[Federal Register Volume 63, Number 165 (Wednesday, August 26, 1998)]
[Notices]
[Pages 45521-45535]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-22766]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and

[[Page 45522]]

make immediately effective any amendment to an operating license upon a 
determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 3, 1998, through August 14, 1998. The 
last biweekly notice was published on August 12, 1998 (63 FR 43200).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve 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. The basis 
for this proposed determination for each amendment request is shown 
below.
    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 
before action is taken. 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 Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, 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, 11545 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. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By September 25, 1998, 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 and at the local public 
document room for the particular facility involved. 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 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 a 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 a 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 request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a

[[Page 45523]]

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 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: July 20, 1998.
    Description of amendments request: The amendment incorporates the 
changes described below into the Technical Specifications (TS) for 
Calvert Cliffs Unit 2. Currently, Calvert Cliffs has four emergency 
diesel generators (EDGs), two per Unit, to provide the onsite emergency 
power supply for both Units. The Unit 2 EDGs rely on the Service Water 
(SRW) System to provide their cooling water. During the Unit 2 1999 
Refueling Outage, Baltimore Gas and Electric Company will replace the 
SRW heat exchangers on Unit 2. During the period of the replacement, no 
SRW cooling will be available for Unit 2. Therefore, both Unit 2 EDGs 
would be inoperable during the replacement work. Unit 1 will continue 
at full power operation during the Unit 2 refueling outage.
    The loss of both EDGs on Unit 2 presents several challenges. First, 
a number of outage activities require an EDG to be operable. BGE 
proposes to provide an alternate cooling water supply to maintain the 
EDGs operable to fulfill the TS requirements. One EDG will be provided 
with cooling water from the Unit 1 SRW System. The other EDG will be 
provided with cooling water from an independent external cooling 
system. Second, Unit 1 is scheduled to be in Mode 1 operation during 
this time. The No. 12 Control Room Emergency Ventilation System, No. 12 
Control Room Emergency Temperature System, and a Hydrogen Analyzer are 
affected by this work because they obtain their emergency power from a 
Unit 2 EDG. These components support Unit 1 continued operation. 
Therefore, the loss of both Unit 2 EDGs would impact operations on both 
units.
    There are several issues associated with this change that create an 
Unreviewed Safety Question (USQ) as defined by 10 CFR 50.59. There is 
an increase in the probability of a malfunction due to the use of an 
independent cooling system that is non-safety-related and unprotected 
from seismic or tornado events. The reliance of a Unit 2 EDG on Unit 1 
SRW results in the increase of the probability of a malfunction, also. 
Additionally, these SRW lineups affect the probability of a malfunction 
for other equipment that relies on SRW during an outage. The approval 
of these USQs, will permit a TS Bases change to the description of an 
operable EDG while Unit 2 is in Modes 5 and 6.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The EDGs are used to mitigate the consequences of an accident. 
They are designed to start and load safety-related loads within a 
specified time period. There are two EDGs for Unit 2. Only one is 
required during the refueling outage, since a single failure 
criterion does not apply during this time. However, it is desirable 
for defense-in-depth and shutdown safety reasons to keep both EDGs 
operable. Additionally, one of the EDGs supports operable equipment 
on Unit 1 that remains at power. We are proposing an amendment that 
would allow the EDGs to continue to be operable with an alternate 
cooling water supply. Other than the change in cooling water supply, 
we are not affecting or modifying the operation of the EDGs. The 
EDGs are not an accident initiator for any previously evaluated 
accident. Therefore, the proposed change does not involve an 
increase in the probability of an accident previously evaluated.
    The EDGs are designed to mitigate the consequences of an 
accident. They will continue to perform that function while being 
supplied with an alternate source of cooling water. The consequences 
of a design basis accident during the period when the alternate 
cooling water is being supplied is not increased because the 
operation of the EDGs has not been adversely affected. Any 
additional electrical loads (such as cooling tower pumps and fans) 
or additional cooling loads (such as additional SRW flow to the No. 
2A EDG) have been evaluated and found to be acceptable under 
conditions postulated to exist during the outage. Therefore, the 
proposed change does not significantly increase the consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The EDGs are not being modified by this proposed change nor will 
any unusual operator actions be required. The EDGs will continue to 
operate in the same manner as before. However, the cooling water 
supplies have been altered and were evaluated under the provisions 
of 10 CFR 50.59 and determined to result in a USQ. These USQs are 
evaluated below.
    The first identified USQ is due to the realignment of a Unit 1 
SRW subsystem to also support a Unit 2 EDG (2A). This alignment will 
rely on two control valves (one to each EDG) to function properly in 
order to provide adequate SRW flow to both EDGs. If one of the 
valves should fail open, it may result in insufficient SRW flow or 
increased SRW temperatures, as the EDGs share the same cooling 
supply. This is an increase in the probability of a malfunction 
because the operability of a EDG relies on both control valves 
performing properly. We believe that this is an acceptable condition 
because the control valves and their air supply are safety-related 
and will be performing their design function. The control valves are 
not being modified by the temporary configuration nor will any 
operator action be required. The control valves will continue to 
operate in the same manner. Therefore, because the malfunction is 
the same as previously identified for these valves and only the 
probability has increased, a new or different type of accident has 
not been created.
    The next USQ identifies a condition where a Unit 2 EDG is 
dependent on a Unit 1 EDG for cooling water. The Unit 1 EDG powers 
the pump for the cooling water system that will now provide cooling 
to both EDGs. Although the consequences of a loss of cooling water 
is the same (i.e., the EDG fails), the probability of a malfunction 
for the Unit 2 EDG has increased because it now depends on the Unit 
1 EDG to maintain its operability. We believe that this is an 
acceptable condition because the Unit 1 EDG is safety-related and is 
proven reliable through testing.

[[Page 45524]]

Additionally, the EDG will not be operated in a manner different 
than it is currently. It is not being modified by the proposed 
change nor will any additional operator actions be required. A 
failure analysis shows that failure of the No. 1B EDG will not 
result in the total loss of any safety function for either unit. 
Therefore, the possibility of a new or different type of accident 
has not been created.
    A USQ has been identified related to the use of a temporary 
cooling system to provide cooling to an EDG. The cooling system what 
is proposed is not safety-related and is not protected from natural 
phenomenon. This leads to an increase in the probability of a 
malfunction because the cooling system is more likely to fail than a 
safety-related, protected system. We believe that this is an 
acceptable condition for the limited time we propose to use the 
cooling system. The consequences of a cooling system failure are no 
different than those of a failure of the SRW System. The events most 
likely to cause the cooling system to fail are seismic events and 
severe weather. Severe weather is not highly probable during this 
time of year. Significant seismic events are not probable on this 
part of the east cost. The cooling tower has been used before at 
Calvert Cliffs to support testing of the EDGs during outages. The 
cooling tower will have enhanced design features that will improve 
its reliability, such as two pumps. The piping provided to and from 
the cooling system will be steel and will be provided with flexible 
joints making it rugged and flexible. Additionally, the cooling 
tower will be placed close to the Auxiliary Building and the makeup 
water piping will be run underground for part of its length. These 
measures help to protect the cooling tower and its piping from 
severe weather events. The EDG is not being altered by this 
temporary configuration. It will continue to operate as before. No 
additional operator action is required for the cooling tower to 
perform its function. Therefore, the possibility of a new or 
different type of accident has not been created.
    This USQ exists because the piping from the cooling tower to the 
EDG is not safety-related and could break, causing a flood in the 
EDG room. This creates an increase in the probability of a 
malfunction because of the increased probability of flooding in the 
room. We believe that this increase is acceptable because the piping 
is constructed from rugged materials and is flexibly connected to 
the EDG. This reduces the chance that flooding will occur. If 
flooding were to occur and the contents of the cooling system were 
spilled into the room, it would not impact safety-related components 
in the room because the water would not be deep enough. Therefore, 
the possibility of a new or different accident has not been created.
    Therefore, the possibility of a new or different type of 
accident from any accident previously evaluated has not been 
created.
    3. Would not involve a significant reduction in a margin of 
safety. The operability of the EDGs in Modes 5 and 6 ensures that 
emergency power is available to mitigate the consequences of a fuel 
handling accident and a boron dilution accident. Additionally, it 
provides emergency power for shutdown cooling and spent fuel pool 
cooling. One of the Unit 2 EDGs provides power to the shared Control 
Room Emergency Ventilation System, Control Room Emergency 
Temperature System, and the Hydrogen Analyzer needed to Support Unit 
1 power operation. The proposed changes do not affect the function 
of the EDGs. Because of the increased probability of a malfunction 
of equipment important to safety (SRW support for the EDGs), the 
margin of safety is reduced. However, the reduction is not 
significant. As described above, each USQ has been evaluated and 
determined to not have a significant impact on safety.
    To provide additional assurance that all reasonable steps have 
been taken to ensure the operability of the Unit 2 EDGs while in the 
temporary configuration, the following actions will be taken in 
addition to the installation of the temporary modifications as 
described above:
    To prevent the loss of the normal power supply to the Control 
Room Emergency Ventilation System and Control Room Emergency 
Temperature System, we will restrict maintenance activities on three 
of the four offsite transmission lines until the Unit 2 EDGs are 
returned to normal configuration.
    To monitor risk, Unit 1 and 2 equipment taken out-of-service 
during this period will be evaluated in the Unit 1 weekly quarterly 
system schedule evaluations.
    To ensure that weather-related events cannot cause a loss of all 
emergency power on Unit 2 during periods of reduced inventory, the 
No. 2A EDG will remain operable during reduced inventory periods.
    To ensure that backup power is available to any of the safety-
related buses, the No. 0C Diesel Generator will not be taken out-of 
service for planned maintenance and will remain available to be 
connected to any of the safety-related buses.
    We believe that the reduction in the margin of safety 
represented by this temporary license amendment is not significant 
based on our evaluation and management of plant risk, the 
reliability of the EDGs, the availability of redundant EDGs, the 
availability of the Station Blackout Diesel Generator and the 
mitigating features described above. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa, Director.

Duke Energy Corporation (DEC or licensee), Docket Nos. 50-369 and 50-
370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: May 27, 1997, as supplemented by letters 
dated March 9, March 20, April 20, June 3, June 24, July 7, July 21, 
and July 22, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) of each unit to conform with 
NUREG-1431, Revision 1, ``Standard Technical Specifications--
Westinghouse Plants.'' The Commission had previously issued a Notice of 
Consideration of Issuance of Amendments in the Federal Register on July 
15, 1997 (62 FR 37940) covering all the proposed changes that were 
indeed within the scope of NUREG-1431. In DEC's May 27, 1997, 
submittal, there are proposed changes that are beyond the scope of 
NUREG-1431, which were, thus, not covered by the staff's July 15, 1997, 
notice. The following description and no significant hazard analysis 
covers a beyond-scope change.
    The licensee proposed to change Section 3.4.6.1 regarding reactor 
coolant leakage detection systems; a system comprising diverse 
instruments such as gaseous radioactivity monitoring, containment floor 
and equipment sump monitoring, etc. In addition to the instruments 
specified by this section, the plant has other installed instruments 
such as monitors for humidity, temperature, etc., which can provide 
indication for reactor coolant leakage. Currently, this specification 
allows operation up to 30 days if the containment floor and equipment 
sump monitoring system is inoperable. The proposed change would impose 
a requirement to perform a precision water balance of the reactor 
coolant system every 24 hours during this period. The proposed change 
would also reduce the number of monitors required operable provided 
compensatory measures are performed or diverse instruments continue to 
be available.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration for each of the above proposed changes. The NRC staff has 
reviewed the licensee's analyses against the standards of 10 CFR 
50.92(c). The NRC staff's analysis is presented below.
    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously

[[Page 45525]]

evaluated? The proposed change will not affect the safety function of 
the subject systems. There will be no direct effect on the design or 
operation of any plant structures, systems, or components. No 
previously analyzed accidents were initiated by the functions of these 
systems, and the systems were not factors in the consequences of 
previously analyzed accidents. Therefore, the proposed change will have 
no impact on the consequences or probabilities of any previously 
evaluated accidents.
    2. Will the change create the possibility of a new or difference 
kind of accident from any accident previously evaluated?
    The proposed change would not lead to any hardware or operating 
procedure change. Therefore, no new equipment failure modes or 
accidents from those previously evaluated will be created.
    3. Will the change involve a significant reduction in a margin of 
safety? Margin of safety is associated with confidence in the design 
and operation of the plant. The proposed change to the TS do not 
involve any change to plant design, operation, or analysis. Thus, the 
margin of safety previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for each of the proposed change. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Project Director: Herbert N. Berkow.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: July 21, 1998.
    Description of amendment request: The proposed change to the 
Technical Specifications would: (1) modify Specification 6.2.2.2(a) to 
provide some flexibility to accommodate unexpected absence of on-duty 
shift crew members, (2) eliminate reference to the Manager, Plant 
Operations in Specification 6.2.2.2(j) as the position has been 
eliminated, (3) reduce the maximum time in which to forward audit 
reports to the responsible manager from 60 days to 30 days, (4) replace 
the term ``Vice President'' with the term ``Corporate Officer'' in 
several places in Section 6, and (5) correct several typographical 
errors.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. State the basis for the determination that the proposed 
activity will or will not increase the probability of occurrence or 
consequences of an accident.
    The activity does not alter the design, function or manner of 
operation of any structures, systems or components. Therefore, this 
activity does not increase the probability or consequences of an 
accident.
    2. State the basis for the determination that the activity does 
or does not create the possibility of an accident or malfunction of 
a different type than any previously identified in the SAR.
    The activity does not alter the design, function, or manner of 
operation of any structures, systems or components. Therefore, this 
activity does not create the possibility of an accident or 
malfunction of a different type than any previously identified in 
the SAR.
    3. State the basis for the determination that the margin of 
safety is not reduced.
    The activity does not alter the design, function or manner of 
operation of any structures, systems or components. In addition, a 
decrease in staff for a short period of time on limited occasions is 
not safety significant and permitted by 10 CFR 50.54 (m). Therefore, 
this activity will not reduce the 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.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: July 23, 1998.
    Description of amendment request: Amend facility license to 
establish that the existing Safety Limit Minimum Critical Power Ratio 
(SLMCPR) contained in Technical Specification 2.1.A is applicable for 
the next operating cycle (Cycle 17).
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The derivation of the Cycle 17 SLMCPR for Oyster Creek for 
incorporation into the TS, and its use to determine cycle-specific 
thermal limits, has been performed using NRC-approved methods. 
Additionally, interim implementing procedures, which incorporate 
cycle-specific parameters, have been used. Based on the use of these 
calculations, the Cycle 17 SLMCPR of 1.09 will not increase the 
probability or consequences of an accident.
    The basis of the MCPR Safety Limit calculation is to ensure that 
greater than 99.9% of all fuel rods in the core avoid transition 
boiling if the limit is not violated. A SLMCPR of 1.09 preserves 
adequate margin to transition boiling and fuel damage in the event 
of a postulated accident. The probability of fuel damage is not 
increased.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The MCPR Safety Limit is a Technical Specification numerical 
value designed to ensure that fuel damage from transition boiling 
does not occur as a result of the limiting postulated accident. The 
limit cannot create the possibility of any new type of accident. The 
Cycle 17 SLMCPR has been calculated using NRC-approved methods. 
Additionally, interim procedures, which incorporate cycle-specific 
parameters, have been used. Therefore, the proposed TS change does 
not create the possibiliy of a new or different kind of accident, 
from any accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The Cycle 17 SLMCPR is calculated using NRC-approved methods, 
which are in accordance with the current fuel design and licensing 
criteria. Additionally, interim implementing procedures, which 
incorporate cycle-specific parameters, have been used. The MCPR 
Safety Limit remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving fuel cladding integrity. 
Therefore, the proposed TS change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 45526]]

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.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: August 6, 1998.
    Description of amendment request: The proposed amendment would 
allow changes to the Updated Safety Analysis Report (USAR) to reflect 
the as-built configuration of the reactor building isolation dampers. 
These changes would clarify the USAR discussion of secondary 
containment isolation and revise the calculated offsite dose 
consequences resulting from a postulated refueling accident. No changes 
to the Technical Specifications (TS) are required; the TS Bases, 
Sec. 3.6.4.2, will be revised under the licensee's Bases control 
program to reflect the changes in the USAR analysis.
    Basis for proposed no significant hazards consideration 
determination: 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:

    The enclosed proposed license amendment for the as-built design 
of the Secondary Containment (Reactor Building) isolation dampers is 
judged to involve no significant hazards based on the following:
    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The existing plant design does not involve a significant 
increase in the probability of an accident previously evaluated in 
the Updated Safety Analysis Report (USAR). The current configuration 
does not affect the performance and reliability of the Secondary 
Containment and the Reactor Building Isolation and Control System or 
any system interface in a way that could lead to an accident 
occurring. The current configuration and analysis do not affect any 
accident precursors or initiators, and therefore, does not increase 
the probability of an accident.
    The present plant configuration also does not involve a 
significant increase in the consequences of an accident previously 
evaluated in the USAR. The current design will require a 
clarification to the Secondary Containment safety design basis as 
described in the USAR to reflect the as-built configuration and 
analysis of the plant by stating that the Reactor Building Isolation 
and Control System is designed to limit the release of fission 
products through the normal ventilation discharge path during a 
postulated Refueling Accident.
    The original analysis determined that the consequences of the 
Refueling Accident were significantly less than 1 Rem to the thyroid 
and whole body (maximum off-site dose). When this analysis was 
revised to account for the 90 second motor-operated damper closure 
time, the calculated whole body off-site dose increased, but was 
still less than 1 Rem; the calculated off-site dose to the thyroid, 
however, increased to 2.7 Rem. While this change in the analysis 
represents an order of magnitude increase in consequences (thyroid 
dose increase from 17 milliRem to 2.7 Rem), the actual increase is 
minimal because this increase in consequences is still less than 1 
percent (1%) of the limits specified in 10 CFR 100. Thus the 
consequences still remain well within the regulatory threshold 
specified in 10 CFR 100 and thus pose no undue hazard to the health 
and safety of the public. This proposed amendment does not alter the 
Control Room dose from that which was submitted to the NRC in 
support of Amendment 167.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This proposed license amendment is administrative in nature in 
that it reflects the effects of a revised analysis for the Refueling 
Accident, which is an accident previously analyzed as a Design Basis 
Accident (DBA) in the SAR, based on the present configuration of the 
plant. The current configuration does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated in the USAR. The proposed license amendment does not 
introduce any new equipment or hardware changes, nor does it require 
existing equipment or systems to perform a different type of 
function than they are presently designed to perform. The as-built 
configuration does not introduce any new mode of plant operation, 
thus there are no new accident failure paths created.
    The as-built configuration does not affect any accident 
precursors or initiators and does not create the possibility of a 
new or different kind of accident.
    3. Does not create a significant reduction in the margin of 
safety.
    The present plant configuration does not involve a significant 
reduction in a margin of safety. Technical Specification Bases 
section 3.2.D.2, Reactor Building Isolation and Standby Gas 
Treatment (SGT) Initiation, states that the trip settings for the 
Reactor Building exhaust plenum radiation monitors are based on 
initiating normal ventilation system isolation and SGT System 
operation so that none of the activity released during the refueling 
accident leaves the Reactor Building via the normal ventilation 
path, but rather all the activity is processed by the SGT System. 
This basis statement remains true unless there is a single failure 
of the air-operated Secondary Containment isolation damper. Under 
single failure conditions there would be the potential for a limited 
release through the normal ventilation system prior to complete 
isolation of the secondary containment and initiation of the SGT 
System.
    The significance of this change is minimal, as Technical 
Specification requirements to isolate Secondary Containment are 
still met. The overall function of the Secondary Containment and 
Reactor Building Isolation and Control System, in conjunction with 
other accident mitigation systems, is to limit fission product 
release during and following postulated DBAs. High radiation in the 
Secondary Containment exhaust is an indication of possible gross 
failure of the fuel cladding, possibly due to a Refueling Accident. 
The trip settings for the Reactor Building (Secondary Containment) 
radiation monitors are such that initiation of secondary containment 
isolation and SGT would still occur in sufficient time (within 90 
seconds of detection) to maintain postulated off-site releases well 
within the limits of 10 CFR 100. As stated previously, the effects 
of the 90 second motor-operated damper closure time on Control Room 
dose have already been taken into consideration in the District's 
submittals supporting Amendment 167.

    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.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Project Director: John N. Hannon.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 4, 1998.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) relating to the Condensate 
Storage Tank (CST) and also add a new TS section that would establish 
requirements for the atmospheric steam dump valves (ASDVs) to assure 
their operability. The applicable TS Bases section for the CST would 
also be changed to reflect the proposed changes and a new TS Bases 
section would be added to discuss the new TS section for the ASDVs.
    Specifically, the proposed changes would modify TS 3.7.1.3, ``Plant

[[Page 45527]]

Systems--Condensate Storage Tank,'' by increasing the minimum required 
CST level from 150,000 gallons to 165,000 gallons to account for the 
discharge nozzle pipe elevation above the tank bottom and vortex 
formation in the CST at the auxiliary feedwater supply piping entrance. 
TS 3.7.1.7, ``Plant Systems--Atmospheric Steam Dump Valves,'' would be 
added to provide the requirements necessary to assure that the ASDVs 
will be available to either maintain the unit in hot standby or cool 
down the unit to shutdown cooling entry conditions if the condenser 
steam dump valves are not available. As previously noted, the TS Bases 
would be modified to reflect the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to increase the minimum required Condensate 
Storage Tank (CST) level of Technical Specification 3.7.1.3 will 
ensure sufficient water is available for the Auxiliary Feedwater 
(AFW) System to function as designed to mitigate design basis 
accidents. There will be no adverse effect on equipment important to 
safety. Therefore, the proposed change will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change to add a Technical Specification for the 
Atmospheric Steam Dump Valves (ASDVs) will provide additional 
assurance that the ASDVs will be available to either maintain the 
unit in hot standby, or cool down the unit to Shutdown Cooling (SDC) 
entry conditions if the condenser steam dump valves are not 
available. The proposed change does not alter the way any structure, 
system, or component functions. There will be no adverse effect on 
any design basis accident previously evaluated or on any equipment 
important to safety. Therefore, the proposed change will not result 
in a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes have no adverse effect on any of the design 
basis accidents previously evaluated. Therefore, the license 
amendment request does not impact the probability of an accident 
previously evaluated nor does it involve a significant increase in 
the consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to increase the minimum required CST level 
will ensure the AFW System will function as designed to mitigate 
design basis accidents. The proposed change to add a Technical 
Specification for the ASDVs will provide additional assurance that 
the ASDVs will be available, if needed. There will be no adverse 
effect on equipment important to safety. Therefore, there will be no 
significant reduction of margin of safety as defined in the Bases 
for Technical Specifications affected by these proposed changes.

    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.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: William M. Dean.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment request: March 18, 1998.
    Description of amendment request: The proposed amendment would 
revise the Bases for Technical Specification (TS) 3/4.6.2.1, 
``Containment Spray System,'' of the combined technical specifications 
for the Diablo Canyon Power Plant, Unit Nos. 1 and 2, to clarify that 
containment spray is not required to be actuated during recirculation, 
but may be actuated at the discretion of the Technical Support Center. 
Additionally, the Bases would be clarified to state that the ability to 
spray containment using the residual heat removal (RHR) system is 
demonstrated by opening the RHR Spray Ring Cross Connect Valve 9003 A 
or B. The Bases will also be clarified to state that flow to the spray 
headers can be established with only one operable RHR pump by closing 
the cold leg discharge valve 8809 A or B.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Containment spray (CS) in the recirculation mode of post-loss-
of-coolant accident (LOCA) safety injection (SI) is used only after 
the accident has already occurred. Its availability or 
unavailability is unrelated to, and is not a precursor for, an 
accident that has already been initiated. The availability or 
unavailability of CS recirculation spray does not involve any 
physical change in plant systems, structures, or components, and 
there is no change in preaccident operating procedures, so there is 
no change to the probability of an accident occurring as a result of 
any such changes. The recirculation mode of emergency core cooling 
is only used following a LOCA; therefore, an evaluation of the 
effects of the use or absence of CS in the recirculation mode 
applies only to a LOCA and not to any other type of accident 
analyzed in the Final Safety Analysis Report (FSAR).
    The peak post-LOCA containment pressure and temperature 
conditions occur prior to the recirculation phase of SI, and are not 
affected by CS operation during the recirculation mode of SI. The 
long term pressure and temperature profiles are slightly increased 
if recirculation spray is unavailable but are still within the dose 
analysis and equipment qualification requirements. There is no 
effect on the offsite dose analysis or on equipment operability.
    If CS is not operated in the recirculation mode, there is no 
reduction in the amount of emergency core cooling system (ECCS) 
water pumped into the reactor vessel. Since the flow to the reactor 
is not reduced, core cooling is not adversely affected if 
recirculation spray is not used. If recirculation spray is used 
under Technical Support Center (TSC) direction with only one train 
of residual heat removal (RHR) in operation, ECCS flow to the 
reactor will be reduced, but analysis has shown that the flow to the 
reactor in this situation is still in excess of that needed to 
supply the required core cooling. Therefore, although it is not 
required, it would still be possible to establish CS in the 
recirculation mode with only one train of RHR in operation, if 
considered desirable by the TSC.
    From the above discussion, it can be seen that the consequences 
of an accident analyzed in the FSAR are not increased because the 
absence of recirculation spray has no effect on the dose analyses 
and the effect on other accident parameters is within limits.
    Therefore, the proposed changes do not involve a significant 
increase in the

[[Page 45528]]

probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The possibility of a malfunction of a different type than 
previously evaluated is not created for the following reasons:
    Data provided in the FSAR can be used to determine that the 
iodine removal function for the CS system is completed in 
approximately 26 minutes, and prior to completing switchover to the 
recirculation mode after a LOCA. The statements in previous 
revisions of the FSAR that recirculation spray will continue for 2 
hours to remove iodine are considered to be descriptive in nature, 
explaining an additional capability of the CS system, but not relied 
upon or evaluated in the FSAR.
    The post-LOCA containment environmental conditions without 
recirculation spray remain bounded by those for which safety-related 
equipment inside containment is qualified; therefore, there is no 
resulting increase in the probability that it will malfunction. 
There is no other new mechanism created by the unavailability of 
recirculation spray that would lead to any greater probability of 
malfunction of safety-related equipment.
    The peak post-LOCA containment pressure and temperature 
conditions occur prior to the recirculation phase of SI, and are not 
affected by CS operation during the recirculation mode of SI. Also, 
the Diablo Canyon Power Plant (DCPP) design bases and accident 
analyses do not assume any contribution to post-accident containment 
hydrogen mixing from recirculation spray. The DCPP design basis has 
always assumed that hydrogen mixing is achieved by containment fan 
cooler unit operation alone.
    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 a margin of safety.
    Technical Specification (TS) 3/4.6.2.1, ``Containment Spray 
System,'' requires the operability of two trains of CS with each 
train capable of taking suction from the refueling water storage 
tank (RWST) and transferring spray function to an RHR train taking 
suction from the containment sump. With the proposed changes, the 
capability to perform the required alignment remains unaffected. 
However, the ability to actually provide CS in the recirculation 
mode of SI is limited by procedure in the event of failure of a 
train of auxiliary saltwater, component cooling water, or RHR. This 
does not affect the margin of safety as defined in the TS Bases. The 
Bases for CS operability are to ensure pressure reduction, cooling 
capability, and iodine removal from the containment atmosphere 
consistent with the assumptions used in the safety analyses.
    All pressure reduction, cooling, and iodine removal parameters 
assumed in the accident analyses continue to bound those resulting 
in the event that recirculation spray is not used. The accident 
analyses require that the peak post-accident pressure does not 
exceed 47 psig, and that post-accident pressure be reduced to less 
than half the peak within 24 hours. These requirements are still 
met, but the long term pressure is slightly higher. Since these 
requirements are based on minimizing leakage rates and on 
environmental qualification concerns, and since the leakage rate in 
the offsite dose analysis and pressures for which safety-related 
equipment inside containment is qualified still bound the analysis 
results, a slightly higher long term pressure has no effect on 
safety margins. Although the long term temperature profile increases 
slightly with no recirculation spray, the equipment is still 
environmentally qualified for these temperatures, so again margin is 
maintained. The use of recirculation spray is not credited in the 
offsite or control room dose analyses since the containment 
atmospheric iodine decontamination factor reaches 1000 prior to the 
time recirculation spray is placed in service, so there is no loss 
of margin in the offsite and control room dose analyses. None of the 
accident analysis limits are exceeded in the absence of 
recirculation spray.
    The function of CS to inject NaOH into the containment 
atmosphere and sump is not affected by the proposed changes. The 
same amount of RWST water will be pumped into the containment via 
the CS system for a given size LOCA with or without recirculation 
spray, so the same amount of NaOH is injected into the containment, 
and hence there is no effect on sump pH, iodine retention, or the 
dose analysis.
    In the event that recirculation spray is established under TSC 
direction with only one train of RHR in operation, there is no 
reduction in the margin of safety from the resulting reduced flow to 
the core since analysis has demonstrated that even with no RHR flow 
to the RCS, the resulting flow to the core will still be greater 
than that required to maintain adequate core cooling and maintain 
peak clad temperatures within limits.
    The functions specified for the CS system in the TS Bases are to 
ensure post-accident pressure reduction, cooling capability, and 
iodine removal from the containment atmosphere consistent with the 
assumptions used in the safety analyses. Since these functions are 
maintained within the limits of the safety analyses even in the 
absence of recirculation spray, the operability of the CS system as 
required by TS 3.6.2.1 is maintained.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room Location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna 
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: August 5, 1998.
    Description of amendment request: The amendment to Unit 2 Technical 
Specifications (TS) involves the addition of a new section entitled 
``Oscillation Power Range Monitoring (OPRM) Instrumentation'' and 
revisions to Section 3.4.1 ``Recirculation Loops Operating'' to remove 
the specifications related to thermal power stability which will not be 
required after the installation of the OPRM instrumentation. Unit 2 is 
currently operating under Interim Corrective Actions (ICAs) defined in 
TS 3.4.1 that specify restrictions on plant operation and actions by 
operators in response to instability events. The OPRM system provides 
an automatic long-term solution to the instability issue and eases the 
burden on the operator.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposal does not involve an increase in the probability or 
consequences of an accident previously evaluated.
    The OPRM most directly affects the [Average Power Range Monitor] 
APRM and [Local Power Range Monitor] LPRM portions of the Power 
Range Neutron Monitoring system. Its installation does not affect 
the operation of these sub-systems. None of the accidents or 
equipment malfunctions affected by these sub-systems are affected by 
the presence or operation of the OPRM.
    The APRM channels provide the primary indication of neutron flux 
within the core and respond almost instantaneously to neutron flux 
changes. The APRM Fixed Neutron Flux-High function is capable of 
generating a trip signal to prevent fuel damage or excessive reactor 
pressure. For the [American Society of Mechanical Engineers] ASME 
overpressurization protection analysis in [Final Safety Analysis 
Report] FSAR Chapter 5, the APRM Fixed Neutron Flux-High function is 
assumed to terminate the main steam isolation valve closure event. 
The high flux trip, along with the safety/relief valves, limit the 
peak reactor pressure vessel

[[Page 45529]]

pressure to less than the ASME Code limits. The control rod drop 
accident (CRDA) analysis in Chapter 15 takes credit for the APRM 
Fixed Neutron Flux-High function to terminate the CRDA. The 
Recirculation Flow Controller Failure event (pump runup) is also 
terminated by the high neutron flux trip. The APRM Fixed Neutron 
Flux-High function is required to be OPERABLE in MODE 1 where the 
potential consequences of the analyzed transients could result in 
the Safety Limits (e.g., [Minimum Critical Power Ratio] MCPR and 
Reactor pressure) being exceeded.
    The installation of the OPRM equipment does not increase the 
consequences of a malfunction of equipment important to safety. The 
APRM and [Reactor Protection System] RPS systems are designed to 
fail in a tripped (fail safe) condition; the OPRM will have no 
affect on the consequence of the failure of either system. An 
inoperative trip signal is received by the RPS any time an APRM mode 
switch is moved to any position other than Operate, an APRM module 
is unplugged, the electronic operating voltage is low, or the APRM 
has too few LPRM inputs. These functions are not specifically 
credited in the accident analysis, but are retained for the RPS as 
required by the NRC approved licensing basis.
    The OPRM allows operation under current operating conditions 
presently restricted by the current Technical Specifications by 
providing automatic suppression functions in the area of concern in 
the event an instability occurs. The consequences of any accident or 
equipment malfunction are not increased by operating under those 
conditions. Although protected by the OPRM from thermal-hydraulic 
core instabilities above 30% core power, operation under natural 
core recirculation conditions is not allowed. No accidents or 
transients of a type not analyzed in the FSAR are created by 
operating under these conditions with the protection of the OPRM 
system.
    This change does not increase the probability of an accident as 
previously evaluated. The OPRM is designed and installed to not 
degrade the existing APRM, LPRM, and RPS systems. These systems will 
still perform all of their intended functions. The new equipment is 
tested and installed to the same or more restrictive environmental 
and seismic envelopes as the existing systems.
    The new equipment has been designed and tested to the 
electromagnetic interference (EMI) requirements of Reference 2, 
which assures correct operation of the existing equipment. The new 
system has been designed to single failure criteria and is 
electrically isolated from equipment of different electrical 
divisions and from non-1E equipment. The electrical loading is 
within the capability of the existing power sources and the heat 
loads are within the capability of existing cooling systems. The 
OPRM allows operation under operating conditions presently forbidden 
or restricted by the current Technical Specifications. No other 
transient or accident analysis assumes these operating restrictions.
    Based upon the analysis presented above, PP&L concludes that the 
proposed action does not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposal does not create the probability of a new or 
different type of accident from any accident previously evaluated. 
The OPRM system is a monitoring and accident mitigation system that 
cannot create the possibility for an accident.
    The OPRM will allow operation in conditions currently restricted 
by the current Technical Specifications. Although protected by the 
OPRM from thermal-hydraulic core instabilities above 30% core power, 
operation under natural circulation conditions is not allowed. No 
accidents or transients of a type not analyzed in the FSAR are 
created by operating under these conditions with the protection of 
the OPRM system. No new failure modes of either the new OPRM 
equipment or of the existing APRM equipment have been introduced. 
Quality software design, testing, implementation and module self-
health testing provides assurance that no new equipment malfunctions 
due to software errors are created. The possibility of an accident 
of a new or different type than any evaluated previously is not 
created.
    The new OPRM equipment is designed and installed to the same 
system requirements as the existing APRM equipment and is designed 
and tested to have no impact on the existing functions of the APRM 
system. Appropriate isolation is provided where new interconnections 
between redundant separation groups are formed. The OPRM modules 
have been designed and tested to assure that no new failure modes 
have been introduced.
    Therefore, the proposed change does 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.
    There has been no reduction in the margin of safety as defined 
in the basis for the Technical Specifications. The OPRM system does 
not negatively impact the existing APRM system. As a result, the 
margins in the Technical Specifications for the APRM system are not 
impacted by this addition.
    Current operation under the ICAs provides an acceptable margin 
of safety in the event of an instability event as the result of 
preventive actions and Technical Specification controlled response 
by the control room operators. The OPRM system provides an increase 
in the reliability of the protection of the margin of safety by 
providing automatic protection of the MCPR safety limit, while the 
protection burden is significantly reduced for the control room 
operators. This protection is demonstrated as described above, and 
in the NRC reviewed and approved Topical Reports NEDO-32465-A and 
CENPD-400-P-A.
    Replacement of the ICA operating restrictions from Technical 
Specifications with the OPRM system does not affect the margin of 
safety associated with any other system or fuel design parameter.
    Therefore, the change does not involve a reduction in the 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.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendments request: July 22, 1998.
    Description of amendments request: The proposed amendments would 
change Technical Specification Tables 3.3.6.1-1 and 3.3.6.2-1 by 
increasing the Allowable Values for the high radiation trip for the 
exhaust monitors for the reactor building and the refueling floor.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The Unit 1 and Unit 2 reactor building and refueling floor 
ventilation exhaust radiation monitors perform no function in 
preventing, or decreasing the probability of, a previously evaluated 
accident. The monitors are designed to monitor ventilation exhaust 
for indications of a release of radioactive material resulting from 
a design basis accident and initiate appropriate protective actions. 
Because the proposed changes affect only the ventilation exhaust 
radiation monitors, the probability of an accident previously 
evaluated remains the same.
    The function of the reactor building and the refueling floor 
ventilation exhaust radiation monitors, in combination with other 
accident mitigation systems, is to limit fission product release 
during and following postulated design basis accidents. The proposed 
new Allowable Values for the high radiation trip will continue to 
ensure the offsite doses resulting from a design basis accident do 
not exceed the NRC-approved

[[Page 45530]]

licensing basis and FSAR [Final Safety Analysis Report] limits. 
Therefore, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes increase the radiation level at which the 
ventilation exhaust monitors actuate; however, the manner in which 
their actuation logic functions and the systems that isolate or 
actuate as a result are unaffected by the proposed changes. 
Furthermore, the ventilation exhaust monitors will continue to 
perform their design function of limiting offsite doses to NRC-
approved licensing basis and FSAR limits at the higher Allowable 
Values. Therefore, the proposed changes cannot create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The Bases for Unit 1 and Unit 2 Technical Specification Tables 
3.3.6.1-1 and 3.3.6.2-1 state that the Allowable Values for the 
reactor building and refueling floor ventilation exhaust radiation 
monitors ``are chosen to ensure radioactive releases do not exceed 
offsite dose limits.'' The proposed Allowable Values ensure the 
radiation monitors actuate at a radiation level sufficient to ensure 
offsite doses are within the NRC-approved licensing basis and FSAR 
limits. The proposed Allowable Values comply with the margin of 
safety defined in the Technical Specifications Bases for the 
ventilation exhaust radiation monitors; therefore, the proposed 
changes do not reduce 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.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: July 7, 1998.
    Description of amendment request: The proposed amendment would 
revise the spent fuel pool criticality analysis and rack utilization 
schemes by allowing credit for spent fuel pool soluble boron.
    Basis for proposed no significant hazards consideration 
determination: 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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The presence of soluble boron in the spent fuel pool water for 
criticality control does not increase the probability of a fuel 
assembly drop accident in the spent fuel pool. The handling of the 
fuel assemblies in the spent fuel pool has always been performed in 
borated water.
    The criticality analysis shows the consequences of a fuel 
assembly drop accident in the spent fuel pool are not affected when 
considering the presence of soluble boron. The rack Keff 
[K effective] remains less than or equal to 0.95.
    There is no increase in the probability of an accident. The 
proposed change does allow a greater number of fuel storage 
configurations in the spent fuel pool. While this could increase the 
probability of a fuel misloading, the presence of sufficient soluble 
boron in the spent fuel pool precludes criticality as a result of 
the misloading. Fuel assembly placement will continue to be 
controlled pursuant to approved fuel handling procedures and will be 
in accordance with the Technical Specification spent fuel rack 
storage configuration limitations.
    There is no increase in the consequences of the accidental 
misloading of spent fuel assemblies into the spent fuel pool racks. 
The criticality analyses demonstrate that the pool Keff 
will remain less than or equal to 0.95 following an accidental 
misloading due to the boron concentration of the pool. The proposed 
Technical Specification limitation will ensure that an adequate 
spent fuel pool boron concentration is maintained.
    There is no increase in the probability of the loss of normal 
cooling to the spent fuel pool water when considering the presence 
of soluble boron in the pool water for subcriticality control since 
a high concentration of soluble boron has always been maintained in 
the spent fuel pool water.
    Reactivity changes due to spent fuel pool temperature changes 
have been evaluated. The basic case criticality analysis covers a 
``normal'' spent fuel pool temperature range of 50 degrees F to 160 
degrees F. Spent fuel pool temperature accidents are considered 
outside the normal temperature range extending from 32 degrees F to 
240 degrees F. In all spent fuel pool temperature accident cases, 
sufficient reactivity margin is available to the 0.95 
Keff limit without requiring additional soluble boron 
above the base case level. Because adequate soluble boron will be 
maintained in the spent fuel pool water to maintain Keff 
less than or equal to 0.95, the consequences of a loss of normal 
cooling to the spent fuel pool will not be increased.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Spent fuel handling accidents are not new or different types of 
accidents, they have been analyzed in Section 15.7.4 of the Updated 
Final Safety Analysis Report.
    Criticality accidents in the spent fuel pool are not new or 
different types of accidents, they have been analyzed in the Updated 
Final Safety Analysis Report and in Criticality Analysis reports 
associated with specific licensing amendments for fuel enrichments 
up to and exceeding the nominal 4.95 weight percent U\235\ [Uranium-
235] that is assumed for the proposed change.
    Current Technical Specifications contain limitations on the 
spent fuel pool boron concentration. The actual boron concentration 
in the spent fuel pool has been maintained at a higher value. The 
proposed changes to the Technical Specifications establish new boron 
concentration requirements for the spent fuel pool water consistent 
with the results of the new criticality analysis (Attachment 2).
    Since soluble boron has always been maintained in the spent fuel 
pool water, and is currently required by Technical Specifications, 
the implementation of this new requirement will have little effect 
on normal pool operations and maintenance. A dilution of the spent 
fuel pool soluble boron has always been a possibility; however, it 
was shown in the spent fuel pool dilution evaluation (Attachment 3) 
that there are no credible dilution events for which the spent fuel 
pool Keff could increase to greater than 0.95. Therefore, 
the implementation of new limitations on the spent fuel pool boron 
concentration will not result in the possibility of a new kind of 
accident.
    The proposed changes to Technical Specifications 3.9.13, 4.9.13, 
and 5.6 continue to specify the requirements for the spent fuel rack 
storage configurations. Since the proposed spent fuel pool storage 
configuration limitations will be similar to the current ones, the 
new limitations will not have any significant effect on normal spent 
fuel pool operations and maintenance and will not create any 
possibility of a new or different kind of accident. Verifications 
will continue to be performed to ensure that the spent fuel pool 
loading configuration meets specified requirements.
    The misloading of a fuel assembly in the required storage 
configuration has been evaluated. In all cases, the rack 
Keff remains less than or equal to 0.95. Removal of an 
[sic] Rod Control Cluster Assembly from a checkboard storage 
configuration has been analyzed and found to be bounded by the 
misloading of a fuel assembly.
    As discussed above, the proposed changes will not create the 
possibility of a new or different kind of accident. There is no 
significant change in plant configuration, equipment design or 
equipment.
    Under the proposed amendment, no changes are being made to the 
racks themselves, any other systems, or to the physical structures 
of the Fuel Handling

[[Page 45531]]

Building itself. Therefore, the proposed change does 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 a margin of safety.
    The Technical Specification changes proposed by this License 
Amendment Request and the resulting spent fuel storage operation 
limits will provide adequate safety margin to ensure that the stored 
fuel assembly array will always remain subcritical. Those limits are 
based on a plant specific criticality analysis (Attachment 2) 
performed in accordance with Westinghouse spent fuel rack 
criticality analysis methodology.
    While the criticality analysis utilized credit for soluble 
boron, storage configurations have been defined using 95/95 
Keff calculations to ensure that the spent fuel rack 
Keff will be less than 1.0 with no soluble boron. Soluble 
boron credit is used to offset uncertainties, tolerances, and off-
normal conditions and to provide subcritical margin such that the 
spent fuel pool Keff is maintained less than or equal to 
0.95.
    The loss of substantial amounts of soluble boron from the spent 
fuel pool which could lead to Keff exceeding 0.95 has 
been evaluated (Attachment 3) and shown to be not credible. A safety 
evaluation has been performed which shows that dilution of the spent 
fuel pool boron concentration from 2500 ppm to 700 ppm is not 
credible. Also, the spent fuel rack Keff will remain less 
than 1.0 (with a 95/95 confidence level) with the spent fuel pool 
flooded with unborated water. These safety analyses demonstrate a 
level of safety comparable to the conservative criticality analysis 
methodology required by Westinghouse WCAP-14416.
    Based on the above evaluation, the South Texas Project concludes 
that the proposed changes to the Technical Specifications involve no 
significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station, 
Unit No. 3, Oconee County, South Carolina

    Date of amendment request: July 20, 1998.
    Description of amendment request: The proposed amendment would 
extend, on a one-time basis, Technical Specification Surveillance 
4.18.3 for hydraulic and mechanical snubber testing. The tests are 
required to be performed at a frequency of 18 months, with a maximum 
allowed frequency of 22 months, 15 days. The amendment would extend 
this to a maximum of 25 months.
    Date of publication of individual notice in Federal Register: July 
27, 1998 (63 FR 40137).
    Expiration date of individual notice: August 26, 1998.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz 
Mill, Pennsylvania

    Date of application for amendment: December 22, 1997 supplemented 
on June 15, 1998.
    Brief description of amendment: This amendment changes the legal 
name of the licensee for the Westinghouse Test Reactor from 
Westinghouse Electric Corporation to CBS Corporation.
    Date of issuance: July 31, 1998.
    Effective Date: July 31, 1998.
    Amendment No: 7.
    Facility Operating License No. TR-2: This amendment changes the 
legal name of the licensee for the Westinghouse Test Reactor from 
Westinghouse Electric Corporation to CBS Corporation.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38207).
    The Commission has issued a Safety Evaluation for this amendment 
dated July 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document: N/A.

Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear Power 
Station, Unit 3, Grundy County, Illinois

    Date of application for amendment: May 6, 1998.
    Brief description of amendment: The proposed amendment would amend 
Technical Specification (TS) 4.6.E to allow a one-time extension of the 
40-month surveillance interval requirement to set pressure test or 
replace all Main Steam Safety Valves (MSSVs) to a maximum interval of 
60 months as currently allowed by the American Society of Mechanical 
Engineers

[[Page 45532]]

(ASME) Boiler and Pressure Vessel Code (Code).
    Date of issuance: August 7, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 163.
    Facility Operating License No. DPR-25: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: June 3, 1998 (63 FR 
30263).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: November 2, 1994, as 
supplemented January 4, 1995, February 19, 1998, April 28, 1998, and 
June 5, 1998.
    Brief description of amendment: The amendment revises the Technical 
Specifications that have become unnecessary due to previous approved 
amendments, make editorial changes, change managerial titles, update 
references and reporting requirements.
    Date of issuance: August 12, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 198.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56365).
    The January 4, 1995, February 19, 1998, April 28, 1998, and June 5, 
1998, letters provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: January 28, 1998 (NRC-98-0011) 
as supplemented March 12 and June 9, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification 3.4.2.1, ``Safety/Relief Valves,'' changing the safety 
relief valve (SRV) setpoint tolerance from plus or minus 1 percent to 
plus or minus 3 percent. An associated footnote is revised to indicate 
that, although the as-found setpoint tolerance is now plus or minus 3 
percent, the as-left settings of the SRVs shall be within plus or minus 
1 percent of the specified setpoints prior to installation of the SRVs 
after testing. Bases section 3/4.4.2 is also revised.
    Date of issuance: July 31, 1998.
    Effective date: July 31, 1998, with full implementation prior to 
restart from the sixth refueling outage.
    Amendment No.: 123.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9600). The March 12 and June 9, 1998, letters provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial proposed no significant 
hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: June 26, 1998 (NRC-98-0040) as 
supplemented July 16, 1998 (NRC-98-0096), and July 23, 1998 (NRC-98-
0117).
    Brief description of amendment: The amendment provides a one-time 
extension of the interval for a number of technical specification (TS) 
surveillance requirements that will be performed during the sixth 
refueling outage. TS 4.0.2 and Index page xxii are revised and TS 
tables 4.0.2-1 and 4.0.2-2 are replaced to reflect the extensions.
    NRC has also granted the request of Detroit Edison Company to 
withdraw a portion of its June 26, 1998, application. By letter dated 
July 16, 1998, the licensee made some editorial changes and withdrew 
the portion of the submittal related to TS 4.0.5 for the inservice 
testing of two valves. A change to the schedule for these valves will 
be handled within the Inservice Testing Program and a TS change is not 
necessary. For further details with respect to this action, see the 
application for amendment dated June 26, 1998, and the licensee's 
letter dated July 16, 1998, which withdrew this portion of the 
application for the license amendment, and the staff's safety 
evaluation enclosed with the amendment. By letter dated July 23, 1998, 
the licensee added an additional surveillance requirement for two 
instruments to the amendment. The above documents are available for 
public inspection at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and at the local public 
document room listed below.
    Date of issuance: August 4, 1998.
    Effective date: August 4, 1998, with full implementation within 30 
days.
    Amendment No.: 124.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1998 (63 FR 
36273). The July 16 and July 23, 1998, letters provided clarifying 
information and updated TS pages that were within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: July 8, 1998.
    Brief description of amendments: The amendments revise TS 
4.5.4.1.b.1 for testing the Penetration Room Ventilation System air 
flow by adding a reference to the following statement that has been 
added to the bottom of the TS page: ``A temporary noncompliance with 
this surveillance requirement is allowed until August 30, 1998, to 
complete necessary modifications to enable flow testing in accordance 
with ANSI N510-1975.'' This action supersedes the Notice of Enforcement 
Discretion that was issued by the staff on July 8, 1998.
    Date of Issuance: August 7, 1998.

[[Page 45533]]

    Effective date: As of the date of issuance.
    Amendment Nos.: Unit 1--231; Unit 2--231; Unit 3--228.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revise the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. (63 FR 38433 dated July 16, 1998). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by August 17, 1998, but indicated that if the Commission makes 
a final no significant hazards consideration determination, any such 
hearing would take place after issuance of the amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and a final no significant hazards consideration 
determination are contained in a Safety Evaluation dated August 7, 
1998.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: May 19, 1995, as supplemented by 
letters dated February 27 and September 30, 1996.
    Brief description of amendment: The amendment modifies the 
technical specifications (TSs) to extend the allowed outage times 
(AOTs) for a single inoperable Safety Injection Tank (SIT) from one 
hour to 24 hours, and for a single inoperable SIT specifically due to 
malfunctioning SIT water level or nitrogen cover pressure 
instrumentation inoperability from one hour to 72 hours.
    Date of issuance: August 7, 1998.
    Effective date: August 7, 1998.
    Amendment No.: 192.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39439).
    The February 27 and September 30, 1996, submittals provided 
clarifying information that did not change the initial proposed NSHC 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: June 18, 1998, and supplemented 
June 30, 1998.
    Brief description of amendment: The amendment proposed to revise 
the Improved Technical Specifications to allow operation with a number 
of indications previously identified as tube end anolmalies and 
multiple tube end anolmalies in the Crystal River Unit 3 Once Through 
Steam Generator tubes.
    Date of issuance: July 30, 1998.
    Effective date: July 30, 1998.
    Amendment No.: 169.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1998 (63 FR 
35615). The June 30, 1998 supplement included clarifying information 
which did not change the original no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: May 27, 1998.
    Brief description of amendments: The amendments revise the 
Administrative Controls, Unit Staff Section 6.2.2.f of TS to authorize 
the use of various controlled shift structures and durations during a 
nominal (36 to 48 hours) work week. This includes the use of up to 12-
hour shifts without heavy use of overtime.
    Date of Issuance: July 30, 1998.
    Effective Date: July 30, 1998.
    Amendment Nos.: 155 and 93.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 1, 1998 (63 FR 
35989).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of application for amendment: October 31, 1996.
    Brief description of amendment: The amendment deletes Table 3.5.2 
which lists automatic primary containment isolation valves. In 
addition, the amendment clarifies the applicability of an action 
statement that applies to several limiting conditions for operation in 
Section 3.5, and deletes closure time requirements for several 
automatic isolation valves in Section 4.5.F.
    Date of Issuance: August 13, 1998.
    Effective date: August 13, 1998, to be implemented within 60 days.
    Amendment No.: 196.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66707).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois

    Date of application for amendment: October 22, 1995.
    Brief description of amendment: The amendment changes Technical 
Specification 5.2.2.e, ``Unit Staff,'' by revising the requirements for 
controls on the working hours of unit staff who perform safety related 
functions.
    Date of issuance: August 13, 1998.
    Effective date: August 13, 1998.
    Amendment No.: 115.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65681).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public

[[Page 45534]]

Library, 120 West Johnson Street, Clinton, IL 61727.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: January 22, 1998, as 
supplemented July 17, 1998.
    Brief description of amendment: The amendment revises the Millstone 
Unit 3 licensing basis to accept the existing use of epoxy coatings on 
safety related components. The revised licensing basis will be 
incorporated into Chapter 9 of the Final Safety Analysis Report.
    Date of issuance: August 7, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 162.
    Facility Operating License No. NPF-49: Amendment revised the Final 
Safety Analysis Report and the Facility Operating License.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9606).
    The July 17, 1998, letter provided clarifying information that did 
not change the scope of the January 22, 1998, application, and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: September 24, 1996, as 
supplemented October 17, 1996, January 3, January 20, and November 10, 
1997, and January 9, June 8, and July 20, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) for the Prairie Island Nuclear 
Generating Plant Units 1 and 2 to allow use of an alternate steam 
generator tube repair criteria (elevated F-star or EF*) in the 
tubesheet region when used with the repair method of additional roll 
expansion. The amendments incorporate revised acceptance criteria for 
tubes with degradation in the tubesheet region and enable the licensee 
to avoid unnecessary plugging and sleeving of steam generator tubes.
    Date of issuance: August 13, 1998.
    Effective date: August 13, 1998, with full implementation within 30 
days.
    Amendment Nos.: 137 and 128.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64388).
    The licensee's submittals dated January 3, January 20, and November 
10, 1997, and January 9, June 8, and July 20, 1998, provided additional 
clarifying information within the scope of the original Federal 
Register notice and did not affect the staff's initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: February 6, 1998.
    Brief description of amendment: The amendment revises the Reactor 
Protection System Normal Supply Electrical Protection Assembly 
Undervoltage Trip Setpoint.
    Date of issuance: July 29, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 245.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19976).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: November 14, 1997.
    Brief description of amendments: The amendments add technical 
specification (TS) surveillance requirements for the service water 
accumulator vessels. Specifically, surveillance requirements are 
provided for vessel level, pressure and temperature, and discharge 
valve response time. The surveillance requirements are included in TS 
3/4.6.1.1 and 3/4.6.2.3, and the applicable Bases sections are expanded 
to provide supporting information.
    Date of issuance: August 6, 1998.
    Effective date: August 6, 1998.
    Amendment Nos.: 213 and 193.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4432).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: May 6, 1998.
    Brief description of amendment: The requested changes would replace 
the two percent penalty addressed in Surveillance Requirement 
3.2.1.2(a) with a burnup-dependent factor to be specified in the WBN 
Core Operating Limits Report and makes associated changes to the 
administrative controls in Technical Specification 5.9.5 and the BASES.
    Date of issuance: August 10, 1998.
    Effective date: August 10, 1998.
    Amendment No.: 11.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33109).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 10, 1998.
    No significant hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

[[Page 45535]]

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: November 5, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Sections 3.9.7, 4.9.7.1, 4.9.7.2, and 3/
4.9.7 for Unit 1, and 3.9.7, 4.9.7.1, 4.9.7.2, and 3/4 .9.7 for Unit 2, 
allowing the movement of the spent fuel pit gate over the irradiated 
fuel.
    Date of issuance: August 3, 1998.
    Effective date: August 3, 1998.
    Amendment Nos.: 213 and 194.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66146).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

    Dated at Rockville, Maryland, this 19th day of August 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-22766 Filed 8-25-98; 8:45 am]
BILLING CODE 7590-01-P