[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 [email protected] or his alternate, Bennett M. Brady, 
telephone (301) 415-6363, e-mail [email protected].

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

[[Page 52991]]

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