[Federal Register Volume 63, Number 47 (Wednesday, March 11, 1998)]
[Notices]
[Pages 11913-11931]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-6085]


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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 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 February 13, 1998, through February 27, 
1998. The last biweekly notice was published on February 25, 1998 (63 
FR 9589).

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

[[Page 11914]]

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 April 10, 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 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.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of amendment request: January 14, 1998, which superseded the 
September 3, 1997, submittal.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to reduce the allowable Unit 1 
Reactor Coolant System Dose Equivalent Iodine-131 from 0.35 
microCuries/gram to 0.05 microCuries/gram thru the end of Unit 1, Cycle 
7.
    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.
    Generic Letter 95-05, ``Voltage-Based Repair Criteria For 
Westinghouse Steam Generator Tubes Affected By Outside Diameter 
Stress Corrosion Cracking,'' allows lowering of the RCS [Reactor 
Coolant System] DE-131 [Dose Equivalent Iodine-131] activity as a 
means for accepting higher projected leak rates if justification for 
equivalent I-131 below 0.35 microCuries/gram is provided. Four 
methods for determining the impact of a release of activity to the 
public were reviewed to provide this justification. These four 
methods are as follows:

Method 1: NRC NUREG 0800, Standard Review Plan (SRP) Methodology

[[Page 11915]]

Method 2: Methodology described in a report by J.P. Adams and C.L. 
Atwood, ``The Iodine Spike Release Rate During a Steam Generator 
Tube Rupture,'' Nuclear Technology, Vol. 94, p. 361 (1991) using 
Braidwood Station reactor trip data.
Method 3: Methodology described in a report by J.P. Adams and C.L. 
Atwood, ``The Iodine Spike Release Rate During a Steam Generator 
Tube Rupture,'' Nuclear Technology, Vol. 94, p. 361 (1991) using 
normalized industry reactor trip data.
Method 4: Methodology described in a draft EPRI Report TR-103680, 
Revision 1, November 1995, ``Empirical Study of Iodine Spiking in 
PWR Plants''.

    The effect of reducing the RCS DE I-131 activity limit on the 
amount of activity released to the environment remains unchanged 
when the maximum site allowable primary-to-secondary leak rate is 
proportionately increased and the iodine release rate spike factor 
is assumed to be 500 in accordance with the SRP. With an RCS DE I-
131 activity limit of 1.0 microCuries/gram, the maximum site 
allowable leakage limit was calculated, in accordance with the NRC 
SRP methodology, to be 6.64 gpm at room temperature and pressure. 
ComEd has evaluated the reduction of the RCS DE I-131 activity to 
0.05 microCuries/gram along with the increase of the allowable 
leakage to 132.8 gpm at room temperature and pressure and has 
concluded:

--assuming a spike factor of 500, the maximum activity released is 
not changed, and
--the offsite dose, including the iodine spiking factor, will be 
less than the 10 CFR 100 limits.

    Based on the NRC SRP methodology for dose assessments and 
assuming the iodine spike factor of 500 is applicable at the new 
0.05 microCuries/gram RCS DE I-131 activity limit, the Control Room 
dose, the Low Population Zone dose, and the dose at the Exclusion 
Area Boundary continue to satisfy the appropriately small fraction 
of the 10 CFR 100 dose limits.
    An evaluation of the Control Room dose, attributed to an MSLB 
accident concurrent with steam generator primary-to-secondary 
leakage at the maximum site allowable limit, was performed in 
support of a license amendment request for application of a 1.0 volt 
Interim Plugging Criteria. This evaluation concluded that the 
activity released to the environment during an eight (8) hour time 
period from an MSLB accident (812 Curies for a Pre-accident iodine 
spike and 888 Curies for an accident-initiated iodine spike) is 
bounded by the activity released to the environment from the Loss of 
Coolant design basis accident (1290 Curies). Therefore, the Control 
Room dose, due to the MSLB accident scenario, is bounded by the 
existing Loss of Coolant Accident (LOCA) analysis. The maximum site 
allowable primary-to-secondary leakage is limited by the offsite 
dose at the Exclusion Area Boundary due to an accident-initiated 
spike.
    The report by J.P. Adams and C.L. Atwood, ``The Iodine Spike 
Release Rate During a Steam Generator Tube Rupture,'' Nuclear 
Technology, Vol. 94, p. 361 (1991), concluded that the NRC SRP 
methodology, which specifies a release rate spike factor of 500 for 
iodine activity from the fuel rod to the RCS, is conservative when 
the RCS DE I-131 concentration is greater than 0.3 microCuries/gram. 
In order to evaluate whether a release rate spike factor of 500 is 
conservative below 0.3 microCuries/gram, actual operating data from 
the previous reactor trips of Braidwood Units 1 and 2, with and 
without fuel defects, were reviewed and analyzed using the 
methodology presented in Section II.C of the Adams and Atwood report 
(Method 2). The same five data screening criteria described in the 
Adams and Atwood report were applied to the Braidwood data to ensure 
consistency and validity when comparing the Braidwood results to the 
data in the Adams and Atwood report. Of the reactor trip events at 
Braidwood Units 1 and 2, seventeen (17) met the five data screening 
criteria.
    Seven (7) of the seventeen (17) Braidwood trips occurred during 
cycles with no fuel defects. In all seven of these instances, the 
calculated spike factor was much less than the spike factor of 500 
assumed in the NRC SRP methodology. Braidwood Unit 1 Cycle 7 is 
currently operating with no fuel defects and an RCS DE I-131 
activity of approximately 3E-4 microCuries/gram. The seven previous 
trips with no fuel defects had steady-state iodine values that are 
reasonably close to the current operating conditions. It is 
therefore reasonable to conclude that, assuming continued operation 
with little to no fuel defects, the calculated spike factors from 
these events would reflect an actual event for Unit 1 Cycle 7, i.e. 
the spike factor will be less than 500.
    Since some of the Braidwood spike factors were greater than 500 
when the RCS DE I-131 activity prior to the accident was less than 
0.3 microCuries/gram, ComEd examined the conservatisms in the 
current release rate calculation. The primary reason for the high 
spiking factors contained in the Adams and Atwood report (up to 
12,000), is not because the absolute post-trip release rate is high 
(factor numerator), but rather because the steady-state release rate 
(factor denominator) is low. The Braidwood specific data resulted in 
six (6) events with a calculated release rate spike factor greater 
than 500. It is not expected based upon the Unit 1 Cycle 7 fuel 
conditions that a spiking factor greater than 500 would occur. The 
revised RCS DE I-131 activity limit will also ensure that the 
operating cycle will not continue if significant fuel defects 
develop.
    In order to evaluate the Braidwood specific data against the NRC 
SRP methodology, the release rate for a steady-state RCS DE I-131 
activity of 1.0 microCuries/gram was calculated. Using the Braidwood 
specific data, the pre-trip steady-state release rate is 27.5 Ci/hr. 
Using a release rate spike factor of 500 for the accident-initiated 
spike, the post-trip maximum release rate would be 13,733 Ci/hr (SRP 
Methodology). The highest post-trip iodine release rate from the 
Braidwood trip data, Event 15, was 1335 Ci/hr, it is important to 
remember that this number is determined by conservatively increasing 
the post-trip RCS DE I-131 activity by a factor of three (3), in 
accordance with the Adams and Atwood report.
    The purpose of this amendment request is to reduce the TS 
[Technical Specification] RCS DE I-131 limit by a factor of twenty 
as compared to the original TS RCS DE I-131 limit of 1.0 
microCuries/gram. By decreasing the TS RCS DE I-131 activity by a 
factor of twenty the maximum iodine release rate is 686.7 Ci/hr, 
(13,733 Ci/hr divided by 20). Two (2) of the seventeen (17) 
Braidwood data points exceed this value. Both occurred during cycles 
with fuel defects. Braidwood Unit 1 is currently operating with no 
fuel defects. Fifteen (15) of the 168 data points in the Adams and 
Atwood report exceed 686.7 Ci/hr. For the combined database of 185 
data points, of which 17 exceeded 686.7 Ci/hr, only two of these 
seventeen (17) data points had a pre-trip RCS DE I-131 activity 
below 0.05 microCuries/gram. The 95% confidence prediction for the 
combined data sets bounded one (1) of these two (2) data points. 
This data indicates that the possibility for a post-trip iodine fuel 
release rate to exceed 686.7 Ci/hr, when the pre-trip RCS DE I-131 
concentration is at or below 0.05 microCuries/gram, is small. The 
conservatisms mentioned in the following sections will reduce the 
possibility of exceeding a small fraction of the 10 CFR 100 limits 
should a fuel release greater than 686.7 Ci/hr occur.
    If the Braidwood data were plotted with the Adams and Atwood 
data, the conclusions of the Adams and Atwood report would not be 
compromised. Where the Braidwood data contains spike factors greater 
than 500, the RCS DE I-131 concentrations are below 0.05 
microCuries/gram. Since the Braidwood data includes very few data 
points near 0.05 microCuries/gram (the requested new TS limit), it 
is appropriate to use the Braidwood database combined with the Adams 
and Atwood database near 0.05 microCuries/gram to determine if a 
spike factor of 500 is appropriate. The combined databases contain 
seventy-nine (79) data points with a Pre-Trip RCS DE I-131 activity 
between 0.01 microCuries/gram and 0.10 microCuries/gram. Sixty-two 
(62) of these seventy-nine (79) data points (78%) have spike factors 
less than 500. Using the entire Braidwood database combined with the 
Adams and Atwood database, 141 of the 185 data points (76%) have an 
iodine spike factor less than 500. Therefore, it is reasonable to 
assume that a spike factor of 500 would not be exceeded for a 
majority of the events if an MSLB accident were to occur while the 
RCS DE I-131 activity is at or below 0.05 microCuries/gram. The 
highest spike factor seen in the Adams and Atwood report near a Pre-
Trip RCS DE I-131 activity of 0.05 microCuries/gram was 773 (at 0.05 
microCuries/gram). The corresponding release rate for this event was 
368 Ci/hr which is less than the calculated Braidwood maximum 
release rate of 686.7 Ci/hr.
    The predominant factors in calculating the offsite dose are the 
post-trip iodine release rate from the fuel and the flowrate at 
which the activity is being released to the environment, not whether 
the spike factor is greater than or less than 500. The post-trip DE 
I-131 release rate will determine the level of activity in the RCS 
that will be released. The flowrate will determine at what rate this

[[Page 11916]]

activity is released to the environment. Method 3, which used an 
approach in the Adams and Atwood report, concluded that, at a 95% 
confidence of a 85 percentile, the post-trip iodine release rate was 
bounded by 0.608 Ci/hr-MWe. For Braidwood Station, which has a MWe 
rating of 1175, the post-trip iodine release rate, at a 95% 
confidence of a 85 percentile, should not exceed 714 Ci/hr. Two (2) 
of the seventeen (17) reactor trips from Braidwood exceeded 714 Ci/
hr. These two (2) reactor trips had post-trip iodine release rates 
of 1335 Ci/hr (spike factor of 3471) and 802 Ci/hr (spike factor of 
1483). Both occurred during cycles with fuel defects. Braidwood Unit 
1 is currently operating with no fuel defects.
    In the fourth method, the results from a Draft Electric Power 
Research Institute (EPRI) Report TR-103680, Rev. 1, November 1995, 
``Empirical Study of Iodine Spiking In PWR Power Plants'' were 
applied. The objective of the EPRI study was to quantify the iodine 
spiking in a postulated Main Steam Line Break/Steam Generator Tube 
Rupture (MSLB/SGTR) accident sequences. In the EPRI report, an 
iodine spike factor between 40 and 150 was determined to match data 
from existing plant trips. The maximum iodine spike factor value of 
150 was applied to a steady-state equilibrium RCS DE I-131 activity 
of 0.33 microCuries/gram. The resulting two-hour average iodine 
concentration for a postulated MSLB/SGTR accident sequence was 
determined to be 3.1 microCuries/gram. Since the EPRI report is 
based on industry data and the EPRI method predicted a post-accident 
iodine activity, which is a small fraction of the activity predicted 
by the NRC SRP methodology, it can be expected that, for the 
proposed 0.05 microCuries/gram limit under an MSLB/SGTR accident 
sequence, the post-accident iodine activity would typically be a 
small fraction of the RCS DE I-131 activity predicted by the NRC SRP 
methodology. For Braidwood, using the SRP methodology with an RCS DE 
I-131 activity of 1.0 microCuries/gram and a spike factor of 500, 
the Post-Trip RCS activity two hours after the event would be near 
38 microCuries/gram. At an RCS DE I-131 activity of 0.05 
microCuries/gram, it would require a spike factor of nearly 10,000 
to obtain a Post-Trip RCS DE I-131 activity near 38 microCuries/
gram. With a Post-Trip RCS DE I-131 activity of 38 microCuries/gram, 
an increase in the allowable leak rate could impact the 10 CFR 100 
limits. To accommodate for an increase in the allowable leak rate by 
a factor of twenty, the resultant activity would need to be below 
1.9 microCuries/gram. Two (2) of the seventeen (17) post-trip data 
points from Braidwood exceeded 1.9 microCuries/gram. Both occurred 
during cycles with fuel defects. Braidwood Unit 1 is currently 
operating with no fuel defects. The conservatisms mentioned below 
will reduce the possibility of exceeding a small fraction of the 10 
CFR 100 limits should the post-trip iodine exceed 1.9 microCuries/
gram.
    Based on evaluations by the four methods above, Braidwood can 
conclude that the current methodology (Method 1) used to predict 
iodine spiking is conservative. Although dose projections indicate 
with confidence that the iodine spiking factor limit will be met, 
the conservatisms in the offsite dose calculation and current 
Braidwood Unit 1 operating conditions listed below, provide added 
assurance that the 10 CFR 100 limits, General Design Criteria (GDC) 
19 criteria, and the requirements of NRC Generic Letter 95-05 will 
be satisfied if the iodine spike factor exceeds 500 or the post-trip 
fuel release rate exceeds 686.7 Ci/hr.
    As further assurance that the 10 CFR 100 and GDC 19 limits are 
not exceeded, several conservatisms are inherent to the offsite dose 
calculation. These conservatisms include, but are not limited to:
    1. The meteorological data used is at the fifth percentile. It 
is expected that the actual dispersion of the iodine would result in 
less exposure at the site boundary than the 30 Rem limit of 10 CFR 
100.
    2. Iodine partitioning is not accounted for in the faulted SG. 
With the high pH of the secondary water, some partitioning is 
expected to occur. An iodine partition factor of 0.1 is more 
realistic (per Table 15.1-3 of Reference 8 [the Braidwood Updated 
Final Safety Analysis Report]) than the 1.0 valued (no partitioning) 
used in the offsite dose calculation. This reduces calculated dose 
by 90%.
    3. The activity in the RCS is not expected to increase 
instantaneously with the spike in iodine released from the defective 
fuel.
    4. The results from the Braidwood tube pull data indicate that 
the projected Interim Plugging Criteria leak rate is conservative.
    In addition, the current Braidwood Unit 1 operating conditions 
provide defense in depth and provide further assurance that the 10 
CFR 100 and GDC 19 limits will not be exceeded:
    1. Braidwood Unit 1 is currently operating with a debris 
resistant fuel design which is less likely to develop fuel defects.
    2. As evidenced by industry data, if debris related fuel 
failures are going to occur they are most likely to be occur early 
in the cycle. Braidwood Unit 1 has operated approximately 6 months 
into its current cycle and has seen no signs of fuel defects. 
Therefore, fuel failure prior to completion of the current cycle is 
not likely.
    3. The RCS DE I-131 activity is likely to be less than the TS 
limit. With the current Braidwood Unit 1 RCS DE I-131 activity near 
3E-4 microCuries/gram with no fuel defects, the spike factor is 
expected to be considerably smaller than the 500 value.
    4. It is unlikely, for the short time period this amendment is 
being requested (remainder of Cycle 7), that an accident-initiated 
iodine spike for Braidwood Unit 1 would be greater than the NRC SRP 
assumed value.
    5. Primary-to-secondary leakage is likely to be less than the TS 
limit (150 gpd) in each of the four SGs prior to the event. 
Currently, minimal primary-to-secondary leakage (less than 5 gpd) 
exists at Braidwood Unit 1.
    These proposed changes do not result in a significant increase 
in the consequences of an accident previously analyzed.
    The RCS DE I-131 activity limit is not considered as a precursor 
to any accident. Therefore, this proposed change does not result in 
a significant increase in the probability of an accident previously 
analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The changes proposed in this amendment request conservatively 
reduce the Unit 1 RCS DE I-131 activity limit at which action needs 
to be taken. The changes do not directly affect plant operation. 
These changes will not result in the installation of any new 
equipment or systems or the modification of any existing equipment 
or systems. No new operating procedures, conditions or 
configurations will be created by this proposed amendment.
    Accordingly, this 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.
    NRC Generic Letter 95-05 allows lowering of the RCS dose 
equivalent iodine as a means for accepting higher projected leakage 
rates provided justification for the RCS DE I-131 activity below 
0.35 microCuries/gram is provided. Four methods for determining the 
fuel rod iodine release rates and spike factors during an accident 
were reviewed. Each of these methods utilized actual industry data, 
including Braidwood Units 1 and 2, for pre-and post-reactor trip RCS 
DE I-131 activities. Each of the methods demonstrated that the 
actual fuel rod iodine release rates are a small fraction of the 
release rate as calculated using the NRC SRP methodology. Although 
these values are a small fraction of that determined by the NRC SRP 
Method, Braidwood is also requesting an increase in the allowable 
primary-to-secondary leak rate during MSLB. By decreasing the TS RCS 
DE I-131 activity limit by a factor of twenty and increasing the 
allowable leak rate by a factor of twenty, the activity released to 
the public would be equal to or less than the activity calculated by 
the SRP method for each of the seventeen reactor trip events 
reviewed at Braidwood. The predicted end-of-cycle 7 leak rate is 
122.3 gpm (Room T/P [temperature and pressure]). The calculated site 
boundary dose due to this leakage is 27.63 Rem. This dose meets the 
requirements of 10 CFR 100 and GDC 19. All design basis and off-site 
dose calculation assumptions remain satisfied. This proposed change 
would not result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.

[[Page 11917]]

    NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of amendment request: September 24, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 4.3.4.2 to 
change the frequency of turbine throttle and governor valve testing 
from monthly to quarterly and incorporate corresponding administrative 
changes. Bases 3/4.3.4 will be changed to update a referenced vendor 
document and incorporate corresponding administrative 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:

    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Bases change is a reference update, which is administrative 
in nature. Additional administrative changes necessitated by a 
change in the presentation of the surveillance requirements are 
proposed. The changes are consistent with Generic Letter 93-05 and 
NUREG-1366. This change reduces the frequency of testing that is 
likely to cause transients or excessive wear of equipment. An 
evaluation of these changes indicates that there will be a benefit 
to plant safety. The evaluation, documented in NUREG-1366, 
considered (1) unavailability of safety equipment due to testing, 
(2) initiation of significant transients due to testing, (3) 
actuation of engineered safety features that unnecessarily cycle 
safety equipment, (4) importance to safety of that system or 
component, (5) failure rate of that system or component, and (6) 
effectiveness of the test in discovering the failure.
    As a result of the decrease in the testing frequencies, the risk 
of testing causing a transient and equipment degradation will be 
decreased, and the reliability of the equipment will not be 
significantly decreased.
    The initial conditions and methodologies used in the accident 
analyses remain unchanged. The proposed changes do not change or 
alter the design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, accident 
analyses results are not impacted. Appropriate testing will continue 
to assure that equipment and systems will be capable of performing 
the intended function. The frequency of testing is not a precursor 
for any analyzed accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes modify allowable intervals between turbine 
throttle and governor valve surveillance tests. The proposed changes 
do not affect the design or operation of any system, structure, or 
component in the plant. The safety functions of the related 
structures, systems, or components are not changed in any manner, 
nor is the reliability of any structure, system, or component 
reduced by the revised surveillance or testing requirements. 
Appropriate testing will continue to assure that the system is 
capable of performing its intended function.
    The changes do not affect the manner by which the facility is 
operated and do not change any facility design feature, structure, 
system, or component. No new or different type of equipment will be 
installed.
    The turbine valve testing surveillances will be changed to 
account for a frequency change from monthly to quarterly for the 
throttle valves and for the governor valves.
    Since there is no change to the facility or operating 
procedures, and the safety functions and reliability of structures, 
systems, or components are not affected, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    All of the proposed Technical Specification changes are 
compatible with plant operating experience and are consistent with 
the guidance provided in Generic Letter 93-05 and NUREG-1366. The 
changes reduce the frequency of testing that increases the risk of 
transients and equipment degradation. There is no impact on safety 
limits or limiting safety system settings. The Bases change is a 
vendor reference update, which is administrative in nature.
    Certain reload designs can be such that power differences 
between the top and bottom of the core are more sensitive to control 
and can develop divergent xenon oscillations when the power 
reduction occurs during the middle of core life. Near the end of 
core life, stabilizing even larger differences in axial power 
distribution becomes more of a problem because of the larger 
temperature coefficient, lower boron concentration and larger 
differential xenon transient. In the Safety Evaluation Report 
related to the Prairie Island Amendment Numbers 86 and 79 in regard 
to the discussion above, the NRC wrote, ``Based on the above, the 
staff has concluded that the margin of safety is reduced when the 
plant is undergoing turbine valve testing.''
    Since this amendment reduces the number of turbine tests while 
still maintaining acceptable equipment reliability, the proposed 
changes result in an increase 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, Monroe County, Michigan

    Date of amendment request: December 15, 1997 (Reference NRC-97-
0115).
    Description of amendment request: The proposed amendment will 
revise License Condition A to delete references to letters dated May 
17, 1985, July 23, 1986, September 15, 1986, September 25, 1987, 
September 15, 1988, and December 22, 1988, and replace them with the 
Enrico Fermi Atomic Power Plant, Unit 1, Safety Analysis Report (F1SAR) 
as the licensing basis.
    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:

    (1) Does the proposed change significantly increase the 
probability or consequences of an accident previously evaluated?
    No, the proposed submittal of the F1SAR as the facility's 
licensing basis document does not significantly increase the 
probability of an accident. The F1SAR is a compilation of previously 
submitted information and other information gathered on the 
condition of the facility. Compilation of current information and 
imposition of the new Fire Protection and Quality Assurance Program 
requirements will not increase the probability of an accident. These 
additional controls would reduce the probability of an event. The 
proposed addition of a hypothetical secondary sodium accident 
scenario identifies one possible previously unidentified potential 
cause of a primary sodium release and/or liquid waste tank release. 
The previous submittal assumed the cause of the primary sodium 
release to be a fire or other catastrophic event. The cause of the 
liquid waste tank rupture was assumed to be an earthquake. 
Recognition of a cause being the reaction of secondary sodium does

[[Page 11918]]

not significantly increase the probability of a primary sodium 
release or liquid waste release. A catastrophic event would still 
need to occur to cause the postulated scenario, so there is no 
discernible increase in the probability of the primary sodium or 
liquid waste accident compared to the existing licensing basis. For 
the reasons discussed above, substituting the F1SAR as the licensing 
basis for Fermi 1 will not significantly increase the probability of 
an accident.
    The proposed submittal of the F1SAR as the Fermi 1 licensing 
basis document will have no impact on the consequences of an 
accident. Consolidating current information on the plant and 
previous submittals does not change the amount of radioactivity at 
the facility or the potential magnitude of any release during an 
accident. Since the potential accident source terms were not updated 
as part of the submittal, the consequences of the accidents 
contained in the F1SAR match the consequences in the previous 
submittal. Though a new postulated hypothetical accident scenario 
was added, the secondary sodium involved in that accident is not 
radioactive, per previous submittals, and so the only potential 
radiological consequences of that scenario occur if the primary 
sodium or liquid waste is released and those consequences have 
already been reviewed in the NRC safety analysis for Amendment No. 9 
to the Fermi 1 license. Therefore, the adoption of the F1SAR as the 
facility's licensing basis will not significantly increase the 
consequences of an accident at Fermi 1.
    (2) Will the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously analyzed?
    No, establishment of the F1SAR as the Fermi 1 licensing basis 
document will not create a new type of accident. The F1SAR is mainly 
a compilation of the previous licensing basis documents, information 
on the facility condition and additional controls. It does not 
involve operating in any new type of mode and so cannot create a new 
or different type of accident. The new hypothetical secondary sodium 
accident contained in the F1SAR is a sodium accident. One of the 
existing licensing basis accidents is the primary sodium accident 
resulting in release of the primary sodium and its activity. The 
hypothetical secondary sodium accident as analyzed may lead to the 
release of the primary sodium or liquid waste and so it is a 
potential precursor of an already identified accident.
    (3) Will the proposed change significantly reduce the margin of 
safety at the facility?
    No, adopting the new F1SAR as the licensing basis document for 
Fermi 1 will not decrease the margin of safety. It will establish an 
up-to-date licensing basis, so future changes can be appropriately 
evaluated against an updated safety analysis report. The F1SAR 
better describes the current condition of the plant. No physical 
changes will be implemented based on the submittal of the F1SAR. 
Some additional administrative requirements will be established in 
the new Quality Assurance program and in the need to keep the F1SAR 
updated biannually. No new types of accidents are discussed in the 
F1SAR--the discussion of the hypothetical secondary sodium event is 
a more detailed discussion of what potentially could happen during a 
catastrophic event leading to a sodium reaction. A total primary 
sodium release was already established as a licensing basis event. 
Because the F1SAR will not, in itself, lead to physical changes, but 
will be the new standard to which future changes are compared, 
establishment of this updated document as the Fermi 1 licensing 
basis will not significantly reduce the margin of safety of the 
facility.

    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, NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Branch Chief: John W. N. Hickey.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: December 19, 1997.
    Description of amendment request: The proposed amendment would 
revise the requirements for the source range neutron flux channels in 
Mode 2 (Below P-6), 3, 4, and 5 to incorporate the guidance provided in 
NUREG-1431, the NRC's Improved Standard Technical Specifications (ISTS) 
with some modifications to address plant-specific design features. This 
change would allow (1) the use of alternate detectors provided the 
required functions are provided, and (2) plant cooldown with inoperable 
detectors provided the shutdown margin accounts for the temperature 
change. This change would also modify the Unit 2 Technical 
Specifications (TS) Table 3.3-1 Channels To Trip and Minimum Channels 
Operable requirements to 0 and 1, respectively. This portion of the 
amendment would make these Unit 2 requirements consistent with the 
current Unit 1 requirements. For both Units 1 and 2, TS Table 4.3-1 
would be modified to include a notation exempting the alternate source 
range detectors from surveillance testing until they are repaired for 
operability.
    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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment would modify the reactor trip system 
instrumentation requirements to permit the use of alternate 
detectors in place of inoperable source range detectors. The 
alternate detectors will be connected to the source range circuits 
to provide the required indications and functions. The alternate 
detectors are not required to be tested to satisfy the surveillance 
requirements until they are connected to the source range circuits 
and required to be operable. The alternate detectors must have the 
accuracy and sensitivity required to adequately monitor changes in 
the core reactivity levels. The alternate detectors will provide 
neutron flux monitoring in place of the source range detectors thus 
assuring core monitoring at a level consistent with the current 
technical specification requirements. Therefore, there is no loss of 
function or need for additional compensatory actions and the 
operators can perform required plant evolutions while relying on the 
alternate detectors.
    Two operable detectors are required when the control rods are 
capable of withdrawal. Rod withdrawal and boron dilution add 
positive reactivity which can significantly affect the reactivity 
condition of the core, therefore, two monitors are required operable 
during startup evolutions. Redundant detectors are required to 
ensure that two source range neutron flux detectors are available to 
detect changes in core reactivity. These changes provide those 
indications and functions consistent with the current technical 
specification requirements where at least two source range detectors 
are operating and capable of providing the required functions. The 
function of the source range detectors is to provide direct neutron 
flux monitoring of the core to detect changes in reactivity which 
would result in a loss of the required shutdown margin.
    One source range or alternate detector is required when the 
control rods are fully inserted and are not capable of withdrawal. 
Plant cooldown is recognized as a positive reactivity addition, 
however, this is accounted for in the shutdown margin calculations. 
The shutdown margin remains essentially unchanged and will be 
available to preclude a criticality event during this evolution. 
Inadvertent control rod withdrawal is not a concern, therefore, one 
source range or alternate detector can adequately monitor the core 
neutron flux. The action statements have been modified to address 
the NUREG-1431 Improved Standard Technical Specification (ISTS) 
requirements along with incorporating the ability to use alternate 
detectors in place of the source range detectors.
    Bases 3/4.3.1 and 3/4.3.2, Protective and Engineered Safety 
Features (ESF) Instrumentation, has been revised to include the 
modifications to the source range detector requirements including 
the use of alternate

[[Page 11919]]

source range detectors. The alternate detectors must provide 
sufficient accuracy and sensitivity to adequately monitor changes in 
core reactivity during Modes 2 (Below P-6), 3, 4, and 5.
    The operability requirements of the source range neutron flux 
instrumentation will continue to be met when using an alternate 
detector in place of a source range neutron flux detector. No 
changes are being incorporated that would act to increase the 
probability of a positive reactivity addition event, therefore, the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The function of the source range detectors is to provide direct 
neutron flux monitoring of the core to detect positive reactivity 
additions which would result in a loss of the required shutdown 
margin. The alternate detectors must provide the accuracy and 
sensitivity required to adequately monitor changes in the core 
reactivity levels during shutdown and startup activities. The 
alternate monitors will be connected to the source range circuits to 
provide the required indications and functions. Therefore, there is 
no loss of function or need for additional compensatory actions and 
plant shutdown and startup activities can be continued while relying 
on the alternate detectors.
    Control rod withdrawal is a method capable of providing rapid 
positive reactivity addition with boron dilution being a much slower 
positive reactivity addition method. With the control rods capable 
of withdrawal, a rod withdrawal event could rapidly initiate core 
criticality so redundant source range detectors are required 
operable. This ensures adequate monitoring capability is available 
to alert the operators of a rapid increase in the core reactivity 
condition. The maximum reactivity addition due to the boron dilution 
is slow enough to allow the operator to determine the cause and take 
corrective action before the shutdown margin is lost. These changes 
will not affect the operability or reliability of the source range 
instrumentation to provide the required indications and functions. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change will continue to ensure the required source 
range instrumentation functions are available during shutdown and 
startup conditions. This change will not reduce the reliability of 
the source range detectors to monitor the core reactivity condition 
and provide the appropriate indications or affect the required 
shutdown margin. Plant operation will continue to be maintained 
within the shutdown margin requirements of [Technical] Specification 
3.1.1.1 and 3.1.1.2. The required indications and functions are 
still maintained in accordance with current technical specification 
requirements and the shutdown margin is unaffected, therefore, the 
proposed change will not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of amendment request: January 29, 1998.
    Description of amendment request: The proposed amendment would 
revise the Beaver Valley Power Station, Unit No. 2, Updated Final 
Safety Analysis Report (UFSAR) calculated doses to address a non-
conversative assumption regarding control room emergency pressurization 
fan flow during the Locked Rotor accident and include new X/Q values in 
calculating the Exclusion Area Boundary (EAB) and Low Population Zone 
(LPZ) doses.
    This change is not the result of hardware changes to the plant or a 
change in operating practices. It reflects corrected analysis results 
only and allows correction of the licensing basis to reflect 
conservative assumptions used in the revised dose analysis for a Locked 
Rotor event.
    The proposed amendment would also revise USFAR Tables 15.0-13, 
15.6-15 and 15.6-16 to modify calculation parameters and UFSAR Section 
15.6.5.5 to include editorial changes to ensure that descriptions of 
the Small Break Loss of Coolant Accident (SBLOCA) radiological 
consequences are clear. The following items in the UFSAR description of 
the SBLOCA radiological consequences analysis were changed: (1) a new 
lower minimum control room emergency pressurization fan flow rate and 
(2) a new lower minimum air bottle discharge rate.
    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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

[Locked Rotor Accident]

    The proposed amendment would revise the calculated control room 
doses for a Locked Rotor accident to address a non-conservative 
assumption for the fan pressurization system flow rate. The proposed 
amendment does not affect the capability of the control room 
habitability system to maintain control room dose within the limits 
of General Design Criterion (GDC) 19 in Appendix A of the Code of 
Federal Regulations Title 10 Part 50. The control room habitability 
system is an accident mitigation system and will continue to operate 
as designed. The system has no accident prevention function nor does 
it interact with systems that have such a function. The proposed 
change does not alter plant systems, structures or components.
    The proposed amendment would also revise calculated offsite 
doses resulting from a locked rotor accident. This change in doses 
is not due to physical plant changes, but results mainly from use of 
more conservative assumptions used in calculating doses.
    The proposed change does not affect the manner in which the 
plant is operated. The physical plant equipment and operating 
practices are not changed; therefore, the probability of an accident 
previously evaluated remains unchanged.
    The performance requirements of the plant systems which are 
required to minimize the radiological consequences of a Locked Rotor 
accident remain unchanged. The proposed change slightly increases 
calculated control room doses due to an analysis input change for 
filtration fan flow rate. This slight increase remains below the 
limits required by GDC 19. The proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated since adequate control room radiation protection continues 
to be provided to ensure actions can be taken to operate the plant 
safely under accident conditions. The radiological consequences to 
the environment from a Locked Rotor accident remain unchanged since 
the performance of plant systems remains unchanged. Although 
slightly increased, revised calculated offsite doses remain less 
than 10 CFR 100 limits.
[SBLOCA]

    The proposed amendment would revise the control room dose 
analysis parameters for a Small Break Loss of Coolant Accident 
(SBLOCA) to include more conservative assumptions for the 
pressurization system flow rate. The proposed amendment does not 
affect the capability of the control room habitability system to 
maintain control room dose within the limits of General Design 
Criterion (GDC) 19 in Appendix A of the Code of Federal Regulations 
Title 10 Part 50. The control room habitability system is an 
accident mitigation system and will continue to operate as designed. 
The system has no accident prevention function nor does it interact 
with systems that have such a function. The proposed change does not 
alter plant systems, structures or components.
    The proposed change does not affect the manner in which the 
plant is operated. The physical plant equipment and operating

[[Page 11920]]

practices are not changed; therefore, the probability of an accident 
previously evaluated remains unchanged.
    The performance requirements of the plant systems which are 
required to minimize the radiological consequences of a SBLOCA 
remain unchanged. The proposed change slightly decreases calculated 
control room doses due to analysis input changes. Calculated doses 
remain below the limits required by GDC 19.
    Based on the above discussion, it is concluded that th[e] 
proposed change[s] [do] not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?

[Locked Rotor Accident]

    The proposed change does not alter the method of operating the 
plant nor does it pose additional challenges to the design or 
function of the control room habitability system. The control room 
habitability system will continue to operate as designed. The 
control room habitability system will continue to maintain the 
control room dose consequences within the limits specified in GDC 
19. Adequate control room radiation protection will continue to be 
provided to ensure actions can be taken to operate the plants safely 
under accident conditions. The proposed change to the control room 
dose is only the result of a change in analysis input parameters. 
Plant performance has not been modified in any way which affects 
doses to the public.

[SBLOCA]

    The proposed change does not alter the method of operating the 
plant nor does it pose additional challenges to the design or 
function of the control room habitability system. The control room 
habitability system will continue to operate as designed. The 
control room habitability system will continue to maintain the 
control room dose consequences within the limits specified in GDC 
19. Adequate control room radiation protection will continue to be 
provided to ensure actions can be taken to operate the plants safely 
under accident conditions. The proposed change to the control room 
dose is only a result of an analysis being revised. Plant 
performance has not been modified in any way which affects doses to 
the public.
    Therefore, the proposed change[s] [do] not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Although no new types of accidents are 
created, the analysis represents a new methodology different than 
any evaluated previously by the NRC.
    3. Does the change involve a significant reduction in a margin 
of safety?

[Locked Rotor Accident]

    The slight increase in calculated control room dose as a result 
of assuming increased fan flow does not result in exceeding the 
limits prescribed in GDC 19. Calculated doses to the public are 
slightly increased, but not as a result of physical changes. The 
proposed change will not result in any additional challenges to 
plant equipment including the fuel and reactor coolant system 
pressure boundary since adequate control room radiation protection 
will continue to be provided. The control room habitability system 
will continue to provide adequate radiation protection to ensure 
actions can be taken to operate the plant safely under accident 
conditions. The offsite doses increase slightly; however, the 
calculated dose results remain less than 10 CFR 100 limits. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

[SBLOCA]

    The slight decrease in calculated control room dose as a result 
of the revised analysis does not result in exceeding the limits 
prescribed in GDC 19. The proposed change will not result in any 
additional challenges to plant equipment including the fuel and 
reactor coolant system pressure boundary since adequate control room 
radiation protection will continue to be provided. The control room 
habitability system will continue to provide adequate radiation 
protection to ensure actions can be taken to operate the plant 
safely under accident conditions. [Therefore, the NRC staff 
concludes that the revision to the SBLOCA analysis does not involve 
a reduction in a margin of safety.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: February 5, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) to update the terminology and 
references to 10 CFR 50.55a(f) and (g) consistent with the 1989 edition 
of Section XI of the American Society of Mechanical Engineer Boiler and 
Pressure Vessel Code (ASME Code). These changes, in effect, provide for 
consistency between (1) the NMP2 TS, (2) the second 10-year interval of 
the Inservice Inspections (ISI) and Inservice Testing (IST) Program 
Plans for NMP2, and (3) the requirement of 10 CFR 50.55a that the ISI/
IST activities conducted during successive 10-year intervals comply 
with the requirements in the latest edition and addenda of Section XI 
of the ASME Code that was in effect 12 months before the start of the 
10-year interval.
    Specifically, TS 4.0.5 would be changed to reference 10 CFR 
50.55a(f) for the second 10-year IST Program and 10 CFR 50.55a(g) for 
the second 10-year ISI Program. The proposed changes to TS Table 
4.3.7.5-1 and TS 4.4.3.2.2 would replace the references to ASME Section 
XI with references to criteria in the IST Program. The changes to TS 
3.4.9.1 and 3.4.9.2 would add the phrase ``system leakage'' to notes 
that identify testing conditions when the shutdown cooling mode loop 
may be removed from service. Changes to TS 4.8.1.1.2.h.2 would correct 
a typographical error for which a reference to ASME Code Section II 
should refer to Section XI. Appropriate changes would be made to the TS 
index. Editoral changes to several other TS (i.e., TS 3/4.4.6.1, TS 
Figure 3.4.6.1-1, TS 3/4.10.7, TS Bases 3/4.4.6, TS Bases 3/4.10.7, and 
TS Table 5.7.1-1) would make references to ``hydrostatic testing'' and 
``leak testing'' conform to the terminology to be used in the second 
10-year ISI/IST Programs.
    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 operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes to the TS will ensure that TS reflect the correct 
10CFR references and the terminology of the second NMP2 10-year ISI/
IST program. The proposed revisions replace references to ASME 
Section XI with references to criteria in the Inservice Testing 
Program. The performance of system leakage testing is added to notes 
that identify conditions when the shutdown cooling mode loop may be 
removed from service. The other changes are editorial changes only 
to ensure that TS reflect the second 10-year ISI/IST program. One of 
the changes corrects a typographical error. These proposed changes 
do not affect the inspections or tests performed under the ISI/IST 
Program and will not result in any changes to the plant. None of the 
precursors of previously evaluated accidents are affected and 
therefore, the probability of an accident previously evaluated is 
not increased.
    The changes will not affect the safety function of any equipment 
covered by the ISI/IST program. Therefore, these changes will not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of

[[Page 11921]]

a new or different kind of accident from any accident previously 
evaluated.
    The changes to the TS will ensure that TS reflect the correct 
10CFR references and the terminology of the second NMP2 10-year ISI/
IST program. One of the changes corrects a typographical error. No 
physical modification of the plant is involved and no changes to the 
methods in which plant systems are operated are required. These 
changes do no affect the inspections or tests performed under the 
ISI/IST Program. The changes do not introduce any new failure modes 
or conditions that may create a new or different accident. 
Therefore, the changes do not by themselves create the possibility 
of a new or different kind of accident [from any accident] 
previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The changes to the TS will ensure that TS reflect the correct 
10CFR references and the terminology of the second NMP2 10-year ISI/
IST program. One of the changes corrects a typographical error. No 
physical modification of the plant is involved and no changes to the 
methods in which plant systems are operated are required. The 
changes do not adversely affect any physical barrier to the release 
of radiation to plant personnel or to the public. These changes do 
not affect the inspections or tests performed under the ISI/IST 
Program. Therefore, these changes do not involve a 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 request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station (LGS), Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: January 27, 1998.
    Description of amendment request: The proposed changes to the LGS, 
Units 1 and 2 Technical Specifications (TS) will revise the TS Table 
3.6.3-1, ``Part A--Primary Containment Isolation Valves,'' by removing 
the numerical maximum stroke time for penetration 210, ``HPCI [High 
Pressure Coolant Injection] Turbine Exhaust,'' and adding a notation 
that the isolation time is not required.
    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 Technical Specifications changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Changes to Technical Specifications regarding the removal of the 
High Pressure Coolant Injection (HPCI) Turbine Exhaust Valve maximum 
stroke times do not change the frequency or consequences of any 
accident previously evaluated.
    The proposed changes do not change the function of the HPCI 
system nor any safety function of the valve as described in the SAR 
[Safety Analysis Report]. The isolation stroke times are not limits 
upon important process variables that are found to be necessary to 
reasonably protect the integrity of certain of the physical barriers 
that guard against the uncontrolled release of radioactivity. The 
stroke times do not detect or indicate an abnormal degradation of 
the reactor coolant pressure boundary. The stroke times are not a 
process variable, design feature, or operating restriction that is 
an initial condition of a design basis accident or transient 
analysis that either assumes the failure of or presents a challenge 
to the integrity of a fission product barrier. The stroke times are 
not part of a component that is part of the primary success path and 
which functions or actuates to mitigate a design basis accident or 
transient that either assumes the failure of or presents a challenge 
to the integrity of a fission product barrier. The stroke times are 
not a structure, system, or component which operating experience or 
probabilistic risk assessment has shown to be significant to public 
health and safety.
    Therefore, the changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed Technical Specifications changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed Technical Specifications changes regarding the 
removal of the High Pressure Coolant Injection (HPCI) Turbine 
Exhaust Valve maximum stroke times do not affect the probability of 
a malfunction of equipment important to safety. Safety related HPCI 
system operation occurs with the subject valve passively open. This 
valve would only be manually closed under events where there was a 
need to isolate the HPCI system from the suppression pool. The 
manual closing of the valve may occur under these events and is 
controlled by station procedures. Given that these procedurally 
mandated valve isolations are all via remote manual means, valve 
isolation time is not a critical parameter requiring specific 
acceptance criteria.
    The Inservice Testing (IST) Program will still maintain an IST 
program basis maximum stroke time for HV-055-1(2)F072 to establish 
action and alert levels for valve performance monitoring. These 
performance based values, in conjunction with diagnostic test 
criteria, are used for motor operated valve material condition 
monitoring and trending. Therefore, eliminating the subject maximum 
isolation time requirement from TS will not increase the probability 
of malfunction of the valve since the principal means of monitoring 
valve performance remains unchanged.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed Technical Specifications changes do not involve 
a significant reduction in a margin of safety.
    There is no defined margin of safety for remote manual valve 
isolation times discussed in Technical Specification Bases. In 
addition, the valve maximum stroke time will be retained in the IST 
program.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: John F. Stolz.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of amendment requests: June 30, 1997.
    Description of amendment requests: The licensee proposes to delete 
SONGS Unit 2 License Condition 2.C.(19)b, ``Shift Manning,'' and revise 
SONGS Units 2 and 3 Technical Specifications (TS) 3.3.1, ``Reactor 
Protective Instrumentation (RPS)-Operating,'' TS 3.3.2, ``Reactor 
Protective Instrumentation (RPS)-Shutdown,'' TS 3.3.5, ``Engineered 
Safety Features Actuation System (ESFAS) Instrumentation,'' TS 3.3.10, 
``Fuel Handling Isolation Signal (FHIS),'' TS 3.3.11, ``Post Accident 
Monitoring Instrumentation,'' TS 3.4.7, ``RCS Loops--Mode 5, Loops 
Filled,'' TS 3.4.12.1, ``Low Temperature Overpressure Protection (LTOP) 
System,'' TS 3.7.5, ``Auxiliary Feedwater (AFW) System,'' TS Section 
5.5.2.10, ``Inservice Testing Program,'' and TS Section 5.5.2.11, 
``Steam

[[Page 11922]]

Generator (SG) Tube Surveillance Program.'' The proposed changes are 
required to either: reinstate provisions of the SONGS Units 2 and 3 TS, 
revised as part of NRC Amendment Numbers 127 and 116, make corrections 
to the TS, or remove information inadvertently added that is not 
applicable.
    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.
    Proposed Technical Specification Change Number NPF-10/15-475 
(PCN-475) addresses modifications to the Technical Specifications 
for San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 
approved by NRC Amendment Nos. 127 and 116. NRC Amendment Numbers 
127 and 116 approved changes to adopt the recommendations of NUREG-
1432, ``Standard Technical Specifications Combustion Engineering 
Plants,'' requested through Proposed Technical Specification Change 
Number NPF-10/15-299 (PCN-299). The proposed changes were identified 
during drafting of the procedure changes required to implement NRC 
Amendment Numbers 127 and 116, and during the self-assessment 
performed by Southern California Edison (SCE).
    The proposed change is required to either: reinstate provisions 
of the SONGS Units 2 and 3 Technical Specifications, revised as part 
of NRC Amendment Numbers 127 and 116, for SONGS Units 2 and 3, make 
corrections to the Technical Specifications, or remove information 
inadvertently added that is not applicable.
    Proposed Change 1 would delete License Condition 2.C.(19)b for 
SONGS Unit 2 only. Presently, overtime restrictions are specified in 
both the license condition and the Topical Report. Through NRC 
Amendment Numbers 127 and 116, the shift manning requirements were 
modified and subsequently moved to the Section 5.5.2.e, with details 
moved to the Topical Report.
    In addition, in the NRC's Safety Evaluation Report related to 
the ``Issuance of Amendment for San Onofre Nuclear Generating 
Station, Unit No. 2 (TAC No. M86191) and Unit No. 3 (TAC No. 
M86192),'' dated February 9, 1996, it is stated that the staff has 
determined on a generic basis, that specific overtime limits need 
not be specified in technical specifications, as they are not 
required by 10 CFR 50.36 (c)(5). The staff also concluded that 
control of this matter through administrative procedures provides 
reasonable assurance that personnel overtime would not jeopardize 
safe plant operation and that specific overtime limits and 
associated procedures could be described in the UFSAR, or other 
licensee controlled documents incorporated in the UFSAR by reference 
for which further changes can be made pursuant to 10 CFR 50.59.
    Retaining a separate license condition provides no function, is 
inconsistent with the Topical Report, and therefore, should be 
deleted. There can be no increase in the probability or consequences 
of any accident previously evaluated as a result of this change, as 
the change does not revise or reduce commitments, it is solely for 
clarity.
    Proposed change 2 would revise TS 3.3.1, ``Reactor Protective 
Instrumentation (RPS)--Operating,'' to delete the exception of the 
power range neutron flux channels from Surveillance Requirement (SR) 
3.3.1.7. TS 3.3.1 requires that four RPS trip and operating bypass 
removal channels for each function covered by this specification be 
operable in the applicable Modes. SR 3.3.1.7 requires that a channel 
functional test be performed on each RPS channel, except the power 
range neutron flux channels. Therefore, the proposed change would 
delete the exception to SR 3.3.1.7 for the power range neutron flux 
channels. Under the former Technical Specifications, the power range 
neutron flux channels were not exempt from the channel functional 
test.
    Proposed change 3 would revise SR 3.3.2.5 of TS 3.3.2, ``Reactor 
Protective Instrumentation (RPS)-Shutdown.'' SR 3.3.2.5 requires 
that the RPS response time be verified within limits every 24 months 
on a staggered test basis. SR 3.3.1.13 of TS 3.3.1 also requires 
that response time tests be performed every 24 months on a staggered 
test basis. However, neutron detectors presently are excluded from 
response time testing in Modes 1 and 2. Therefore, the proposed 
change will add a note to SR 3.3.2.5 to allow exclusion of neutron 
detectors from response time testing. Under the former Technical 
Specifications, the neutron detectors were exempt from response time 
testing.
    Proposed change 4 would revise SR 3.3.5.4. SR 3.3.5.4 requires 
that a channel calibration of the Recirculation Actuation Signal 
(RAS), including the bypass removal function, be performed. However, 
a bypass removal function is not part of the RAS design. A change is 
required therefore, to delete the bypass removal function, as it is 
not a part of the RAS function. Because the RAS function does not 
utilize the bypass removal function, eliminating the words from the 
SR cannot increase the probability or consequences of any accident 
previously evaluated as a result of this change.
    Proposed change 5 would revise Technical Specification (TS) 
3.3.10, ``Fuel Handling Isolation Signal (FHIS).'' Specifically, the 
proposed change would revise the allowable value specified in SR 
3.3.10.2 for the required FHIS monitor, from ``less than or equal to 
6E4 cpm above background,'' to ``Sufficiently high to prevent 
spurious alarms/trips, yet sufficiently low to assure an alarm/trip 
should an inadvertent release occur.''
    The 6E4 cpm setpoint does not provide adequate margin above and 
beyond background during a normal refueling outage. Thus, the 
proposed setpoint, which can be set greater than the highest ambient 
background level, but remains well below the calculated monitor 
response to a fuel handling accident, would provide that margin, and 
was previously specified in the former Technical Specifications.
    The proposed change would permit relocation of the allowable 
value for the monitors from the Technical Specifications to the 
administrative control procedures. This change is consistent with 
the existing Containment Airborne Radiation Monitor Specification. 
This change will not prevent the radiation monitors from performing 
their intended function following a design basis accident.
    The consequences of a Fuel Handling Accident inside the FHB have 
been evaluated, assuming no FHB isolation. The results of the 
calculation indicated off-site, and control room doses with control 
room isolation within three minutes, are well within the limits 
established by the NRC guidelines.
    Compliance with this statement would provide suitable 
confirmation that the monitors will be capable of performing their 
intended function, and is further justified by the fact that no 
credit was given to the monitors in the radiological dose analysis.
    This change will not involve a significant increase in the 
probability of any accident previously evaluated because the 
setpoint is not an accident initiator. The consequences of an 
accident would not be increased either as the administrative value 
would be set sufficiently low to assure an alarm/trip should an 
inadvertent release occur. The actual values would be 
administratively controlled by quality-affecting procedures (i.e., 
changes to procedures will be evaluated under 10 CFR 50.59).
    In addition, a typographical error in SR 3.3.10.3 would be 
corrected. The SR Note would be revised to refer to ``initiation 
relay,'' not ``ignition relay.'' This change will not involve a 
significant increase in the probability of any accident previously 
evaluated because it corrects a typographical error only.
    Proposed change 6 would revise Function 6 of Table 3.3.11-1. 
Currently, Function 6 refers to Containment Sump Water Level (wide 
range). However, Function 6 is the combined function of the wide 
range emergency sump level transmitters, and the containment area 
level transmitters. Therefore, the description of the combination 
should not be the description of the function of the single 
transmitter. There can be no increase in the probability or 
consequences of any accident previously evaluated as a result of 
this change, as the change does not revise or reduce commitments, it 
is solely for clarity.
    Proposed change 7 would revise Surveillance Requirement 3.4.7.2 
of TS 3.4.7. The change would remove an inconsistency between what 
is specified in the Limiting Condition for Operation (LCO), and what 
is required to be verified by the SR. The proposed change 
conservatively removes the inconsistency by revising SR 3.4.7.2 to 
specify that the required steam generator secondary side water level 
be verified greater than 50% (wide range). This change is for 
clarity only, and is consistent with existing station procedures and 
operation of the facility.
    Proposed change 8 would revise TS 3.4.12.1, ``Low Temperature 
Overpressure

[[Page 11923]]

Protection (LTOP) System.'' Specifically, the Applicability would be 
revised to clarify the Mode 6 applicability. The Applicability 
should read ``Mode 6 when the head is on the reactor vessel and the 
RCS is not vented.'' This change is intended to clarify the 
Applicability of TS 3.4.12.1 in Mode 6, and also reflects the 
previous requirements of former TS 3/4.4.8.3.1, ``Overpressure 
Protection Systems RCS Temperature less than or equal to 256'F.'' 
This change is editorial only and there can be no increase in the 
probability or consequences of any accident previously evaluated as 
a result of this change.
    Proposed change 9 would revise SR 3.7.5.3 and SR 3.7.5.4 of TS 
3.7.5, ``Auxiliary Feedwater (AFW) System.'' Presently, SR 3.7.5.3 
requires that AFW automatic valves actuate to their correct position 
on an actual or simulated signal when in Mode 1, 2, or 3 (except 
valves HV-8200 and HV-8201) and SR 3.7.5.4 requires that each AFW 
pump starts automatically on an actual or simulated signal when in 
Mode 1, 2, or 3. The Bases, however, for these SRs makes it clear 
that the tests are a refueling surveillance which should be 
performed in Mode 5. The proposed change will delete the reference 
to Modes 1, 2, and 3 from both SR 3.7.5.3 and 3.7.5.4.
    The intent of the wording for the SR is to perform the test in 
Mode 5 in order to demonstrate the operability of the system in 
Modes 1, 2, and 3. This change would also be consistent with the 
former SRs which previously specified that the surveillances were 
required to be performed at least once per refueling interval during 
shutdown. Therefore, there can be no increase in the probability or 
consequences of any accident previously evaluated as a result of 
this change.
    Proposed change 10 would revise Section 5.5.2.10, ``Inservice 
Testing Program.'' The change will clarify that this section applies 
not only to the Inservice Testing Program, but includes the 
Inservice Inspection Program as well. This change is editorial in 
that it correctly identifies the intent of this section. As this is 
an editorial change only, there can be no increase in the 
probability or consequences of any accident previously evaluated as 
a result of this change.
    Proposed change 11 would revise Section 5.5.2.11 to correct 
typographical errors. A table is provided that identifies 
supplemental sampling requirements for steam generator tube 
inspections. However, the table is numbered incorrectly. The 
proposed change would correct the table number.
    In addition, under the table heading ``Action Required'' for 
both the first ``1st Sample Inspection'' and ``2nd Sample 
Inspection,'' for result C-3, notification is to be made to the NRC, 
and an incorrect reference to 10 CFR 50.72 is made. The proper 
notification is pursuant to 10 CFR 50.73. The proposed change would 
correct this reference. Also under the ``Action Required'' heading 
for the ``1st Sample Inspection'' for Result C2, is a typographical 
error. It is currently written, ``Plug defective tubes and inspect 
an additional 25 tubes in this SG.'' However, the statement should 
read, ``Plug defective tubes and inspect an additional 2S tubes in 
this SG.'' The proposed requirement is consistent with the 
requirement of the former TS 3/4.4.4, ``Steam Generators.''
    Operation of the facility would remain unchanged as a result of 
the proposed changes as the changes correct typographical errors. 
Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes would either: reinstate provisions of the 
former SONGS Units 2 and 3 Technical Specifications, make 
corrections to the Technical Specifications, or remove information 
inadvertently added that is not applicable to SONGS Units 2 and 3.
    Proposed change 1 deletes the SONGS Unit 2 license condition 
regarding shift manning requirements as it conflicts with the 
requirements contained in the revised Technical Specifications and 
the Topical Report. Operation of the facility would remain unchanged 
as a result of the proposed changes and could not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Proposed change 2 would revise TS 3.3.1, ``Reactor Protective 
Instrumentation (RPS)-Operating,'' to delete the exception of the 
power range neutron flux channels from Surveillance Requirement (SR) 
3.3.1.7. SR 3.3.1.7 requires that a channel functional test be 
performed on each RPS channel, except the power range neutron flux 
channels. Therefore, the proposed change would delete the exception 
to SR 3.3.1.7 for the power range neutron flux channels. This change 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Proposed change 3 would revise SR 3.3.2.5 of TS 3.3.2, ``Reactor 
Protective Instrumentation (RPS)-Shutdown.'' SR 3.3.2.5 requires 
that the RPS response time be verified within limits every 24 months 
on a staggered test basis. SR 3.3.1.13 of TS 3.3.1 also requires 
that response time tests be performed every 24 months on a staggered 
test basis. However, neutron detectors presently are excluded from 
response time testing in Modes 1 and 2. Therefore, the proposed 
change will add a note to SR 3.3.2.5 to allow exclusion of neutron 
detectors from response time testing. The proposed change will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Proposed change 4 would revise Surveillance Requirement (SR) 
3.3.5.4. A change is required to delete the bypass removal function, 
as it is not a part of the RAS function. Because the RAS function 
does not utilize the bypass removal function, eliminating the words 
from the SR cannot create the possibility of a new or different kind 
of accident from any previously evaluated.
    Proposed change 5 revises the FHIS the monitor allowable value. 
The value would be controlled by administrative procedures. This 
change would not alter the design and operational interface between 
the FHIS and existing plant equipment. As such, the monitors would 
continue to operate and perform their intended safety function to 
isolate the FHB following a design basis accident as before. In 
addition, the Note to SR 3.3.10.3 would be corrected to read ``* * * 
verification of the proper operation of each initiation relay.'' 
Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Proposed change 6 revises the name of Function 6 of Table 
3.3.11-1. Currently, Function 6 refers to Containment Sump Water 
Level (wide range), and is more correctly specified as the 
Containment Water Level (wide range). The proposed change cannot 
create the possibility of a new or different kind of accident from 
any accident previously evaluated as the change only revises the 
name of an instrument and is solely for clarity.
    Proposed change 7 would remove an inconsistency between what is 
specified in the LCO, and what is required to be verified by the SR. 
The proposed change conservatively removes the inconsistency by 
revising SR 3.4.7.2 to specify that the required steam generator 
secondary side water level be verified greater than 50% (wide 
range). This change is for clarity only, is consistent with existing 
station procedures, and consistent with operation of the facility. 
The proposed change cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Proposed change 8 would revise TS 3.4.12.1, ``Low Temperature 
Overpressure Protection (LTOP) System.'' Specifically, the 
Applicability would be revised to clarify the Mode 6 applicability. 
The Applicability should read ``Mode 6 when the head is on the 
reactor vessel and the RCS is not vented.'' This change is intended 
to clarify the Applicability of TS 3.4.12.1 in Mode 6, and also 
reflects the previous requirements of former TS 3/4.4.8.3.1, 
``Overpressure Protection Systems RCS Temperature less than or equal 
to 256 deg.F.'' This change is editorial only and cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Proposed change 9 would revise SR 3.7.5.3 and SR 3.7.5.4 of TS 
3.7.5, ``Auxiliary Feedwater (AFW) System,'' to delete the 
requirements that the SRs be performed in Mode 1, 2, or 3. The 
intent of the wording for the SR is to perform the test in Mode 5 in 
order to demonstrate the operability of the system in Modes 1, 2, 
and 3. This change would also be consistent with the former SRs 
which previously specified that the surveillances were required to 
be performed at least once per refueling interval during shutdown. 
Therefore, the proposed change cannot create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Proposed change 10 would revise Section 5.5.2.10, ``Inservice 
Testing Program.'' The change will clarify that this section applies

[[Page 11924]]

not only to the Inservice Testing Program, but includes the 
Inservice Inspection Program as well. This change is editorial in 
that it correctly identifies the intent of this section. As this is 
an editorial change only, and cannot create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Proposed change 11 would revise Section 5.5.2.11 to correct 
typographical errors. A table is provided that identifies 
supplemental sampling requirements for steam generator tube 
inspections. Operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes will either: reinstate provisions of the 
SONGS Units 2 and 3 Technical Specifications, make corrections to 
the Technical Specifications, or remove information inadvertently 
added that is not applicable to SONGS Units 2 and 3. Operation of 
the facility would remain unchanged as a result of the proposed 
change. Therefore, the proposed change will not involve a 
significant reduction in a margin of safety.
    Proposed change 1 deletes the SONGS Unit 2 license condition 
regarding shift manning requirements as it conflicts with the 
requirements contained in the revised Technical Specifications and 
the Topical Report. The NRC staff has concluded that control of 
overtime restrictions through administrative procedures provides 
reasonable assurance that personnel overtime would not jeopardize 
safe plant operation and that specific overtime limits and 
associated procedures could be described in the UFSAR, or other 
licensee controlled documents incorporated in the UFSAR by reference 
for which further changes can be made pursuant to 10 CFR 50.59. 
Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    Proposed change 2 would revise TS 3.3.1, ``Reactor Protective 
Instrumentation (RPS)--Operating,'' to delete the exception of the 
power range neutron flux channels from Surveillance Requirement (SR) 
3.3.1.7. SR 3.3.1.7 requires that a channel functional test be 
performed on each RPS channel, except the power range neutron flux 
channels. Therefore, the proposed change would delete the exception 
to SR 3.3.1.7 for the power range neutron flux channels. This change 
will not involve a significant reduction in a margin of safety.
    Proposed change 3 would revise SR 3.3.2.5 of TS 3.3.2, ``Reactor 
Protective Instrumentation (RPS)-Shutdown.'' SR 3.3.2.5 requires 
that the RPS response time be verified within limits every 24 months 
on a staggered test basis. SR 3.3.1.13 of TS 3.3.1 also requires 
that response time tests be performed every 24 months on a staggered 
test basis. However, neutron detectors presently are excluded from 
response time testing in Modes 1 and 2. Therefore, the proposed 
change will add a note to SR 3.3.2.5 to allow exclusion of neutron 
detectors from response time testing. The proposed change will not 
involve a significant reduction in a margin of safety.
    Proposed change 4 would delete the bypass removal function, as 
it is not a part of the RAS function. Because the RAS function does 
not utilize the bypass removal function, eliminating the words from 
the SR cannot involve a significant reduction in a margin of safety.
    Proposed change 5 would revise the FHIS monitor allowable values 
and would not alter the existing margin of safety. The change would 
only relinquish control of the allowable values from the TSs to 
quality-affecting (changes will require a 10 CFR 50.59 evaluation) 
procedures. In addition, the proposed change would correct a 
typographical error in the Note to SR 3.3.10.3. Therefore, operation 
of the facility will not involve a significant reduction in a margin 
of safety.
    Proposed change 6 revises the name of Function 6 of Table 
3.3.11-1. Currently, Function 6 refers to Containment Sump Water 
Level (wide range), and is more correctly specified as the 
Containment Water Level (wide range). The proposed change cannot 
involve a significant reduction in a margin of safety.
    Proposed change 7 would remove an inconsistency between what is 
specified in the LCO, and what is required to be verified by the SR. 
The proposed change conservatively removes the inconsistency by 
revising SR 3.4.7.2 to specify that the required steam generator 
secondary side water level be verified greater than 50% (wide 
range). This change is consistent with existing station procedures, 
and consistent with operation of the facility. The proposed change 
cannot involve a significant reduction in a margin of safety.
    Proposed change 8 would revise TS 3.4.12.1, ``Low Temperature 
Overpressure Protection (LTOP) System.'' Specifically, the 
Applicability would be revised to clarify the Mode 6 applicability. 
The Applicability should read ``Mode 6 when the head is on the 
reactor vessel and the RCS is not vented.'' This change is intended 
to clarify the Applicability of TS 3.4.12.1 in Mode 6, and also 
reflects the previous requirements of former TS 3/4.4.8.3.1, 
``Overpressure Protection Systems RCS Temperature less than or equal 
to 256 deg.F.''
    Proposed change 9 would revise SR 3.7.5.3 and SR 3.7.5.4 of TS 
3.7.5, ``Auxiliary Feedwater (AFW) System,'' to delete the 
requirements that the SRs be performed in Mode 1, 2, or 3. The 
intent of the wording for the SR is to perform the test in Mode 5 in 
order to demonstrate the operability of the system in Modes 1, 2, 
and 3. Therefore, the proposed change cannot involve a significant 
reduction in a margin of safety.
    Proposed change 10 would revise Section 5.5.2.10, ``Inservice 
Testing Program.'' The change will clarify that this section applies 
not only to the Inservice Testing Program, but includes the 
Inservice Inspection Program as well. This change is editorial in 
that it correctly identifies the intent of this section. This is an 
editorial change only.
    Proposed change 11 would revise Section 5.5.2.11 to correct 
typographical errors. Operation of the facility would remain 
unchanged as a result of the proposed changes and could not create 
the possibility of a new or different kind of accident from any 
previously evaluated.

    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: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T.E. Oubre, Esquire, Southern California 
Edison Company, P.O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: February 3, 1998.
    Description of amendment request: The proposed changes will replace 
the augmented inspection requirements for the Reactor Coolant Pump 
flywheels specified by Regulatory Guide 1.14, ``Reactor Coolant Pump 
Integrity,'' Revision 1, dated August 1975, with those established by 
WCAP-14535A, ``Topical Report on Reactor Coolant Pump Flywheel 
Inspection Elimination,'' dated November 1996, and will eliminate the 
inspection requirements for the flow straighteners.
    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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed changes 
for the North Anna Units 1 and 2 and determined that a significant 
hazards consideration is not involved.
    (a) The elimination of the inspection requirements for the flow 
straighteners, and the reduction of the inspection requirements for 
the reactor coolant pump flywheels as granted by the NRC and 
supported by WCAP-14535A do not significantly increase the 
probability of an accident previously evaluated in the safety 
analysis report.
    The surveillance frequency changes for the reactor coolant pump 
flywheels are based upon the technical basis of the Westinghouse 
Energy Systems Topical Report WCAP-14535A. The results of WCAP-
14535A have been reviewed, evaluated, and accepted for referencing 
in license applications by the NRC in their letter entitled 
``Acceptance for Referencing of Topical Report WCAP-14535, Topical 
Report on Reactor Coolant Pump

[[Page 11925]]

Flywheel Inspection Elimination'' dated September 12, 1996.
    The proposed surveillance (inspection) requirements only reduce 
the inspection frequency for the reactor coolant pump flywheels and 
eliminate the inspection requirements for the flow [straighteners]. 
There is no change in the method of plant operation or system 
design. Therefore, the proposed changes do not increase the 
probability of occurrence or the consequences of any previously 
analyzed accident.
    (b) The proposed changes for the elimination of the inspection 
requirements for the flow straighteners, and for the reduction in 
inspection requirements for the reactor coolant pump flywheels as 
granted by the NRC and supported by WCAP-14535A do not create the 
possibility of an accident or malfunction of a different type than 
any evaluated previously in the safety analysis report.
    The proposed surveillance (inspection) requirements only reduce 
the inspection frequency for the reactor coolant pump flywheels and 
eliminate the inspection requirements for the flow [straighteners] 
in Unit 1. There is no change in the method of plant operation or 
system design. Therefore, there are no new or different kinds of 
accident or malfunction from any accidents previously evaluated.
    (c) The proposed changes for the elimination of the inspection 
requirements for the flow straighteners, and for the reduction in 
inspection requirements for the reactor coolant pump flywheels as 
granted by the NRC and supported by WCAP-14535A do not impact the 
accident analysis assumptions or the basis of any Technical 
Specification. The revised inspection requirements only reduce the 
examination frequency for the reactor coolant pump flywheels and 
eliminate the inspection requirements for the flow [straightener] in 
Unit 1. Therefore, the proposed changes in surveillance (inspection) 
frequency do not result in 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Gordon E. Edison, Acting.

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

    Date of amendment request: February 3, 1998.
    Description of amendment request: The proposed changes will allow 
the reactor trip bypass breakers to be tested in the racked-in 
position. This change will continue to ensure the operability of the 
breakers and eliminate unnecessary movement caused by racking the 
breakers, thus reducing the wear and tear on the breakers and the 
possibility of a reactor trip. The operation of the Reactor Protection 
System and the reactor trip and the reactor trip bypass breakers are 
not being changed. The proposed changes in the test sequence for the 
reactor trip bypass breakers continue to provide assurance that the 
reactor trip bypass breakers will operate as designed to mitigate the 
consequence of any unsafe or improper reactor operation during steady-
state or transient power operations when the bypass breakers are placed 
in service for reactor trip system testing or trip breaker maintenance.
    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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed changes 
for the North Anna Units 1 and 2 and determined that a significant 
hazards consideration is not involved.
    (a) Operation and testing of the reactor trip breakers does not 
increase the probability of an accident or malfunction of equipment 
important to safety previously evaluated in the safety analysis 
report.
    The testing sequence will continue to ensure that the reactor 
trip system will be operable to mitigate the consequences of any 
unsafe or improper reactor operation during steady state or 
transient power operations. Although the breaker is placed in 
service before it is tested, the breaker is tested as soon as 
practicable to reestablish operability prior to performing testing 
of the reactor trip system or maintenance on the reactor trip 
breakers. During the short period of time the breaker is closed 
before the local shunt trip device test, the operability of the 
breaker is established based on satisfactory breaker testing 
conducted during the previous surveillance interval. Changing the 
minimum channels operable requirement for the reactor trip bypass 
breakers does not affect the operation of the reactor trip system 
since only one reactor trip breaker can be inservice for testing or 
maintenance of the reactor protection system. Therefore, the 
proposed test sequence does not significantly increase the 
probability of occurrence or the consequences of any previously 
analyzed accident.
    (b) The proposed Technical Specifications do not create the 
possibility of an accident or malfunction of a different type than 
any evaluated previously in the safety analysis report.
    The proposed test sequence change does not alter the actual test 
performed to establish operability of the reactor trip bypass 
breakers. The bypass breakers will be proven operable prior to 
reactor trip system testing or reactor trip breaker maintenance. 
Although the breaker is placed in service before it is tested, the 
breaker is tested as soon as practicable to reestablish operability 
prior to performing testing of the reactor trip system or 
maintenance on the reactor trip breakers. During the short period of 
time the breaker is closed before the local shunt trip device test, 
the operability of the breaker is established based on satisfactory 
breaker testing conducted during the previous surveillance interval. 
Changing the minimum channels operable requirement for the reactor 
trip bypass breakers does not affect the operation of the reactor 
trip system since only one reactor trip bypass breaker can be 
inservice for testing or maintenance of the reactor protection 
system. Therefore, it is concluded that no new or different kind of 
accident or malfunction from any previously evaluated has been 
created.
    (c) The proposed Technical Specifications change does not result 
in a significant reduction in margin of safety.
    The proposed change in the reactor trip bypass breaker test 
sequence provides assurance that the reactor trip system remains 
operable during normal operations or during reactor trip system 
testing and reactor trip breaker maintenance to mitigate the 
consequences of any unsafe or improper reactor operation. Changing 
the minimum channels operable requirement for the reactor trip 
bypass breakers does not affect the operation of the reactor trip 
system since only one reactor trip bypass breaker can be inservice 
for testing or maintenance of the reactor protection system. 
Therefore, the proposed change in the test sequence for the reactor 
trip bypass breaker does not significantly 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 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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Gordon E. Edison, Acting.

[[Page 11926]]

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: October 13, 1997, as supplemented by a 
letter dated February 10, 1998.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Technical Specifications (TS) to denote several 
changes. The proposed changes are: Relocating information to the 
Updated Safety Analysis Report (USAR), deleting redundant information, 
incorporating new references and deleting incorrect references, 
correcting errors, and augmenting existing requirements.
    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 proposed changes were revised in accordance with the provision 
of 10 CFR 50.92 to show no significant hazards exist. The proposed 
changes will not:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The likelihood that an accident will occur is neither increased 
nor decreased by these TS changes. The TS changes will not impact 
the function or method of operation of plant equipment. Thus, there 
is not a significant increase in the probability of a previously 
analyzed accident due to the changes. Since no plant practices have 
changed and no physical changes are being made, no systems, 
equipment, or components are affected by the proposed changes. Thus, 
the consequences of the malfunction of equipment important to safety 
previously evaluated in the Updated Safety Analysis Report (USAR) 
are not increased by the changes.
    The proposed changes are administrative in nature and, 
therefore, have no impact on accident initiators or plant equipment, 
and thus, do not affect the probabilities or consequences of an 
accident.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Operation of the facility in accordance with the proposed TS 
changes would not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve changes to the physical plant or operations. 
Since these administrative changes do not contribute to accident 
initiation, they do not produce a new accident scenario or produce a 
new type of equipment malfunction. Also, these changes do not alter 
any existing accident scenarios; they do not affect equipment or its 
operation, and thus, do not create the possibility of a new or 
different kind of accident.
    (3) Involve a significant reduction in the margin of safety.
    Changes in the proposed amendment include relocating information 
to the USAR, deleting redundant information, incorporating new 
references, deleting incorrect references, correcting errors, and 
augmenting existing requirements. Operation of the facility in 
accordance with the proposed TS would not involve a significant 
reduction in a margin of safety. The proposed changes do not affect 
plant equipment or operation. Safety limits and limiting safety 
system settings are not 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: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Project Director: Richard P. Savio.

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. 50-270, Oconee Nuclear Station, 
Unit 2, Oconee County, South Carolina

    Date of amendment request: January 15, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Table 4.1-1 and TS 4.5.2.1.2 to 
allow a one-time extension for specified Unit 2 refueling outage 
surveillances during operating cycle 16.
    Date of publication of individual notice in the Federal Register: 
January 23, 1998 (63 FR 3593).
    Expiration date of individual notice: February 23, 1998.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of application for amendment: February 3, 1998.
    Brief description of amendment request: The proposed amendment 
would change the operability requirement for the Standby Liquid Control 
system to Run/Power Operations and Startup.
    Date of individual notice in Federal Register: February 26, 1998 
(63 FR 9872).
    Expiration date of individual notice: March 30, 1998.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of application for amendment: February 3, 1998.
    Brief description of amendment request: The proposed amendment 
would revise the definitions of Cold Condition and Cold Shutdown and 
add a new section, 3.17, Vessel Hydrostatic Pressure and Leak Testing, 
to the Technical Specifications to specifically allow reactor vessel 
hydrostatic pressure testing to be performed during plant shutdown.
    Date of individual notice in Federal Register: February 26, 1998 
(63 FR 9874).
    Expiration date of individual notice: March 30, 1998.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: July 23, 1997, as supplemented September 
30, October 27, and December 18, 1997, and February 12, 1998.
    Description of amendment request: The July 23, 1997, application 
was previously noticed in the Federal Register on September 10, 1997 
(62 FR 47699). In addition, the December 18, 1997, supplement provided 
additional information that revised the original licensee's evaluation 
of the no

[[Page 11927]]

significant hazards consideration and, therefore, was noticed in the 
Federal Register on January 14, 1998 (63 FR 2281). The February 12, 
1998, supplement provided additional information that revised the 
licensee's evaluation of the no significant hazards consideration. 
Therefore, renotification of the Commission's proposed determination of 
no significant hazards is necessary.
    The proposed amendments would revise the Technical Specifications 
(TSs) by relocating the reactor coolant system (RCS) pressure and 
temperature limits from the TSs to the proposed Pressure Temperature 
Limits Report in accordance with the guidance provided by Generic 
Letter 96-03, ``Relocation of the Pressure Temperature Limit Curves and 
Low Temperature Overpressure Protection System Limits.'' TS 3.4.10.3 
would be revised to require that two residual heat removal system 
suction relief valves be operable or that the RCS be vented at RCS 
indicated cold leg temperatures less than or equal to 325  deg.F. In 
addition, a new TS would be added to limit the operation of more than 
one reactor coolant pump below 110  deg.F.
    Date of publication of individual notice in the Federal Register: 
February 23, 1998 (63 FR 9020).
    Expiration date of individual notice: March 25, 1998.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: November 7, 1997.
    Brief description of amendments: The amendments remove the 24/48 
Volt direct current (Vdc) batteries and associated charger and 
distribution systems from the Unit 2 Technical Specifications. All 
safety-related loads associated with the 24/48 Vdc batteries for Unit 2 
will be connected to other safety related battery systems which are in 
the TS.
    Date of issuance: February 25, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 165 and 160.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2277).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: October 3, 1996.
    Brief description of amendments: The amendments will correct a 
typographical error that was introduced into the Technical 
Specifications with the issuance of Amendment Nos. 150 and 145 issued 
on June 28, 1996.
    Date of issuance: February 25, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 166 and 161.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 Fr 
2273).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: October 15, 1997.
    Brief description of amendments: The amendments eliminate 
unnecessary detail from the Accident Monitoring Instrumentation 
Surveillance Requirements (TS Table 4.3.7.5-1).
    Date of issuance: February 17, 1998.
    Effective date: Immediately, to be implemented prior to startup 
from L1F35 for Unit 1 and L2R07 for Unit 2.
    Amendment Nos.: 123 and 108.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61841).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station, 
Unit 2, Oconee County, South Carolina

    Date of application for amendment: January 15, 1998.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Table 4.1-1 and Specification 4.5.2.1.2 to allow a 
one-time extension for specified Unit 2 refueling outage surveillances 
during operating cycle 16.

[[Page 11928]]

    Date of issuance: February 23, 1998.
    Effective date: As of the date of issuance to be implemented upon 
receipt.
    Amendment No.: 228.
    Facility Operating License No. DPR-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 23, 1998 (63 FR 
3593).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

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 for amendments: February 2, 1998, as 
supplemented February 18, 1998.
    Brief description of amendments: The amendments revise the wording 
used to specify refueling outage surveillances.
    Date of issuance: February 26, 1998
    Effective date: As of the date of issuance and will be implemented 
within 30 days.
    Amendment Nos.: Unit 1-228; Unit 2-229; Unit 3-225.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. (63 FR 6784 dated February 10, 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 March 12, 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 February 
18, 1998, letter provided clarifying information that did not change 
the scope of the February 2, 1998, application and the no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and a final no significant hazards consideration 
determination are contained in a Safety Evaluation dated February 26, 
1998.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: August 22, 1997.
    Brief description of amendment: Changes to the Technical 
Specifications (TS) to relocate the inservice testing program 
requirements from TS 4.0.5 to the Administrative Controls Section in 
the Unit 1 and 2 TS.
    Date of Issuance: February 25, 1998.
    Effective Date: February 25, 1998.
    Amendment Nos.: 153 and 91.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50006).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: October 21, 1997, as 
supplemented by letter dated February 3, 1998. The application 
superseded a previous application of May 16, 1997.
    Brief description of amendment: This amendment revised 
administrative requirements regarding the unit staff positions of 
General Supervisor Operation and Manager Operations as stated in TS 
6.2.2.i and 6.3.1.
    Date of issuance: February 19, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 160.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59916).
    The February 3, 1998, letter provided clarifying information that 
did not change the no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 19, 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.

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

    Date of application for amendment: March 27, 1997, as supplemented 
on September 25, 1997.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Limiting Condition for Operation (LCO) 3.7.11 and 
Surveillance Requirement (SR) 4.7.11 for the ultimate heat sink. TS LCO 
3.7.11 is changed to indicate that the ultimate heat sink is operable 
at a water temperature of less than or equal to 75  deg.F instead of an 
average value. The use of average when verifying the water temperature 
and the reference to a specific monitoring location are deleted in TS 
SR 4.7.11.a and .b. The TS Bases Section 3/4.7.11 is also modified to 
reflect the above changes.
    A license condition was also included in Appendix B of the 
Operating license, which is a list of additional license conditions. 
This license condition was discussed with NNECO in a conference call on 
December 15, 1997, and NNECO agreed to the inclusion of the license 
condition for approving the amendment.
    Date of issuance: February 9, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 213.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications and Appendix B of Operating License.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19831).
    The September 25, 1997, letter provided clarifying information that 
did not change the scope of the March 27, 1997, 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 February 9, 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,

[[Page 11929]]

49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: August 29, 1997, as supplemented 
by letters dated September 25 and November 14, 1997.
    Brief description of amendment: Based on a review and subsequent 
calculations of the cold overpressurization protection (COPS) enabling 
temperature and the emergency core cooling system (ECCS)/charging 
system mode 3 requirements, NNECO proposes to reduce the COPS enabling 
temperature. As a result, NNECO proposed the following Technical 
Specifications (TS) changes: add new heatup and cooldown pressure/
temperature limit curves and their associated requirements; add new 
power operated relief valve (PORV) setpoint curves and their associated 
requirements; revise the reactor coolant loops and coolant circulation, 
ECCS, boration systems, and COPS to incorporate the lower enabling 
temperature and new restrictions for cold overpressure protection 
system, PORV undershoot, and residual heat removal (RHR) relief valve 
bellows; add a footnote to allow a reactor coolant pump to substitute 
for an RHR pump during heatup from Mode 5 to 4, which is consistent 
with the improved standard technical specification (STS); reword TS 3/
4.4.9.3 and its surveillance requirement to be consistent with the 
improved STS; and revise the affected Bases sections to be consistent 
with the proposed changes.
    Date of issuance: February 12, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days of issuance.
    Amendment No.: 157.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1997 (62 FR 
52583).
    The September 25 and November 14, 1997, letters provided clarifying 
information that did not change the August 27, 1997, 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 February 12, 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.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: November 20, 1996, as supplemented by 
letter dated February 20, 1997, and submittal dated March 25, 1997.
    Brief description of amendment: The amendment revised the technical 
specifications to reflect organizational changes and correct editorial 
and typographical inaccuracies. It also removed paragraph 3.D of the 
facility operating license that described the modification that 
increased the spent fuel pool storage capacity.
    Date of issuance: February 3, 1998.
    Effective date: February 3, 1998.
    Amendment No.: 184.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications and Facility Operating License No. DPR-40.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
131) and April 9, 1997 (62 FR 17238). The March 25, 1997, submittal did 
not change the staff's original no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 3, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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 application for amendments: October 4, 1995, as 
supplemented by letters dated July 17, 1996, August 20, 1996, and June 
2, 1997.
    Brief description of amendments: The amendments revise the 
technical specifications to relocate the requirements in 10 subsections 
of the technical specifications to licensee-controlled documents.
    Date of issuance: February 3, 1998.
    Effective date: February 3, 1998, to be implemented within 90 days 
of issuance.
    Amendment Nos.: Unit 1--120; Unit 2--118.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Operating Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58404). The July 17, 1996, August 20, 1996, and June 2, 1997, 
supplemental letters provided additional clarifying information and did 
not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 3, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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 application for amendments: May 14, 1997, as supplemented 
by letter dated December 15, 1997.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Unit Nos. 1 and 2 to revise Technical Specification (TS) 
6.9.1.8.b.5 to replace reference WCAP-10266-P-A with WCAP-12945-P for 
best estimate loss-of-coolant accident (LOCA) analysis. The amendment 
also revises TS Bases 3/4.2.2 and 3/4.2.3 to change the emergency core 
cooling system (ECCS) acceptance criteria limit to state that there is 
a high level of probability that the ECCS acceptance criteria limits 
are not exceeded.
    Date of issuance: February 13, 1998.
    Effective date: February 13, 1998, to be implemented within 90 days 
of issuance.
    Amendment Nos.: Unit 1--121; Unit 2--119.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40855).
    The December 15, 1997, supplemental letter provided additional 
clarifying information and did not change the staff's initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated February 13, 1998.

[[Page 11930]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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 application for amendments: December 9, 1996.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP), Unit Nos. 1 and 2 to revise the surveillance frequencies 
from at least once every 18 months to at least once per refueling 
interval (nominally 24 months) for the reactor trip system (RTS) and 
engineering safety features actuation systems (ESFAS) instrumentation 
channels, and make certain changes in trip setpoints and allowance 
values due to a setpoint methodology change in support of the 
calibration extensions. Channel operational tests (COTs) and trip 
actuating device operational tests (TADOTs) associated with these 
channels are also being extended. Revisions to the appropriate TS Bases 
are being revised to support the TS revisions.
    Date of issuance: February 17, 1998.
    Effective date: February 17, 1998, to be implemented within 90 days 
of issuance.
    Amendment Nos.: Unit 1--Amendment No. 122; Unit 2--Amendment No. 
120.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6577)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: May 31, 1996.
    Brief description of amendments: These amendments delete, from the 
Technical Specifications, Section 4.7.2.d.2, the surveillance 
requirement for chlorine detection for the control room emergency 
outside air supply system as a result of the removal of bulk quantities 
of gaseous chlorine from the Susquehanna Steam Electric Station.
    Date of issuance: February 19, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 172 and 145.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38137).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 19, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Southern Nuclear Power Company, Inc., Georgia Power Company, Oglethorpe 
Power Corporation, Municipal Electric Authority of Georgia, City of 
Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric 
Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of application for amendments: August 8, 1997, as supplemented 
October 10, 1997, January 16, 23, and 27, 1998.
    Brief description of amendments: The amendment changes Vogtle 
Electric Generating Plant, Units 1 and 2, Technical Specifications (TS) 
3.7.17, ``Fuel Storage Pool Boron Concentration,'' TS 3.7.18, ``Fuel 
Assembly Storage in the Fuel Storage Pool,'' and TS 4.3, ``Fuel 
Storage,'' to allow credit for soluble boron, in the spent fuel pool, 
for maintenance of subcriticality associated with spent fuel storage.
    Date of issuance: February 20, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 99--Unit 1; 77--Unit 2
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68136).
    The January 16, 23, and 27, 1998, letters provided clarifying 
information that did not change the scope of the August 8, 1997, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 20, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: September 17, 1997 (TS 97-02).
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by modifying Surveillance Requirements 
(SRs) 4.6.2.1.1.b., 4.6.2.1.1.c,. 4.6.2.1.1.d, and 4.6.2.1.2.b to 
account for a plant modification to the containment spray system and to 
make the SRs more consistent with the Westinghouse Standard TS (NUREG-
1431).
    Date of issuance: February 20, 1998.
    Effective date: February 20, 1998.
    Amendment Nos.: 231 and 221.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise TS.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52589).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 20, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: February 23, 1996, as 
supplemented by letters dated April 24, 1996, and November 15, 1996.
    Brief description of amendment: The amendment revises the Callaway 
Plant, Unit 1 operating license to reflect Union Electric Company (UEC) 
as a wholly-owned operating subsidiary of Ameren Corporation at the 
closing of the contemplated merger between UEC and CIPSCO Incorporated.
    Date of issuance: February 13, 1998.

[[Page 11931]]

    Effective date: February 13, 1998.
    Amendment No.: 120.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25713) The November 15, 1996, supplemental letter provided only 
clarifying information and did not change the original no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri.

    Date of application for amendment: August 8, 1997.
    Brief description of amendment: The amendment revises the Callaway 
Plant, Unit 1 surveillance requirements of Technical Specification 3/
4.7.4, ``Essential Service Water System'' by removing the requirement 
to perform 4.7.4.b, 4.7.4.b.2 and 4.7.4.c during shutdown.
    Date of issuance: February 24, 1998.
    Effective date: February 24, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 121.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66143) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

    Dated at Rockville, Maryland, this 4th day of March 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV Office of Nuclear 
Reactor Regulation.
[FR Doc. 98-6085 Filed 3-10-98; 8:45 am]
BILLING CODE 7590-01-P