[Federal Register Volume 60, Number 175 (Monday, September 11, 1995)]
[Pages 47187-47193]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-22461]


[Docket No. 50-293]

Boston Edison Company, Pilgrim Nuclear Power Station; Issuance of 
Director Decision Under 10 CFR 2.206

    Notice is hereby given that the Director, Office of Nuclear Reactor 
Regulation, granted in part and denied in part a Petition dated March 
10, 1995 (Petition), filed pursuant to 10 CFR 2.206 by Ms. Mary 
Elizabeth Lampert and 62 other persons (Petitioners).
    The Petition requested that during the March 25, 1995, refueling 
outage and In-Vessel Visual Inspection conducted by the licensee, 
certain technical concerns be addressed, and that before Pilgrim goes 
back on-line, appropriate repairs be made or corrective action be 
taken, and that the U.S. Nuclear Regulatory Commission (NRC or 
Commission) discuss the status of such repairs or corrective actions 
with the public in Plymouth, Massachusetts. The Petition also requested 
that the NRC terminate its policy of issuing Notices of Enforcement 
Discretion (NOEDs) and asserted that the NRC has not been enforcing its 
    On April 19, 1995, the Director informed the Petitioner that the 
NRC management and staff was meeting with the Boston Edison Company 
(licensee) on May 11, 1995, and they would hold a meeting to receive 
public input on the evening of May 11, 1995. The Petitioner's request 
to discuss the status of repairs or corrective actions was granted by 
virtue of the public meeting.
    The Director of the Office of Nuclear Reactor Regulation has denied 
the Petitioners' requests to require repairs and corrective actions 
before permitting the Pilgrim plant to resume operation, and to 
terminate the use of NOEDs.
    The reasons for this decision are explained in the ``Director's 
Decision Under 10 CFR 2.206,'' (DD-95-19) which is available for public 
inspection in the Commission's Public Document Room, in the Gelman 
Building, Lower Level, 2120 L Street, NW., Washington, DC 20555 and at 
the Local Public Document Room for the Pilgrim facility at Plymouth 
Public Library, 11 North Street, Plymouth, Massachusetts 02360.
    A copy of the Decision will be filed with the Office of the 
Secretary for the Commission's review in accordance with 10 CFR 
2.206(c). As provided by this regulation, the Director's Decision will 
constitute the final action of the Commission 25 days after date of 
issuance of the Decision unless the Commission, on its own motion, 
institutes review of the Decision within that time period.

    Dated at Rockville, Maryland, this 31st day of August 1995.

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

Appendix A to this Document--Director's Decision Under 10 CFR 2.206: 
DD-95-19; Boston Edison Company, License No. DRP-35

I. Introduction

    Ms. Mary Elizabeth Lampert and 62 other individuals 
(Petitioners) submitted a Petition dated March 10, 1995, pursuant to 
10 CFR 2.206 requesting action with regard to the Pilgrim Nuclear 
Power Station (Pilgrim), operated by the Boston Edison Company 
    The Petition requested that: (1) during the refueling outage and 
In-Vessel Visual Inspection scheduled for March 25, 1995, by the 
licensee, certain technical concerns be addressed, and that before 
Pilgrim goes back on-line, appropriate repairs be made or corrective 
action be taken; (2) the U.S. Nuclear Regulatory Commission (NRC or 
Commission) discuss the status of such repairs or corrective actions 
with the public in Plymouth, Massachusetts; and (3) the NRC 
terminate its policy of issuing Notices of Enforcement Discretion 
(NOEDs) and begin enforcing the regulations again.
    As the bases for these requests, the Petitioners identified 
three groups of technical concerns: (1) age-related deterioration of 
25 safety related reactor internals; (2) parts and components 
``known to be a problem at Pilgrim,'' including the core shroud, 
water level indicators, quality assurance for fuel pool cooling 
system during loss-of-coolant accident/loss of offsite power, motor-
operated valves, containment integrity, drywell liner corrosion 
vulnerability, station blackout vulnerability, and Rosemount 
transmitters; and (3) parts and components ``potentially a problem 
at Pilgrim,'' including potential fuel rod corrosion and substandard 
and/or counterfeit parts. The Petitioners contend that allowing the 
reactor to operate under a NOED cannot pose less risk to the public 
health and safety than keeping the reactor shut down until NRC 
regulations are met.

II. Background

    By letter dated April 19, 1995, the NRC acknowledged receipt of 
the Petition and offered a public meeting, which was held in 
Plymouth, Massachusetts on May 11, 1995. At that meeting, the 
results of the licensee's inspections conducted during the outage 
were discussed.
    I have completed my evaluation of the Petition. As explained 
below, Petitioners have failed to raise any safety concern which 
would warrant delaying restart of the Pilgrim Nuclear Power Station 
(which occurred on June 2, 1995), and the Petitioners' request that 
the NRC terminate the use of NOEDs is denied.

III. Discussion

A. Age-Related Deterioration of Reactor Internals

    Many components inside boiling-water reactor (BWR) vessels 
(i.e., internals) are made of materials such as stainless steel and 
various alloys that are susceptible to corrosion and cracking. As 
materials age, they degrade. This degradation can be accelerated by 
stresses from temperature and pressure changes, irradiation effects 
on material properties, chemical interactions, and other corrosive 
environments. As BWRs age, the amount of cracking is expected to 
increase. Several cases of internals cracking and degradation have 
been reported to the NRC over the years. In a number of cases, the 
NRC has concluded that full power operation of the reactor with 
time-dependent degradation, related to the operating environment, of 
reactor vessel internals is acceptable as long as the American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code 
(ASME Code) safety margins are satisfied and maintained. In the 
remaining cases, replacement or repairs were performed on the 
degraded components or internals. The NRC has met with industry 
every year since 1988 to review the generic safety implications of 
reactor internals potentially susceptible to age-related cracking. 
Additionally, a special industry review group, the Boiling Water 
Reactor Vessels and Internals Project (BWRVIP), was formed to focus 
on resolution of reactor vessel and internals degradation.
    Several industry standards and regulatory requirements and 
guidelines are in place to address inservice inspections (ISIs) of 
reactor components. Moreover, the NRC and industry have responded as 
new issues emerge. For example, issued Generic Letter (GL) 94-03, 
``Intergranular Stress Corrosion Cracking of Core Shrouds (IGSCC) in 
Boiling Water Reactors,'' ``in July 1994 requesting Licensees to 
inspect their shrouds and provide an analysis justifying continued 
operation until inspections could be completed. General Electric 
issued Services Information Letter (SIL) No. 588, ``Top Guide and 
Core Plate Cracking,'' in February 1995 providing specific 
recommendations for inspections of BWR top guides and core plates. 
In addition to addressing emerging the BWRVIP is working on a 
comprehensive plan that will provide detailed guidance on managing 
cracking in all BWR internals. The plan will address cracking 
susceptibility, safety consequences, inspection scope and 
methodology, flaw evaluation, repair strategies, and mitigation of 
degradation. Several top level executives and technical staff of the 
Licensee are on the various BWRVIP committees that are developing 
generic standards for ISI and repairs.
    Petitioners request that 25 components be inspected during the 
1995 refueling outage (RFO No. 10), and that they be free of any 
signs of IGSCC or other kind of fatigue. 

[[Page 47188]]
During RFO No. 10, the licensee indicated completion of the ISI 
examinations for the third period of the second Pilgrim 10-year 
inspection interval in accordance with Section XI of the ASME Code, 
1980 Edition with Winter 1980 Addenda. This included all 25 
components requested by the Petitioners, except the steam separator, 
neutron source holder and surveillance sample holders which are not 
safety-related components. The in-core neutron flux monitor 
components, in-housings, guide tubes, dry tubes, the vessel head 
cooling spray nozzle, and the fuel supports are not required by NRC 
regulations to be inspected. The NRC inspected Pilgrim's ISI program 
and related activities during the 1994 RFO No. 9 and concluded that 
the second interval program plan was sufficiently comprehensive to 
ensure safety and met the requirements of the ASME Code, and thus 10 
CFR 50.55a(a)(2). The ISI examinations conducted in RFO No. 10 
included the core support structure, control rod drive housing, core 
spray internal piping and spargers, and feedwater spargers.
    Augmented examinations were also conducted in which various 
internals were examined, including the shroud support and access 
hole covers, jet pump riser braces, shroud head bolts, jet pump 
sensing lines, steam dryer support, steam dryer baffle plate, top 
guide, core plate, and control rod stub tubes.
    Control blades (control rods for BWRs) are replaced at specified 
intervals. The licensee also implemented a preemptive repair of its 
core shroud due to the high susceptibility to IGSCC. See Section 
III.B.(1), below. As discussed during the May 11, 1995, meeting 
between the NRC and the public, the inspection results from RFO No. 
10 did not reveal any indications of significant time-dependent 
deterioration of the reactor internals.
    The NRC staff concludes that the inspections, examinations, and 
repairs performed by the licensee during RFO No. 10 and previous 
outages are sufficient to provide reasonable assurance that no age-
related failure of components or internals would occur during the 
next operating cycle, which is scheduled to end March 21, 1997. 
Design features, plant procedures, and operator training are 
developed to ensure safety in the unlikely event that a failure were 
to occur. The NRC will continue to take regulatory action on a 
plant-specific or generic basis, as may be appropriate, when time-
dependent degradation issues are identified. During the next 
refueling outage, the licensee will again conduct an in-vessel 
inspection of safety-related interval components.
    Accordingly, Petitioners have not raised a safety concern 
regarding age-related degradation of reactor internals at Pilgrim 
which would have warranted prohibiting restart after RFO No. 10.
    B. Parts and Components Known To Be a Problem at Pilgrim

(1) Core Shroud

    Petitioners express concern about the type of repairs that would 
be done to the core shroud during RFO No. 10, based on ``the 
different approach taken in Germany at the Wuergassen NPS and at the 
Oyster Creek NPS in NJ.'' Petitioners state that German nuclear 
regulators required replacement of shrouds with cracking, rather 
than repair of the shroud. Petitioners state that at Oyster Creek, 
ten tie rods are attached to holes in Type 304 stainless steel, 
which is subject to IGSCC and is welded to the bottom of the core 
shroud assembly. Petitioners are concerned that if the same approach 
were used at Pilgrim, there would be problems with the structural 
integrity of the materials the tie rods are welded to and with 
``loose parts.''
    Officials of PreussenElektra AG, the owner of Wuergassen, 
initially intended to replace the core shroud at Wuergassen, as 
reported in Nucleonics Week on November 24, 1994. Differences in the 
design of Wuergassen and NRC-licensed BWRs exist which would make 
replacement of the core shroud at Wuergassen less complicated than 
at NRC-licensed plants. For example, the shroud at Wuergassen is 
bolted on to the shroud support, whereas shrouds of NRC licensees 
are welded. However, in a press release issued June 1, 1995, 
PreussenElektra AG decided to decommission the Wuergassen NPS based 
on economic considerations. As a result, replacement of a BWR core 
shroud, foreign or domestic, has yet to be undertaken.
    By letter dated November 25, 1994, the NRC staff issued the 
``Safety Evaluation Regarding the Oyster Creek Core Shroud Repair,'' 
which approved the scheduled repair as an acceptable alternative to 
the standards of the ASME Boiler and Pressure Vessel Code. See 10 
C.F.R. Sec. 50.55a(a)(2) and 50.55a(a)(3)(i). Oyster Creek and 
Pilgrim are utilizing similar tie-rod assemblies to structurally 
replace the core shroud during normal and accident conditions. The 
difference in the number of tie-rod assemblies used, i.e., ten tie-
rod assemblies at Oyster Creek and four tie-rod assemblies at 
Pilgrim, is related to the contracted vendor's loading distribution 
design and the associated hardware on the tie-rod assembly. The NRC 
staff has thoroughly reviewed the Pilgrim repair design and 
conducted inspections during the core shroud repair process. The 
staff issued the ``Safety Evaluation Regarding Pilgrim Nuclear Power 
Station Core Shroud Repair,'' dated May 12, 1995. A synopsis of our 
review follows.
    The design of the Pilgrim shroud repair consists of four (4) 
stabilizer assemblies, which are installed 90 deg. apart in the 
shroud/reactor vessel annulus, between attachment points at the top 
of the shroud and the gusset assemblies on the lower shroud support 
plate. Each stabilizer assembly consists of a tie rod, and upper 
spring, a lower spring, an upper bracket and other smaller parts. 
The tie rod provides the vertical load transfer from the upper 
bracket to the reactor pressure vessel (RPV) gusset attachment and 
supports the springs. The upper spring provides radial load transfer 
at the top guide elevation from the shroud to the RPV. The lower 
spring provides radial load transfer from the shroud at the core 
plate elevation to the RPV. The upper bracket provides an attachment 
to the top of the shroud and restrains the upper shroud weld. Upper-
mid and lower-mid supports along the tie rod length provide radial 
load transfer for the mid sections of the shroud and increase the 
natural frequency of the tie rods to reduce flow-induced vibration. 
Two wedges between the core support plate and the shroud are also 
installed at each stabilizer location to prevent relative motion of 
the core plate to the shroud. Each cylindrical section of the shroud 
between welds H1 through H9 is prevented from unacceptable lateral 
motion by the stabilizers. The section between H9 and H10 is 
prevented from unacceptable motion by the existing gussets. The 
lower end of the stabilizers are attached to pins which are placed 
in holes cut into gusset plates at the bottom. The gusset assemblies 
and their welds are Inconel and are not considered subject to 
cracking by industry and the NRC staff. Inconel is a nickel based 
alloy which is less likely to corrode and degrade than stainless 
steel, which is an iron based alloy. However, these welds, including 
those attaching the gussets to the vessel and to the lower shroud 
support plate (which must resist the vertical stabilizer loads) have 
been inspected for cracks during this outage, and no crack 
indications were found. Together, the tie rods and lateral 
restraints resist both vertical and lateral loads resulting from 
normal operation and design accident loads, including seismic loads 
and postulated pipe ruptures.
    The NRC staff found that the proposed repair does not affect the 
ability of operators to insert control rods, the performance of the 
ECCS, particularly the core spray system, or the ability to reflood 
and cool the core. The staff concluded that the proposed repair does 
not pose adverse consequences to plant safety; therefore, plant 
operation is acceptable with the proposed core shroud repair 
    In compliance with 10 CFR 50.55a(a)(3)(i), the core shroud 
repair has been designed as an alternative to the requirements of 
the ASME Code. Based on a review of the shroud modification hardware 
from structural, systems, materials, and fabrication considerations, 
the NRC staff concludes that the proposed modifications of the 
Pilgrim core shroud would provide an acceptable level of quality and 
safety. The staff has determined that the licensee's repair of the 
core shroud will not result in any increased risk to the public 
health and safety and is, therefore, acceptable.

(2) Water Level Indicators

    Petitioners assert that because of a pipe design deficiency, 
water level indicators at Pilgrim are not fully operable due to 
high-pressured gas in the water, and that operator training is not 
the appropriate solution.
    Level anomalies were observed in reactor vessel water level 
indication at several BWRs during controlled depressurization, while 
commencing plant outages or following reactor trips. These anomalies 
consisted of ``spiking'' or ``notching'' of level indication, and in 
one instance, a sustained error in level indication. The root cause 
of these level indication anomalies is the effect of non-condensible 
gas dissolved in the reference leg of ``cold reference leg'' type 
water level instruments. Under rapid depressurization conditions, 
non-condensible gases can cause significant errors in the level 
    Cold reference leg water level instruments measure reactor 
vessel water level by 

[[Page 47189]]
measuring the differential pressure of two columns of water, i.e., the 
variable leg and the constant height reference leg. The reference 
leg is maintained filled to a constant height of water by the 
condensate chamber. Steam is condensed in the condensate chamber and 
keeps the reference leg full. Excess condensate is returned to the 
vessel through the steam supply line. Non-condensible gases, such as 
hydrogen and oxygen, formed by radiolysis in the reactor vessel, are 
present in the steam supplied to the condensate chamber. The gases 
can collect in the condensate chamber and can accumulate to high 
partial pressures. The gases then become dissolved in the water at 
the top of the reference leg, and the dissolved gases can be 
transported down the reference leg by small leaks in valves and 
fittings at the bottom of the reference leg, diffusion, and/or 
thermal convection.
    Dissolved gases in the reference leg do not present a problem 
unless the instrument is depressurized. When depressurized, the 
gases come out of solution and form bubbles that travel up the 
reference leg. During slow depressurization, level indication has 
been seen to temporarily ``spike'' or ``notch'' while a bubble moves 
through the vertical sections of the piping. Significant spiking may 
automatically actuate such systems as the primary containment 
isolation system (PCIS). This occurred at the Pilgrim plant. After 
spiking, which is of short duration, the indicated water level 
returns to actual level. Level spiking is of little significance. 
Bubbling of the gases may eject a significant amount of water from 
the reference leg. Loss of reference leg inventory will cause an 
erroneously high level indication. This occurred during a normal 
plant cooldown on January 21, 1993, at Washington Nuclear Power Unit 
2 (WNP-2), resulting in a 32-inch error in level indication that 
gradually recovered over a period of 2 hours. If the reactor is 
rapidly depressurized, as would occur during a design basis loss-of-
coolant accident (LOCA) or opening of the automatic depressurization 
system (ADS) valves, even larger errors in the level indication 
could result. However, analyses presented by the industry indicated 
that significant errors would not be expected until the reactor is 
depressurized below approximately 450 psi.
    The NRC staff has taken several actions to address this problem. 
The BWR Owners Group (BWROG) Regulatory Response Group (RRG) was 
activated during July 1992. The staff also issued Information Notice 
92-54 in July 1992, GL 92-04 in August 1992, and Information Notice 
93-27 in March 1993 to alert licensees to the potential problem and 
to request information concerning actions taken or planned by 
licensees in response to potential errors in level indication. The 
BWROG conducted a test program to support their efforts to resolve 
this issue. The results of the BWROG reference leg de-gas test 
program confirmed that no significant errors in level indication 
will occur until the reactor is depressurized below 450 psig, and 
that large errors in level indication are possible once the reactor 
is depressurized to lower pressures.
    The NRC staff received additional information from the BWROG 
pertaining to reactor vessel water level instrumentation 
inaccuracies during normal depressurization due to the effects of 
non-condensible gas. At the staff's request, the BWROG submitted a 
report on May 20, 1993, discussing the impact of level errors on 
automatic safety system response and operator actions during 
transients and accidents initiated from reduced pressure conditions 
during plant cooldown (shutdown mode). Based on this information, in 
addition to the January 21, 1993, WNP-2 event, and data from the 
reference leg de-gas testing that was conducted by the BWROG, the 
staff concluded that additional short-term actions needed to be 
taken for protection against potential events occurring during 
normal cooldown. On May 28, 1993, NRC Bulletin (NRCB) 93-03,
    ``Resolution of Issues Related to Reactor Vessel Water Level 
Instrumentation,'' was issued, in which the staff requested each BWR 
licensee to implement additional short-term compensatory actions, 
and to implement a hardware modification to resolve this issue at 
the next cold shutdown after July 30, 1993.
    The staff has received responses to NRC Bulletin 93-03 from all 
licensees. All licensees completed short-term compensatory actions 
and committed to install hardware modifications. Licensees for all 
affected plants have either completed installation of hardware 
modifications or are currently shutdown and will install the 
hardware modifications prior to restart.
    To solve the problem identified in NRC Bulletin 93-03, Pilgrim 
installed a backfill modification to all safety-related water level 
instrumentation in July 1993. Non-safety-related control 
instrumentation was not modified by Pilgrim, because such 
instrumentation was not covered by the actions requested in NRC 
Bulletin 93-03.
    As Petitioners note, an event occurred at Pilgrim on November 8, 
1993, involving the non-safety-related water level instrumentation. 
This event was caused by failure of the licensee to back flush the 
feedwater control instrumentation reference legs prior to restart 
due to procedural inadequacy and failure to cross-check multiple 
indications of reactor vessel water level during startup due to 
operator error. This event is not safety significant for the 
following reasons:
    (a) event initiation was the result of two independent errors 
which are not expected to have a high frequency of recurrence;
    (b) safety systems and non-safety systems are separated by 
design; thus, the availability and capability of the safety systems 
should not be impacted by errors in the non-safety instrumentation 
and the ability of safety systems to protect the plant should not be 
compromised; and
    (c) the safety systems responded to the event as expected.
    This issue is closed because the licensee took adequate 
corrective actions in response to the November 8, 1993, event. See 
NRC Inspection Report 50-293/93-20, dated January 11, 1994.
    Based on the above, Petitioners have not raised a substantial 
safety concern regarding safety-related water level instrumentation 
at Pilgrim.
    (3) Quality Assurance for Fuel Pool Cooling System During LOCA/
    The Petitioners asserted that workers would be exposed to fatal 
levels of radiation while manually activating the backup cooling 
system during a LOCA.
    In November 1992 two engineers working under contract at 
Susquehanna Steam Electric Station filed a 10 CFR 21.21 report. The 
report detailed design concerns at Susquehanna that could lead to 
the sustained loss of forced cooling for the stored spent fuel under 
certain accident or abnormal conditions. The engineers postulated 
that the environmental conditions developed following a loss of 
forced cooling would adversely affect equipment necessary for safe-
shutdown and accident mitigation. The engineers concluded that these 
issues had generic implications.
    Between November 1992 and October 1994, the NRC staff performed 
an extensive evaluation of the Susquehanna spent fuel pool cooling 
design concerns. The staff concluded that these concerns were of low 
safety significance in the ``Final Safety Evaluation By the Office 
of Nuclear Reactor Regulation Regarding Loss of Spent Fuel Pool 
Cooling Events,'' dated June 19, 1995. This conclusion was based on 
the fact that the probability of recovering forced cooling of the 
stored spent fuel with access to the necessary equipment was high, 
and the probability of experiencing a severe core damage accident, 
which may prevent access to systems need to cool the spent fuel 
pool, was low.
    The staff issued Information Notice 93-83, ``Potential Loss Of 
Spent Fuel Pool Cooling Following A Loss Of Coolant Accident,'' 
(October 7, 1993), describing the Section 21.21 report related to 
Susquehanna. The information notice did not require specific action 
by licensees. Recognizing the plant-specific design features and 
operational controls of most spent fuel pool cooling system designs, 
the staff concluded that further evaluation of spent fuel pool 
storage safety issues at other plants was warranted to determine the 
need for further generic action.1

    \1\ In the near future, the staff will issue an additional 
information notice describing the results of its detailed evaluation 
of the Susquehanna facility. This information notice will be an 
interim communication and will not represent the end of the staff's 
generic review.

    The staff has developed and begun implementing a generic action 
plan to evaluate generic issues. On-site safety assessments of spent 
fuel storage at selected reactor facilities have been completed. 
Monticello Nuclear Power Plant is similar to Pilgrim and was one of 
the nuclear facilities assessed during the week of March 27, 1995. 
The assessment team concluded that the potential for a sustained 
loss of spent fuel pool cooling or a significant loss of spent fuel 
pool coolant inventory at the site visited was remote based on 
observed design features and operational controls. Based on the 
above, the NRC staff has concluded that the Petitioners have not 
identified any safety concerns at Pilgrim regarding spent fuel pool 
cooling during a LOCA/LOOP. 

[[Page 47190]]

(4) Motor-Operated Valves

    Petitioners request information on the status of the motor-
operated valve (MOV) program at Pilgrim, and inquire why Pilgrim has 
not been required to fix all MOVs during the March 1995 outage.
    The NRC issued GL 89-10, ``Safety-Related Motor-Operated Valve 
Testing and Surveillance'' (June 28, 1989) to request that licensees 
verify the capability of all safety-related MOVs to perform their 
design basis functions. GL 89-10 requested that licensees complete 
differential pressure and flow testing for the verification of MOV 
design basis capability within 5 years after the issuance of GL 89-
10 or three refueling outages after December 1989, whichever was 
    Pilgrim is scheduled to complete its MOV Design Basis Capability 
Verification by April 1997. Although this is somewhat later than 
some other plants, the licensee is being given the same number of 
outages (three outages with 24 month cycles) as other licensees to 
complete the verification, and the program commenced somewhat later 
at Pilgrim due to the 1990 restart from an extended outage.
    During the implementation of GL 89-10, licensees have discovered 
more MOV concerns and experienced greater difficulty in conducting 
MOV tests at full design basis differential pressure and flow than 
envisioned when the GL 89-10 schedule was established. Where 
significant MOV problems are identified, the NRC ensures that 
licensees resolve these problems promptly. Further, when the 
evaluation of NRC-sponsored MOV test results indicated potential 
problems with specific MOVs in high pressure systems at boiling-
water reactor (BWR) nuclear power plants, the NRC issued Supplement 
3 to GL 89-10 in October 1990. Supplement 3 requested that BWR 
licensees promptly evaluate the capability of MOVs used for 
containment isolation in the steam lines of the high-pressure 
coolant injection and reactor core isolation cooling systems and in 
the supply line to the reactor water cleanup system. Further, the 
staff issued Supplement 5 to GL 89-10 in June 1993, requesting that 
licensees ensure that new information on the increased inaccuracy of 
MOV diagnostic equipment be addressed. These two actions were 
satisfactorily completed by Pilgrim.
    The NRC staff has been monitoring the progress of the GL 89-10 
program at Pilgrim closely. From December 13 to 17, 1993, and March 
22 to 25, 1994, the NRC staff conducted an inspection of the GL 89-
10 program at Pilgrim. As stated in NRC Inspection Report 50-293/92-
80, the NRC staff had the following findings as a result of the 
March 1992 inspection:
    (a) The method used to set the MOV torque switches using 
diagnostic testing equipment was inadequate;
    (b) the torque switch settings on several safety-related MOVs 
were not set in accordance with the plant design documents;
    (c) corrective actions taken in response to an internal audit of 
the GL 89-10 Program regarding the torque switch settings of safety-
related valves were inadequate;
    (d) the GL Supplement 3 response for the reactor water cleanup 
system isolation valve 1202-5 was inadequate;
    (e) plans for conducting design-basis differential pressure 
testing have not been clearly established;
    (f) the current work instructions for performing design basis 
reviews and switch setting calculations lack adequate detail; and
    (g) a considerable effort remains to implement the GL 89-10 
program in a timely manner.
    The NRC staff found considerable progress in the licensee's MOV 
program since the initial NRC team inspection in March 1992. 
Particularly, the staff concluded that the findings from the March 
1992 inspection had been satisfactorily addressed. See Inspection 
Report No. 50-293/93-22 (April 14, 1994). In addition, the testing 
of differential pressure and/or static pressure of all of the 
Priority 1 (highest risk) MOVs that can be tested was completed by 
the end of RFO No. 10. Additionally, the licensee has evaluated all 
of the GL 89-10 MOVs for susceptibility to pressure locking and 
thermal binding and, by the end of RFO No. 10, completed 
modifications on the few valves that were considered susceptible. 
The staff concludes that the licensee is on schedule to meet its 
April 1997 completion date.
    Based on the progress made to date by the licensee in 
implementing its GL 89-10 program at Pilgrim, the NRC staff did not 
consider it necessary that the licensee complete its GL 89-10 
program during RFO No. 10. In addition to review of the licensee's 
submittals in response to GL 89-10 and its supplements, the NRC 
staff is conducting an extensive inspection program to evaluate the 
MOV program implemented in response to GL 89-10 at Pilgrim, as well 
as at other nuclear power plants. The NRC staff concludes that the 
licensee has substantially reduced the concerns with MOV operation 
under design basis conditions and is progressing significantly 
toward completing the GL 89-10 program. Nevertheless, if significant 
MOV problems are identified at Pilgrim, the licensee will be 
responsible for addressing those problems in accordance with their 
safety significance, irrespective of the GL 89-10 completion 
schedule. Further, the NRC will continue to take regulatory action 
on a plant-specific or generic basis, as appropriate, when MOV 
problems are identified.
    Based upon the actions taken to date by the licensee to address 
safety-related MOV issues and the NRC's inspections regarding the 
licensee's actions on the GL 89-10 program, the NRC staff concludes 
that no corrective actions are required.

(5) Containment Integrity

    Petitioners ask whether the hardened wetwell vent system 
(HWWVS), referred to as the ``Torus Vent'', which ``allows venting 
of radioactive effluents directly into our atmosphere,'' will be 
corrected in RFO No. 10.
    The licensee installed the HWWVS modification during the 1986-
1988 outage, thus providing the capability to establish alternate 
containment decay heat removal if RHR torus cooling capability is 
lost. The direct torus venting minimizes the potential for core 
damage and containment failure. The HWWVS has the capability of 
mitigating a wide range of events including many that are beyond the 
Design Basis Accidents for the facility. Its installation, along 
with the procedures for its use, will reduce the likelihood of a 
core melt from accident sequences involving the loss of long-term 
decay heat removal. This accomplished by preventing any further 
damage to safety equipment in the reactor building by ensuring that 
the piping from the containment to the venting stack will not fail. 
Further, as a mitigation measure, the vent pathway is located in the 
wetwell air space. This location ensures that the vented non-
condensible gases will pass through the suppression pool thereby 
significantly scrubbing the fission products. The HWWVS is an 
improvement that the NRC staff recommended in its Mark I Containment 
Performance Improvement Program, which identified plant 
modifications that could enhance the capability to both prevent and 
mitigate the consequences of severe accidents.
    The HWWVS has valves that are kept closed during plant 
operation, assuring containment integrity. Additionally, the HWWVS 
design incorporates a device called a rupture disc, which provides 
an additional leak-tight barrier to further prevent the transport of 
the containment atmosphere in the wetwell to the atmosphere. The 
HWWVS is not in use during normal plant operation, nor is it 
expected to be used during anticipated transient conditions. 
Petitioners have not demonstrated any basis why this system should 
be ``corrected.''

(6) Drywell Liner Corrosion

    Petitioners request information on the status of drywell liner 
corrosion vulnerability and asks whether it would be corrected 
during RFO No. 10.
    The NRC issued GL 87-05, ``Request For Additional Information-
Assessment of Licensee Measures to Mitigate and/or Identify 
Potential Degradation of Mark I Drywells,'' as a result of the 
November 1986 discovery of corrosion of the Oyster Creek steel 
drywell in the area of the sand cushion. GL 87-05 did not establish 
any regulatory requirements other than for Mark I licensees to 
provide the staff with information as to what actions, if any, were 
being taken as a result of the Oyster Creek finding. The licensee 
responded to GL 87-05 by letter dated May 11, 1987. The licensee 
implemented a surveillance program to detect whether a corrosive 
environment exists on the external surface of the drywell. This is 
done by checking the drywell liner air gap drain lines for the 
presence of water during every refueling outage.
    In January 1987, prior to issuance of GL 87-05, the licensee 
conducted ultrasonic inspections of the interior of the drywell 
liner in the area of the sand drains, which confirmed liner 
integrity. In January 1988, the drain lines were verified not to be 
blocked by using a boroscope. As of the last surveillance, conducted 
on March 31, 1995, no water leakage had been detected. Petitioners 
have not demonstrated any basis for correcting this system.

(7) Station Blackout

    Petitioners request information on station blackout 
vulnerability and ask whether it would be corrected during RFO No. 

[[Page 47191]]

    On December 23, 1993, the NRC issued ``NRC Pilot Station 
Blackout Team Inspection,'' a report concerning the Pilgrim plant, 
Inspection Report 50-293/93-80. The purpose of that inspection was 
to review Pilgrim's programs, procedures, training, equipment and 
systems, and supporting documentation for implementing the Station 
Blackout (SBO) Rule, 10 CFR 50.63. The actions taken to implement 
the station blackout rule are important because many of the systems 
required for decay heat removal and containment cooling are 
dependent on the availability of alternating current (ac) power. In 
the event of a station blackout, relatively few systems that do not 
require ac power are depended upon to remove decay heat, until ac 
power is restored.
    The staff concluded in Inspection Report 50-293/93-80 that:
    (a) Pilgrim had sufficient condensate inventory to cope with an 
8-hour SBO duration;
    (b) all areas which contained equipment needed for SBO coping 
had proper cooling;
    (c) there was sufficient evidence that the torus temperature and 
the reactor vessel conditions would be maintained according to the 
plant TSs;
    (d) the overall communications capability available during an 
SBO were adequate;
    (e) adequate emergency lighting was available to support plant 
personnel operations during a station blackout; and
    (f) plant modifications were properly installed, and post-
modification and pre-operational tests were conducted in accordance 
with proper test procedures. Quality assurance and maintenance 
practices, operator training, and staffing levels were appropriate 
to cope with an SBO.
    Accordingly, the Pilgrim plant is in compliance with Section 
50.63 and the plant does not have a SBO vulnerability requiring 
``correction'' during RFO No. 10.

(8) Rosemount Transmitters

    Petitioners request information on the status of Rosemount 
transmitters at Pilgrim, and ask whether all would be inspected and 
corrected during RFO No. 10.
    On December 22, 1992, the NRC staff issued Bulletin 90-01, 
Supplement 1, ``Loss of Fill-Oil in Transmitters Manufactured by 
Rosemount,'' which requested that licensees take appropriate 
corrective actions for Model 1153, Series B and D, and Model 1154 
Rosemount transmitters manufactured before July 11, 1989, and used 
in safety-related applications or Anticipated Transient Without 
Scram (ATWS) systems. The performance of a transmitter that is 
leaking fill-oil gradually deteriorates and may eventually lead to 
failure. Although some failed transmitters have shown symptoms of 
loss of fill-oil prior to failure, it has been reported that in some 
cases the failure of a transmitter that is leaking fill-oil may be 
difficult to detect during operation. Transmitter failures that are 
not readily detectable increase the potential for common mode 
failure and may result in the affected safety system not performing 
its intended safety function. Supplement 1 identified specific 
actions for replacement or enhanced surveillance monitoring of the 
these transmitters, used in high pressure (greater than 1500 psi), 
medium pressure (greater than 500 psi and less than 1500 psi), and 
low pressure (less than 500 psi) applications.
    The licensee responded to the requested actions of Bulletin 90-
01, Supplement 1, on March 5, 1993 and August 30, 1993. There are a 
total of 40 Model 1153B transmitters currently in service, 14 medium 
pressure transmitters and 26 low pressure transmitters. The licensee 
committed to include each of these transmitters in its enhanced 
surveillance monitoring program. The licensee stated that there were 
no Model 1153D or 1154 transmitters currently in service.
    The licensee also stated that there were 33 Model 1153B 
transmitters, manufactured after July 1989, in service. Such 
transmitters are not subject to the Bulletin 90-01, Supplement 1, 
requested actions because Rosemount corrected the oil leakage 
problem by an improved manufacturing and quality assurance process. 
Although Supplement 1 does not require these transmitters to be 
included in an enhanced surveillance monitoring program, the 
licensee has chosen to include them in its program. The licensee's 
enhanced surveillance program is based on both the trending of 
operating drift data and calibration drift data, and is in 
accordance with Rosemount Technical Bulletin No. 4.
    The NRC, with assistance from its contractor, reviewed the 
licensee's response to Supplement 1, and in a letter dated November 
29, 1994, concluded that the licensee satisfied the reporting 
requirements and conformed to the requested actions of Bulletin 90-
01, Supplement 1. Accordingly, no further actions by the licensee 
were required with respect to this Rosemount Issue during RFO No. 

C. Parts and Components Potentially a Problem at Pilgrim

(1) Fuel Rod Corrosion

    Petitioners request information regarding the status of 
zirconium alloy tubes installed at Pilgrim, and asks if their 
susceptibility to nodular corrosion would be corrected during RFO 
No. 10.
    Nodular corrosion is a phenomena seen in plants that have copper 
in the reactor water at a concentration in the 20-30 parts per 
billion (ppb) range. Pilgrim systems design limits copper levels to 
less than 1 ppb in the reactor water. Additionally, all fuel rod 
cladding in use at Pilgrim has been subject to the GE Nuclear Energy 
in-process heat treatment (IPHT) process2, which is a heat 
treatment process that evenly distributes the composition of the 
alloy thus lowering the susceptibility to nodular corrosion. Pilgrim 
has not experienced nodular corrosion, and failure of fuel rods is 
not expected from this phenomenon.

    \2\ TWCA does not produce fuel clad tubing, but supplies an 
intermediate product form to customers that do, including GE Nuclear 
Energy, who performs the IPHT on the forms.

    The NRC staff conducted two inspections of Teledyne Wah Chang 
Albany (TWCA), the manufacturer of zirconium alloy tubes. In April 
1990, an employee of Teledyne Wah Chang Albany (TWCA) raised two 
concerns regarding the efficacy of TWCA's ``beta quench'' process, a 
step in the manufacture of zircaloy tube shells which improves the 
corrosion resistance of that product: (1) the accuracy of 
temperature indicating devices as a predictor of the temperature of 
the bulk profile of the zircaloy billet the beta quench process was 
measuring, and (2) even if the profiles of the induction furnaces 
are accurate, the induction furnaces cannot reproduce the profile 
conditions for each production zircaloy billet as the heating in the 
furnace is very sensitive to the position of the billet in the 
    Neither of the two NRC inspections substantiated the employee's 
concerns. See Inspection Reports 99901229/91-01 (November 27, 1991) 
and 99901229/94-01 (January 31, 1995). These inspection reports are 
available in the NRC Public Document Room, the Gelman Building, 2120 
L Street, NW., Washington, DC. TWCA also investigated these 
concerns. In a letter to the NRC dated January 10, 1991, TWCA 
forwarded the results of its investigation, concluding that these 
concerns were unfounded, although the employee continued to have 
    Based on the above, Petitioners have not demonstrated any basis 
for fuel rod corrosion corrective actions.

(2) Substandard and/or Counterfeit Parts

    Petitioners state that Pilgrim was one of several plants 
identified in a 1990 study by the United States Government 
Accounting Office as using parts which did not meet government 
standards, but that the NRC has not asked plants such as Pilgrim to 
replace those parts. Petitioners request information on the status 
of substandard or counterfeit parts at Pilgrim, such as nuts, bolts, 
pipe fittings, circuit breakers and fuses, and whether corrective 
action would be required during RFO No. 10.
    The NRC has been pursuing the issue of counterfeit and 
substandard parts as a two prong process for a number of years. The 
first process is reactive, directly addressing the possibility that 
substandard or counterfeit parts may have been supplied to nuclear 
power plants, assessing the safety significance and, if needed, 
replacing the parts. The second process is a proactive approach of 
improving the assurance that parts are of a high quality before they 
are put into use.
    Since 1988, the NRC has performed over 200 inspections of 
vendors. During these inspections, the staff occasionally identified 
suspect practices and referred those cases to the Office of 
Investigations to determine if wrongdoing had been committed. The 
NRC also quickly published and disseminated the information to the 
entire nuclear industry. Over the past several years, the NRC has 
issued numerous Bulletins and Information Notices having to do with 
potential counterfeit and/or substandard parts and material. 
However, the staff has not yet identified an issue that, from a 
safety standpoint, resulted in any plant shutdowns. Nonetheless, the 
NRC determined that several issues could potentially reduce the 
margin of safety in some plants and requested some actions by 
licensees, usually through a Bulletin. 

[[Page 47192]]

    If the NRC obtains information that some licensees are 
identified as potential customers of a vendor suspected of supplying 
counterfeit or substandard parts, an Information Notice is issued. 
The issuance of an Information Notice does not mean that the 
identified licensee(s) did, in fact, receive the questionable parts, 
but rather that they were potential customers. The licensees are 
responsible for reviewing their own procurement records to identify 
if they received the suspect parts. Their actions are subject to NRC 
review and inspection.
    The 1990 GAO report, ``Nuclear Safety and Health: Counterfeit 
and Substandard Products Are a Governmentwide Concern,'' lists a 
wide range of products as having been received or suspected of 
having been received by nuclear plants. The information provided by 
the GAO report regarding products used in nuclear operations was 
obtained from the NRC and all of the information was made public 
through various NRC Information Notices and Bulletins. The Pilgrim 
station was listed in the GAO report as having received counterfeit 
or substandard fasteners and circuit breakers. Pilgrim was also 
listed as being suspected of receiving counterfeit or substandard 
pipe fittings/flanges and fuses.
    On November 6, 1987, the NRC issued Bulletin 87-02, ``Fastener 
Testing to Determine Compliance With Applicable Material 
Specifications.'' The Bulletin requested all licensees to review 
their receipt inspection requirements and internal controls for 
fasteners and to determine, through testing, whether fasteners in 
stores at their facilities met required mechanical and chemical 
material specification requirements. Licensee responses were 
summarized in NUREG-1349, ``Compilation of Fastener Testing Data 
Received in Response to NRC Compliance Bulletin 87-02.'' NUREG-1349 
identified that, of over 3500 fasteners tested, 8 percent of safety-
related and 12 percent of nonsafety-related fasteners were found to 
be nonconforming. However, only 2 percent of the safety-related 
fasteners were found to be sufficiently out of specification to 
cause a concern regarding their ability to perform their intended 
safety function. As a result of the licensees' responses to Bulletin 
87-02, the NRC issued a temporary inspection instruction to ensure 
that licensees verified that fasteners used in nuclear plants met 
the requisite specifications and that operability of safety-related 
components was not affected.
    In response to Bulletin 87-02, Pilgrim tested 35 safety-related 
and 29 non-safety-related fasteners. Three safety-related and 6 non-
safety-related fasteners were identified as having hardness values 
slightly out of specification. These slight deviations were not 
considered safety significant since the hardness deviations 
consisted of only 1 to 2 Rockwell points which is very close to the 
test accuracy of  1.0 Rockwell point. Furthermore, it is 
commonly recognized in the industry that this property is most 
easily influenced by variations in chemistry, heat treatment, and 
surface treatments.
    On May 6, 1988, the NRC issued Bulletin 88-05, ``Nonconforming 
Materials Supplied by Piping Supplies, Inc. at Folsom, New Jersey 
and West Jersey Manufacturing Company at Williamstown, New Jersey.'' 
That Bulletin required NRC licensees to submit information regarding 
materials supplied by the named companies and requested the 
licensees to assure that the materials complied with ASME Code 
Section III, Subarticle NCA-3800 and design specifications 
requirements, or were suitable for their intended use, or to replace 
the materials. Following the issuance of that Bulletin and actions 
taken by licensees, the NRC met with representatives of the Nuclear 
Management and Resources Council (NUMARC) to discuss the status of 
licensee actions. NUMARC presented information on licensee and 
NUMARC/Electric Power Research Institute (EPRI) testing and 
evaluation methodology of numerous flanges. The information 
presented at that meeting showed that the material in question had 
acceptable strength and that continued use of the fittings and 
flanges did not present a safety problem. Therefore, the NRC issued 
Supplement 2 to Bulletin 88-05 on August 3, 1988, announcing that it 
was appropriate to suspend the actions requested by the Bulletin. 
NUMARC follow-up reports were analyzed by the staff and judged 
acceptable. Therefore, no further actions were required.
    In response to Bulletin 88-05, Pilgrim identified and tested a 
number of suspect flanges. All were found to be satisfactory, with 
the exception of one which tested low in hardness. An engineering 
evaluation performed by Pilgrim determined the flange was acceptable 
and did not need to be replaced.
    On July 8, 1988, the NRC issued Information Notice 88-46, 
``Licensee Report of Defective Refurbished Circuit Breakers,'' which 
alerted licensees to the possibility of defective circuit breakers 
being supplied to the nuclear industry. Following the issuance of 
the notice, the NRC issued Bulletin 88-10, ``Nonconforming Molded-
Case Circuit Breakers,'' which requested licensees to take action to 
provide reasonable assurance that those molded-case circuit breakers 
that did not have verifiable traceability to the circuit breaker 
manufacturer were able to perform their safety function. In response 
to the Bulletin, Pilgrim identified only one of 978 circuit breakers 
in its warehouse as not being traceable to the original equipment 
manufacturer. That breaker was the only one purchased on its 
purchase order and was subsequently discarded.
    On April 26, 1988, the NRC issued Information Notice 88-19, 
``Questionable Certification of Class 1E Components,'' to alert 
licensees to a possible problem with the certification of Class 1E 
components by Planned Maintenance Systems (PMS) of Mt. Vernon, 
Illinois. Information provided to the NRC by a licensee raised 
questions regarding the validity of certifications issued by PMS for 
Class 1E fuses PMS supplied. In response to Information Notice 88-
19, the licensee reviewed its procurement/QAD documents. There was 
no indication that the licensee had procured any material from PMS 
directly or through Bechtel or General Electric. Furthermore, the 
NRC review of PMS records indicated that PMS did not supply material 
or services through intermediate suppliers to the Pilgrim station.
    In addition to the Information Notices and Bulletins which 
identified specifics about potential counterfeit or substandard 
materials, the NRC staff has issued two generic letters providing 
information to the industry regarding procurement program 
improvements to help prevent the acceptance and use of counterfeit 
and/or substandard material. The industry, through the efforts of 
the Nuclear Energy Institute (NEI, successor to NUMARC), has also 
taken a strong approach to improve procurement programs by means of 
a Comprehensive Procurement Initiative, which addressed five areas 
which included general procurement, vendor audits, tests and/or 
inspections, obsolescent, and information exchanges. The 
Comprehensive Procurement Initiative has greatly reduced the 
incidence of substandard and/or counterfeit parts in the industry.
    In view of the above, no action regarding substandard or 
counterfeit parts needed to be taken by the licensee before start-up 
of the Pilgrim plant following RFO No. 10.

D. NRC Oversight and Enforcement Discretion

    Petitioners state that since September 1989, the NRC has either 
waived or chosen not to enforce regulations at nuclear reactors more 
than 340 times, and that of the last 100 industry requests for 
enforcement discretion, the Commission has granted every one. 
Petitioners also state that the NRC has granted at least seven NOEDs 
to Pilgrim since 1989. Petitioners assert that permitting a reactor 
to operate cannot pose less risk to public health and safety than 
keeping the reactor shut down until it meets regulations.
    The NRC Enforcement Policy, Section VII.C., permits the staff to 
exercise discretion not to enforce applicable TSs or license 
conditions by issuance of a NOED. Such enforcement discretion may be 
exercised only if the NRC staff is clearly satisfied that the action 
is consistent with protecting the public health and safety, in cases 
when a licensee's compliance with a TS Limiting Condition for 
Operation or other license condition would involve:
    (a) an unnecessary plant transient; or
    (b) performance of testing, inspection or system realignment 
that is inappropriate with the specific plant conditions; or
    (c) unnecessary delays in plant startup without a corresponding 
health and safety benefit.
    For an operating plant, the NOED is intended to (1) avoid 
undesirable transients as a result of forcing compliance with the 
license condition and, thus, minimize potential safety consequences 
and operational risks or (2) eliminate testing, inspection, or 
system realignment that is inappropriate for the particular plant 
conditions. For plants in a shutdown condition, the NOED is intended 
to reduce shutdown risk by avoiding testing, inspection, or system 
realignment that is inappropriate for the particular plant 
conditions, in that it does not provide an overall safety benefit, 
or may, in fact, be detrimental to safety in the particular plant 
    For plants attempting to start up, the need for exercising 
enforcement discretion is expected to occur less often than for 

[[Page 47193]]
operating plants, because delaying startup does not usually leave a 
plant in a condition in which it could experience undesirable 
transients. Thus, the issuance of NOEDs for plants attempting to 
start up must meet a higher threshold.
    The use of enforcement discretion does not change the fact that 
a violation of a license requirement will occur, nor does it imply 
that enforcement discretion is being exercised for any violation 
that may have led to the violation for which the licensee requests 
issuance of a NOED. Where the NRC staff has chosen to issue a NOED, 
enforcement action is normally considered for the root causes, to 
the extent violations led to the noncompliance for which enforcement 
discretion was used.
    Petitioners have provided no basis warranting a change in the 
Commission's policy regarding the exercise of enforcement discretion 
pursuant to Section VII.C. of the Enforcement Policy.

IV. Conclusion

    The institution of proceedings in accordance with Section 2.206, 
as requested by the Petitioner, is appropriate only where 
substantial safety issues have been raised. See Consolidated Edison 
Co. of New York (Indian Point Units 1, 2, and 3), CLI-75-8, NRC 173, 
175 (1975), and Washington Public Power Supply System (WPPSS Nuclear 
Project No. 2), DD-84-7, 19 NRC 899, 923 (1984). This is the 
standard I have applied to the Petition. Petitioners have not raised 
any substantial safety concerns regarding age-related deterioration 
of reactor internals, or with other parts and components at Pilgrim. 
To the contrary, all potential problems identified by Petitioners 
regarding reactor internals and components have been satisfactorily 
addressed by the licensee at Pilgrim. Therefore, Petitioner's 
request to delay startup of the Pilgrim plant is denied. 
Additionally, for the reasons discussed above, Petitioners request 
to terminate the NRC policy of issuing notices of enforcement 
discretion to reactor licensees is denied. Petitioner's request for 
a public meeting was granted.
    A copy of the Director's Decision will be filed with the Office 
of the Secretary for the Commission to review in accordance with 10 
CFR 2.206(c). As provided by Section 2.206(c), this Decision will 
constitute the final action of the Commission 25 days after 
issuance, unless the Commission, on its own motion, institutes a 
review of the decision within that time.

    For the Nuclear Regulatory Commission.
    Dated at Rockville, Maryland, this 31st day of August 1995,
William T. Russell,
Director, Office of Nuclear Reactor Regulation
[FR Doc. 95-22461 Filed 9-8-95; 8:45 am]