[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
[[Page 30644]]
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
[[Page 30645]]
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
