[Federal Register Volume 61, Number 227 (Friday, November 22, 1996)]
[Notices]
[Pages 59469-59472]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-29898]


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NUCLEAR REGULATORY COMMISSION
[Dockets Nos. 50-335 and 50-389]


Florida Power & Light Co., St. Lucie, Units 1 and 2; 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 for action under 
10 CFR 2.206 dated June 12, 1996, by Mr. Thomas J. Saporito, Jr. and on 
behalf of the National Litigation Consultants. The Petition pertains to 
St. Lucie, Units 1 and 2.
    The Petitioners requested the Commission (1) to issue a 
confirmatory order requiring that the Florida Power and Light Company 
(Licensee) not operate the St. Lucie Nuclear Station, Unit 1 above 50% 
of its power level

[[Page 59470]]

capacity, (2) to require the Licensee to specifically identify the 
``root cause'' for the premature failure of the steam generator tubing, 
and (3) to require the Licensee to specifically state what corrective 
measures will be implemented to prevent recurrence of steam generator 
tube failures in all the steam generators in Unit 1 and Unit 2.
    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-
96-19), 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 18th day of November 1996.

    For the Nuclear Regulatory Commission.

Frank J. Miraglia, Jr.,
Acting Director, Office of Nuclear Reactor Regulation.

Director's Decision Under 10 CFR 2.206

I. Introduction

    On June 12, 1996, Mr. Thomas J. Saporito, Jr., on behalf of himself 
and the National Litigation Consultants (Petitioners), filed a Petition 
with the U.S. Nuclear Regulatory Commission (NRC or Commission) 
pursuant to 10 CFR 2.206. The Petitioners requested the Commission (1) 
to issue a confirmatory order requiring that the Florida Power & Light 
Company (FP&L or licensee) not operate St. Lucie Plant, Unit 1, above 
50 percent of its power-level capacity, (2) to require the Licensee to 
specifically identify the ``root cause'' for the premature failure of 
the steam generator tubing, and (3) to require the licensee to 
specifically state what corrective measures will be implemented to 
prevent recurrence of steam generator tube failures in all the steam 
generators in Unit 1 and Unit 2.
    The Petitioners' requests are based on assertions that (1) the 
licensee's Unit 1 steam generator tubes have degraded to the extent 
that more than 2,500 of the tubes have been plugged, (2) the licensee 
has not identified the root cause for the premature failure of the 
steam generator tubing, (3) the licensee will most likely experience 
similar tube ruptures on other steam generators at the station, and (4) 
the licensee's ``FSAR's [Final Safety Analysis Reports] and the NRC's 
CFR's [Code of Federal Regulations] require that the integrity of the 
primary systems on Unit 1 and Unit 2 not be breached.
    The Petition has been referred to my office pursuant to 10 CFR 
2.206 of the Commission's regulations. By letter dated July 8, 1996, an 
acknowledgement of receipt of the Petition was sent to the Petitioners. 
In that letter, the Petitioners were informed that the NRC would take 
appropriate action within a reasonable time. I have completed my 
evaluation of the matters raised by the Petitioners and have determined 
that, for the reasons stated below, the Petition is denied.

II. Discussion

    The NRC staff's evaluation of the Petitioners' requests follows.
    (a) Issue a confirmatory order requiring that the licensee not 
operate Unit 1 above 50 percent of its power-level capacity.
    In a meeting held at NRC Headquarters on July 3, 1996, the licensee 
presented the inspection and repair history for the Unit 1 steam 
generator tubes.1 The licensee has performed 15 inspections since 
commercial operation began in December 1976. For the most recent 
inspection, completed in June 1996, the licensee inspected the full 
length of all active tubes using a bobbin coil.2 In addition, the 
licensee used a motorized rotating pancake coil 3 (MRPC) to 
inspect all expansion transition joints and drilled support 
intersections in the hot and cold legs, all free-span locations having 
bobbin coil indications,4 and free-span tube regions in the upper 
two support areas in the hot legs. The inspection was based on the 
Electric Power Research Institute (EPRI) report ``PWR Steam Generator 
Examination Guidelines,'' dated November 1992. Defective tubes having 
circumferential indications, axial indications, or volumetric 
indications 5 were plugged and removed from service.
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    \1\ NRC Meeting Summary, Subject: ``Steam Generator Inspection, 
Repair and Operating Issues--St. Lucie Unit 1,'' dated July 16, 
1996.
    \2\ The bobbin coil is used for a general screening of tubes for 
indications of possible defects, while the motorized rotating 
pancake coil (MRPC) probe is used to further characterize bobbin 
coil indications. The MRPC is also used to inspect regions 
susceptible to circumferentially orientated degradation.
    \3\ See note 2.
    \4\ See note 2.
    \5\ Circumferential indications are crack-like indications 
orientated on the diameter of the tube. Axial indications are crack-
like indications orientated on the long axis of the tube. Volumetric 
indications are areas of general reduction in tube wall thickness 
with no specific orientation.
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    Including tubes plugged during earlier outages, 2,159 of 8,519 
tubes (25.3 percent) in the ``A'' steam generator and 1,834 of 8,519 
tubes (21.5 percent) in the ``B'' steam generator have been plugged and 
removed from service. The licensee performed an evaluation that showed 
that the plant could be safely operated at full power with the reduced 
reactor coolant flow resulting from the increased number of plugged 
tubes.6 The NRC reviewed the licensee's evaluation and concluded 
that it was acceptable and that the units could be operated at full 
power. The staff's evaluation is documented in a safety evaluation 
dated July 9, 1996.
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    \6\  FP&L letter, ``Thermal Margin and RCS Flow Limits,'' dated 
June 1, 1996.
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    In the meeting on July 3, 1996, the licensee presented a 
preliminary run-time analysis for Unit 1, which was used to determine 
the length of steam generator operation before the need for further 
tube inspections to ensure adequate tube integrity. The licensee stated 
that the preliminary results of its analysis support a tube inspection 
interval of 15 months for the current Unit 1 cycle that started in July 
1996. The licensee also stated that in situ pressure testing of the 
steam generator tubes during the spring 1996 outage indicated that the 
most severely degraded tubes had adequate structural integrity and 
satisfied the safety margins in NRC's Regulatory Guide 1.121, ``Bases 
for Plugging Degraded PWR Steam Generator Tubes.'' On the basis of the 
results of the in situ pressure tests, the staff concluded that 
adequate assurance of tube integrity existed to allow operation pending 
completion of the licensee's run-time analysis. The NRC is currently 
reviewing the licensee's analysis, which was submitted October 24, 
1996.
    The plant Technical Specifications for each of the units specify 
leakage limits for the reactor coolant pressure boundary, including 
steam generator tube leakage. If a tube leaks beyond the allowed 
limits, the unit must be shut down. The plant off-normal operating 
procedures for St. Lucie Units 1 and 2 also include criteria for 
shutdown based on EPRI TR-104788, ``PWR Primary to Secondary Leak 
Guidelines,'' dated May 1995, which are more conservative than the 
limits in the plant Technical Specifications. Finally, if a tube fails, 
the plant's Emergency Operating Procedures contain the specific actions 
necessary for the operators to shut down

[[Page 59471]]

and cool down the plant to mitigate the consequences of the event.
    Thus, as required, the licensee has implemented measures for both 
units to protect public health and safety in the unlikely event that 
tube integrity is compromised. These measures include a primary-to-
secondary leakage monitoring program and emergency operating 
procedures. The leakage monitoring program provides early warning of 
tube leakage. The steam generator blowdown monitor and condenser air 
ejector monitor at each of the units continuously monitors the 
radioactivity level in the main steamline. A significant increase in 
the instrument readings, which would result from a relatively small 
tube leak, will cause an alarm to alert the operators to the change in 
radioactivity levels and potential tube leakage.
    On the basis of the information submitted, the NRC staff has 
concluded that the operation of the Unit 1 steam generators at full 
power poses no undue risk to public health and safety.
    (b) Require the licensee to specifically identify the ``root 
cause'' for the premature failure of the steam generator tubing.
    It is not clear how the Petitioners define ``premature failure''; 
however, since there have not been any steam generator tube ruptures at 
St. Lucie Units 1 or 2, it is assumed the reference is to tube 
degradation. Many of the tubes in the Unit 1 steam generators have 
degraded as a result of corrosion and/or mechanical conditions. The 
root cause of tube degradation in steam generators is the interaction 
of water chemistry, thermal-hydraulic design, materials selection, 
fabrication methods, and operating conditions. The causes of tube 
degradation are well understood by the industry and are documented in 
the public record. The root causes for the St. Lucie steam generator 
tube degradations were presented to the NRC staff in a meeting on 
August 27, 1986.7
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    \7\ NRC Meeting Summary, Subject: ``Summary of August 27, 1986 
Meeting with FP&L and NRC Staff Regarding Steam Generator Tube 
Degradation Mechanism,'' dated September 12, 1986.
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    The licensee has identified to the NRC modes of degradation that 
have affected the steam generator tubes in both St. Lucie Units 1 and 2 
in its response of June 23, 1995, to NRC Generic Letter 95-03, 
``Circumferential Cracking of Steam Generator Tubes,'' and in the 
meeting of July 3, 1996. The degradation modes identified include 
intergranular attack, stress-corrosion cracking and denting. 
Intergranular attack refers to localized attack at and adjacent to 
grain boundaries of tube material, with relatively little corrosion of 
the grains. Intergranular stress-corrosion cracking refers to cracking 
caused by the simultaneous presence of stress and a specific corrosive 
medium. Denting is the accumulation of corrosion products at the tube-
to-tube support plate that causes plastic deformation of the tube. The 
licensee has identified locations of these degradations in the tubes 
during the most recent steam generator inspection of St. Lucie Unit 
1.8 They include egg crate and drilled tube support plates, free 
spans, expansion transition regions, and sludge pile areas. In every 
case, the root cause of tube degradation can be attributed to material 
selection, water chemistry, fabrication methods, or residual stresses 
at the affected location.
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    \8\  See note 1.
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    The staff concludes that the licensee understands and has 
identified the root cause of tube degradation at St. Lucie Units 1 and 
2.
    (c) Require the licensee to specifically state what corrective 
measures will be implemented to prevent recurrence of steam generator 
tube failures in all the steam generators in Unit 1 and Unit 2.
    As previously discussed, degradation of the steam generator tubing 
is caused by the interaction of water chemistry, thermal-hydraulic 
design, materials selection, fabrication methods, and operating 
conditions. The licensee has applied corrective measures in order to 
reduce the rate of tube degradation. For example, the rate of tube 
degradation may be reduced through improvements in water chemistry. The 
licensee follows industry guidelines 9 on secondary water 
chemistry for both units, and these guidelines represent a significant 
improvement over the guidelines followed when Unit 1 began operating. 
The guidelines have stringent requirements and limitations on specific 
types and amounts of chemicals in the primary and secondary water to 
mitigate corrosion. Replacement steam generators having improved 
design, for example, better material selection and tube support 
configuration, have had much better operating experience than the 
earlier steam generators, such as those at St. Lucie. The licensee 
plans to replace the Unit 1 steam generators in October 1997 with steam 
generators that incorporate these design improvements.
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    \9\  FP&L letter, ``Generic Letter 95-03 Response,'' dated June 
23, 1995.
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    The NRC staff focuses on ensuring adequate tube integrity by 
requiring licensee compliance with applicable regulations and Technical 
Specification requirements. The staff uses its field inspections, 
meetings with the licensee, and licensing reviews to ensure that the 
licensee satisfies the regulations 10 and plant Technical 
Specifications as they apply to steam generator tube integrity and that 
appropriate inspection methods and repair criteria are used to address 
specific forms of degradation. Plant Technical Specifications define 
degraded and defective tubes, specify the scope of inspections and 
reporting requirements and set forth tube plugging criteria and limits 
for allowable leakage in the reactor coolant system. NRC regulations 
and plant Technical Specifications require that steam generator tube 
degradation be managed through a combination of inservice inspection, 
repair of tubes exceeding the plugging criteria in the plant Technical 
Specifications, primary-to-secondary leakage monitoring, and structural 
and run-time analyses to ensure that safety objectives are met. On the 
basis of the information provided by the licensee in the meeting on 
July 3, 1996, and the staff's onsite inspection, the staff has 
concluded that the licensee is in compliance with these requirements.
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    \10\  The NRC regulations that require steam generator tube 
integrity be maintained include 10 CFR Part 50, Appendix A, General 
Design Criteria for Nuclear Power Plants, Criterion 1--Quality 
Standards and Records, Criterion 14--Reactor Coolant Pressure 
Boundary, Criterion 30--Quality of Reactor Coolant Pressure 
Boundary, Criterion 31--Fracture Prevention of Reactor Coolant 
Pressure Boundary, and Criterion 32--Inspection of Reactor Coolant 
Pressure Boundary; 10 CFR Part 50, Appendix B, Quality Assurance 
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants; and 
10 CFR Part 50.55a, which specifies codes and standards for nuclear 
power plants.
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    In summary, the licensee's corrective measures to reduce the rate 
of steam generator tube degradation and continued compliance with NRC 
regulations and plant Technical Specification requirements provide 
reasonable assurance that steam generator tube integrity at St. Lucie 
Units 1 and 2 will be maintained.

III. Conclussion

    On the basis of the fact that (1) the licensee has performed 
adequate steam generator tube inspections that identified areas of 
degradation, (2) the licensee has completed analyses and repairs of 
degraded tubes, (3) the licensee's in situ pressure testing of degraded 
tubes indicated adequate structural integrity remains, (4) the licensee 
is monitoring primary-to-secondary leakage on a continuing basis, and 
(5) the licensee is complying with NRC regulations and plant Technical 
Specifications, I have concluded that a confirmatory order limiting St. 
Lucie Unit 1 to 50 percent of its power-level capacity is not warranted 
and that the

[[Page 59472]]

licensee has identified the root cause of tube degradation and 
implemented adequate corrective measures to provide reasonable 
assurance that steam generator tube integrity will be maintained at St. 
Lucie Units 1 and 2.
    For the reasons previously discussed, no basis exists for taking 
any further action in response to the Petition. 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 18th day of November 1996.

    For the Nuclear Regulatory Commission.
Frank J. Miraglia, Jr.,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-29898 Filed 11-21-96; 8:45 am]
BILLING CODE 7590-01-P