwater system.
(6) Containment systems including: containment and reactor vessel
isolation systems (general containment isolation signals affecting
numerous valves and main steam isolation valve [MSIV] closure signals
in BWRs) and containment heat removal and depressurization systems,
including containment spray and fan cooler systems.
(7) Emergency ac electrical power systems, including: emergency
diesel generators (EDGs) and their associated support systems;
hydroelectric facilities used in lieu of EDGs at the Oconee Station;
safety related gas turbine generators; BWR dedicated Division 3 EDGs
and their associated support systems; and station blackout diesel
generators (and black-start gas turbines that serve a similar purpose)
which are started from the control room and included in the plant's
operating and emergency procedures.
(8) Anticipated transient without scram (ATWS) mitigating systems.
(9) Service water (standby emergency service water systems that do
not normally run).
(v) Any event or condition occurring within three years of the date
of discovery that 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 (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 this paragraph 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) 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)(1) The Energy Industry Identification System component function
identifier and system name of each component or system referred to in
the LER.
(i) 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.
(ii) IEEE Std 803-1983 has been approved for incorporation by
reference by the Director of the Federal Register.
(2) 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,
345 East 47th Street, New York, NY 10017. 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; and at the Office of the Federal Register, 1100 L
Street, NW, Washington, DC.
* * * * *
(J) For each human performance related problem that contributed to
the event, 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 the availability of systems or
components that:
[[Page 36307]]
(i) Could have performed the same function as the components and
systems that failed during the event, or
(ii) 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.
* * * * *
(d) Submission of reports. Licensee Event Reports must be prepared
on Form NRC 366 and submitted within 60 days of discovery of a
reportable event or situation 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]
(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 25th day of June, 1999.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99-16934 Filed 7-2-99; 8:45 am]
BILLING CODE 7590-01-P
