Reporting Requirements for Nuclear Power Reactors; Meeting

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

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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

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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]

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