[Federal Register Volume 61, Number 162 (Tuesday, August 20, 1996)]
[Notices]
[Pages 43087-43091]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-21162]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-245, 50-336, and 50-423; License Nos. DPR-21, DPR-65,
and NPF-49]
Northeast Nuclear Energy Company, (Millstone Nuclear Power
Station Units 1, 2, and 3); Confirmatory Order Establishing Independent
Corrective Action Verification Program (Effective Immediately)
I
Northeast Nuclear Energy Company (Licensee) is the holder of
Facility Operating License Nos. DPR-21, DPR-65, and NPF-49 issued by
the Nuclear Regulatory Commission (NRC or Commission) pursuant to Title
10 of the Code of Federal Regulations (10 CFR) Part 50 on October 31,
1986,1 September 26, 1975, and January 31, 1986 respectively. The
licenses authorize the operation of Millstone Units 1, 2 and 3 in
accordance with conditions specified therein. All three facilities are
located on the Licensee's site in Waterford, Connecticut.
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\1\ Millstone Unit 1 was issued its provisional operating
license on October 7, 1970 and commenced operation on March 1, 1971.
This unit received a full term operating license on October 31,
1986.
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II
On August 21, 1995, as supplemented August 28, 1995, the NRC
received a petition under 10 CFR 2.206 which requested that NRC shut
down Millstone Unit 1 and take enforcement action based upon alleged
violations of NRC requirements related to operation of the spent fuel
pool cooling systems and refueling practices. On November 4, 1995, the
Licensee shut down Millstone Unit 1 for a planned 50-day refueling
outage. During the fall of 1995, an NRC investigation of licensed
activities at Millstone Unit 1 identified potential violations
regarding refueling practices and the operation of the spent fuel pool
cooling systems of Millstone Unit 1. On December 13, 1995, the NRC
issued a letter to the Licensee requiring that it inform the NRC,
pursuant to Section 182a of the Atomic Energy Act of 1954, as amended,
and 10 CFR 50.54(f), with regard to Millstone Unit 1, of the actions it
would be taking to ensure that future operation of that facility would
be conducted in accordance with the terms and conditions of the plant's
operating license, the Commission's regulations, including 10 CFR
50.59, and the plant's Updated Final Safety Analysis Report (UFSAR).
On February 20, 1996, the Licensee shut down Millstone Unit 2 when
both trains of the high pressure safety injection (HPSI) system were
declared inoperable due to the potential to clog the HPSI discharge
throttle valves during the recirculation phase following a loss-of-
coolant accident (LOCA). On February 22, 1996, the Licensee issued
Adverse Condition Report (ACR) 7007--Event Response Team Report, which
describes in detail the underlying causes for numerous inaccuracies
contained in Millstone Unit 1's UFSAR. Those causes, as determined by
the Licensee, include the following: (1) Errors and omissions in the
original 1986/87 UFSAR; (2) failure of the administrative control
programs to address fully NRC requirements; (3) failure of the Licensee
to implement fully those administrative programs; (4) a pattern of
failure of Licensee management to correct identified weaknesses and
risks associated with the UFSAR and design bases; and (5) failure of
Licensee oversight to identify this pattern to management, the
significance of the pattern itself, or the ineffectiveness of
corrective actions to prevent its recurrence. The report acknowledged
that, due to the nature of these identified causes, the potential
existed for the presence of similar configuration management problems
at Connecticut Yankee and Millstone Units 2 and 3.
In response to the Licensee's ACR 7007 and the NRC's own ongoing
inspections, evaluations and investigations, on March 7, 1996, the NRC
issued a letter to the Licensee requiring that it inform the NRC,
pursuant to Section 182a of the Atomic Energy Act of 1954, as amended,
and 10 CFR 50.54(f), with regard to Millstone Unit 2, of the actions it
would be taking to ensure that future operation of that facility would
be conducted in
[[Page 43088]]
accordance with the terms and conditions of the plant's operating
license, the Commission's regulations, including 10 CFR 50.59, and the
plant's UFSAR. The letter stated that this information was to be
submitted no later than 7 days prior to the Unit's restart (prior to
criticality) from its current outage. The Millstone Unit 2 letter also
described findings the NRC had made in recent inspections of that
facility which suggested that significant operability and design
concerns remained, including the HPSI issue identified above, as well
as inadequate containment sump screen mesh and a flawed post-accident
containment hydrogen monitor design.
On March 7, 1996, the NRC also issued a 50.54(f) letter to the
Licensee regarding the Millstone Unit 3 plant, which was then operating
at full power. In that letter, the NRC noted that it did not have an
inspection history at Millstone Unit 3 that revealed design
deficiencies similar in number and nature to that of Millstone Units 1
and 2. Nonetheless, the NRC concluded that it required additional
information, within 30 days of the date of the letter, including the
Licensee's plans and actions to address the implications of ACR 7007
for Millstone Unit 3, as well as the Licensee's plans and schedules to
ensure that future operation of the unit would be conducted in
accordance with the Commission's regulations, the terms and conditions
of the operating license, and the facility UFSAR.
Following the March 7 letter, the NRC conducted a special
inspection at Millstone Unit 3 that identified design and other
deficiencies similar to those reported in ACR 7007 and by the NRC at
the other Millstone units. On March 30, 1996, Unit 3 was shut down
after it was determined that containment isolation valves for the
auxiliary feedwater (AFW) turbine-driven pump were inoperable due to
the valves' noncompliance with NRC requirements. Shortly thereafter,
while still shut down, the Licensee discovered that the facility had
been operating in a condition outside its design basis due to the
Licensee's failure to adequately address design temperature conditions
in the stress calculations for the Containment Recirculation Spray
System (RSS) piping and supports. Both of these deficiencies had
existed for over ten years, since initial operation of the facility.
All three Millstone Units remain shut down.
On April 4, 1996, the NRC issued a second letter to the Licensee,
pursuant to 10 CFR 50.54(f), with regard to Millstone Unit 3, similar
to those issued for Millstone Units 1 and 2. The letter described
programmatic issues and design deficiencies identified during the NRC's
ongoing special inspection of the plant that were similar in nature to
those present at Millstone Units 1 and 2. These included the
inoperability of the turbine-driven AFW pump during startup and
shutdown, the failure to remove plastic shipping plugs from Rosemount
transmitters, the failure to correct a degraded non-safety battery,
inadequate control of the modification of the service water system, and
the potential for introduction of foreign material into the containment
sump. In addition, the letter noted Licensee-identified design
deficiencies in the AFW containment isolation valves and RSS that had
existed for more than 10 years. As in the case of the Millstone Unit 1
and 2 letters, as described above, the Licensee was required to provide
the NRC, no later than 7 days prior to the Unit's restart, with
information necessary to assure the NRC that the plant will be operated
in conformance with the terms and conditions of the plant's operating
license, the Commission's regulations, including 10 CFR 50.59, and the
plant's UFSAR.
On May 21, 1996, pursuant to 10 CFR 50.54(f), the NRC issued a
letter to the Licensee requiring specific information regarding design
and configuration deficiencies identified at each of the Millstone
units as well as a detailed description of the Licensee's plans for
completion of the work required to respond to the NRC's letters of
December 13, 1995, March 7, 1996, and April 4, 1996. The NRC required
this information to be submitted within 30 days of the date of the
letter for the first unit that the Licensee proposed to restart and not
later than 60 days prior to the Licensee's proposed restart for the
remaining Millstone units.
Based upon the Licensee's assessment of the extent and scope of
identified design control problems at Millstone Station, the Licensee
decided to focus its near-term efforts on restart of Millstone Unit 3.
In a letter dated June 20, 1996, the Licensee responded to the NRC's
May 21, 1996, letter and informed NRC that Millstone Unit 3 would be
the first Millstone unit the Licensee proposed to restart. In
Attachment 1 to its June 20 response, the Licensee listed 881 design
and configuration deficiencies identified since issuance of ACR 7007
and entered into the Licensee's Deficiency Review Team Report database
as of June 13, 1996. The Licensee designated 378 items to be corrected
prior to restart of Millstone Unit 3. The Licensee determined that the
items it had designated for correction prior to restart, if not
corrected, could impact upon operability of required equipment, raise
unreviewed safety questions, or indicate discrepancies between the
plant's UFSAR and the as-built plant or operating procedures.
In the June 20 letter, the Licensee also described its own
Configuration Management Plan (CMP), intended to provide reasonable
assurance that the future operation of Millstone Units 1, 2, and 3 will
be conducted in accordance with the terms and conditions of their
applicable operating licenses, UFSARs and NRC regulations. The CMP
includes efforts to understand licensing and design basis issues which
led to issuance of the 50.54(f) letters and actions to prevent those
issues' recurrence. Additionally, the Licensee described its CMP
objective to clearly document and meet the units' licensing and design
basis requirements, and its intention to ensure that adequate programs
and processes exist to maintain control of licensing and design basis
requirements.
On July 2, 1996, the Licensee supplemented its June 20, 1996
response to NRC's May 21, 1996 50.54(f) letter. The Licensee provided
additional information on Millstone Unit 3 deficiencies previously
reported, identified revisions to its plans and committed to complete a
review to identify and correct, as necessary, Millstone Unit 3 UFSAR
deficiencies prior to restart. The Licensee reported a substantial
increase in the total number of identified design and configuration
management discrepancies (1187 items), and an increase in those
proposed by the Licensee for corrective action prior to restart (597
items).
As the Licensee's own submissions and NRC inspections indicate,
significant design control deficiencies and degraded and non-conforming
conditions have been identified at Millstone Units 1, 2, and 3. The
staff has identified three major types of design control problems which
exist at all three Millstone plants. Specific examples of deficiencies
at each plant in each of the categories are provided below.
1. Errors in Licensing/Design Basis Documentation
The NRC identified errors in the UFSARs for Millstone Units 1,
2, and 3. For example, at Millstone Unit 3, the protective relay
settings and calculations for 4kv safety-related motor feeders were
not set consistent with the UFSAR. At Millstone Unit 2, the UFSAR
indicated that certain non-essential loads of the reactor building
closed cooling water (RBCCW) system inside containment were
automatically isolated during a sump recirculation actuation signal
when in fact the associated isolation valves received no
[[Page 43089]]
automatic isolation signal. Additionally, the RBCCW flow rates
assumed in the accident analyses were non-conservative with respect
to the actual system flow rates.
In addition, the NRC found instances of modifications that were
completed without implementing required revisions to the UFSAR. For
example, the Licensee revised the Millstone Unit 3 Technical
Specifications (TS) in January 1995 to change the testing frequency
of the auxiliary feed pumps from monthly to quarterly, but did not
update the UFSAR to reflect the change.
At Unit 1, the Licensee failed to perform and document a safety
evaluation for an electrical separation deficiency associated with a
feedwater regulating valve interlock. This deficiency was not
corrected and constituted a change to the design of the facility as
described in the UFSAR. Also, the Licensee's assessment of the need
for upgrades to the intake structure ventilation system was
inadequate. Specifically, insufficient heat removal capability
existed under several postulated scenarios.
At Unit 2, the NRC found that the UFSAR had not been updated to
reflect that the intake structure design temperature could not be
met following a loss of non-vital exhaust fans.
Furthermore, while the Millstone Unit 3 UFSAR documented that
the design bases for the containment heat removal systems had been
established in accordance with specific general design and code
criteria, portions of these systems were found to violate certain
analytical stress considerations. Specifically, the recirculation
spray system (RSS) pipe supports inside containment were not
designed to withstand a single failure of a supporting service water
train. Also, both the RSS and quench spray systems were found to
contain pipe supports for which ASME Code stress allowables would be
exceeded during design basis accident temperature conditions within
the Unit 3 containment building.
2. Failure To Translate Design Bases to Procedures and Hardware
The NRC found instances where the Licensee did not adequately
translate design basis information into procedures, practices,
hardware and drawings. For example, at Millstone Unit 1, the reactor
pressure assumed as an initial condition in the accident analyses
was exceeded during reactor power operation. At Unit 3, a
modification that installed the service water intake structure sump
pump called for specific periodic testing, but such testing was
never performed. In another case at Unit 3, prelubrication of the
AFW pump was not performed every 40 days as required by the vendor.
As noted in the NRC's letter of December 13, 1995, at Millstone
Unit 1, the Licensee's core offload practices were not consistent
with the Unit's UFSAR. Specifically the heat load assumptions were
not maintained as a result of full core offloads performed sooner
than the required delay time after reactor shutdown.
Also at Unit 1, measures established to ensure that the design
bases were satisfied for control room habitability were not adequate
in that the means for maintaining viable self-contained breathing
apparatus capability for each person in the control room were not
translated into procedures. In addition, the Licensee failed to
translate the design bases for the Unit 1 standby gas treatment
system (SGTS) into design specifications, and failed to perform
comprehensive pre-operational testing of the SGTS to ensure that it
met its design specifications.
At Millstone Unit 2, the Licensee failed to adequately update
the surveillance requirements to reflect modifications to contact
positions in the anticipated transient without scram (ATWS)
mitigating system actuating circuitry. Also at Unit 2, the procedure
requirements for the time of initiation of hydrogen monitoring
following a LOCA were not consistent with the licensing and design
bases.
In addition, there were a number of instances where the original
design basis was inadequate or the original installation was
incorrect. For example, at Units 2 and 3, the Licensee failed to
remove plastic shipping plugs from Rosemount transmitters prior to
installation, notwithstanding the vendor's instructions which
required those plugs' replacement with stainless steel plugs. At
Unit 2, the NRC found that nuclear instrumentation and post-LOCA
hydrogen monitors were not single-failure proof.
At Millstone Unit 2, the Licensee's inspection of the
containment sump screen mesh revealed that debris larger than the
size specified in the design basis could pass through with potential
adverse consequences to the operability of the emergency core
cooling systems. The NRC also identified that the post-accident
containment hydrogen monitor design at Millstone Unit 2 was flawed
in that insufficient sample flow would be available at low
containment pressures when the monitor must be operable.
Also at Unit 2, when it was found that postulated failures of
the non-vital intake structure ventilation systems could cause the
intake structure ambient temperature to exceed the design basis, the
Licensee did not perform appropriate evaluations relative to the
design basis before concluding that no modifications to equipment or
the design basis were needed.
3. Inadequate Engineering and Modifications
The NRC identified a number of instances in which a modification
was not installed in accordance with the design, a modification was
inadequate, or a modification was based on incorrect design
assumptions. In one example at Millstone Unit 1, the Licensee failed
to maintain the design bases for the loss of normal power (LNP)
logic. Specifically, a modification resulted in a single failure
vulnerability of the LNP logic that would have prevented both
emergency power sources from properly starting and sequencing the
required loads. The Licensee also revised the Unit 1 maximum spent
fuel pool temperature through an amendment to the Technical
Specifications but failed to evaluate the impact of the change on
the SGTS.
At Millstone Unit 2, both trains of service water were rendered
inoperable when the strainer backwash line froze due to an
undocumented modification that extended the backwash line through an
opening under the wall to a point just outside the intake structure.
Also at Millstone Unit 2, the NRC identified that both trains of
the post-accident sampling system have been inoperable since the
steam generator replacement modification because higher containment
pressures would have delayed taking a containment sample for 24
hours.
At Millstone Unit 3, the Licensee prepared a modification
package for the high pressure safety injection thermal relief valves
which relied on incorrect design assumptions because a previous
modification had revised the design. In addition, the Licensee had
no approved calculation to demonstrate the adequacy of the station
blackout diesel generator battery at Millstone Unit 3.
Although the Licensee's own programs, such as the CMP, are intended
to correct existing and prevent future deficiencies at the facilities,
I have concluded that these programs by themselves are not sufficient,
given the Licensee's history of poor performance in ensuring complete
implementation of corrective action for both known degraded and non-
conforming conditions and past violations of NRC requirements. In
addition, the magnitude and scope of the design and configuration
deficiencies currently being identified indicate multiple significant
failures to comply with NRC regulations (e.g., 50.59, 50.71(e), etc.)
The Licensee's history of poor performance, coupled with the magnitude
and scope of its failure to maintain and control conformance of
Millstone Units 1, 2, and 3 to their design bases, require resolution
prior to plant restarts.
The extent and duration of the deficiencies identified also
indicate ineffective implementation of the Licensee's oversight
programs, including the NRC-approved quality assurance (QA) program.
Effective oversight activities should have identified and led to
corrective measures for design control deficiencies. One conclusion of
ACR 7007 was that the Licensee's oversight organizations (Review
Boards, Quality Assessment Section (QAS), Independent Safety
Engineering Group, and Operating Experience) did not identify the
pattern of Millstone Unit 1 UFSAR discrepancies to management; nor did
they identify the significance of the pattern, or the effectiveness of
corrective actions to prevent recurrence. In a July 2, 1996 letter to
the NRC, the Licensee provided the preliminary findings of an
independent Root Cause Evaluation Team chartered to determine the
causes for these oversight failures. The team
[[Page 43090]]
found that there was no history of escalating issues effectively and
that QAS operated in an environment that did not lend itself to
resolution of QAS-identified problems. Such findings of program
weaknesses that represent poor oversight functions are not recent. It
is apparent that the Licensee was aware of significant weaknesses in
its oversight functions as early as 1991 and took no effective actions
to correct those weaknesses. The Licensee's Performance Task Group
Final Report, issued in September 1991, and Procedure Compliance Task
Force Final Report, issued in October 1991, identified significant
programmatic weaknesses affecting configuration management that either
went unnoticed or were not corrected by the Licensee oversight
functions.
It is necessary to ensure that the Licensee's programs to correct
design control failures at Millstone Units 1, 2 and 3 are effective and
that identification of degraded and non-conforming conditions and
implementation of corrective actions are satisfactory and can
effectively preclude repetition of these failures. For this reason, the
NRC requires an independent verification of the adequacy of the results
of the programs currently being implemented by the Licensee which are
directed at resolving existing design and configuration management
deficiencies. Accordingly, the Commission in this Order directs the
Licensee to obtain the services of an organization, independent of the
Licensee and its design contractors, to conduct a multi-disciplinary
review of Millstone Units 1, 2, and 3. The review is to provide
independent verification that, for the selected systems, the Licensee's
CMP has identified and resolved existing problems, documented and
utilized licensing and design bases, and established programs,
processes and procedures for effective configuration management in the
future. This review must be comprehensive, incorporating appropriate
engineering disciplines, such that the NRC can be confident that the
Licensee has been thorough in identification and resolution of
problems.
III
On August 12, 1996, a transcribed meeting was conducted between the
Licensee and the NRC staff regarding this matter. In response to the
staff's concerns, the Licensee subsequently submitted a letter dated
August 13, 1996, in which it agreed and committed to take a number of
actions with respect to Millstone Units 1, 2, and 3. Specifically, the
Licensee committed to have an independent team conduct an Independent
Corrective Action Verification Program (ICAVP) at Millstone Units 1, 2,
and 3. The Licensee committed that the corrective action verification
program will include: (1) Conduct of an in-depth review of selected
systems which will address control of the design and design basis since
issuance of the operating license for each unit; (2) selection of
systems for review based on risk/safety based criteria similar to those
used in implementing the Maintenance Rule (10 CFR 50.65); (3)
development and documentation of an audit plan that will provide
assurance that the quality of results of the Licensee's problem
identification and corrective action programs on the selected systems
is representative of and consistent with that of other systems; (4)
procedures and schedules for parallel reporting of findings and
recommendations by the ICAVP team to both the NRC and the Licensee; and
(5) procedures for the ICAVP team to comment on the Licensee's proposed
resolution of the findings and recommendations. The Licensee also
committed to the scope of the ICAVP review, encompassing modifications
to the selected systems since initial licensing, including: (1) A
review of engineering design and configuration control processes; (2)
verification of current, as-modified plant conditions against design
basis and licensing basis documentation; (3) verification that design
and licensing bases requirements are translated into operating
procedures, and maintenance and test procedures; (4) verification of
system performance through review of specific test records and/or
observation of selected testing of particular systems; and (5) review
of proposed and implemented corrective actions for Licensee-identified
design deficiencies.
I find that the Licensee's agreements and commitments as set forth
in its letter of August 13, 1996 are acceptable and necessary.
In view of the foregoing, I have determined that public health and
safety require that the Licensee's agreements and commitments in its
August 13, 1996 letter be confirmed by this Order. The Licensee has
agreed to this action. Pursuant to 10 CFR 2.202, I have also
determined, based on the significance of the matters described above,
as well as on the Licensee's consent, that the public health and safety
require that this Order be immediately effective.
IV
Accordingly, pursuant to Sections 103, 104, 161b, 161i, 161o, 182
and 186 of the Atomic Energy Act of 1954, as amended, and the
Commission's regulations in 10 CFR 2.202 and 10 CFR Part 50, It is
hereby ordered, effective immediately, That:
1. The Licensee shall implement an Independent Corrective Action
Verification Program (ICAVP) for each Millstone Unit to confirm that
the plant's physical and functional characteristics are in conformance
with its licensing and design bases. The ICAVP review shall begin after
the Licensee has completed the problem identification phase of the CMP,
including the activities of the QA organization. The ICAVP shall be
performed and completed for each Unit, to the satisfaction of the NRC,
prior to the Unit's restart.
2. The ICAVP is to be conducted by an independent verification team
whose selection must be approved by the NRC. The ICAVP team shall
provide input on its findings on an ongoing basis concurrently to both
the Licensee and the NRC. The ICAVP team shall also periodically
provide to the NRC its comments on the Licensee's proposed resolution
of the team's findings and recommendations.
3. The ICAVP team shall provide for NRC review and approval, prior
to implementation, a plan for the conduct of the team's review. The
plan must describe (a) the conduct of an in-depth review of selected
systems' design and design bases since issuance of the facilities'
operating licenses; (b) risk/safety based criteria for selection of
systems for review; (c) a description of the audit plan to provide
assurance that the quality of results of the Licensee's problem
identification and corrective action programs on the selected systems
is representative of and consistent with that of other systems; (d)
procedures and schedules for parallel reporting of findings of the
ICAVP team to both the NRC and the Licensee; and (e) procedures for the
ICAVP team to comment on the Licensee's proposed resolution of the
team's findings and recommendations. The scope of the ICAVP effort
shall encompass all modifications made to the selected systems since
initial licensing, and shall include: (1) Review of engineering design
and configuration control processes, (2) verification of current, as-
modified conditions against design and licensing basis documentation,
(3) verification that the design and licensing bases requirements have
been translated into operating procedures, and maintenance and test
procedures, (4) verification of system performance through review of
specific test records and/or observation of selected testing, and (5)
review of proposed and
[[Page 43091]]
implemented corrective actions for licensee-identified design
deficiencies.
4. The Licensee shall provide written replies to the Regional
Administrator, Region I and the Director, Office of Nuclear Reactor
Regulation, addressing ICAVP team findings and recommendations
discussed in reports made pursuant to item 3(d) above. The Licensee's
written replies to ICAVP team findings and recommendations shall
include a statement of agreement or disagreement with reasons for each
ICAVP finding or recommendation, and of the status of implementation of
corrective actions. Subsequent written replies shall be made until all
corrective actions are implemented.
The Director, Office of Nuclear Reactor Regulation, may, in
writing, relax or rescind this order upon demonstration by the Licensee
of good cause.
V
The Licensee has, as described above, consented to the issuance of
this Order and waived its right to request a hearing. Thus, any person
adversely affected by this Order, other than the Licensee, may request
a hearing within 20 days of its issuance. Where good cause is shown,
consideration will be given to extending the time to request a hearing.
A request for extension of time must be made in writing to the
Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and include a statement of good cause for the
extension. Any request for a hearing shall be submitted to the
Secretary, U.S. Nuclear Regulatory Commission, ATTN: Chief, Docketing
and Service Section, Washington, DC 20555. Copies of the hearing
request shall also be sent to the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, to the Assistant
General Counsel for Hearings and Enforcement at the same address, to
the Regional Administrator, NRC Region I, 475 Allendale Road, King of
Prussia, PA 19406-1415, and to the Licensee. If such a person requests
a hearing, that person shall set forth with particularity the manner in
which his interest is adversely affected by this Order and shall
address the criteria set forth in 10 CFR 2.714(d).
If a hearing is requested by a person whose interest is adversely
affected, the Commission will issue an Order designating the time and
place of any hearings. If a hearing is held, the issue to be considered
at such hearing shall be whether this Confirmatory Order should be
sustained.
In the absence of any request for hearing, or written approval of
an extension of time in which to request a hearing, the provisions
specified in Section IV above shall be final 20 days from the date of
this Order without further order or proceedings. If an extension of
time for requesting a hearing has been approved, the provisions
specified in Section IV shall be final when the extension expires if a
hearing request has not been received. AN ANSWER OR A REQUEST FOR
HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.
Dated at Rockville, Maryland, this 14th day of August, 1996.
For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-21162 Filed 8-19-96; 8:45 am]
BILLING CODE 7590-01-P