[Federal Register Volume 61, Number 117 (Monday, June 17, 1996)]
[Notices]
[Pages 30643-30645]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-15262]



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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-247 and 50-286]


Consolidated Edison Company of New York; Indian Point Nuclear 
Generating Units 2 and 3; Issuance of Director's Decision Under 10 CFR 
2.206

    Notice is hereby given that the Director, Office of Nuclear Reactor 
Regulation, has taken action with regard to a Petition dated May 18, 
1995, by Ms. Connie Hogarth (Petition for action under 10 CFR 2.206). 
The Petition pertains to Indian Point Nuclear Generating Units 2 and 3.
    In the Petition, the Petitioner requested that the operating 
licenses for Indian Point Units 2 and 3 be suspended until the 
licensees have completed the actions requested by Generic Letter 95-03. 
The Petitioner also requested that the U.S. Nuclear Regulatory 
Commission hold a public meeting in the vicinity of the plant to 
explain its response to this request.
    The Director, Office of Nuclear Reactor Regulation, has determined 
to deny the Petition. The reasons for this denial are explained in the 
``Director's Decision Pursuant to 10 CFR 2.206'' (DD-96-06), the 
complete text of which follows this notice, and is available for public 
inspection at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, D.C.
    A copy of the Decision will be filed with the Secretary of the 
Commission for the Commission's review in accordance with 10 CFR 
2.206(c) of the Commission's regulations. As provided by this 
regulation, the Decision will constitute the final action of the 
Commission 25 days after the date of issuance unless the Commission, on 
its own motion, institutes a review of the Decision within that time.

    Dated at Rockville, Maryland, this 10th day of June 1996.

    For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
ATTACHMENT TO ISSUANCE OF DIRECTOR'S DECISION UNDER 10 CFR 2.206-96-06

Director's Decision Under 10 CFR 2.206

I. Introduction

    On May 18, 1995, Ms. Connie Hogarth (Petitioner) filed a Petition 
with the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 
2.206. The Petitioner requested that the operating licenses for Indian 
Point Nuclear Generating Units 2 and 3 be suspended until the licensees 
have completed the actions requested by Generic Letter (GL) 95-03, 
``Circumferential Cracking of Steam Generator Tubes.'' The Petitioner 
also requested that the NRC hold a public meeting to explain its 
response to the suspension request.
    The Petitioner stated that the impetus for GL 95-03 was the 
discovery at the Maine Yankee plant of steam generator tube cracks that 
had previously gone undetected due to inadequate inspection procedures. 
The Petitioner also stated that while GL 95-03 calls for comprehensive 
examination of steam generator tubes, it appears to allow licensees to 
postpone their evaluations until the next scheduled inspection.
    On June 16, 1995, I informed the Petitioner that the Petition had 
been

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referred to my office for preparation of a Director's Decision. I 
informed the Petitioner that her request for immediate suspension of 
the operating licenses of Indian Point Nuclear Generating Units 2 and 3 
was denied because the continued operation of these units posed no 
undue risk to public health and safety. I further informed the 
Petitioner that her request for a public meeting to explain the denial 
of her request for license suspension was denied, primarily because the 
NRC assessment of risk associated with steam generator tube rupture 
events has already been articulated in public documents.

II. Discussion

    The Petitioner requested that the operating licenses for Indian 
Point Nuclear Generating Units 2 and 3 be suspended until the licensees 
have completed the actions required by GL 95-03. The Petitioner's 
request appears to be based on her belief that without the immediate 
completion of the requested actions of GL 95-03, the steam generators 
in Indian Point Nuclear Generating Units 2 and 3 could be susceptible 
to one or more steam generator tube ruptures brought about by existing 
circumferential cracks.
    Generic Letter 95-03 was issued on April 28, 1995, after Maine 
Yankee shut down due to primary-to-secondary leakage through 
theretofore undetected circumferential steam generator tube cracks. The 
generic letter was intended to alert licensees to the importance of 
performing steam generator inspections with equipment capable of 
detecting degeneration to which the steam generator tubes are 
susceptible. GL-95-03 requested three actions of licensees of 
pressurized water reactors. It requested (1) that they evaluate their 
operating experience to determine whether or not they could have a 
circumferential cracking problem, (2) that based on this evaluation 
they develop a safety assessment justifying continued operation until 
the next scheduled steam generator tube inspection, and (3) that they 
develop a plan for inspecting for circumferential cracking during the 
next steam generator tube inspection.
    Stress corrosion cracking of the Indian Point Unit 2 steam 
generator tubes was first detected during the 1993 refueling outage. 
During the 1995 refueling outage Unit 2 conducted a steam generator 
inspection as required by their technical specifications; this 
inspection included a complete examination of all areas deemed most 
susceptible to circumferential cracking. This inspection, which used 
enhanced techniques and eddy current probes sensitive to indications of 
circumferential cracking, identified 114 tubes with potential 
circumferential crack indications; however, these may actually have 
been closely spaced axial indications. Since the licensee could not 
conclusively determine that these 114 tubes did not contain indications 
of circumferential cracks the worst case was assumed, that is, that the 
indications were in fact circumferential. The indications were logged 
as circumferential and all of these tubes were removed from service 
before the unit was restarted. All of the logged circumferential 
indications were deep within the tubesheet. The fact that the 
indications were all within the tubesheet is significant since, if a 
circumferential failure were to occur at this location, the structural 
strength lent to the tubes by the tubesheet would reduce the amount of 
primary-to secondary leakage. The licensee for Indian Point Unit 2 will 
continue to use inspection techniques capable of detecting 
circumferentially oriented tube degradation.
    Because pitting corrosion had caused deterioration of the Indian 
Point Unit 3 steam generators, they were replaced in 1989 with steam 
generators designed and fabricated to reduce the possibility of 
corrosion-related problems; specifically, the new generators have tubes 
made of thermally treated Alloy 690. Four other nuclear plants in the 
United States have thermally treated Alloy 690 tubes and to date 
neither Indian Point Unit 3 nor any of the other four units have 
experienced tube cracks.
    Circumferential cracking of steam generator tubes is accompanied by 
other forms of tube degradation that are readily detected by bobbin 
coil inspections. Since the bobbin coil inspections at Indian Point 3 
have detected no service induced tube degradation, the staff has 
concluded that Indian Point 3 does not have a circumferential tube 
cracking problem. Indian Point 3 has not yet experienced steam 
generator tube degradation; nevertheless, the licensee has committed to 
performing an augmented inspection for indications of circumferential 
cracking during the next scheduled steam generator inspection. Unit 3 
is currently operating and this inspection is required by May 1997.
    The requirements placed on licensees to ensure steam generator tube 
integrity go beyond the requested actions of GL-95-03. Steam generator 
tube degradation is dealt with through a combination of inservice 
inspection, tube plugging and repair criteria, primary-to-secondary 
leak rate monitoring, and water chemistry analysis. In addition to the 
steam generator inspections required by their technical specifications, 
both Indian Point Nuclear Generating Units 2 and 3 are required to 
monitor primary-to-secondary leakage to ensure that, in the event that 
steam generator tubes begin to leak, operators will be able to bring 
the plant to a depressurized condition before a tube ruptures. In 
addition, both units are required to implement secondary water 
chemistry management programs that are designed to minimize steam 
generator tube corrosion.
    The layers of protection that licensees are required to implement 
make multiple steam generator tube ruptures unlikely events. The NRC 
issued the results of its study of the risk and potential consequences 
of a range of steam generator tube rupture events in NUREG-0844, ``NRC 
Integrated Program for the Resolution of Unresolved Safety Issues A-3, 
A-4, and A-5 Regarding Steam Generator Tube Integrity'' dated September 
1988. The staff estimated the risk contribution due to the potential 
for multiple steam generator tube ruptures. A combination of 
circumstances is required to produce such failures, specifically: (1) A 
main steam line break or other loss of secondary system integrity, (2) 
the existence of a large number of tubes susceptible to rupture in a 
particular steam generator, (3) the failure of operators to take action 
to avoid high differential pressure, and (4) the actual simultaneous 
rupture of a large number of tubes. In the NUREG-0844 assessment, the 
staff concluded that the probability of simultaneous multiple tube 
failure was small (approximately 10-5), and the risk resulting 
from releases during steam generator tube ruptures with loss of 
secondary system integrity was also small.

III. Conclusion

    Based on the facts that (1) adequate steam generator tube 
inspections have been performed at both Indian Point Nuclear Generating 
Units 2 and 3, (2) Unit 2 steam generator tubes that showed signs of 
circumferential cracking have been removed from service, (3) Unit 3 
steam generator tubes show no sign of service induced corrosion, (4) 
Items (1), (2), and (3) above collectively constitute an acceptable 
response to the requested actions of GL-95-03 for both units, (5) 
operational limits are placed on primary to secondary leakage, (6) the 
risk of multiple steam generator tube rupture events is small, and (7) 
the NRC assessment of risk associated with steam generator tube rupture 
events has already been articulated in public

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documents (NUREG-0844 and GL 95-03), I have concluded that neither the 
suspension of the licenses of Indian Point Nuclear Generating Units 2 
and 3 nor the holding of a public meeting to explain this decision is 
warranted.
    The Petitioner's request for action pursuant to 10 CFR 2.206 is 
denied. As provided in 10 CFR 2.206(c), a copy of the Decision will be 
filed with the Secretary of the Commission for the Commission's review. 
This Decision will constitute the final action of the Commission 25 
days after issuance unless the Commission, on its own motion, 
institutes a review of the Decision within that time.

    Dated at Rockville, Maryland, this 10th day of June 1996.

    For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-15262 Filed 6-14-96; 8:45 am]
BILLING CODE 7590-01-P