[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