[Federal Register Volume 61, Number 116 (Friday, June 14, 1996)]
[Notices]
[Pages 30456-30470]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-15149]



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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-277 and 50-278 (10 CFR 2.206)]


PECO Energy Company, (Peach Bottom Atomic Power Station, Unit 
Nos. 2 and 3; Final Director's Decision Under 10 CFR 2.206

I. Introduction

    On October 6, 1994, the Maryland Safe Energy Coalition (Petitioner) 
issued a press release describing its concerns with the operation of 
PECO Energy Company's Peach Bottom Atomic Power Station (PBAPS). In the 
press release, the Petitioner requested that the U.S. Nuclear 
Regulatory Commission (NRC) take action to address those concerns. The 
Petitioner requested the NRC, among other things, to immediately shut 
down both reactors at Peach Bottom and keep them shut down until 
certain conditions are corrected. Specifically, the Petitioner stated 
that (1) the risk of fire near electrical control cables due to 
combustible insulation could cause a catastrophic meltdown; (2) cracks 
were discovered in the structural support (core shroud) of the reactor 
fuel in Peach Bottom Unit 3, indicating possible cracks in other parts 
of the reactor vessel; (3) the NRC discovered that both reactors had no 
emergency cooling water for an hour on August 3, 1994; and (4) other 
chronic problems exist at Peach Bottom according to an August 16, 1994, 
NRC report.
    The Petitioner seeks relief from the risk of fire (Request 1) due 
to cable insulation on the basis of a September 30, 1994, article in 
the Baltimore Sun that described the indictment of Thermal Sciences, 
Inc., on charges of falsifying laboratory records related to Thermo-
Lag. Thermo-Lag is a material used to insulate electrical cables and 
other equipment from fire damage. The Petition states that a fire in 
combustible insulation near electrical control cables could cause a 
catastrophic meltdown.
    The Petition also seeks the correction of cracks that were 
discovered in the structural support (core shroud) of the reactor fuel 
in Peach Bottom Unit 3, indicating possible cracks in other parts of 
the reactor vessel (Request 2). In support of this request, the 
Petitioner also references an earlier demand by the Nuclear Information 
and Resource Service (NIRS) 1 that all safety class component 
parts in both reactor vessels, including the cooling system, the heat 
transfer system, and the reactor core, be inspected and that an 
analysis be conducted of the synergistic effects of cracks in multiple 
parts. The Maryland Safe Energy Coalition did not, however, provide any 
information to support the application of the NIRS Petition to PBAPS.
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    \1\ On September 19, 1994, NIRS sought relief, pursuant to 10 
CFR 2.206, regarding safety class reactor internal components at 
Oyster Creek Nuclear Generating Station (OCNGS) on the following 
premises: (a) the core shroud in General Electric boiling water-
reactors (BWRs) is vulnerable to age-related deterioration; (b) 12 
domestic and foreign BWR owners have found extensive cracking on 
welds of the core shroud; (c) only 10 of 36 U.S. BWR owners have 
inspected their core shrouds and 9 of the 10 core shrouds had cracks 
at the time of the NIRS Petition; (d) 19 of 25 selected BWR internal 
components are susceptible to stress corrosion cracking and 6 of 19 
are susceptible to irradiation-assisted stress corrosion cracking; 
(e) as the oldest operating General Electric Mark I BWR and the 
third oldest operating reactor in the United States, OCNGS has been 
subjected for the longest period to operational conditions that 
cause embrittlement and cracking; (f) according to the BWR Owners 
Group (BWROG), cracking of the core shroud is a warning signal that 
additional safety class reactor internals are increasingly 
susceptible to age-related deterioration; (g) cracking of any single 
part or multiple components jeopardizes safe operation of that 
nuclear station; (h) Oyster Creek did not inspect for core shroud 
cracking prior to the current refueling outage and other safety-
class reactor internals have not been adequately inspected for 
cracking; and (i) a safety analysis has not been performed on the 
potential synergistic effects of multiple-component cracking. The 
relief sought in the Petition based upon these concerns was denied 
in a Partial Director's Decision issued on August 4, 1995 (See 
General Public Utilites Nuclear Corporation (Oyster Creek Nuclear 
generating Station), DD-95-18, 42 NRC 67 (1995)).
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    The Petitioner also raises equipment problems at PBAPS, stating 
that: (a) the NRC discovered both reactors at PBAPS had no emergency 
cooling water for approximately one hour on August 3,

[[Page 30457]]

1994 (Request 3), and (b) an NRC inspection report dated August 16, 
1994, which the Petitioner asserts described numerous chronic problems 
at PBAPS 2 (Request 4).
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    \2\ The Petitioner stated that the problems described in the 
August 16, 1994, NRC report included: cooling tower leaks, coolant 
injection system vibration, injection valve failures, feedwater 
vibrations and leakage, fuel pool hot spots, incore probe failures, 
auxiliary boiler unreliability, valve failures, air solenoid 
failure, and hydraulic leaks and malfunctions.
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    In a letter dated December 2, 1994, I acknowledged receipt of the 
October 6, 1994, Petition and denied the Petitioner's requests for 
immediate relief. In the acknowledgement letter I informed the 
Petitioner that the remaining requests were being evaluated under 10 
CFR 2.206 of the Commission's regulations and that action would be 
taken in a reasonable time.
    The issues raised by the Petitioner concerning the use of Thermo-
Lag fire barriers raised by Request 1 of the October 6, 1994, Petition 
have been previously considered. A Director's Decision (DD-96-03) (see 
attachment) addressing this specific request as well as the requests of 
other Petitioners with concerns regarding the use of Thermo-Lag by 
reactor licensees, was issued on April 3, 1996.3 The NRC staff's 
review of the issues related to cracking of reactor internal components 
and concerns regarding equipment problems raised by Requests 2, 3 and 4 
of the October 6, 1994, Petition is now complete. Accordingly, I am 
issuing a Final Director's Decision with regard to Requests 2, 3, and 
4. A discussion of the Final Director's Decision follows.
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    \3\ All Reactor Licensees with Installed Thermo-Lag Fire Barrier 
Material, DD-96-03, 43 NRC (1996). In addition to the Maryland Safe 
Energy Coalition, Petitioners with concerns about the use of Thermo-
Lag included the Citizens for Fair Utility Regulation and the 
Nuclear Information and Resource Service, the GE Stockholder's 
Alliance and Dr. D.K. Cinquemani, the Toledo Coalition for Safe 
Energy, R. Benjan, B. DeBolt and the Oyster Creek Nuclear Watch. In 
the Decision under 10 CFR 2.206, the Director of the Office of 
Nuclear Reactor Regulation determined that the Petitioners' requests 
concerning the use of Thermo-Lag should be denied.
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II. Discussion

A. Correction of Cracks in the Core Shroud and Assertion of Possible 
Cracks in Other Parts of the Reactor Vessel (Request 2)

    Nuclear power reactor licensees, including PECO, are required by 10 
CFR 50.55a to implement inservice inspection programs that meet the 
requirements set forth in the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code (ASME Code). The scope of the inservice 
inspection programs for reactor pressure vessels and their internal 
components are prescribed by ASME Code, Section XI, Division 1, 
Subsections IWA and IWB. Licensees are also required by ASME Code, 
Section XI, Article IWA-6000, to submit the results of these 
inspections to the NRC within 90 days of completion. The NRC staff 
performs periodic audits of licensee-implemented inservice inspection 
programs to determine compliance with applicable codes and regulations. 
These audits are documented in NRC inspection reports, which are 
publicly available at the NRC Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC. Inspection reports 
related to PBAPS are also available at the local public document room 
for PBAPS located at the State Library of Pennsylvania (REGIONAL 
DEPOSITORY), Government Publications Section, Education Building, 
Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    The licensee's inservice inspection program contains provisions for 
the periodic inspection of the PBAPS reactor vessel internal 
components, including such components as the top guides, core shroud 
welds, shroud support plate access hole covers, incore instrument 
tubes, steam dryer drain channels, core spray piping, and jet pump 
assemblies. By letter dated April 8, 1986, the NRC found the Inservice 
Inspection Program for the Second Ten-Year Interval at PBAPS Units 2 
and 3 to be satisfactory (September 1986-November 1997 and December 
1985-August 1997, for Units 2 and 3, respectively).
    In addition to the ASME Code design and inservice inspection 
program requirements, the NRC provides information to the nuclear power 
industry on various emerging phenomena that may potentially affect the 
safe operation of nuclear power plants. For example, intergranular 
stress corrosion cracking (IGSCC) of BWR internal components has been 
identified as a technical issue of concern by both the NRC staff and 
the nuclear industry. The core shroud is among the internal reactor 
components susceptible to IGSCC. Identification of cracking at the 
circumferential beltline region welds in several plants during 1993 led 
to the publication of NRC Information Notice (IN) 93-79, ``Core Shroud 
Cracking at Beltline Region Welds in Boiling-Water Reactors,'' issued 
on September 30, 1993. Several licensees inspected their core shrouds 
during planned outages in the spring of 1994 and found cracking at the 
circumferential welds. To disseminate this information to nuclear power 
plant licensees, the NRC issued IN 94-42, ``Cracking in the Lower 
Region of the Core Shroud in Boiling-Water Reactors,'' on June 7, 1994, 
and Supplement 1 to IN 94-42, on July 19, 1994, concerning cracking 
found in the core shrouds at Dresden Unit 3 and Quad Cities Unit 1. On 
July 25, 1994, the NRC issued GL 94-03, ``Intergranular Stress 
Corrosion Cracking of Core Shrouds in Boiling Water Reactors,'' 
requesting that BWR licensees inspect their core shrouds by the next 
refueling outage and justify continued operation until inspections 
could be completed. The NRC has been closely monitoring these 
inspection activities. Additional examples of NRC action regarding 
reactor vessel internal component reliability issues are the issuance 
of Bulletin 80-13, ``Cracking in Core Spray Spargers'', on May 12, 
1980, after the detection of cracks in core spray system sparger piping 
at several operating BWRs and the issuance of IN 95-17, ``Reactor 
Vessel Top Guide and Core Plate Cracking,'' issued on March 10, 1995, 
that concerned reactor vessel top guide and core plate cracking.
Core Shroud Cracks
    The licensee submitted letters dated March 14, 1994, November 7, 
1994 and November 3, 1995, regarding the results of its inspections of 
the PBAPS Unit 2 and 3 core shrouds. The inspections revealed a 
moderate amount of crack indications in the Unit 2 and Unit 3 core 
shrouds, totaling 5 percent of the weld length examined in Unit 2 and 
12 percent of the weld length examined in Unit 3. Along with the 
inspection results, the licensee presented an analysis of the impact of 
the crack indications on the structural strength of the core shrouds 
for Unit 2 and Unit 3. For both the Unit 2 and Unit 3 core shroud, the 
staff reviewed the licensee's analysis of structural loading of the as-
found shroud weld which showed that the loadings were less than ASME 
Code allowable values. In a letter dated February 6, 1995, the NRC 
staff issued a safety evaluation of the 1994 Unit 2 core shroud 
inspection concluding that sufficient structural margin remained in the 
Unit 2 shroud to justify operation of PBAPS 2 for another operating 
cycle (current operating cycle 11 that ends in September 1996) without 
modification to the shroud. In a letter dated January 29, 1996, the NRC 
staff issued a safety evaluation of the 1995 Unit 3 core shroud 
inspection concluding that sufficient structural margin remained in the 
Unit 3 shroud to justify operation of PBAPS 3 for another operating 
cycle (current operating cycle 11 that ends in

[[Page 30458]]

September 1997) without modification to the shroud.
Reactor Vessel Internals Cracking
    In addition to the inspection of core shrouds, PECO performs 
inspections of the PBAPS Unit 2 and 3 reactor vessel internals and 
other internal safety-related components in accordance with the PBAPS 
inservice inspection program, as set forth in 10 CFR 50.55a and ASME 
Code, Section XI. By letter dated January 17, 1995, PECO submitted, in 
accordance with 10 CFR 50.55a(g)(3), a report on its inservice 
inspection activities conducted during the September 1994, Unit 2, 
refueling outage. In the report PECO listed the inspections performed 
and discussed the disposition of indications in certain components. In 
addition to the core shroud flaws described above, the licensee 
discovered some minor defects, such as a crack in a jet pump assembly 
restrainer adjustment screw tack weld, and performed an engineering 
evaluation to determine if a repair was needed. In the case of the jet 
pump restrainer adjustment screw tack weld crack, a second existing 
weld was found intact and no repair was necessary. The NRC staff 
conducted an inspection of the licensee's inservice inspection 
activities during the PBAPS Unit 2 refueling outage. The results of 
that inspection are documented in Inspection Report 50-277/94-28 and 
50-278/94-28 (IR 94-28). The staff concluded that PBAPS inservice 
inspection programs and nondestructive examination programs were well 
planned, controlled, and executed for both PBAPS 2 and PBAPS 3. 
Therefore, the requirements of 10 CFR 50.55a and the ASME Code have 
been met in this area, and the results confirm that satisfactory 
material conditions exist for the safe operation of both units.
    The NRC staff has reviewed the content and results of other 
licensee inspection activities, as discussed below.
    NRC Bulletin 80-13, issued on May 12, 1980, requested that BWR 
licensees visually inspect core spray piping inside the reactor vessel 
at each subsequent refueling outage. During inspections conducted as 
requested by the staff in Bulletin 80-13, PECO detected cracks in core 
spray piping inside the reactor vessel in Unit 2 and Unit 3 in 1982 and 
1985, respectively. In both instances, the licensee installed clamps on 
the affected piping to mitigate the consequences of the cracks. In 
letters dated June 10, 1982, and November 21, 1985, the NRC staff 
reviewed the licensee's analysis of the crack consequences and repair 
plans 4 and found them acceptable for PBAPS Units 2 and 3, 
respectively.
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    \4\ Correspondence regarding these cracks, including letters 
from PECO to the NRC dated April 29, 1982, May 11, 1982, June 4, 
1982, and November 8, 1985 are available in the local public 
document room.
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    In November 1993, during subsequent inspections, PECO identified 
cracking in the downcomer portion of the Unit 3 core spray piping. By 
letters dated November 5 and November 10, 1993, the licensee provided 
an analysis which demonstrated that this downcomer piping had 
sufficient structural integrity to justify operation without repair for 
the subsequent operating cycle. In a letter dated November 16, 1993, 
the NRC found PECO's proposal to operate for one operating cycle 
without repairing the core spray downcomer cracks acceptable. During 
the September 1995 refueling outage for PBAPS Unit 3, PECO performed 
additional inspections of the core spray piping within the reactor 
vessel. As documented in its letter dated October 9, 1995, PECO stated 
that this inspection revealed additional cracking. In its letter of 
October 9, 1995, as supplemented by a letter dated October 12, 1995, 
PECO proposed to repair the core spray piping by installing mechanical 
clamps over the affected cracked welds. The NRC staff reviewed the 
design of the proposed clamps and found that the clamps provided the 
required structural integrity for the piping. The NRC staff also 
approved restart of the Peach Bottom Unit 3 based on PECO's 
installation of the clamps.5
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    \5\ The NRC staff's review of the clamp design is addressed in 
Inspection Report 50-277/95-18; 50-278/95-18 and in a letter dated 
October 13, 1995.
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    Although cracking of the top guide has not been detected at PBAPS, 
the licensee has implemented a program to inspect the top guide and has 
included the top guide inspection into the PBAPS inservice inspection 
program.
Analysis Regarding Synergistic Effects of Cracking of Multiple 
Components
    The Petitioner raises a concern about the lack of an analysis of 
the synergistic effects of cracks in multiple reactor vessel 
components.
    Most reactor internals are fabricated from high-toughness materials 
such as stainless steel and were designed with significant margins on 
allowable stresses. Cracking must be severe to adversely impact plant 
safety. It is unlikely that licensee inspections would not find such 
severe degradation. In fact, the PECO inspections, using qualified 
inspectors and procedures, have been effective in identifying and 
sizing of the cracks in the Peach Bottom Unit 2 and Unit 3 core 
shrouds. In addition, after evaluating the results from internals 
inspections performed to date at PBAPS, the NRC staff has concluded 
that ASME Code structural margins have been maintained to meet ASME 
design requirements. Thus, these components will perform their function 
in the safe operation of the plants.
    Implementation of an effective inservice inspection program serves 
to detect cracking. Upon detection of cracking, proper actions by the 
licensee to maintain component integrity will prevent cracks, large 
enough to affect operability, from existing in multiple components at 
the same time. Nevertheless, the NRC has asked the BWR Vessel Internals 
Project (BWRVIP), an industry group, to develop an assessment to 
address this unlikely situation. A report from the BWRVIP on this 
issue, ``Reactor Pressure Vessel and Internals Examination Guidelines 
(BWRVIP-03; EPRI Report TR-105696,'' dated November 10, 1995, is 
currently under NRC staff review. In addition, the NRC has undertaken a 
longer term evaluation of the effects of cracking in multiple internal 
components. This evaluation will involve appropriate probabilistic 
treatment of the key variables (such as material susceptibility, 
loading and environment).
    Moreover, the licensee is not required by 10 CFR 50.55a or the ASME 
Code to perform an analysis that addresses the synergistic effects of 
cracking in multiple safety-class components. Since the NRC staff has 
found during reviews of the initial plant design and reviews of the 
licensee's response to subsequently identified cracks, as described 
above, that each affected component has been shown to meet the ASME 
design margins; the NRC staff is satisfied that these components will 
perform their intended function in the safe operation of the 
facilities. Because of this and the inspection requirements that 
pertain to reactor internals and the results of the inspections 
performed to date, the NRC staff does not consider the lack of an 
analysis of the synergistic effects of cracks in multiple reactor 
components for PBAPS, to be a substantial safety concern.
    In summary, on the basis of the NRC inspections and the evaluations 
of the licensee inspections required by 10 CFR 50.55a and the ASME 
Code, the NRC staff has concluded that the licensee has taken 
appropriate actions to ensure the structural integrity of the PBAPS 
reactor vessel internal components. The NRC staff, however, continues 
to overview PECO's inspections, evaluations, and

[[Page 30459]]

repairs as necessary to meet these requirements. At this time, the NRC 
staff has not found any reason to question the safe operation of PBAPS. 
Therefore, the NRC staff has concluded that the Petitioner has not 
presented a substantial health or safety issue to warrant taking the 
actions requested in the Petition.

B. Correction of Equipment Problems Identified in Recent NRC Inspection 
Reports (Requests 3 and 4)

Emergency Core Cooling
    The Petition referred to a situation on August 3, 1994, wherein the 
PBAPS emergency service water (ESW) system was placed in a degraded 
condition. The Petitioner asserted that both reactors at PBAPS had no 
emergency cooling water for about one hour. The NRC resident inspectors 
at the Peach Bottom site conducted an inspection of this event and 
documented their findings in Inspection Report 50-277/94-24 and 50-278/
94-24, dated September 29, 1994 (IR 94-24). In the report the NRC 
inspectors concluded that the discharge valve from the ESW system back 
to the Susquehanna River was shut and left unattended for approximately 
fifty minutes after maintenance and testing on the valve. In the 
report, the NRC staff concluded that, if an accident requiring the use 
of safety equipment (including emergency diesel generators and 
emergency core cooling equipment) had occurred during that fifty minute 
period, the operation of that safety equipment could have been 
jeopardized.
    By letter dated November 21, 1994, the NRC issued a Notice of 
Violation and Proposed Imposition of Civil Penalty (EA-94-197) to PECO 
Energy Company regarding the circumstances surrounding the August 3, 
1994, event. The NRC staff cited the licensee for failure to implement 
maintenance and testing procedures that were adequate to ensure that 
the ESW system could perform its intended function while maintenance 
activities were being performed. The staff noted that since the August 
3, 1994, event, the licensee had restored the ESW to its intended 
configuration and had initiated steps to assure that future maintenance 
activities would not lead to a degraded ESW system. Notwithstanding the 
specific corrective actions implemented by the licensee, the staff 
imposed a civil penalty in the amount of $87,500. On December 21, 1994, 
PECO Energy paid the civil penalty.
    Because appropriate NRC action has been taken and the licensee has 
restored the ESW system to its intended configuration and has 
implemented corrective actions to prevent recurrence of the 
deficiencies that occurred on August 3, 1994, no specific concern about 
the ability of the ESW system to perform its intended function 
currently exists.
Chronic Equipment Problems
    The Petition also referenced a list of chronic equipment problems 
at PBAPS.6 The Petition referenced an NRC report dated August 16, 
1994 (NRC Inspection Report 50-277/94-17; 50-278/94-17 (IR 94-17)), as 
the source of the chronic problems.
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    \6\ See footnote 2.
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    In this inspection report the NRC assessed the performance of the 
licensee's engineering and technical support organization at Peach 
Bottom. The NRC inspector reviewed various facets of PECO's engineering 
department's performance in order to identify potential organizational 
weaknesses and deficiencies. The NRC uses the inspection findings to 
maintain a close understanding of the licensee's performance in areas 
that can affect safe plant operation. As such, the NRC reviews the 
licensee's program for identifying, addressing, and resolving recurring 
or ``chronic'' equipment problems. At the time that IR 94-17 was 
issued, the basis document for the licensee's program was the ``Chronic 
Equipment/System Problems'' list. This was a list of recurring problems 
for which the licensee had either identified the need for engineering 
department review and action or had determined a method for resolving 
the problem but had not yet implemented the solution.
    The ``Chronic Equipment/System Problems'' list included equipment 
problems with potential safety impact as well as obvious non-safety-
related problems. In assessing the management of recurring problems, 
the NRC evaluates the licensee's ability to address and resolve 
problems in a timely manner and the licensee's ability to evaluate the 
safety significance of each problem. The existence of a list of issues 
does not in itself indicate poor engineering department performance. As 
noted in IR 94-17, the licensee had developed solutions for a number of 
the problems on the list and had developed plans to implement these 
solutions. Further, the NRC staff assessed the PBAPS Chronic Equipment/
System Problem list as a positive management feature and a commitment 
on the part of the licensee to improve overall plant performance.
    The NRC staff, including the resident inspectors and the Region I 
inspection staff, periodically reevaluate the performance of the 
licensee's engineering department. In addition, NRC inspectors continue 
to review the licensee's action on many of the individual problems on 
the PBAPS Chronic Equipment/System Problem list. Accordingly, the NRC 
performed a follow-up inspection to IR 94-17. In the follow-up 
inspection, documented in Inspection Report 50-277/94-21; 50-278/94-21 
(IR 94-21), dated November 4, 1994, the NRC staff examined the safety 
significance of those items that were on the Chronic Equipment/System 
Problem List as of September 13, 1994. The staff concluded that none of 
the items on the list was a significant current safety concern. The 
inspectors concluded that the licensee had initiated appropriate action 
to evaluate and correct those items detailed in IR 94-21. The staff 
concluded that the licensee used the Chronic Equipment/System Problem 
list to appropriately focus long-term engineering and management 
attention to known reliability problems.
    In summary, the staff considers proper management of recurring 
equipment problems important to the continued safe operation of a 
nuclear power plant. Accordingly, the NRC staff views positively the 
licensee's activities such as the formulation of the Chronic Equipment/
Systems Problem list, which was cited in the Petition. On the basis of 
the review efforts by the NRC staff, I conclude that no substantial 
health or safety issues have been raised by the Petitioner.

IV. Conclusion

    The institution of proceedings in response to a request pursuant to 
Section 2.206 is appropriate only when substantial health or safety 
issues have been raised. See Consolidated Edison Co. of New York 
(Indian Point Units 1, 2, and 3), CLI-75-8, 2 NRC 173, 176 (1975) and 
Washington Public Power Supply System (WPPSS Nuclear Project No. 2), 
DD-84-7 19 NRC 899, 923 (1984). This standard has been applied to the 
concerns raised by the Petitioner to determine whether the action 
requested by the Petitioner is warranted. With regard to the specific 
requests made by the Petitioner discussed herein, the NRC staff finds 
no basis for taking any additional actions. Rather, as explained above, 
the NRC staff considers that no substantial health or safety issues 
have been raised by the Petitioner. Accordingly, the Petitioner's 
requests for additional action pursuant to Section 2.206, specifically 
requests 2, 3, and 4, are denied. Accordingly, no action pursuant to 
Section 2.206 is being taken in this matter.

[[Page 30460]]

    A copy of this Final Director's Decision will be filed with the 
Secretary of the Commission for review in accordance with 10 CFR 
2.206(c). This Decision will become the final action of the Commission 
25 days after issuance unless the Commission, on its own motion, 
institutes 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.
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NUCLEAR REGULATORY COMMISSION
Office of Nuclear Reactor Regulation

William T. Russell, Director
    In the Matter of: All Reactor Licensees With Installed Thermo-
Lag Fire Barrier Material.

Director's Decision Under 10 CFR 2.206

I. Introduction

    By letter dated September 26, 1994, the Citizens for Fair Utility 
Regulation and the Nuclear Information and Resource Service (NIRS); by 
press release dated October 6, 1994, the Maryland Safe Energy 
Coalition; by separate letters dated October 21, 1994, the GE 
Stockholders' Alliance and Dr. D. K. Cinquemani; by letter dated 
October 25, 1994, the Toledo Coalition for Safe Energy; by letter dated 
October 26, 1994, R. Benjan; by letter dated November 14, 1994, B. 
DeBolt; and by letter dated December 8, 1994, NIRS and the Oyster Creek 
Nuclear Watch (the Petitioners), requested that the U.S. Nuclear 
Regulatory Commission (NRC) take action with regard to the use of 
Thermo-Lag by reactor licensees and that their letters be treated as 
Petitions pursuant to Section 2.206 of Title 10 of the Code of Federal 
Regulations (10 CFR 2.206).

    The Citizens for Fair Utility Regulation and NIRS requested that 
(1) Texas Utilities Electric Company (TU Electric), licensee of 
Comanche Peak Steam Electric Station, Unit 1, perform additional 
destructive analysis for Thermo-Lag configurations in proportion to the 
total installed amount of Thermo-Lag to determine the degree of ``dry 
joint'' occurrence, (2) the licensee perform fire tests on upgraded 
``dry joint'' Thermo-Lag configurations for conduit and cable trays to 
rate the barrier as a tested configuration in compliance with fire 
protection regulations, and (3) the NRC immediately suspend the 
Comanche Peak Unit 1 license until the above corrective actions are 
taken. The Maryland Safe Energy Coalition requested immediate shutdown 
of both reactors at the Peach Bottom plant until the risk of fire near 
electrical control cables due to combustible insulation is 
corrected.1 Dr. Cinquemani and the Toledo Coalition for Safe 
Energy requested that the NRC immediately shut down all reactors where 
Thermo-Lag is used until it has been removed and replaced. The GE 
Stockholders' Alliance requested shutdown of all reactors where Thermo-
Lag is used until it has been removed and replaced with fire-retardant 
material meeting NRC standards. R. Benjan requested immediate shutdown 
of all reactors where Thermo-Lag is used. B. DeBolt requested shutdown 
of all reactors in which Thermo-Lag is used until it has been removed 
and replaced. NIRS and the Oyster Creek Nuclear Watch requested that 
NRC immediately suspend GPU Nuclear Corporation's (GPUN's) operating 
license for Oyster Creek Nuclear Generating Station (OCNGS) until GPUN 
removes Thermo-Lag fire barrier material and replaces it with a 
competitive product that meets current NRC fire protection regulations.
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    \1\ The Petition submitted by the Maryland Safe Energy Coalition 
expressed several concerns in addition to the fire hazard issue. 
These other issues, that is other than the fire hazard issue, will 
be the subject of a separate Director's Decision.
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    As a basis for their requests concerning Thermo-Lag 330-1 fire 
barrier upgrades, the Citizens for Fair Utility Regulation and NIRS 
Petitioners stated that (1) the licensee's records on the original 
installation of Thermo-Lag fire barriers on conduits and cable trays 
indicate that its contractor followed specifications for pre-buttering 
all joints; (2) NRC Inspection Reports 50-455/93-42 and 50-446/93-42 
found, based on destructive analysis documents, that a concern did 
exist where Thermo-Lag conduit joints fell apart easily and did not 
appear to have any residual material of a buttered surface, indicative 
of a joint that had not been pre-buttered; (3) the ``dry joint'' 
deficiency appeared in Room 115A and other areas of the unit; (4) the 
licensee directly contradicts an NRC inspector's findings that were 
determined in part by destructive analysis; (5) the ``dry joint'' or 
absence of pre-buttering of Thermo-Lag panels can be determined only by 
destructive analysis and cannot be determined by a walkdown visual 
inspection; (6) the findings reported in the Comanche Peak Unit 1 
Region IV Inspection Reports 50-455/93-42 and 50-446/93-42, based on 
the limited amount of destructive analysis conducted at the unit, 
constitute a substantial documentation of installation deficiencies 
found in Thermo-Lag fire barriers as documented in NRC Information 
Notice (IN) 91-79, ``Deficiencies in the Procedures for Installing 
Thermo-Lag Fire Barrier Materials,'' December 6, 1991, and IN 91-79, 
Supplement 1, ``Deficiencies Found in Thermo-Lag Fire Barrier 
Installation,'' August 4, 1994; (7) neither the NRC nor the industry, 
by its agent Nuclear Energy Institute (NEI), nor a utility, have 
conducted fire tests on dry-fitted or ``dry joint'' upgraded 
configurations of Thermo-Lag 330-1; and (8) the presence of ``dry 
joint'' upgraded configurations in Comanche Peak Unit 1 constitutes an 
untested application of Thermo-Lag fire barriers.
    As a basis for the requests concerning Thermo-Lag 330-1 fire 
barrier upgrades, the Maryland Safe Energy Coalition stated that the 
manufacturer of the flame retardant (Thermo-Lag insulation) was 
indicted on criminal charges (of falsifying tests of the effectiveness 
of the insulation as a fire barrier), and fire near the electrical 
control cables, due to combustible Thermo-Lag insulation, could cause a 
catastrophic meltdown.
    As the bases for their requests, Dr. Cinquemani, the Toledo 
Coalition for Safe Energy, the GE Stockholders' Alliance, and R. Benjan 
stated either individually or collectively that (1) the widespread use 
of Thermo-Lag in more than 70 reactors presents a safety crisis; (2) 
the NRC has known since 1982 that Thermo-Lag fails NRC performance 
standards for material that protects vital electrical cables for 
ampacity rating and fire resistance; (3) Thermo-Lag has failed not only 
NRC tests, but almost all other independent tests; (4) Thermo-Lag is 
combustible, contrary to NRC regulations, and is an ineffective fire 
barrier; (5) the use of Thermo-Lag could lead to shorts, to failure of 
the cables in an emergency, and to fire; (6) Thermo-Lag is faulty in 
that fraudulent ampacity ratings allowed utilities to use smaller cable 
than permitted by design requirements, causing the cable to overheat 
and its insulation to deteriorate; (7) the NRC has stated that fire at 
some nuclear power plants can contribute as much as 50 percent of the 
risk to a core meltdown, and a typical reactor will have three to four 
significant fires during its licensed lifetime; (8) Thermal Science, 
Inc. (TSI), the manufacturer of Thermo-Lag, and its President were 
indicted by a Federal grand jury on seven criminal charges related to 
conspiracy to defraud the U.S. Government in regard to the 
effectiveness of Thermo-Lag; and (9) the hourly fire watches at the 
Davis-Besse

[[Page 30461]]

Nuclear Power Plant operated by Toledo Edison do not replace fire 
barrier material and do not prevent fires.
    As the bases for his request, B. DeBolt stated that Thermo-Lag 
fails to meet NRC regulations concerning combustibility and that the 
manufacturer of Thermo-Lag was indicted for defrauding the Government 
and the utilities. Among the many bases for their request, NIRS and the 
Oyster Creek Nuclear Watch stated that (1) Southwest Research Institute 
(SwRI) conducted fire tests on Thermo-Lag 330-1 specimens for GPUN and 
reported that all specimens ignited approximately 2 seconds after it 
was inserted into the furnace and failed specified criteria because of 
flaming after the first 30 seconds of testing, an outside temperature 
rise higher than 30  deg.C, and a weight loss of 50 percent; (2) GPUN's 
operation of OCNGS with knowledge of the SwRI report is an example of 
GPUN's reckless disregard for fire protection and public safety; (3) in 
the event of fire, Thermo-Lag is likely to fail its intended function 
of protecting vital electrical cables running from the control room to 
plant safety systems used to shut down the reactor; (4) current 
installations of Thermo-Lag are likely to fail in less time than 1 hour 
(when smoke detectors and automatic sprinkler systems are present) or 3 
hours (when there are no fire detection and suppression systems) that 
NRC regulations require for fire barriers to withstand fire; (5) the 
NRC Inspector General issued a report in August 1992 condemning NRC's 
handling of the Thermo-Lag issue and documenting the NRC staff's 
failure to understand the scope of the problem; (6) in April 1994, 
Industrial Testing Laboratories and its President pleaded guilty to 
five felony counts of aiding and abetting the distribution of falsified 
test data; (7) on September 29, 1994, the U.S. Department of Justice 
issued a seven-count indictment against the manufacturer of Thermo-Lag 
and its Chief Executive Officer for willful violations of the Atomic 
Energy Act, conspiracy to conceal material facts, and making false 
statements to defraud the United States in connection with $58 million 
in fire barrier material; (8) GPUN has known since at least August 11, 
1992, that Thermo-Lag 330-1 as a structural base material is 
combustible and that GPUN was in violation of Appendices A and R to 10 
CFR Part 50 and the NRC Standard Review Plan, NUREG-0800; (9) GPUN 
failed to report the SwRI test results in response to a request for 
additional information regarding Generic Letter (GL) 92-08 (``Thermo-
Lag 330-1 Fire Barriers'') of February 10, 1994, when asked to describe 
the Thermo-Lag 330-1 fire barriers installed as required to meet 10 CFR 
Part 50, Appendix R; and (10) continued reliance on fire watches at 
OCNGS is an unreasonable and unnecessary hazard to the public health 
and safety because of an inoperable fire protection system for safe 
shutdown of the reactor and installed combustible material on the 
shutdown systems.
    On November 7, 1994, I informed the Citizens for Fair Utility 
Regulation and NIRS that the request for an immediate suspension of the 
Comanche Peak Unit 1 operating license was denied. On December 2, 1994, 
I informed the Maryland Safe Energy Coalition that the request for an 
immediate shutdown of the Peach Bottom plant and for an immediate 
suspension of the Peach Bottom license was denied. On December 15, 
1994, I informed the GE Stockholders Alliance, Dr. D. K. Cinquemani, 
the Toledo Coalition for Safe Energy, and R. Benjan that the immediate 
suspension of the operating licenses of all reactors where Thermo-Lag 
is used was denied. On January 3, 1995, I informed NIRS and the Oyster 
Creek Nuclear Watch that the immediate suspension of the OCNGS 
operating license was denied. On January 19, 1995, I informed B. DeBolt 
that the request for immediate suspension of the operating licenses of 
all reactors in which Thermo-Lag is used was denied. The decisions were 
based on the following: (1) the staff is addressing deficiencies in 
fire barriers constructed with Thermo-Lag material as part of a 
Commission-approved action plan and has issued several bulletins and a 
generic letter to the nuclear industry to provide information and 
guidance, (2) fire barrier systems constructed with Thermo-Lag have 
been identified and declared inoperable, and (3) compensatory measures 
(fire watches) approved by the NRC have been instituted. Additionally 
in the above correspondence, all Petitioners were informed that the 
Petitions were being treated pursuant to 10 CFR 2.206 and had been 
referred to this office for action pursuant to 10 CFR 2.206 of the 
Commission's regulations and that appropriate action would be taken 
within a reasonable time.
    For the reasons stated below, the Petitions have been denied.

II. Background

    The picture painted by the Petitioners of inaction by the NRC staff 
in responding to the issues presented by the use of Thermo-Lag is at 
odds with the facts. A review of the chronological development of the 
issues shows that the NRC staff has been working diligently to resolve 
the issues and has consistently sought to ensure that there is adequate 
protection of the public health and safety. It is also inaccurate to 
contend that Thermo-Lag generic deficiencies have been known since 
1982. As can be seen from the following information, the development of 
the Thermo-Lag issue has been evolutionary. Reports of problems 
regarding Thermo-Lag began to surface in the late 1980s when Gulf 
States Utilities, the licensee for River Bend Station, discovered some 
cracks and wear damage due to installation deficiencies (Licensee Event 
Report 87-005, March 25, 1987) and declared the material inoperable as 
a fire barrier. The licensee further discovered that stress skin was 
missing on all 3-hour Thermo-Lag fire barriers in the turbine building 
as a result of an installation error. In a series of plant-specific 
tests performed by Gulf States Utilities in 1989, Thermo-Lag barriers 
failed to meet the fire endurance test acceptance criteria. Gulf States 
Utilities categorized all 1-hour and 3-hour barriers as indeterminate 
and implemented compensatory measures in the form of fire watches. 
Other isolated plant-specific fire protection problems had been found 
during NRC inspections at various utilities as early as 1982 and had 
been acted on by the NRC staff. These problems were treated as plant-
specific issues and were not considered as indications of generic 
problems.
    In February 1991, the NRC received allegations that Thermo-Lag did 
not provide fire protection for electrical cables as claimed by the 
vendor. In response, in May 1991, the NRC visited River Bend Station to 
review the installation procedures and the failed fire endurance tests 
and concluded that a generic concern existed with 30-inch-wide cable 
trays. The NRC alerted the industry of the results of the test failures 
in IN 91-47, ``Failure of Thermo-Lag Fire Barrier Material To Pass Fire 
Endurance Test,'' August 6, 1991.
    In June 1991, the Office of Nuclear Reactor Regulation (NRR) 
established a special review team to investigate the safety 
significance and generic applicability of technical issues regarding 
allegations and operating experience concerning Thermo-Lag fire 
barriers. In its final report, which was issued with IN 92-46, 
``Thermo-Lag Fire Barrier Material Special Review Team Final Report 
Findings, Current Fire Endurance Testing, and Ampacity Calculation 
Errors,'' June 23, 1992, the special review team reached the following 
conclusions:

[[Page 30462]]

     The fire-resistive ratings and the ampacity derating 
factors for the Thermo-Lag fire barrier system were indeterminate.
     Some licensees had not reviewed and evaluated the fire 
endurance test results and the ampacity derating test results used as 
the licensing basis for their Thermo-Lag barriers to determine the 
validity of the tests and the applicability of the test results to 
their plant designs.
     Some licensees had not reviewed the Thermo-Lag fire 
barriers installed in their plants to ensure that they met NRC 
requirements and guidance, such as that provided in GL 86-10, 
``Implementation of Fire Protection Requirements,'' April 24, 1986.
     Some licensees used inadequate or incomplete installation 
procedures during the construction of their Thermo-Lag barriers.
    After the special review team completed its charter, the NRC staff 
prepared an action plan that provided a process to resolve technical 
issues identified with Thermo-Lag fire barrier systems. The NEI, 
formerly the Nuclear Management and Resources Council (NUMARC), agreed 
to coordinate industry efforts to resolve the issues.
    In regard to the Petitioners' allegations of NRC's inaction in 
responding to the issues presented by the use of Thermo-Lag, the 
significant progress made by the NRC staff and the nuclear reactor 
licensees in resolving Thermo-Lag issues speaks to the contrary. The 
NRC staff has issued a number of generic communications related to 
Thermo-Lag, which include the following: (1) two bulletins: BUL 92-01, 
``Failure of Thermo-Lag 330 Fire Barrier System To Maintain Cabling in 
Wide Cable Trays and Small Conduits Free From Fire Damage,'' June 24, 
1992, and BUL 92-01, Supplement 1, ``Failure of Thermo-Lag 330 Fire 
Barrier System To Perform Its Specified Fire Endurance Function,'' 
August 28, 1992; (2) two generic letters: GL 92-08, ``Thermo-Lag 330-1 
Fire Barriers,'' December 17, 1992, and GL 86-10, Supplement 1, ``Fire 
Endurance Test Acceptance Criteria for Fire Barrier Systems Used To 
Separate Redundant Safe Shutdown Trains Within the Same Fire Area,'' 
March 25, 1994; and (3) 12 information notices: IN 91-47; IN 91-79; IN 
91-79, Supplement 1; IN 92-46; IN 92-55, ``Current Fire Endurance Test 
Results for Thermo-Lag Fire Barrier Material,'' July 27, 1992; IN 92-
82, ``Results of Thermo-Lag 330-1 Combustibility Testing,'' December 
15, 1992; IN 94-22, ``Fire Endurance and Ampacity Derating Test Results 
for 3-Hour Fire-Rated Thermo-Lag 330-1 Fire Barriers,'' March 16, 1994; 
IN 94-86, ``Legal Actions Against Thermal Science, Inc., Manufacturer 
of Thermo-Lag,'' December 22, 1994; IN 95-27, ``NRC Review of Nuclear 
Energy Institute, Thermo-Lag 330-1 Combustibility Evaluation 
Methodology Plant Screening Guide,'' May 31, 1995; IN 95-32, ``Thermo-
Lag 330-1 Flame Spread Test Results,'' August 10, 1995; IN 95-49, 
``Seismic Adequacy of Thermo-Lag Panels,'' October 27, 1995, and IN 94-
86, Supplement 1, ``Legal Actions Against Thermal Science, Inc., 
Manufacturer of Thermo-Lag,'' November 15, 1995.
    The NRC staff, the nuclear industry, and others have expended much 
time and many resources to address and resolve the Thermo-Lag issues. 
The NRC staff developed comprehensive fire test guidance and acceptance 
criteria and worked with industry to improve existing ampacity test 
procedures. The NRC staff and industry performed about 100 fire 
endurance and ampacity derating tests of Thermo-Lag fire barrier 
materials and full-scale test assemblies. The fire endurance tests 
established the limitations and the true fire-resistive capabilities of 
certain Thermo-Lag fire barrier configurations, without relying on the 
fire endurance test data supplied by TSI, the manufacturer of Thermo-
Lag. On the basis of some of these tests, the NRC staff concluded that 
existing Thermo-Lag barriers could be upgraded with some additional 
Thermo-Lag material to satisfy NRC regulations. Precluding all use of 
Thermo-Lag materials for current and future fire barrier installations 
would remove a realistic option for resolving safety issues. Therefore, 
the NRC staff does not object to the use of Thermo-Lag in specific 
applications, where, through upgrades, NRC requirements are satisfied. 
The NRC staff issued three requests for additional information (RAIs) 
regarding GL 92-08 to each licensee using Thermo-Lag to obtain 
information on the specific Thermo-Lag material installed at each 
plant. The NRC staff reviewed and approved comprehensive Thermo-Lag 
fire barrier programs proposed by TU Electric for Comanche Peak Steam 
Electric Station, Unit 2, and by Tennessee Valley Authority (TVA) for 
Watts Bar Nuclear Power Plant, Unit 1, which attests to the fact that 
Thermo-Lag barriers can meet NRC fire protection guidelines and 
requirements. The NRC staff completed toxicity tests of Thermo-Lag 
material. The NRC staff and the industry completed chemical 
composition, combustibility, and flame spread tests of Thermo-Lag 
materials. Finally, the NRC staff reassessed previous technical 
conclusions to determine the extent to which the NRC staff and industry 
relied on information supplied by TSI to reach these conclusions. The 
staff had concerns about the reliability of information and data 
supplied by TSI that have been or could be used to make judgments 
regarding Thermo-Lag materials. The NRC staff identified and 
categorized the issues and previous conclusions and used the results of 
the industry-wide testing program regarding the chemical composition of 
Thermo-Lag, as discussed below, to determine if the in-plant Thermo-Lag 
materials were consistent. The results of this reassessment indicated 
that previous technical conclusions were valid independent of the 
information provided by TSI. The staff therefore concluded that 
additional action to reassess the issues or reverify the previous 
conclusions was not needed.
    The NEI testing program on the chemical composition of Thermo-Lag 
analyzed samples from 18 utilities representing 25 nuclear power 
plants. The samples represented Thermo-Lag material manufactured 
between 1984 and 1995. NEI performed pyrolysis gas chromatography 
evaluation of 169 samples to assess organic chemical composition and 
performed energy-dispersive X-ray spectroscopy of 33 samples to assess 
inorganic chemical composition. On the basis of the tests, NEI 
concluded that (1) all of the samples contained the constituents 
identified by TSI as essential to fire barrier performance; (2) the 
composition of the samples was consistent; and (3) the test results 
provided a basis on which to close NRC questions about chemical 
composition and product consistency and for utility use of generic test 
data relative to fire endurance ratings, flame spread, heat release, 
ampacity derating, and other material properties.
    The NRC staff test program on the chemical composition of Thermo-
Lag was conducted by the National Institute of Standards and Technology 
(NIST) during 1992 and 1995. NIST analyzed 21 samples that were either 
collected by the staff during site visits to plants and test 
laboratories or provided by TVA, Gulf States Utilities, Commonwealth 
Edison Company, and NEI. The analysis included elemental and ammonia 
analysis, pyrolysis, gas chromatography, mass spectrometry, and X-ray 
fluorescence. These analytical techniques indicated that all of the 
samples were similar in their bulk chemical composition. These results 
were consistent with the results of the NEI chemical testing program 
pertaining

[[Page 30463]]

to the chemical composition and uniformity of Thermo-Lag.
    Industry-wide progress has generally been commensurate with the 
complexity of the plant-specific issues and the amounts of Thermo-Lag 
installed at the individual plants. Several licensees have initiated 
programs to replace Thermo-Lag and are performing plant-specific tests 
of other fire barrier materials such as Mecatiss (Florida Power & Light 
for Crystal River Unit 3) and Darmatt KM-1 (Carolina Power & Light for 
Brunswick, IES Utilities, Inc., for Duane Arnold Energy Center, 
Commonwealth Edison Company for LaSalle County Station, and Northern 
States Power Company for Prairie Island Nuclear Generating Plant). The 
NRC staff is reviewing the plant-specific fire endurance test programs 
and has recently approved the plant-specific application of Darmatt KM-
1 fire barrier at the LaSalle plant. The remaining licensees have 
submitted to the NRC staff detailed plans and schedules for resolving 
the issues at their plants. Most licensees are pursuing a combination 
of such options as upgrading existing Thermo-Lag fire barriers to meet 
NRC fire barrier requirements, replacing Thermo-Lag fire barriers with 
another type of fire barrier, reducing or eliminating reliance on 
Thermo-lag fire barriers by relocating equipment and cables and by 
post-fire safe-shutdown reanalysis, installing additional fire 
protection features such as automatic sprinkler systems, and requesting 
configuration-specific exemptions when such exemptions are allowed by 
NRC regulations and are technically justified to provide a level of 
safety equivalent to that prescribed by the regulations. The NRC staff 
has completed its review of the plans for resolving fire protection 
issues that were proposed by most of the licensees. As with any issues 
as technically complex, challenging, and resource intensive as those 
presented by Thermo-Lag barriers, some plant-specific questions remain. 
However, the number of issues has steadily declined. The NRC staff and 
the licensees will continue to address the residual questions on a 
case-by-case basis as they arise, and the NRC staff will continue to 
follow up with individual licensees on their corrective actions, as 
appropriate. Every licensee with Thermo-Lag fire barriers will continue 
to maintain NRC-approved compensatory measures, such as fire watches, 
until its permanent corrective actions are implemented. Therefore, the 
public health and safety are protected.
    The NRC's ``defense-in-depth'' fire protection concept relies on 
protecting safe shutdown functions by achieving a balance among three 
echelons or levels of protection, which are (1) fire prevention 
activities; (2) the ability to rapidly detect, control, and suppress a 
fire; and (3) physical separation of redundant safe shutdown functions. 
Weaknesses found in one area may be dealt with by enhancing the 
protection capabilities of the remaining areas.2 The NRC foresaw 
cases in which fire protection features would be inoperable and 
required licensees, through technical specifications or approved fire 
protection plans controlled by license conditions, to provide 
compensation for the deficient condition. The concept of allowing 
alternative actions to compensate for an inoperable condition or 
component is used in various programs associated with the operation of 
nuclear power plants and has long been an integral part of NRC 
regulatory requirements.3
---------------------------------------------------------------------------

    \2\ The ``defense-in-depth'' concept is detailed in the ``NRC 
Standard Review Plan,'' NUREG-0800, Section 9.5.1, ``Fire Protection 
Program,'' page 9.5.1-10.
    \3\ NRC GL 91-18, ``Information to Licensees Regarding Two NRC 
Manual Sections on Resolution of Degraded and Nonconforming 
Conditions and Operability,'' issued November 7, 1991, and NRC 
Inspection Manual, Part 9900, ``Resolution of Degraded and 
Nonconforming Conditions,'' issued October 31, 1991.
---------------------------------------------------------------------------

    The fire endurance test results contained in NRC BUL 92-01 and NRC 
BUL 92-01, Supplement 1, confirmed that certain Thermo-Lag fire barrier 
configurations compromise one facet of the fire protection defense-in-
depth concept. In response to NRC BUL 92-01 and its supplement, the 
licensees for plants using Thermo-Lag fire barriers established fire 
watches in accordance with their technical specifications or license 
conditions as a compensatory measure. Fire watches are personnel 
trained by the licensees to inspect for the control of ignition 
sources, fire hazards, and combustible materials; to look for signs of 
incipient fires; to provide prompt notification of fire hazards and 
fires; and to take appropriate actions to begin fire suppression 
activities. Generally, therefore, by providing additional fire 
prevention activities through enhanced detection capabilities to find 
fire hazards and in the case of a fire, augmented suppression 
activities before a barrier's ability to endure a fire is challenged, 
fire watches compensate for degraded fire barriers.
    The NRC staff has carefully evaluated the issues associated with 
continued use of Thermo-Lag material, including the use of fire watches 
to compensate for any degradation in the effectiveness of required fire 
barriers. Such compensatory actions provide an adequate level of fire 
protection without an undue risk to the health and safety of the 
public. Licensees have established fire watches to compensate for 
degraded and possibly inoperable fire barriers. Also, licensees rely on 
a defense-in-depth concept that incorporates multiple safety measures. 
Automatic fire detection and suppression systems are provided in most 
areas that have safe shutdown equipment. Trained fire brigades are 
required 24 hours a day at all plants. All areas that have safe 
shutdown equipment have manual fire suppression features. Fuels that 
can feed a fire and ignition sources to start a fire are controlled. 
The combination of fire watches and the defense-in-depth fire 
protection features provides an adequate level of fire protection until 
licensees implement permanent corrective actions.
    Taken together, these factors represent an adequate means of fire 
protection at the plants using Thermo-Lag to ensure, with margin,4 
that operation can be conducted without an undue risk to the health and 
safety of the public. Nevertheless, with these considerations in mind, 
the NRC staff addressed below the Petitioners' specific concerns to 
demonstrate that no substantial health and safety issue has been 
raised.
---------------------------------------------------------------------------

    \4\ The fact that Thermo-Lag barriers, as installed, will 
provide protection for some period of time is supported by, among 
others, the fire endurance test results documented in IN 92-55.
---------------------------------------------------------------------------

III. Response to Specific Concerns

    The Petitioners alleged that (1) the NRC has been slow to enforce 
its own regulations, (2) fire watches do not replace fire barriers and 
continued reliance on fire watches is an unreasonable and unnecessary 
hazard to the public health and safety because of an inoperable fire 
protection system for safe shutdown of the reactor and installed 
combustible material on the shutdown systems, (3) utilities are in 
violation of NRC requirements because Thermo-Lag is combustible and 
could contribute to a fire instead of protecting from it, and, in spite 
of the danger, the NRC allows continued use of Thermo-Lag, (4) faulty 
ampacity ratings could result in the use of inappropriate cables, 
which, if undersized, could overheat and cause its insulation to 
deteriorate, (5) the licensee for Oyster Creek did not report to the 
NRC its findings regarding the combustibility of Thermo-Lag and, (6) 
the Thermo-Lag barriers have been improperly installed at Comanche Peak 
Unit 1, which contributes further to the poor performance of Thermo-
Lag.

[[Page 30464]]

    The NRC staff acknowledged and has stated that certain Thermo-Lag 
fire barrier configurations have failed to demonstrate the ability to 
perform their fire resistance functions. In this regard, the NRC staff, 
in BUL 92-01, Supplement 1, has stated that Thermo-Lag fire barriers 
should be treated as inoperable until licensees can declare the fire 
barriers operable on the basis of successful, applicable tests. Given 
the foregoing deficiencies identified for Thermo-Lag, the NRC staff 
concluded that compensatory measures are necessary until a licensee can 
declare fire barriers operable on the basis of applicable tests that 
demonstrate successful barrier performance.
    The Petitioners also asserted that (1) the NRC should have 
protected the public and not Rubin Feldman, the President of the 
company manufacturing Thermo-Lag, and (2) public safety has been 
compromised by NRC's seeming complicity with utilities.5
---------------------------------------------------------------------------

    \5\ These statements could be interpreted as the appearance of 
unwarranted favoritism toward the manufacturer of Thermo-Lag and 
complicity with utilities. Therefore, the Petitions were referred to 
the NRC Office of the Inspector General.
---------------------------------------------------------------------------

A. Regulatory Compliance

    The NRC staff acknowledges that certain fire endurance tests have 
demonstrated that Thermo-Lag barriers may not meet the fire endurance 
rating criteria set forth in Section III.G. of Appendix R to 10 CFR 
Part 50. This acknowledgment does not mean, however, that there no 
longer is reasonable assurance of protection of the public health and 
safety or that such actions as the shutdown of all reactors using 
Thermo-Lag and the suspension of Comanche Peak, Peach Bottom, and 
Oyster Creek operating licenses are warranted.
    It should first be noted that Appendix R, which sets forth criteria 
for specific fire protection features to protect safe shutdown systems, 
is applicable only to facilities that commenced operation prior to 
1979. Facilities commencing operation on or after January 1, 1979, 
although not bound by Appendix R, generally are bound by licensing 
commitments to follow the criteria set forth in Appendix R through 
license conditions.6
---------------------------------------------------------------------------

    \6\ In addition, there are a very limited number of plants which 
commenced operation on or after January 1, 1979, that are not 
subject to specific license conditions but whose licensees have made 
commitments to comply with NRC fire protection requirements, 
including Section III.G. of Appendix R. The NRC is elevating these 
commitments to license conditions.
---------------------------------------------------------------------------

    Even assuming that all of the plants in which Thermo-Lag is 
installed and that commenced operation prior to 1979 are not in 
compliance with Appendix R, it does not follow that the failure to 
comply with a regulation indicates the absence of adequate protection. 
The Commission has explained that--

    [W]hile it is true that compliance with all NRC regulations 
provides reasonable assurance of adequate protection of the public 
health and safety, the converse is not correct, that failure to 
comply with one regulation or another is an indication of the 
absence of adequate protection, at least in a situation where the 
Commission has reviewed the noncompliance and found that it does not 
pose an ``undue risk'' to the public health and safety.

(Ohio Citizens for Responsible Energy, DPRM 88-4, 28 NRC 411 
(1988).)

    All the plants using Thermo-Lag have instituted fire watches as 
required by their action statements regarding inoperable barriers 
contained in their technical specifications or fire protection programs 
subject to license conditions. Generally, action statements provide 
alternative remedial actions to shutting down a plant when limiting 
conditions for operation are not met. Compliance with the required 
remedial actions provides reasonable assurance that the public health 
and safety is protected notwithstanding the plant's continued operation 
and its failure to meet the respective limiting condition for 
operation. Here, since all of the plants using Thermo-Lag have 
implemented the required fire watches in accordance with plant-specific 
requirements, their continued operation does not pose an undue risk to 
the public health and safety.
    The Petitioners assert that fire watches do not replace fire 
barriers and continued reliance on fire watches is a hazard to public 
safety. The NRC staff acknowledges that fire watches do not replace 
fire barriers. However, as will be discussed in greater detail later in 
this Decision, fire watches are judged by the NRC to be acceptable 
compensatory measures and are legally sanctioned remedial actions based 
on 10 CFR 50.36(c)(2).7
---------------------------------------------------------------------------

    \7\ In instances in which fire protection programs have been 
moved from technical specifications and are now subject to license 
conditions, the NRC's approval of the fire protection programs 
subject to license conditions provides the legal basis for the 
implementation of fire watches as a remedial measure.
---------------------------------------------------------------------------

    In sum, notwithstanding the failure to have operable fire barriers 
meeting the fire endurance rating criteria specified by Section III.G. 
of Appendix R, a plant is not necessarily unsafe to continue operation. 
To the contrary, fire watches are judged by the NRC to be adequate 
remedial measures that provide reasonable assurance that the public 
health and safety is protected. By reason of compliance by all 
facilities using Thermo-Lag with their technical specifications or fire 
protection program action statements requiring the implementation of 
fire watches, protection of the public health and safety is still 
reasonably ensured for such plants. Because the Commission has 
discretion regarding enforcement of its regulations, and given the 
circumstances here in which no significant health and safety issues 
have been raised, enforcement action of the nature requested by the 
Petitioners is not warranted.

B. Ability of Fire Watches To Compensate for a Degraded Barrier

    One of the Petitioners' allegations is that the measures taken by 
licensees to compensate for degraded barrier conditions, specifically 
fire watches, are not adequate to protect the public health and safety. 
The Petitioners have questioned the continued reliance on fire watches 
in the light of an inoperable fire protection system for safe plant 
shutdown and the combustibility of Thermo-Lag. In addition, the 
Petitioners claim that a fire watch does not replace a fire barrier in 
that fire watches are not preventive.
    Despite the acknowledged shortcomings identified with certain 
Thermo-Lag fire barriers and after fully considering the arguments 
presented by the Petitioners regarding the ability of fire watches to 
provide adequate compensation, the NRC staff has determined that 
compensatory measures using fire watches are adequate and acceptable to 
ensure public health and safety until permanent corrective measures are 
implemented.
    The use of fire watches in instances of degraded or inoperable 
barriers is an integral part of NRC-approved fire protection programs. 
In general, these NRC staff-approved compensatory measures specify the 
establishment of a continuous fire watch or an hourly fire watch in 
cases in which automatic detection systems protect the affected 
components. Although it is true that Thermo-Lag is intended as a 
barrier and fire watch personnel cannot act as physical shields, a fire 
watch provides more than simply a detection function. Personnel 
assigned to fire watches are trained by the licensee to inspect for the 
control of ignition sources, fire hazards, and combustible materials; 
to look for signs of incipient fires; to provide prompt notification of 
fire hazards and fires; and to take appropriate action to begin fire 
suppression activities. Fire watch personnel are capable of

[[Page 30465]]

determining the size, the actual location, the source, and the type of 
fire--valuable information that cannot be provided by an automatic fire 
detection system.
    During a plant fire, compartment temperatures are likely to be less 
severe at the early stages. On the basis of enhanced capabilities 
provided by fire watches and notwithstanding that the level of barrier-
type protection may be reduced, the NRC staff has determined that there 
is an adequate margin of safety to ensure protection in cases in which 
fire watches are approved.
    The goal of the NRC staff's Thermo-Lag Action Plan is directed 
towards restoring the functional capability of fire barriers as soon as 
practicable. There is not a time limit associated with the use of fire 
watches as a compensatory measure. Given the margin of safety a fire 
watch brings to a fire protection program, as discussed above, the NRC 
staff has determined that continuing the use of fire watches while 
barriers are inoperable is acceptable. However, the NRC believes that 
notwithstanding interim reliance on compensatory measures, appropriate 
actions must be taken by licensees to restore operability of Thermo-Lag 
barriers. Individual licensees have provided schedules for restoring 
operability and these are being tracked by the NRC staff.
    The NRC staff has carefully evaluated the use of fire watches to 
compensate for any degradation in the effectiveness of required fire 
barriers and has concluded that fire watches continue to ensure 
protection of the public health and safety. Therefore, the Petitioners' 
assertion that the measures taken by licensees to compensate for 
degraded fire barrier conditions, specifically fire watches, are a 
hazard is without merit.

C. Combustibility

    The Petitioners alleged that, contrary to NRC regulations, Thermo-
Lag is combustible.
    The NRC staff recognizes that Thermo-Lag is combustible. To assess 
Thermo-Lag combustibility, the NRC staff conducted a testing program at 
the National Institute of Standards and Technology (NIST) based on the 
American Society for Testing and Materials (ASTM) Standard E-136. Under 
this testing standard, the material is considered to be ``combustible'' 
if three out of four samples tested exceed the following criteria: (1) 
the recorded temperature of the specimen's surface and interior 
thermocouples, during the test, rises 54  deg.F (30  deg.C) above the 
initial furnace temperature; (2) there is flaming from the specimen 
after the first 30 seconds of irradiance; and (3) the weight loss of 
the specimen, due to combustion during the testing, exceeds 50 percent. 
Of the four Thermo-Lag specimens tested, all experienced a weight loss 
of greater than 50 percent and flaming continued in excess of 30 
seconds. IN 92-82, which provided licensees with the results of the E-
136 tests and confirmed the combustibility of Thermo-Lag, restated the 
NRC fire protection requirements of Section III.G. of Appendix R to 10 
CFR Part 50 and asked that licensees review the information for 
applicability to their facilities.
    The NRC's basic fire protection regulation for commercial nuclear 
power plants is Section 50.48 of 10 CFR Part 50 ``Fire protection.'' 
Section 50.48 references General Design Criterion (GDC) 3 of Appendix A 
to 10 CFR Part 50, ``Fire protection,'' Appendix R to 10 CFR Part 50 
``Fire Protection Program for Nuclear Power Facilities Operating Prior 
to January 1, 1979,'' and various NRC fire protection guidance 
documents. Specifically, Section 50.48(a) states that each operating 
nuclear power plant must have a fire protection plan that satisfies GDC 
3, and Section 50.48(b) states that Appendix R to 10 CFR Part 50 
establishes fire protection features required to satisfy GDC 3 with 
respect to certain generic issues for nuclear power plants licensed to 
operate prior to January 1, 1979.8 These issues are addressed in 
Section III.G, ``Fire protection of safe shutdown capability,'' Section 
III.J, ``Emergency lighting,'' and Section III.O, ``Oil collection 
system,'' of Appendix R. Of these three sections of Appendix R, Section 
III.G addresses the use of fire barriers to protect one train of 
systems necessary to achieve and maintain hot shutdown conditions in 
the event of a fire and, therefore, is the regulation of interest here.
---------------------------------------------------------------------------

    \8\ While Appendix R is applicable only to facilities that 
commenced operation prior to January 1, 1979, as discussed earlier 
in this Director's Decision, facilities commencing operation on or 
after January 1, 1979, are bound to satisfy the criteria of Appendix 
R through license conditions or licensing commitments.
---------------------------------------------------------------------------

    Section 50.48(a) notes that fire protection guidance for nuclear 
power plants is contained in two NRC documents. These are (1) Branch 
Technical Position (BTP) Auxiliary Power Conversion Systems Branch 
(APCSB) 9.5-1, ``Guidelines for Fire Protection for Nuclear Power 
Plants,'' for new plants docketed after July 1, 1976, and (2) Appendix 
A to BTP APCSB 9.5-1, ``Guidelines for Fire Protection for Nuclear 
Power Plants Docketed Prior to July 1, 1976.'' These two NRC documents 
specify preferred methods for fire protection program design including 
the use of fire barriers to satisfy Section III.G of Appendix R. Fire 
barriers that meet the criteria of Section III.G of Appendix R to 10 
CFR Part 50 and these NRC guidance documents satisfy GDC 3. NUREG-0800, 
``Standard Review Plan,'' (SRP) Section 9.5-1, ``Fire Protection 
Program,'' incorporates the guidance of BTP APCSB 9.5-1 and Appendix A 
to BTP APCSB 9.5-1 and the criteria of Section III.G of Appendix R to 
10 CFR Part 50. Therefore, fire barriers that meet the guidelines of 
SRP Section 9.5-1 also satisfy 10 CFR 50.48 and GDC 3.
    As stated in 10 CFR 50.48(a), the purpose of the fire protection 
plan is ``to limit fire damage to structures, systems, or components 
important to safety so that the capability to safely shut down the 
plant is ensured.'' In general, a fire protection plan consists of 
administrative controls and procedures, personnel for implementing the 
plan and for fire prevention and manual fire suppression activities, 
fire detection systems, automatic and manually operated fire 
suppression systems and equipment, and fire barriers.
    Section III.G of Appendix R to 10 CFR Part 50 is the only part of 
the fire protection regulations that addresses the use of fire 
barriers. It addresses the use of fire barriers to protect one train of 
systems necessary to achieve and maintain hot shutdown conditions in 
the event of a fire. Fire barriers are required to have either a 1-hour 
or 3-hour rating depending on the specific requirement. However, 
Section III.G does not provide acceptance criteria for fire barriers, 
nor does it address the combustibility of fire barrier materials. The 
criteria are set out in BTP APCSB 9.5-1, Appendix A to BTP APCSB 9.5-1, 
and SRP Section 9.5-1. These NRC documents do not preclude the use of 
combustible materials for construction of fire barriers required to 
have a 1-hour or 3-hour rating. On March 25, 1994, the staff 
consolidated and clarified in Supplement 1 to Generic Letter (GL) 86-
10, the fire barrier criteria specified in the BTPs and the SRP. This 
GL supplement provides detailed staff guidelines for assessing the 
combustibility of fire barrier materials, but it does not preclude the 
use of combustible materials for fire barriers required to satisfy a 1-
hour or 3-hour rating. In fact, the fire barrier criteria are 
appropriately focused on the performance of the fire barrier and its 
ability to achieve its intended design function, that is, its ability 
to limit temperature rise within the barrier enclosure and to prevent 
the passage of flame or gasses hot enough to adversely

[[Page 30466]]

affect the functionality of the safe shutdown components (e.g., cables) 
enclosed within the fire barrier.
    Thermo-Lag 330-1 is a sacrificial material. When it is exposed to 
elevated temperatures, such as those experienced during a fully-
developed room fire, it sublimes and transitions from a solid to a 
vapor. The vapors go through an endothermic decomposition process 
(pyrolysis) which absorbs heat from the fire. As a result of the 
pyrolysis, the unreacted Thermo-Lag material is replaced by an 
insulating char layer which is composed of small interconnecting cells 
having a large surface area. The char layer re-radiates energy and 
limits heat transfer through the Thermo-Lag material. The low thermal 
conductivity of the char layer provides additional thermal insulation. 
Therefore, even though Thermo-Lag is classified as a combustible 
material when testing in accordance with the guidance of Supplement 1 
to GL 86-10, properly designed, qualified, and installed Thermo-Lag can 
yield fire barriers with a 1-hour or 3-hour rating which will protect 
safe shutdown components from the effects of the fire. Therefore, such 
barriers can satisfy the requirements of 10 CFR 50.48 and GDC 3.
    To provide reasonable assurance that Thermo-Lag fire barriers 
installed in the nuclear power plants can meet their intended function, 
representative Thermo-Lag fire barrier assemblies have been subjected 
to full-scale qualification-type fire endurance tests conducted in 
accordance with the guidance of Supplement 1 to GL 86-10. This guidance 
provides standard and uniform test methods and acceptance criteria for 
assessing the fire-resistive capabilities of these barriers. The staff 
has found the use of Thermo-Lag acceptable as a fire barrier material 
when it is used in accordance with existing NRC regulations and 
guidance and where supported by appropriate tests and analyses.
    However, there are two types of applications where the use of 
Thermo-Lag material is not appropriate. These are (1) Enclosing 
combustible materials (e.g., insulated cables) within Thermo-Lag fire 
barriers to eliminate the combustible materials as a fire hazard and 
(2) using Thermo-Lag as radiant energy heat shields inside noninerted 
containments.
    Section III.G of Appendix R (and the equivalent SRP guidance) 
specifies three options for protecting redundant trains of systems 
necessary to achieve and maintain hot shutdown conditions located 
within the same fire area outside of containment. Two of the three 
options (Sections III.G.2.a and c) rely on the use of fire barriers 
with a 1-hour or 3-hour rating, as discussed above. The third option, 
Section III.G.2.b, specifies the separation of redundant safe shutdown 
trains by a horizontal distance of more than 20 feet with no 
intervening combustibles or fire hazards. (A typical example of 
intervening combustibles is a cable tray loaded with cables, because 
cable jacket materials are combustible.) Therefore, spacial separation, 
and not fire barriers, are used to meet Section III.G.2.b. However, to 
meet this requirement, some licensees have enclosed combustibles that 
are installed between redundant shutdown trains within a fire barrier. 
In theory, the fire barrier prevents an exposure fire from igniting the 
intervening combustible materials and spreading along them from one 
redundant train to the other. Thus the fire barrier effectively 
eliminates the intervening combustible as a fire hazard. If the fire 
barrier itself is noncombustible and the redundant safe shutdown trains 
are separated by a horizontal distance of more than 20 feet, then the 
configuration meets Section III.G.2.b of Appendix R. However, if the 
fire barrier material used to enclose the intervening combustibles is 
also combustible, such as Thermo-Lag, then the licensee has simply 
installed one combustible material over another and has not eliminated 
the intervening fire hazard. In a limited number of cases, licensees 
have enclosed intervening combustibles within Thermo-Lag fire barriers 
under the incorrect assumption that the Thermo-Lag fire barrier would 
eliminate the intervening combustibles as a fire hazard. Corrective 
actions will be required in these cases.
    As an alternative to the three options discussed above, Section 
III.G.2.f of Appendix R (and the equivalent SRP guidance) provides a 
fourth option for noninerted containments, that is, the separation of 
redundant safe shutdown components with noncombustible radiant energy 
heat shields. Thermo-Lag is classified as a combustible material when 
tested in accordance with the guidance of Supplement 1 to GL 86-10. 
Therefore, it does not meet the criteria for radiant energy heat 
shields. Licensees using Thermo-Lag in this fashion will also be 
required to take corrective action.
    To assure that corrective actions are taken in these cases, the NRC 
staff issued IN 95-27. In that IN, the staff addressed enclosing 
combustible materials within Thermo-Lag fire barriers in an attempt to 
eliminate the combustible materials as a fire hazard and using Thermo-
Lag to construct radiant energy heat shields inside noninerted 
containments. The staff identified such solutions for reevaluating the 
use of Thermo-Lag for these applications as: (1) Reanalyzing post-fire 
safe shutdown circuits inside containment and their separation to 
determine if the Thermo-Lag radiant energy shields are needed, (2) 
replacing Thermo-Lag barriers installed inside the containment with 
noncombustible barrier materials, (3) replacing Thermo-Lag barriers 
used to create combustible-free zones with noncombustible barrier 
materials, (4) rerouting cables or relocating other protected 
components, or (5) requesting plant-specific exemptions where 
technically justified.
    One of the Petitioners also asserted that subsection 5a(3) of 
Section 9.5-1 of the SRP states that fire barrier designs ``should 
utilize only non-combustible materials.'' This section of the SRP does 
not apply to fire barriers which are used to separate redundant safe 
shutdown components located within a nuclear power plant fire area. 
Rather, it applies to fire barrier penetration seals, which are 
typically installed in fire area boundaries. Thermo-Lag 330-1 is not 
used in such applications.
    The principal consideration for 1-hour and 3-hour rated fire 
barriers installed to meet NRC fire protection requirements and 
guidelines is that they can achieve their intended design function. 
That is, that they can limit temperature rise within the barrier 
enclosure and prevent the passage of flame or gasses hot enough to 
adversely affect the functionality of the safe shutdown components 
enclosed within the fire barriers. The fact that Thermo-Lag material is 
combustible does not preclude Thermo-Lag fire barriers from achieving 
the intended function of preventing fire damage if the fire barriers 
are properly designed, qualified, and installed. The Petitioners' 
contention that Thermo-Lag material should not be used because it is 
combustible is without basis.

D. Ampacity Derating

    The Petitioners assert that Thermo-Lag could contribute to starting 
a fire instead of protecting from it. They further alleged that faulty 
ampacity derating factors could result in the use of inappropriate 
cables that, if undersized, could overheat and cause its insulation to 
deteriorate.
    Ampacity derating is the lowering (derating) of the current-
carrying capacity of power cables enclosed in electrical raceways 
protected with fire barrier materials because of the insulating effect 
of the fire barrier material. This insulating effect may

[[Page 30467]]

reduce the ability of the cable insulation to dissipate heat. If not 
accounted for in the plant design, the increased cable insulation 
temperature could lead to premature insulation failure. Other factors 
also affect ampacity derating, including the extent of cable fill in 
the raceway, cable type, raceway construction, and ambient temperature. 
The National Electrical Code, Insulated Cable Engineers Association 
(ICEA) publications, and other industry standards provide ampacity 
derating factors for open air installations. These standards do not 
provide derating factors for fire barrier systems. Although a national 
standard test method is in the process of being developed but has not 
yet been established, ampacity derating factors for raceways enclosed 
with fire barrier material are determined by testing for the specific 
installation configurations.
    TSI, the manufacturer of Thermo-Lag, has documented a wide range of 
ampacity derating factors that were determined by testing, for raceways 
enclosed within Thermo-Lag fire barrier materials. On October 2, 1986, 
TSI informed its customers that, while conducting tests in September 
1986 at Underwriters Laboratories, Inc. (UL), it found that the 
ampacity derating factors for Thermo-Lag barriers were greater than 
previous tests indicated. However, the cable fill and tray 
configurations were different for each test than those tested 
previously. In addition, the NRC staff learned that UL performed a 
duplicate cable tray test that resulted in an even higher derating 
factor. The NRC staff also learned of the determination of other 
derating factors during its review of other tests conducted at 
Southwest Research Institute (SwRI).9
---------------------------------------------------------------------------

    \9\ The test procedures and test configurations differed among 
the testing laboratories. Therefore, the results from the different 
ampacity tests may not be directly comparable to each other.
    The NRC staff is concerned that the ampacity derating factors, 
as determined in UL tests for Thermo-Lag barrier designs, are 
inconsistent with TSI results for similar designs because different 
times were allowed for the temperature to stabilize before taking 
current measurements. Inconsistent stabilization times would call 
into question the validity of previous TSI results. The NRC also 
noticed during the review of the Industrial Testing Laboratories 
(ITL) test reports that ambient temperature and maximum cable 
temperature were allowed to vary widely for some tests. Therefore, 
those tests in which the ambient and maximum cable temperatures were 
not maintained within specified limits may be questionable. 
Additionally, a licensee discovered a mathematical error for the 
ampacity derating factor published in an ITL test report. A 
preliminary assessment of the use of a lower-than-actual ampacity 
derating factor indicates that higher-than-rated cable temperatures 
are possible for Thermo-Lag installations. Higher-than-rated cable 
temperatures could accelerate the aging effects experienced by the 
cable.
---------------------------------------------------------------------------

    The NRC special review team concluded that the ampacity derating 
test results completed at the time of the review, including the UL test 
results, were indeterminate. This conclusion was based on observed 
inconsistencies in the derating test results of the various testing 
laboratories. The special review team found that there was no national 
consensus test standard (e.g., Institute of Electrical and Electronics 
Engineers (IEEE) or American National Standards Institute (ANSI)) for 
conducting these tests, and that some licensees had not adequately 
reviewed ampacity derating test results to determine the validity of 
the tests and the applicability of those test results to their plant 
design. The special review team recognized that, in hypothetical cases, 
nonconservative ampacity derating factors could have been instrumental 
in the installation of inappropriate cables, which as a result, could 
suffer premature cable jacket and cable insulation failures over a 
period of time. However, since that time, the NRC staff has determined 
that in practice the ampacity derating factor resulting from Thermo-Lag 
insulating properties represents only one of many variables used in 
determining the design ampacity for power cable systems and that, as 
discussed below, sufficient margin exists in this area to preclude any 
immediate safety concerns.
    For actual installations, various derating factors are typically 
applied to the ICEA ampacity values provided for each cable size. In 
general, the cables typically used in actual installations have higher 
current-carrying capacity than the ICEA ampacity values.10 Also, 
cables are sized based on full-load current plus a 25 percent margin to 
account for starting current requirements of the load. Given the short 
duration of typical equipment starts, this margin is available to 
compensate for any errors in ampacity derating. Further, use of a cable 
size larger than normal may be required as a result of voltage drop 
considerations for long circuit lengths. In typical applications this 
also provides additional current-carrying capacity. Given these 
conservatisms inherent in the design ampacity of cable systems and in 
addition the fact that most power cables required for safe shutdown are 
not normally energized, but are typically operated during surveillance 
testing for short time periods, the likelihood that cables could ignite 
as a result of Thermo-Lag ampacity derating errors has been judged by 
the NRC staff to be unlikely. In addition, based on these conservatisms 
and the currently available information on existing plants, ampacity 
design, and operating history, the NRC staff believes that the ampacity 
derating issue is not an immediate safety issue but rather is an aging 
issue to be resolved over the long term.11
---------------------------------------------------------------------------

    \10\ ICEA ampacity values include conservatisms to compensate 
for skin and proximity effects and shield and/or sheath losses which 
may or may not apply in specific situations.
    \11\ Generic Letter 92-08 requires licensees to review the 
ampacity derating factors used for all raceways protected by Thermo-
Lag 330-1 (for fire protection of safe shutdown capability or to 
achieve physical independence of electrical systems) and to 
determine whether the ampacity derating test results relied upon are 
correct and applicable to the plant design. Presently, the staff is 
conducting reviews of followup actions to close out ampacity 
derating concerns with licensees pursuant to GL 92-08.
---------------------------------------------------------------------------

E. Oyster Creek Failed To Report Test Results on Combustibility to the 
NRC

    The Petitioners requested that Oyster Creek's license be suspended 
based on the following: (1) SwRI conducted fire tests on Thermo-Lag 
330-1 specimens for GPUN, the licensee for Oyster Creek, and reported 
that all specimens ignited approximately 2 seconds after they were 
inserted into the furnace and failed specified criteria because of 
flaming after the first 30 seconds of testing, an outside temperature 
rise higher than 30  deg.C, and a weight loss of 50 percent; (2) GPUN's 
operation of Oyster Creek with knowledge of the SwRI report is an 
example of GPUN's reckless disregard for fire protection and public 
safety; (3) in the event of fire, Thermo-Lag is likely to fail its 
intended function of protecting vital electrical cables running from 
the control room to plant safety systems used to shut down the reactor; 
(4) current installations of Thermo-Lag are likely to fail in less time 
than the 1 hour (when smoke detectors and automatic sprinkler systems 
are present) or 3 hours (when there are no fire detection and 
suppression systems) that NRC regulations require for fire barriers to 
withstand fire; (5) the NRC Inspector General issued a report in August 
1992 condemning NRC's handling of the Thermo-Lag issue and documenting 
the NRC staff's failure to understand the scope of the problem; (6) in 
April 1994, ITL and its President pleaded guilty to five felony counts 
of aiding and abetting the distribution of falsified test data; (7) on 
September 29, 1994, the U.S. Department of Justice issued a seven-count 
indictment against the manufacturer of Thermo-Lag and its Chief 
Executive Officer for willful violations of the Atomic Energy Act, 
conspiracy to conceal material facts, and making false statements to 
defraud the United States, in connection with $58 million in fire 
barrier material; (8)

[[Page 30468]]

GPUN has known since at least August 11, 1992, that Thermo-Lag 330-1 as 
a structural base material is combustible and that it was in violation 
of Appendices A and R to Part 50 of Title 10 of the Code of Federal 
Regulations (10 CFR) and the NRC Standard Review Plan, NUREG-0800; (9) 
GPUN failed to report the SwRI test results in response to GL 92-08 of 
February 10, 1994, when asked to describe the Thermo-Lag 330-1 fire 
barriers installed as required to meet 10 CFR Part 50, Appendix R; and 
(10) continued reliance on fire watches at Oyster Creek is an 
unreasonable and unnecessary hazard to the public health and safety 
because of an inoperable fire protection system for safe shutdown of 
the reactor and installed combustible material on the shutdown systems.
    Several of the issues listed above have been addressed earlier in 
this decision. Therefore, the NRC staff will only address below the 
remaining plant-specific issues. As discussed earlier in this decision, 
the NRC issued IN 92-82 to inform the industry of the results of 
combustibility tests performed by NIST in early August 1992. These 
tests confirmed the combustibility of Thermo-Lag. As a result of 
discussions with the NRC staff on the subject of Thermo-Lag 
combustibility, GPUN decided to independently verify the results of the 
E-136 tests performed by NIST and contracted SwRI to perform the E-136 
tests. The results of these tests, as documented by the telecopy 
transmittal sheet submitted with the Petition, confirmed the 
combustibility of Thermo-Lag. Contrary to the Petitioners' allegations, 
the NRC staff does not require that licensees report the results of 
their independent testing. It should be noted here that, prior to the 
SwRI testing that confirmed combustibility, the NRC was aware of the 
combustibility of Thermo-Lag and that the NRC was also well aware of 
the results of the E-136 tests performed by GPUN through telephone 
conversations with GPUN personnel, even though there was no requirement 
for GPUN to report these test results.
    The Petitioners also alleged that GPUN did not report to NRC its 
findings of the SwRI test results in its ``Response to Request for 
Additional Information Regarding Generic Letter 92-08, `Thermo-Lag Fire 
Barriers,' '' (RAI) dated February 10, 1994.
    The RAI quoted by the Petitioners did not request that GPUN report 
to NRC its findings of the SwRI test results and, in addition, the NRC 
staff does not require that licensees report the results of their 
independent testing. Therefore the NRC staff has concluded that, 
contrary to the Petitioners' allegation, GPUN did not have to report to 
the NRC its findings of the SwRI test results.
    For the reasons stated above, the suspension of Oyster Creek's 
license, as requested by the Petitioners, is not warranted.

F. Dry-Joint Issue at Comanche Peak Unit 1

    The Petitioners requested that (a) the Comanche Peak Unit 1 license 
be suspended, (b) the licensee perform additional destructive analysis 
for Thermo-Lag configurations, and, (c) the licensee perform fire tests 
on upgraded ``dry-joint'' Thermo-Lag configurations based on the 
following: (1) the licensee's records on the original installation of 
Thermo-Lag fire barriers on conduits and cable trays indicate that its 
contractor followed specifications for pre-buttering all joints; (2) 
NRC Inspection Report Nos. 50-445/93-42; 50-446/93-42 found, based on 
destructive analysis documents, that a concern did exist where Thermo-
Lag conduit joints fell apart easily and did not appear to have any 
residual material of a buttered surface, indicative of a joint that had 
not been pre-buttered; (3) the ``dry joint'' deficiency appeared in 
Room 115A and other areas of the unit; (4) the licensee directly 
contradicts an NRC inspector's findings that were determined in part by 
destructive analysis; (5) the ``dry joint'' or absence of pre-buttering 
of Thermo-Lag panels can be determined only by destructive analysis and 
cannot be determined by a walk down visual inspection; (6) the findings 
reported in the Comanche Peak Unit 1 Region IV Inspection Reports 50-
445/93-42 and 50-446/93-42, based on the limited amount of destructive 
analysis conducted at the unit, constitute a substantial documentation 
of installation deficiencies found in Thermo-Lag fire barriers as 
documented in NRC IN 91-79 and Supplement 1; (7) neither the NRC nor 
the industry, by its agent NEI, nor a utility, have conducted fire 
tests on dry fitted or ``dry joint'' upgraded configurations of Thermo-
Lag 330-1; and (8) the presence of ``dry joint'' upgraded 
configurations in Comanche Peak Unit 1 constitutes an untested 
application of Thermo-Lag fire barriers.
    These allegations were based on the Petitioners' interpretation of 
NRC Inspection Report 93-42 issued on February 21, 1994. By letter of 
November 29, 1994, TU Electric, the licensee for Comanche Peak Unit 1, 
sent a letter to the NRC staff responding to the Petition.
    The term ``joint'' refers to the interface between two adjacent 
Thermo-Lag surfaces. Comanche Peak Unit 1 installation procedures for 
Thermo-Lag fire barriers specify that, during the initial installation 
process, the joints should be pre-buttered (or covered) with Thermo-Lag 
trowel grade material before the mating surfaces are joined to ensure 
adhesion of the surfaces. The term ``dry joint'' refers to the lack of 
Thermo-Lag trowel grade material in a joint. The failure to pre-butter 
a joint with trowel grade Thermo-Lag could result in a weakening of the 
joint during a potential fire exposure and could provide an exposure 
path in the fire barrier envelope. The NRC performed an inspection at 
Comanche Peak Unit 1 on November 2-5, and 23-24, 1993, and January 26-
28, 1994, to compare the Thermo-Lag test specimens with the upgraded 
Thermo-Lag configurations on site. The results of this inspection are 
documented in NRC Inspection Report 93-42. The report stated that there 
appeared to be a large number of deficiencies with the installed fire 
barriers and that an example of these deficiencies involved dry joints 
on conduit overlays installed on pedestal hangers. The NRC inspector 
did not personally observe the dry joints in question. His statements 
were based on observations made by TU Electric and documented in an 
Operations Notification and Evaluation (ONE) form. However, the ONE 
form in question did not identify a dry joint. Instead, the ONE form 
identified a condition that was conservatively reported as an apparent 
dry joint. Upon further evaluation of the ONE form, TU Electric 
determined that the joint in question had in fact been pre-buttered 
with trowel grade Thermo-Lag. These facts are discussed in more detail 
below.
    On November 25, 1992, a speed memo was written by a TU Electric 
contractor identifying ``apparent unsatisfactorily conditions on Unit 1 
commodities.'' This memorandum identified ``an apparent'' dry joint on 
an oversize coupling section (on top of a pedestal hanger). The speed 
memo also stated that, ``we have decided that the best vehicle to call 
attention to these apparent deficiencies would be a letter to your 
attention for further evaluation of the situation. * * *'' The letter 
was forwarded to the appropriate TU Electric engineering section.
    The cognizant TU Electric engineer performed a walkdown of the 
described areas and evaluated the commodities. He conservatively 
initiated a ONE form (the process used by TU Electric to report 
problems and develop resolution for the identified problems). A 
comprehensive evaluation of this condition determined that the joint 
had been pre-buttered. Therefore, the

[[Page 30469]]

engineering resolution for this condition was that ``this is not a 
deficient condition, and there are no generic implications.''
    The originator of the speed memo initially believed that the 
condition in question was a dry joint because of the appearance of the 
joint. During alignment of Thermo-Lag panels, the leading edge of one 
panel contacts the outer edge of a preceding panel and forces most of 
the trowel grade along the initial contact edge toward the inside of 
the Thermo-Lag envelope. Subsequent shrinkage of the trowel grade in 
the joint can give the appearance of a dry joint because the trowel 
grade material is not visible. Therefore, contrary to the Petitioners' 
allegation, there was no ``dry joint'' deficiency on the pedestal 
hanger.
    The Petitioners also alleged that dry joints appear in other 
Thermo-Lag installations at Comanche Peak Unit 1. In response to the 
Petition, TU Electric performed an electronic search of its ONE form 
data base. The search did identify additional ONE forms related to dry 
joints. However, Thermo-Lag rework crews and the quality control 
inspectors at Comanche Peak Unit 1 have used the term ``dry joints'' 
and ``no visible trowel grade material'' synonymously. Upon further 
investigation of these ONE forms, it was determined that trowel grade 
material had in fact been applied to the joints in question. Therefore, 
these ONE forms were also dispositioned as ``not a nonconforming 
condition.'' These findings support the NRC staff's conclusion that, 
contrary to the Petitioners' allegations, there is no evidence of dry 
joints at Comanche Peak Unit 1. The Petitioners' allegations regarding 
dry joints at Comanche Peak Unit 1 are based on premises that are 
faulty and contrary to the information contained in Inspection Report 
93-42.
    In regard to the Petitioners' request that the licensee perform 
fire tests on upgraded ``dry joint'' Thermo-Lag configurations and 
additional destructive analysis, the NRC staff has reviewed the 
documentation provided by the licensee in response to the RAIs 
regarding GL 92-08 and concluded that the licensee's quality assurance 
program gave adequate confidence that the as-installed Thermo-Lag 
configurations at Comanche Peak Unit 1 conform with NRC specification 
requirements for both material and installation attributes.
    Accordingly, suspension of the Comanche Peak Unit 1 license, as 
requested by the Petitioners, is not warranted.

G. Protection of Rubin Feldman

    The Petitioners assert that, rather than protecting the public, the 
NRC is protecting Rubin Feldman, President of the company that 
manufactures Thermo-Lag.
    As discussed earlier, the NRC received allegations in 1991 that 
questioned the adequacy of Thermo-Lag fire barriers. In response (1) 
the Office of the Inspector General (OIG) and the Office of 
Investigations (OI) formed a joint task force to investigate the 
allegations and (2) the Office of Nuclear Reactor Regulation (NRR) 
established a special team to review the safety issues raised by the 
allegations. Throughout its review, the special team gave expert 
technical advice and assistance to the OIG/OI task force. The Director 
of NRR tasked the NRR staff to resolve the technical issues raised by 
the special team. The NRC staff continued to cooperate fully with the 
investigative task force. Further, the NRR staff carried out a full-
scale test program and developed other technical data and information 
for the investigative task force. These NRC staff efforts contributed 
significantly to a referral to the Department of Justice of possible 
wrongdoing by TSI. The referral resulted in a seven-count criminal 
indictment of TSI, the manufacturer and supplier of Thermo-Lag fire 
barriers and of its President, Rubin Feldman, by a Federal Grand Jury. 
The NRC staff continued to support the Department of Justice throughout 
the criminal case.12 In addition, throughout the trial, the NRC 
staff continued to pursue corrective actions consistent with its action 
plan for the resolution of the Thermo-Lag issues. The above facts 
contradict the Petitioners' assertion that the NRC was protecting Rubin 
Feldman.
---------------------------------------------------------------------------

    \12\ The jury returned a verdict of ``not guilty'' on all counts 
of the indictment against TSI and Mr. Feldman.
---------------------------------------------------------------------------

H. NRC Seeming Complicity With Utilities

    The Petitioners also assert that there is seeming complicity 
between the NRC and the licensees and that licensees seek to avoid 
costly replacement of the Thermo-Lag.
    In May 1991, the NRC Office of the Inspector General performed an 
inspection of the NRC's staff performance in regard to Thermo-Lag 
barriers and found indications of inadequate performance by the NRC 
staff in the acceptance and review of Thermo-Lag barriers. 
Subsequently, the NRC staff initiated an aggressive program of 
corrective actions to rectify the deficiencies identified in the review 
and response process, as summarized earlier in this decision.
    In addition, the staff has expended considerable time and effort to 
address and resolve Thermo-Lag issues to ensure that licensees return 
to compliance with existing NRC fire protection requirements. The NRC 
staff issued three requests for additional information regarding GL 92-
08 to each licensee using Thermo-Lag to obtain information on the 
specific Thermo-Lag material installed at each plant, details about the 
corrective actions each licensee intended to take to return to 
compliance with NRC fire protection requirements, and schedules for the 
implementation of these corrective actions. The response of each 
licensee was evaluated by the NRC staff. As a consequence of this 
substantial NRC staff effort, a number of licensees have already 
returned to compliance with NRC requirements by a variety of means 
which include replacing, rerouting, or upgrading existing Thermo-Lag 
barriers, performing post-fire safe shutdown reanalysis, and installing 
additional fire detection and suppression features. All of these 
measures involve some burden on licensees. In addition, some licensees 
have initiated costly programs to perform plant-specific fire endurance 
tests of other fire barriers with the intention of replacing Thermo-Lag 
with these barriers. All licensees who utilize Thermo-Lag will need to 
expend resources commensurate with their reliance on Thermo-Lag to come 
into compliance with NRC fire protection requirements. NRC staff 
oversight will ensure that this is the case.
    The Petitioners' assertion of seeming complicity with utilities on 
the part of the NRC staff is unfounded in the light of the significant 
NRC staff efforts to ensure that licensees expend the resources 
necessary to return to compliance with NRC requirements.

IV. Conclusion

    The Petitioners request that the NRC order the immediate shutdown 
of all reactors using Thermo-Lag and the suspension of Oyster Creek, 
Peach Bottom Units 1 and 2, and Comanche Peak Unit 1 operating 
licenses.
    For the reasons discussed above, I find no basis for taking such 
actions. Rather, on the basis of the review efforts by the NRC staff, I 
conclude that the issues raised by the Petitioners are being addressed 
by licensees in a manner which assures adequate protection of the 
public health and safety. Accordingly, the Petitioners' requests for 
action pursuant to 10 CFR 2.206 are denied.

[[Page 30470]]

    A copy of this Decision will be placed in the Commission's Public 
Document Room, Gelman Building, 2120 L Street, N.W., Washington, D.C., 
and at the Local Public Document Room for the named facilities. A copy 
of this Decision will also be filed with the Secretary for the 
Commission's review as provided in 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 issuance, unless the 
Commission, on its own motion, institutes a review of the Decision 
within that time.

    Dated at Rockville, Maryland this 3rd day of April 1996.

    For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-15149 Filed 6-13-96; 8:45 am]
BILLING CODE 7590-01-P