[Federal Register Volume 59, Number 80 (Tuesday, April 26, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-10008]


[[Page Unknown]]

[Federal Register: April 26, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-458]

 

Entergy Operations, Inc.; Consideration of Issuance of Amendment 
to Facility Operating License, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an amendment to Facility Operating License No. 
NPF-47 issued to Entergy Operations, Inc. (the licensee) for operation 
of the River Bend Station, Unit 1, located in West Feliciana Parish, 
Louisiana.
    The proposed amendment would revise various instrumentation 
technical specifications by extending the allowable outage times (AOTs) 
of the instruments, and by increasing their channel functional 
surveillance test intervals (STIs) to quarterly. The amendment also 
revises certain technical specification actions to address loss-of-
function concerns associated with the AOT and STI changes.
    Before issuance of the proposed license amendment, the Commission 
will have made findings required by the Atomic Energy Act of 1954, as 
amended (the Act) and the Commission's regulations.
    The Commission has made a proposed determination that the amendment 
request involves no significant hazards consideration. Under the 
Commission's regulations in 10 CFR 50.92, this means that operation of 
the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

Reactor Protection System (RPS)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These proposed changes do not involve a change to the plant 
design or operation, they simply involve the frequency at which 
testing of the RPS instrumentation is performed and the allowable 
outage time (AOT) for instruments. Failure of the RPS 
instrumentation itself cannot create an accident. As a result, these 
proposed changes cannot increase the probability of occurrence of 
any design basis accident previously evaluated.
    As identified in NEDC-30851P, these proposed changes increase 
the average RPS failure frequency from 4.6 x 10-6/year to 
5.4 x 10-6/year. This increase (8 x 10-7/year) is 
considered to be insignificant. As identified in the NRC Staff's 
Safety Evaluation Report of NEDC-30851P, this increase in average 
RPS failure frequency would contribute to a very small increase in 
core-melt frequency. The small increase in average RPS failure 
frequency is offset by safety benefits such as a reduction in the 
number of inadvertent test-induced scrams, a reduction in wear due 
to excessive equipment test cycling, and better optimization of 
plant personnel resources. Hence, the net change in risk resulting 
from these proposed changes would be insignificant. Therefore, these 
proposed changes do not result in a significant increase in either 
the probability or the consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not result in any change to the plant 
design or operation, only to the AOT and frequency at which testing 
of the RPS instrumentation is performed. Since failure of the RPS 
instrumentation itself cannot create an accident, these proposed 
changes can at most affect only accidents which have been previously 
evaluated. Therefore, these proposed changes cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    As identified above, these proposed changes increase the average 
RPS failure frequency from 4.6x10-6/year to 5.4x10-6/year. 
The NRC Staff's Safety Evaluation Report of NEDC-30851P concluded 
that this small average RPS failure frequency increase would 
contribute to a very small increase in core-melt frequency. This 
small increase in average RPS failure frequency would be offset by 
safety benefits such as a reduction in the number of inadvertent 
test-induced scrams, a reduction on wear due to excessive equipment 
test cycling, and better optimization of plant personnel resources. 
Hence, the net change in risk resulting from these proposed changes 
would be insignificant. In addition, RBS has confirmed that the 
proposed changes to the functional test intervals will not result in 
excessive instrument drift relative to the current established 
setpoints. Therefore, these proposed changes do not result in a 
significant reduction in a margin of safety.

Emergency Core Cooling System (ECCS)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These proposed changes do not involve a change to the plant 
design or operation, they simply involve the frequency at which 
testing of the ECCS actuation Instrumentation is performed and the 
allowable outage time (AOT) for instruments. Failure of the ECCS 
actuation instrumentation itself cannot create an accident. As a 
result, these proposed changes cannot increase the probability of 
occurrence of any design basis accident previously evaluated.
    As identified in NEDC-30936P (Part 2), these proposed changes 
increase the calculated average water injection failure frequency 
from 1.952x10-5 to 1.992x10-5 per year for Case 5B and 
from 1.386x10-4 to 1.401x10-4 per year for Case 5C. This 
represents an increase of 4x10-7 for Case 5B (2.0%) and 
1.5x10-6 for Case 5C (1.1%), which are well within the 
acceptance criteria (4%) provided in NEDC-30936P (Part 2). The small 
increase in average water injection failure frequency is offset by 
safety benefits such as a reduction in the number of inadvertent 
test-induced scrams, a reduction in wear due to excessive equipment 
test cycling, and better optimization of plant personnel resources. 
Hence, the net change in risk resulting from these proposed changes 
would be insignificant. Therefore, these proposed changes do not 
result in a significant increase in either the probability or the 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not result in any change to the plant 
design or operation, only to the AOT and frequency at which testing 
of the ECCS actuation instrumentation is performed. Since failure of 
the ECCS actuation instrumentation itself cannot create an accident, 
these proposed changes can at most affect only accidents which have 
been previously evaluated. Therefore, these proposed changes cannot 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    As identified above, these proposed changes increase the 
calculated average water injection failure frequency from 
1.952x10-5 to 1.992x10-5 per year for Case 5B and from 
1.386x10-4 to 1.401x10-4 per year for Case 5C. This 
increase is well within the acceptance criteria found acceptable in 
the NRC Staff's Safety Evaluation Report for NEDC-30936P (Part 2). 
This small increase in average ECCS actuation failure frequency 
would be offset by safety benefits such as a reduction in the number 
of inadvertent test-induced scrams, a reduction on wear due to 
excessive equipment test cycling, and better optimization of plant 
personnel resources. Hence, the net change in risk resulting from 
these proposed changes would be insignificant. In addition, RBS has 
confirmed that the proposed changes to the functional test intervals 
will not result in excessive instrument drift relative to the 
current, established setpoints. Therefore, the proposed changes do 
not result in a significant reduction in a margin of safety.

Control Rod Block Instrumentation

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These proposed changes do not involve a change to the plant 
design or operation, only the Allowable Outage Time (AOT) and 
frequency at which testing of the Control Rod Block Instrumentation 
is performed. Failure of the Control Rod Block instrumentation 
itself cannot create an accident. As a result, these proposed 
changes cannot increase the probability of occurrence of any design 
basis accident previously evaluated.
    As identified in NEDC-30851P, Supplement 1, these proposed 
changes increase the average Control Rod Block failure frequency 
less than 0.06%. As provided in the NRC Staff's Safety Evaluation 
Report of NEDC-30851P, Supplement 1, this increase is very slight 
and is offset by the safety benefits associated with the proposed 
changes to the RPS and Control Rod Block Instrumentation. As a 
result, the combined effect of the changes proposed for the RPS and 
Control Rod Block Instrumentation requirements should result in an 
overall improvement in plant safety. Therefore, these proposed 
changes do not result in a significant increase in either the 
probability or the consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not result in any change to the plant 
design or operation, only to the AOT and frequency at which testing 
of the Control Rod Block instrumentation is performed. Since failure 
of the Control Rod Block instrumentation itself cannot create an 
accident, these proposed changes can at most affect only accidents 
which have been previously evaluated. Therefore, these proposed 
changes cannot create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    As identified above, these proposed changes increase the average 
Control Rod Block failure frequency less than 0.06%. This increase 
is very slight and is offset by the safety benefits associated with 
the proposed changes to the RPS and Control Rod Block 
Instrumentation. As a result, the combined effect of the changes 
proposed for the RPS and Control Rod Block Instrumentation 
requirements should result in an overall improvement in plant 
safety. In addition, RBS has confirmed that the proposed changes to 
the functional test intervals will not result in excessive 
instrument drift relative to the current, established setpoints. 
Therefore, the proposed changes do not result in a significant 
reduction in a margin of safety.

Isolation Actuation Instrumentation

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These proposed changes do not involve a change to the plant 
design or operation, only the Allowable Outage Time (AOT) and 
frequency at which testing of the Isolation Actuation 
instrumentation is performed. Failure of the Isolation Actuation 
instrumentation itself cannot create an accident. As a result, these 
proposed changes cannot increase the probability of occurrence of 
any design basis accident previously evaluated.
    As identified in NEDC-30851P, Supplement 2, these proposed 
changes to the surveillance test interval requirements for the 
Isolation Actuation instruments which are common to RPS or ECCS have 
a negligible (less than 1%) impact on the average isolation 
unavailability when combined with the individual valve failure 
probability, and that the changes to the AOTs has [have] less than a 
2% impact. The analyses demonstrate that the individual valve 
failure probabilities dominate the overall isolation failure 
probability. As provided in the NRC Staff's Safety Evaluation Report 
of NEDC-30851P, Supplement 2, these proposed changes would have a 
very small impact on plant risk. As a result, overall plant safety 
is not reduced by these proposed changes.
    As identified in NEDC-31677P, the proposed changes to the 
requirements for Isolation Actuation instrumentation not common to 
RPS or ECCS result in a small decrease of 1.97x10-8/year in the 
average isolation failure frequency. As identified in the NRC 
Staff's Safety Evaluation Report of NEDC-31677P, the NRC agreed that 
these proposed changes are acceptable because the failure frequency 
impact is minimal. As a result, overall plant safety is not reduced 
by these proposed changes.
    The small increase in the average failure frequency of the 
instruments common to RPS or ECCS due to the proposed changes to the 
Isolation Actuation instrumentation requirements is offset by safety 
benefits such as a reduction on the number of inadvertent test-
induced scrams and engineered safety feature actuations, a reduction 
in wear due to excessive test cycling, and better optimization of 
plant personnel resources. Hence, the net change in risk resulting 
from these proposed changes would be insignificant. Therefore, these 
proposed changes do not result in a significant increase in either 
the probability or the consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not result in any change to the plant 
design or operation, only to the AOT and frequency at which testing 
of the Isolation Actuation instrumentation is performed. Since 
failure of the Isolation Actuation instrumentation itself cannot 
create an accident, these proposed changes can at most affect only 
accidents which have been previously evaluated.
    Therefore, these proposed changes cannot create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    As identified above, the proposed changes to the requirements 
for Isolation Actuation instruments common to RPS or ECCS have a 
negligible impact on the average isolation unavailability when 
combined with the individual valve failure probability. The analyses 
demonstrate that the individual valve failure probabilities dominate 
the overall isolation failure probability.
    The proposed changes to the requirements for Isolation Actuation 
instruments not common to RPS or ECCS decrease their average 
isolation failure frequency approximately 1.97x10-8/year.
    The small increase in average Isolation Actuation 
instrumentation failure frequency of the instruments common to RPS 
or ECCS are offset by the safety benefits such as a reduction on the 
number of inadvertent test-induced scrams and engineered safety 
feature actuations, a reduction in wear due to excessive test 
cycling, and better optimization of plant personnel resources. As a 
result, the NRC Staff's Safety Evaluation Reports for these BWROG 
report concluded that these proposed changes would have a very small 
impact on plant risk. In addition, RBS has confirmed that the 
proposed changes to the functional test intervals will not result in 
excessive instrument drift relative to the current, established 
setpoints. Therefore, the proposed changes do not result in a 
significant reduction in a margin of safety.

Other Technical Specification Instrumentation

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These proposed changes do not involve a change to the plant 
design or operation, only the Allowable Outage Time (AOT) and 
frequency at which testing of the associated instrumentation is 
performed. These instruments are designed to mitigate the 
consequences of previously analyzed accidents. Failure of these 
instruments cannot increase, and is independent of, the probability 
of occurrence of such accidents. As a result, these proposed changes 
cannot increase the probability of any accident previously 
evaluated. As identified in GENE-770-06-01, although not 
specifically analyzed, these proposed changes are bounded by the 
results of the analyses discussed in Parts I through IV of this 
request. Such analyses have shown that the safety function failure 
frequency is not significantly impacted by similar proposed changes. 
In addition, any increase in the probability of failure of these 
instruments to perform their required functions would be offset by 
safety benefits such as a reduction in the number of inadvertent 
test-induced scrams and engineered safety features actuations, a 
reduction in wear due to excessive equipment test cycling, and 
better optimization of plant personnel resources. Therefore, these 
proposed changes do not result in a significant increase in the 
probability or the consequences of any accident previously evaluated
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not result in any change to the plant 
design or operation, only to the AOT and frequency at which testing 
of the associated instrumentation is performed. As a result, these 
proposed changes can at most affect only accidents which have been 
previously evaluated. Therefore, these proposed changes cannot 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    As identified in GENE-770-06-01, although not specifically 
analyzed, these proposed changes are bounded by the results of the 
analyses discussed in Parts I through IV of this request. Such 
analyses have shown that the safety function failure frequency is 
not significantly impacted by similar proposed changes. In addition, 
any increase in the probability of failure of these instruments to 
perform their required functions would be offset by safety benefits 
such as a reduction in the number of inadvertent test-induced scrams 
and engineered safety features actuations, a reduction in wear due 
to excessive equipment test cycling, and better optimization of 
plant personnel resources. As a result, these proposed changes will 
reduce overall plant risk. In addition, RBS has confirmed that the 
proposed changes to the functional test intervals will not result in 
excessive instrument drift relative to the current, established 
setpoints. Therefore, these proposed changes do not involve a 
significant reduction in a margin of safety.

Technical Specification Changes Relating to Loss-of-Function Issues

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes associated with the ``loss-of-function'' 
checks ensure a plant configuration which would have the capability 
to automatically actuate the respective system/valve(s). These 
instruments are designated to mitigate the consequences of 
previously analyzed accidents. Failure of these instruments cannot 
increase, and is independent of, the probability of occurrence of 
such accidents. As a result, the proposed changes cannot increase 
the probability of any accident previously evaluated. The proposed 
changes do not involve a change to the plant design or operation and 
do not degrade the capability of the system(s) to perform its 
required function. Further, these functions or tripped channel(s) in 
an isolation logic are not considered as initiators for any 
accidents previously analyzed. Therefore, these changes do not 
significantly increase the consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not result in any change to the plant 
design and no new mode of plant operation is introduced. As a 
result, the proposed changes can at most affect only accidents which 
have been previously evaluated. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety since the required safety function of the 
inoperable channel(s) will be fulfilled. The allowable Outage Time 
(AOT) for several trip functions have been increased but only in 
conjunction with the incorporation of the loss-of-function check 
which ensures a plant configuration which would have the capability 
to automatically actuate the respective system/valve(s). Therefore, 
the proposed changes do not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received. 
Should the Commission take this action, it will publish in the Federal 
Register a notice of issuance and provide for opportunity for a hearing 
after issuance. The Commission expects that the need to take this 
action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to room 6D22, Two White Flint North, 11555 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 
20555.
    The filing requests for hearing and petitions for leave to 
intervene is discussed below.
    By May 26, 1994, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room located at the Government Documents Department, 
Louisiana State University, Baton Rouge, Louisiana 70803. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment requests involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to William D. Beckner, Director, Project 
Directorate IV-1: petitioner's name and telephone number, date petition 
was mailed, plant name, and publication date and page number of the 
Federal Register notice. A copy of the petition should also be sent to 
the Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and to Mark J. Wetterhahn, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005, attorney for the 
licensee.
    Nontimely filings of petitioners for leave to intervene, amended 
petitions, supplemental petitions and/or requests for hearing will not 
be entertained absent a determination by the Commission, the presiding 
officer or the presiding Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(l)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment dated January 14, 1994, which is available 
for public inspection at the Commission's Public Document Room, the 
Gelman Building, 2120 L Street, NW., Washington, DC 20555 and at the 
local public document room located at the Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

    Dated at Rockville, Maryland, this 20th day of April, 1994.

    For the Nuclear Regulatory Commission.
Robert G. Schaaf,
Acting Project Manager, Project Directorate IV-1, Division of Reactor 
Projects III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 94-10008 Filed 4-25-94; 8:45 am]
BILLING CODE 7590-01-M