Northeast Nuclear Energy Company, (Millstone Nuclear Power Station Units 1, 2, and 3); Confirmatory Order Establishing Independent Corrective Action Verification Program (Effective Immediately)

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