[Federal Register Volume 63, Number 191 (Friday, October 2, 1998)]
[Proposed Rules]
[Pages 52990-52992]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-26421]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AF98
Reporting Requirements for Nuclear Power Reactors; Meeting
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of public meeting.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is announcing a public
meeting on November 13, 1998 to discuss rulemaking to modify power
reactor reporting requirements.
DATES: Friday, November 13, 1998.
ADDRESSES: The public meeting will be held in the auditorium of NRC's
headquarters at Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Dennis P. Allison, Office for Analysis
and Evaluation of Operational Data, Washington DC 20555-0001, telephone
(301) 415-6835, e-mail dpa@nrc.gov or his alternate, Bennett M. Brady,
telephone (301) 415-6363, e-mail bmb1@nrc.gov.
SUPPLEMENTARY INFORMATION:
Background
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
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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.
This meeting is being conducted in response to that comment.
Purpose
The purpose of the meeting is to test key aspects of the
contemplated amendments to 10 CFR 50.72 and 50.73 for clarity and
consistency, early in the process of drafting them, by discussing how
reportability decisions could be made for example events. This
discussion will provide insights to NRC staff, which can then be used
in drafting the proposed requirements and associated guidance.
Topics
The following topics will be discussed:
Loss of function: As discussed in the ANPR, any design or analysis
defect or deviation that results in a system not being capable of
performing its specified safety functions would be reported pursuant to
10 CFR 50.72(b)(2)(iii) and 50.73(a)(2)(v), ``Any event or condition
that alone could have prevented the fulfillment of the safety function
of structures or systems that are needed to: (A) Shut down the reactor
and maintain it in a safe shutdown condition; (B) Remove residual heat;
(C) Control the release of radioactive material; or (D) Mitigate the
consequences of an accident.'' Comments have raised questions about how
to determine when a system is ``not capable of performing.''
An example relevant to this issue is provided in LER #28997001,
Three Mile Island 1, ``Potential Overpressurization of Piping Between
Closed Reactor Building Isolation Valves Due to Inadequate Design Code
Guidance.'' Stresses for postulated accident conditions would exceed
the allowable values in the design code (ANSI B 31.1-1967). However,
they would remain within the limits of ASME Section III, Appendix F,
which demonstrates that the piping is capable of maintaining
containment integrity (and, as a result, the piping was considered
operable).
Partial loss of function: As discussed in the ANPR, any design or
analysis defect or deviation that results in one train of a multi-train
system not being capable of performing its specified safety functions
for a period of time in excess of that allowed by the plant's TS would
be reported pursuant to 10 CFR 50.73(a)(2)(i)(B), ``Any operation or
condition prohibited by the plant's Technical Specifications.''
Comments have raised questions about how to determine the ``specified
safety function.''
An example relevant to this issue is provided in LER #26697014,
Point Beach 1, ``Auxiliary Feedwater System Inoperability Due to Loss
of Instrument Air.'' It was found that a loss of offsite power could
cause a loss of instrument air and, as a result, auxiliary feedwater
(AFW) flow control valves could fail open. Then for low steam generator
pressure, such as could occur for certain main steam line breaks, high
AFW flow rates could result in tripping the motor driven AFW pumps on
thermal overload. The single turbine driven AFW pump would not be
affected.
Unanalyzed condition that significantly compromised plant safety:
No changes were proposed in the ANPR. However, comments have questioned
the clarity of the current requirement with regard to the meaning of
the term significant.
The two examples relevant to this issue that are provided in the
current guidance in NUREG-1022, Revision 1 are: (a) Accumulation of
voids that could inhibit the ability to adequately remove heat from the
reactor core, particularly under natural circulation conditions and (b)
voiding in instrument lines that results in an erroneous indication
causing the operator to misunderstand the true condition of the plant.
Another relevant example would be an unanalyzed condition that
warrants declaration of an emergency class, such as an unplanned loss
of most or all safety system annunciators for longer than 15 minutes.
Also, a relevant example is provided in LER #24797006, Indian Point
2, ``Open Electric Penetration Area Door Creates Unanalyzed
Condition.'' Equipment in the electrical penetration area was not
qualified on the basis that a closed door would protect the area from a
harsh environment. The door was improperly left open during plant
operation; however, the condition lasted less than 6 hours before it
was discovered and corrected.
Compliance with technical specification surveillance requirements:
As proposed in the ANPR, reporting would be eliminated for events that
consist of late TS required surveillance tests provided there is no
systematic breakdown of compliance with the TS, the oversight is
corrected, the testing is performed, and the equipment is still
functional or, alternately, the requirements of the TS are implemented.
Comments have questioned whether the proposed conditions (i.e.,
``provided there is no systematic non-compliance * * *'') are clear and
appropriate.
One example of an event relevant to this issue would be a case
where review of a surveillance procedure indicates inadequate circuit
overlap, so that a relay has not been included in the testing for some
time. When tested, the relay is functional.
Another relevant example would be a case where review of a
surveillance procedure indicates that a component has not been tested
for some time. When tested, the component is not functional; however,
upon discovery that the component is not operable, the TS action
statements are met by correcting the condition within the allowed time.
A third relevant example would be a case where, because of an
oversight, a surveillance test was not performed within the time
required. This is the third case of a similar oversight in one calendar
quarter.
Condition that alone could prevent fulfillment of a safety
function: In the ANPR it was proposed to clarify this criterion by
revising it to require reporting any event or condition that alone or
in combination with other existing condition(s) 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, etc. However, comments have suggested that the proposed
change would detract from clarity.
An example relevant to this issue is provided in NUREG-1022,
Revision 1. While one EDG was out of service for maintenance, the
second EDG failed its surveillance test (and, as a result, was declared
inoperable).
Nuclear power plant, including its principal barriers, being in a
seriously degraded condition: No changes were proposed in the ANPR.
However, comments have indicated that this criterion is redundant and
should be deleted.
The following guidance and examples are relevant to this issue. The
current guidance in NUREG-1022, Revision 1 states that this criterion
includes material (e.g., metallurgical or chemical) problems that cause
abnormal degradation of the principal safety
[[Page 52992]]
barriers (i.e., the fuel cladding, reactor coolant system pressure
boundary, or the containment) such as:
(a) 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, and would involve a release of
significant quantities of fission products.
(b) Cracks and breaks in the piping or reactor vessel (steel or
prestressed concrete) or major components in the primary coolant
circuit that have safety relevance (steam generators, reactor coolant
pumps, valves, etc).
(c) Significant welding or material defects in the primary coolant
system, such as items 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.''
(d) Serious temperature or pressure transients, such as 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).
(e) Loss of relief and/or safety valve functions during operation.
(f) Loss of containment function or integrity including: (A)
Containment leakage rates exceeding the authorized limits, 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, (B) loss of containment isolation valve
function during tests or operation, (C) loss of main steam isolation
valve function during test or operation, or (D) loss of containment
cooling capability.
Participation
The meeting is scheduled for 9 a.m. to 3:15 p.m. and is open to the
general public. Interested individuals may address relevant remarks or
comments to the NRC staff at the meeting. To facilitate the scheduling
of available time for and orderly conduct of the meeting, members of
the public who wish to request the opportunity to speak and/or
introduce particular examples for discussion should contact the
cognizant NRC staff member listed in the For Further Information
Contact section before the meeting. Indicate as specifically as
possible the topic(s) of your comment and/or the example(s) you wish to
introduce. Provide your name and a telephone number at which you can be
reached, if necessary, before the meeting.
Agenda for November 13, 1998
9:00 a.m.-9:30 a.m. Introductory remarks by NRC staff members
9:30 a.m.-10:00 a.m. Introductory comments by industry representatives
and members of the general public
10:00 a.m.-12:00 noon Discussion among NRC staff members and public on
how reportability decisions could be made for example events
12:00 noon-1:00 p.m. Lunch Break
1:00 p.m.-3:00 p.m. Continued discussion on how reportability
decisions could be made for example events
3:00 p.m.-3:15 p.m. Concluding remarks
Note that the discussions may be completed earlier than indicated
and, if so, the meeting will be concluded earlier.
Dated at Rockville, Maryland, this 25th day of September, 1998.
For the Nuclear Regulatory Commission.
Patrick W. Baranowsky,
Acting Director, Safety Programs Division, Office for Analysis and
Evaluation of Operational Data.
[FR Doc. 98-26421 Filed 10-1-98; 8:45 am]
BILLING CODE 7590-01-P
