[Federal Register Volume 60, Number 238 (Tuesday, December 12, 1995)]
[Notices]
[Pages 63737-63739]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-30176]



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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-309, 50-285, 50-317, 50-318, 50-336, and 50-335]


Maine Yankee Atomic Power Co., Omaha Public Power District, 
Baltimore Gas and Electric Co., Northeast Nuclear Energy Co., and 
Florida Power & Light Co.; Maine Yankee, Fort Calhoun Unit 1, Calvert 
Cliffs Units 1 and 2, Millstone Unit 2, and St. Lucie Unit 1; 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 2, 
1995, by Mr. John F. Doherty, J.D. (Petition for action under 10 CFR 
2.206). The Petition pertains to the following plants: Maine Yankee, 
Fort Calhoun Unit 1, Calvert Cliffs Units 1 and 2, Millstone Unit 2, 
and St. Lucie Unit 1.
    In the Petition, Petitioner requested that the following six 
pressurized-water reactors be immediately shut down: Maine Yankee, Fort 
Calhoun Unit 1, Calvert Cliffs Units 1 and 2, Millstone Unit 2, and St. 
Lucie Unit 1. In addition, the Petitioner requested that steam 
generator tubes be inspected immediately at those plants.
    The Director of the 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-
95-22), 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, DC.
    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 6th day of December, 1995.

    For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.

Office of Nuclear Reactor Regulation, William T. Russell, Director

    In the Matter of: Maine Yankee Atomic Power Co., Omaha Public 
Power District, Baltimore Gas and Electric Co., Northeast 

[[Page 63738]]
Nuclear Energy Co., Florida Power & Light Co. (Maine Yankee, Fort 
Calhoun Unit 1, Calvert Cliffs Units 1 and 2, Millstone Unit 2, and 
St. Lucie Unit 1). Docket Nos. 50-309, 50-285, 50-317, 50-318, 50-
336, and 50-335. License Nos. DPR-36, DPR-40, DPR-53, DPR-69, DPR-
65, DPR-67.

I. Introduction

    On May 2, 1995, Mr. John F. Doherty, J.D. (Petitioner), filed a 
Petition with the U.S. Nuclear Regulatory Commission (NRC) pursuant to 
10 CFR 2.206. The Petitioner requested that the following six 
pressurized-water reactors be immediately shut down: Maine Yankee, Fort 
Calhoun Unit 1, Calvert Cliffs Units 1 and 2, Millstone Unit 2, and St. 
Lucie Unit 1. In addition, the Petitioner requested that steam 
generator tubes be inspected immediately at those plants. The 
Petitioner stated that an inspection by the license in April 1995 of 
the Maine Yankee plant using the newly developed Point Plus system 
revealed that the steam generator tubes are on the verge of rupture, 
threatening the release of radioactive liquid and gaseous material into 
the environment and consequent harm to human health and safety. Because 
the other plants the Petitioner identified were built by the same 
manufacturer (Combustion Engineering) and are of similar operating age, 
the Petitioner asked that they, along with the Maine Yankee, be 
immediately shut down and that all steam generator tubes be immediately 
inspected using the Point Plus Probe system.
    On June 28, 1995, I informed the Petitioner that the Petition had 
been referred to my office for preparation of a Director's Decision. I 
further informed the Petitioner that his request for immediate shutdown 
and inspection was denied because continued operation of these units 
until their next scheduled outage posed no undue risk to public health 
and safety. I also informed the Petitioner that the NRC would take 
appropriate action within a reasonable time.

II. Discussion

    The Petitioner requested that six CE-designed plants be shut down 
and their steam generator tubes inspected with the Plus Point 
inspection probe. The request appears to be based on concerns that 
without inspections using the Plus Point probe, the steam generators in 
these plants may be susceptible to one or more steam generator tube 
ruptures (SGTRs). However, the results of examinations of tubes removed 
from the Maine Yankee steam generators and in situ pressure tests of 
the most severely degraded tubes in the Maine Yankee steam generators 
have demonstrated that the tubes, although severely degraded, still had 
a significant margin before failure even under postulated accident 
conditions. Furthermore, the NRC has taken actions to ensure that other 
plants have performed appropriate steam generator tube inspections to 
assure tube integrity. These important actions are discussed below in 
greater detail.
    The NRC applies a defense-in-depth approach toward protecting 
public health and safety from the potential consequences of events 
involving the rupture of steam generator tubes. Steam generator tube 
degradation is managed through a combination of several different 
elements, including inservice inspection, tube repair criteria, 
primary-to-secondary leak rate monitoring, water chemistry, and 
analyses to ensure safety objectives are met.
    The primary means for assessing steam generator tube degradation is 
through inservice inspections. Plant technical specifications require a 
periodic inspection of the steam generator tubes. Any tubes with 
identified degradation in excess of the repair criteria are repaired or 
removed from service. In order to assess the condition of steam 
generator tubing, the industry primarily relies on eddy current 
inspection techniques, which includes the motorized rotating pancake 
coil (MRPC) test. Circumferential cracking in steam generator tubing 
has been identified at expansion transitions, small radius U-bends, 
dented tube support plate intersections, and sleeved joints. Based on 
the utilities' responses to GL 95-03, the inservice CE steam generators 
(i.e., not including retired CE steam generators) have been inspected 
in these areas with techniques capable of detecting circumferential 
cracking and, to date, such cracking was found only at the expansion 
transitions.
    Experience to date, including experience at the Maine Yankee plant, 
shows that the standard MRPC probe is a reliable means for detecting 
structurally significant cracking in steam generator tubes. The use of 
an MRPC probe in conjunction with adequate inspection procedures is a 
reliable means for detecting circumferential cracking in steam 
generator tubes. As discussed above, metallographic examinations of 
removed tubing and in situ pressure testing of degraded tubes continue 
to support the staff's conclusion that properly conducted MRPC 
inspections can identify circumferential cracking before the cracking 
exceeds the structural limits.
    In addition to requiring periodic steam generator tube inspections, 
the NRC requires an operational leak rate limit to provide reasonable 
assurance that should a primary-to-secondary leak be experienced during 
service, it will be detected and the plant will be shut down in a 
timely manner before rupture occurs and with no undue risk to public 
health or safety. Requiring operation within these limits decreases the 
possibility that steam generators may be vulnerable to tube ruptures 
during postulated accidents such as a main steamline break or a loss-
of-coolant accident.
    Inspection findings at Maine Yankee in 1994 revealed indications of 
large circumferential cracks that had been missed in previous 
inspections because of inadequacies in MRPC test and analysis 
procedures. The test and analysis procedures were upgraded accordingly. 
However, subsequent inspections at Maine Yankee performed with the MRPC 
in early 1995 revealed circumferential indications that were more 
numerous and larger than expected based on the short operating interval 
since the previous inspection. The 100-percent MRPC inspection of the 
expansion transitions were supplemented by inspections with the 
recently developed Plus Point probe and a specially wound high-
frequency MRPC coil. These latter probes offer improved sensitivity to 
inner-diameter-initiated circumferential cracks of the type present at 
the Maine Yankee expansion transitions and identified substantial 
numbers of relatively small circumferential cracks not detected with 
the conventional MRPC.
    Three tubes were removed from these steam generators in early 1995. 
Before the tubes were removed, they were tested by ultrasonic, visual 
(fluorescent penetrant dye), and eddy current techniques to confirm the 
nature of the indications. Eddy current methods included examination 
with a standard rotating pancake coil, a Plus Point coil, and a high-
frequency pancake coil. The indications were sized with various 
techniques and the tubes were then destructively examined so that the 
actual size of the indications could be determined. The results of the 
destructive examinations are provided in NRC Information Notice 95-40, 
``Supplemental Information Pertaining to Generic Letter 95-03, 
`Circumferential Cracking of Steam Generator Tubes.' '' The destructive 
examination results and data obtained with a high-frequency pancake 
coil suggest that many of the indications may not have been as 
structurally significant as the standard pancake coil appeared to 
indicate.

[[Page 63739]]

    In situ pressure tests were conducted on the tubes with the largest 
MRPC indications and the results indicate acceptable margins against 
burst under normal operating and postulated accident conditions. The 
NRC had a review conducted by an independent contractor of the in situ 
test method used at Maine Yankee and determined that it provides a 
reasonable simulation of the hydraulic pressure loads induced during a 
postulated main steamline break.
    Thus, it has been demonstrated that the tubes with the largest 
indications at Maine Yankee continued to exhibit adequate structural 
integrity at the time they were found. This finding is attributable to 
the morphology of the cracks as determined from metallographic 
examinations of pulled tube specimens from Maine Yankee. This 
morphology consists of cracks that were not coplanar but rather of 
short circumferential length and staggered around the circumference 
over a short axial region with ligaments of material between the 
cracks. These ligaments add considerably to the strength of the tube, 
but these ligaments are generally not detectable by the MRPC.
    The findings at Maine Yankee nevertheless raised concern that large 
undetected circumferential cracks could possibly exist at other plants. 
Therefore, the NRC issued Generic Letter (GL) 95-03, ``Circumferential 
Cracking of Steam Generator Tubes,'' on April 28, 1995, notifying 
licensees of the Maine Yankee experience and requesting that they 
evaluate recent operating experience concerning the detection and 
sizing of circumferential cracks and the potential applicability of 
this experience to their plants. On the basis of the results of this 
evaluation, past inspections and the results thereof, and other 
relevant factors, licensees were requested to develop a safety 
assessment justifying continued operation until the next scheduled 
steam generator tube inspections were to be performed. The generic 
letter also requested that licensees develop and submit their plans for 
the next steam generator tube inspection as they pertain to the 
detection of circumferential cracks. The utilities were required to 
respond to GL 95-03 within 60 days. By now, the utilities that own the 
six plants listed in the Petition have responded to GL 95-03 and the 
responses have been evaluated by the staff.
    Based on the utilities' responses to GL 95-03, with the exception 
of Millstone Unit 2, the CE plants listed in the Petition have been 
inspected in those areas susceptible to circumferential cracking with 
improved eddy current inspection probes equally capable as the Point 
Plus system of detecting circumferential cracking. All tubes with 
detected cracks have been removed from service. The licensee for 
Millstone Unit 2 replaced the original CE steam generators during an 
outage that ended in January 1993. The new steam generators 
incorporated many new design features that are expected to eliminate or 
greatly reduce the potential for circumferential tube cracking. These 
include the use of Inconel 690, a material that has significantly 
greater resistance to cracking and hydraulic expansion of tubes, which 
reduces the potential for cracking in the expansion transitions. The 
limited operational time, improvements in design, and favorable plant 
operating conditions minimize the potential for the development of 
circumferential cracking in the Millstone Unit 2 steam generators. 
Millstone Unit 2 steam generators will continue to be inspected during 
refueling outages.
    The NRC has studied the risk and potential consequences of a range 
of SGTR 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.'' The staff estimated the risk 
contribution due to the potential for single and multiple SGTRs. The 
study also examined the expected consequences of SGTR scenarios, 
including beyond-design-basis situations, such as the potential for 
release as a result of containment bypass because of failed tubes 
concurrent with a breach of secondary system integrity. A combination 
of circumstances and conditions is required to produce such 
simultaneous failures: (1) Main steamline break or other less severe 
loss of secondary system integrity, (2) the potential that a population 
of tubes susceptible to rupture exists in a particular steam generator, 
(3) the potential that operators would not take actions to avoid high 
differential pressures, and (4) the probability that a large number of 
tubes would actually fail simultaneously. In the NUREG -0844 
assessment, the staff concluded that the probability of simultaneous 
multiple tube failure was small (approximately 10<SUP>-5), and that the 
risk resulting from releases during SGTRs with loss of secondary system 
integrity was small (about 10<SUP>-7 latent fatalities per reactor 
year).

III. Conclusion

    Based on the fact that (1) adequate steam generator tube 
inspections have been performed, (2) primary-to-secondary leakage is 
being monitored on a continuing basis, and (3) the risk of multiple 
SGTR events is low, I have concluded that an immediate shutdown and 
Plus Point probe inspection of Maine Yankee, Fort Calhoun Unit 1, 
Calvert Cliffs Units 1 and 2, St. Lucie Unit 1, and Millstone Unit 2 is 
not 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 6th day of December , 1995.

    For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 95-30176 Filed 12-11-95; 8:45 am]
BILLING CODE 7590-01-P-M