[Federal Register Volume 64, Number 115 (Wednesday, June 16, 1999)]
[Notices]
[Pages 32284-32296]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-15098]


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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 May 21, 1999, through June 4, 1999. The last 
biweekly notice was published on June 2, 1999 (64 FR 29707).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By July 19, 1999, 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

[[Page 32285]]

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: May 23, 1997, as revised by letters 
dated September 27, 1998, and May 26, 1999.
    Description of amendments request: The proposed amendments would 
revise Technical Specification (TS) Limiting Condition of Operation 
(LCO) 3.4.14 and TS Sections 5.5.9 and 5.6.8 to allow the use of steam 
generator (SG) tube sleeves as an alternative to plugging defective SG 
tubes. The May 26, 1999, letter completely revised the May 23, 1997, 
request for amendments, and this notice supersedes the original Federal 
Register notice dated July 30, 1997 (62 FR 40845).
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to TS LCO 3.4.14.d and e will replace the 
leakage limits of 1 gallon per minute (gpm) primary to secondary 
leakage through all SGs and 720 gallon per day (gpd) through any one 
SG with a new limit of 150 gpd through any one SG. This is a more 
restrictive change. A TS limit of 150 gpd primary to secondary 
Leakage through any one steam generator is significantly less than 
the initial conditions assumed in the safety analyses. The 150 gpd 
limit is based on operating experience as an indication of one or 
more propagating tube leak mechanisms. The Steam Generator Tube 
Surveillance Program described in TS Section 5.5.9 ensures that the 
structural integrity of the SG tubes is maintained. The leakage rate 
limit of 150 gpd for any one SG provides additional assurance 
against tube rupture at normal and faulted conditions and provides 
additional assurance that cracks will not propagate to burst prior 
to detection by leakage monitoring methods and commencement of plant 
shutdown. Therefore, this change to TS LCO 3.4.14.e will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to TS 5.5.9 will add inservice inspection 
requirements for SG tube sleeves. These requirements will ensure 
that all installed SG tube sleeves will be inspected prior to 
initial operation and routinely thereafter, to assure the capability 
of each sleeve to perform its design function during each operating 
cycle. The tube sleeves will be the Combustion Engineering, Inc. (CE 
or ABB-CE) Leak Tight sleeves, as described in CE report CEN-630-P, 
``Repair of \3/4\'' O.D. Steam Generator Tubes Using Leak Tight 
Sleeves,'' Revision 02, dated June 1997. (This proprietary report is 
provided as Enclosure 4 with this submittal.) The tube sleeve 
dimensions, materials and joints are designed to the applicable ASME 
[American Society of Mechanical Engineers] Boiler and Pressure 
Vessel code requirements. An extensive test program was performed 
that demonstrated that the sleeves will fulfill their intended 
function as leak tight structural members. Evaluation of sleeved 
tubes indicates no detrimental effects on the sleeve-tube assembly 
resulting from reactor coolant system flow, coolant chemistries, or 
thermal and pressure conditions. Structural analyses of the sleeve-
tube assembly have established its integrity under normal and 
accident conditions. Mechanical testing using ASME code stress 
allowables was performed to support the analyses. Also, corrosion 
tests were performed and revealed no evidence of sleeve or tube 
corrosion considered detrimental under anticipated service 
conditions. A sleeved tube will exhibit greater hydraulic resistance 
and reduced heat transfer capability than an un-sleeved tube. 
However, these effects are much less than would be imposed by taking 
the tube out of service by plugging. Section 10.0 of CE report CEN-
630-P describes the analyses to determine the hydraulic and heat 
transfer effects. Calculations using plant-specific information will 
identify sleeve-to-plug equivalency ratios. The proposed changes to 
the SG inservice inspection program will assure that sleeved SG 
tubes will meet the structural requirements of tubes that are not 
defective. The proposed sleeve plugging limit of 35% of nominal wall 
will ensure that the sleeves remaining in service will perform their 
design function. Also, installation of

[[Page 32286]]

sleeves will not significantly [a]ffect the primary system flow rate 
or the heat transfer capability of the SGs. Therefore, this change 
to TS section 5.5.9 will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The change to the SG reporting requirements in TS section 5.6.8 
will ensure that the number of sleeved SG tubes will be reported to 
the NRC along with the number of plugged tubes. This is an 
administrative change that has no effect on the operation or 
maintenance of the plant and will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to TS LCO 3.4.14.d and e will replace the 
leakage limits of 1 gpm primary to secondary leakage through all SGs 
and 720 gpd through any one SG with a new limit of 150 gpd through 
any one SG. This is a more restrictive change that will provide 
added assurance against steam generator tube ruptures. Since the 
current allowable primary to secondary leakage is being reduced, 
this change will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes to TS section 5.5.9 for the SG inservice 
inspection program will assure that sleeved SG tubes will meet the 
structural requirements of tubes that are not defective. Also, 
installation of sleeves will not significantly [a]ffect the primary 
system flow rate or the heat transfer capability of the SGs. 
Therefore, this change to TS section 5.5.9 will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The change to the SG reporting requirements in TS section 5.6.8 
will ensure that the number of sleeved SG tubes will be reported to 
the NRC along with the number of plugged tubes. This is an 
administrative change that has no effect on the operation or 
maintenance of the plant and will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change to TS LCO 3.4.14.d and e will replace the 
leakage limits of 1 gpm primary to secondary leakage through all SGs 
and 720 gpd through any one SG with a new limit of 150 gpd through 
any one SG. This is a more restrictive change that will provide 
added assurance against steam generator tube ruptures. Since the 
current allowable primary to secondary leakage is being reduced, 
this change will not involve a significant reduction in a margin of 
safety.
    The proposed changes to TS section 5.5.9 for the SG inservice 
inspection program will assure that sleeved SG tubes will meet the 
structural requirements of tubes that are not defective. Also, 
installation of sleeves will not significantly [a]ffect the primary 
system flow rate or the heat transfer capability of the SGs. 
Therefore, this change to TS section 5.5.9 will not involve a 
significant reduction in a margin of safety.
    The change to the SG reporting requirements in TS section 5.6.8 
will ensure that the number of sleeved SG tubes will be reported to 
the NRC along with the number of plugged tubes. This is an 
administrative change that has no effect on the operation or 
maintenance of the plant and will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: May 5, 1999.
    Description of amendment request: The proposed amendments would 
revise the basis for evaluation of the reactor building ventilation 
(VR) system exhaust plenum masonry walls. Specifically, the amendment 
would approve the use of different methodology and acceptance criteria 
for the reassessment of certain masonry walls subjected to transient 
pressurization loads resulting from a high energy line break.
    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:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The change involves reassessment of the VR exhaust plenum due to 
a transient pressurization during a Main Steam Line Break (MSLB). 
Since the transient pressurization is a result of the MSLB, and the 
block walls and the dampers are not initiators of any accident, the 
probability of an accident previously evaluated is not affected.
    This analysis does not affect the total amount of radioactive 
release due to the MSLB Outside of the Primary Containment, so the 
total offsite dose consequences does not change. A small portion of 
the release, which passes the dampers prior to closure, will now be 
an elevated release via the plant ventilation stack instead of a 
ground level release. The original analysis assumed the entire 
release was a ground level release, and thus remains bounding for 
the MSLB accident.
    The Control Room and Auxiliary Electric Equipment Room (AEER) 
dose consequences are impacted only slightly due to the small amount 
of steam/air mixture released from the new pressure relief damper. 
The steam/air mixture becomes mixed with the air volume in that area 
of the Auxiliary Building but was all assumed to be available for 
inleakage to the Control Room and AEER. The dose increase for the 
Control Room and AEER is less than or equal to 0.05 Rem thyroid and 
negligible change to the whole body dose, such that the dose due to 
the MSLB accident remains much less than the DBA LOCA dose and 
General Design Criteria 19. The MSLB accident dose consequences 
remain bounded by the Design Basis Loss of Coolant Accident.
    The effects of the steam released by the pressure relief damper 
into the Auxiliary Building has been evaluated for environmental 
qualification impact on systems, structures and components (SSCs) in 
the area of the Auxiliary Building affected for both radiation and 
steam/temperature affects. The effect on area temperature is about 4 
 deg.F and is above initial temperature for not more than 24 hours. 
The change in humidity is negligible, and radiation dose impact is 
small and bounded by previous calculations.
    These consequences assume that the VR exhaust plenum masonry 
walls do not rupture based on the design changes being made in 
conjunction with the masonry wall reevaluation for each LaSalle Unit 
that will prevent the failure of the VR exhaust plenum masonry 
walls.
    Therefore this proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The MSLB accident is previously analyzed but considered only 
instantaneous closure of installed dampers. The reevaluation and 
design changes extend the previous accident analysis to assure that 
structures previously considered unaffected by the MSLB will 
maintain their structural integrity. The block walls are static and 
the dampers function in response to an accident, thus the analysis 
method and design changes are not accident initiators. Therefore the 
change does not create the possibility of a new [or] different kind 
of accident from any accident previously evaluated.
    The design changes being made in conjunction with the masonry 
wall reevaluation for each LaSalle Unit that will prevent the 
failure of the VR exhaust plenum masonry walls are as follows:
    (1) Installation of a pressure relief damper,
    (2) An excess-flow check damper, and
    (3) Required masonry wall support improvements in the reactor 
building ventilation exhaust plenum for each Unit.
    The reevaluation of the masonry walls uses different load 
factors and load combinations

[[Page 32287]]

as well as reduced acceptance criteria than previously used for 
these walls. The change in the evaluation does not cause the rupture 
or failure of the effected masonry walls, since the evaluation shows 
the walls remain intact.
    The installation of the above design changes, in conjunction 
with masonry wall analysis assure that the subject masonry walls 
will not rupture or fail. Therefore, SSCs that would be affected by 
wall rupture can fulfill their intended function, maintaining the 
consequences of previously evaluated accident the same.
    The new pressure relief damper and excess-flow check damper are 
safety-related and are analyzed to function under the conditions 
created by the MSLB. In addition, the dampers and the duct they are 
installed in have been analyzed to assure no failure will occur 
during an Operating Basis Earthquake (OBE) or Safe Shutdown 
Earthquake (SSE).
    Based on an analysis of potential failure modes in accordance 
with ANSI/ANS-58.9-1981, ``Single Failure Criteria for Light Water 
Reactor Safety-Related Fluid Systems,'' Paragraph 4.1, the active 
function of the pressure relief damper and excess flow check damper 
are considered exempted from consideration of single failure. The 
principles governing operation of the dampers are simple and direct 
and not subject to change or deterioration with time, similar to the 
function of a code safety relief valve and a swing check valve. With 
periodic testing of the dampers, continued reliable performance is 
assured.
    The dampers are designed and set so that the pressures created 
by normal ventilation flow changes do not cycle the dampers, and 
thus the new dampers do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Administrative controls will be in place prior to implementation 
of this change to assure the testing and maintenance is periodically 
performed in accordance with vendor recommendations. These dampers 
will be included as equipment required to be monitored/maintained, 
because the function performed by the dampers is within the scope of 
the Maintenance Rule, 10 CFR 50.65.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    Originally, no masonry walls were evaluated for HELB 
pressurization effects, because the walls were considered protected 
by the isolation dampers. However, the original design methodology 
for masonry did include load combinations including Pa:

Abnormal
1.0D + 1.0L + 1.5Pa

Abnormal/Severe Environment
1.0D + 1.0L + 1.25Pa + 1.25Eo

Abnormal/Extreme Environment
1.0D + 1.0L + 1.0Pa + 1.0Ess,

    Where D is Dead Load; L is Live Load; Pa is 
pressurization due to HELB; Eo is Loads generated by the 
Operating Basis Earthquake (OBE); and Ess is Loads 
generated by the Safe Shutdown Earthquake (SSE).
    The current reevaluation was required due to determination that 
some block walls in the LaSalle Auxiliary Building are affected by a 
transient pressurization due to a MSLB. The specific changes from 
the original analyses involve the following for loads and load 
combinations.
    1. Abnormal:

1.0D + 1.0L + 1.0PHELB

    2. Abnormal/severe environmental:

1.0D + 1.0L + [(1.1Eo)2 + 
1.0PHELB2]\1/2\

    3. Abnormal/extreme environmental:

1.0D + 1.0L + [1.0Ess2 + 
1.0PHELB2]\1/2\

Where:
(1) PHELB is the short-term differential pressurization 
load on the VR plenum masonry walls resulting from non-instantaneous 
opening/closure of the protection dampers.
(2) The Load Factor on pressure due to HELB is 1.0 for all cases.
(3) The Loading Combination of pressure and seismic is the Square 
Root of the Sum of Squares (SRSS).

    LaSalle has selected the proposed load combinations in 
consideration of the following:
    Isolation, check, and relief dampers protect the walls; 
therefore the pressurization effects are not sustained, but are 
transient in nature.
    The transient pressurization effect (PHELB) is 
derived from a conservative detailed analysis of an instantaneous 
HELB combined with non-instantaneous damper opening/closure. Due to 
the precise nature and conservatism of this HELB analysis, there is 
little uncertainty in PHELB .
    Therefore a load factor of 1.0 is used for all abnormal load 
combinations.
    PHELB is a short duration, dynamic load. Accordingly, 
the seismic and transient HELB pressurization loads are combined 
using the Square Root of Sum of the Squares (SRSS) method because 
the peak effects of these dynamic loads are unlikely to occur 
simultaneously. This combination method is used in the analysis of 
other components such as component supports.
    The proposed load combinations accordingly provide a 
conservative basis for reassessment of the VR exhaust plenum masonry 
wall systems.
    In regards to the masonry acceptance criteria, the original 
acceptance criteria used for this condition are the National 
Concrete Masonry Associations (NCMA) ``Specification for the Design 
and Construction of Load Bearing Masonry--1979'' allowable stresses 
times a 1.67 factor. These allowable stresses correspond to stress 
equal to the modulus of rupture (fr) of the masonry 
divided by a factor of safety of 3.35. During reviews to address 
masonry wall issues per NRC IE Bulletin 80-11, six walls did not 
meet this acceptance criteria. The acceptance criteria used for 
these walls was for fr values determined from testing at 
Clinton Power Station divided by a factor of safety of 2.5. This 
acceptance criteria was accepted by the NRC for LaSalle in 
Supplement 5 of NUREG 0519, Safety Evaluation Report related to the 
Operation of LaSalle County Station, Units 1 and 2. The VR exhaust 
plenum walls will use the same acceptance criteria for the transient 
HELB pressurization cases.
    The minimum masonry safety factor for the LaSalle Unit 2 walls 
affected by the HELB loads range from 2.6 to 3.1 with one wall 
having a safety factor of 4.9.
    Masonry wall steel support members were originally designed for 
this condition elastically to the American Institute of Steel 
Construction's (AISC) ``Steel Construction Manual--Seventh Edition'' 
allowable stresses times a 1.6 factor. In the reassessment of these 
members due to the transient HELB pressurization, elasto-plastic 
behavior is allowed (with a ductility ratio limit of 10). It is 
appropriate to consider them similar to high-energy line break 
systems that will maintain their integrity as they absorb the energy 
of the incidental pressure excursion.
    High-energy line breaks are discussed in Section 3.6 of the 
UFSAR. The discussion in this section focuses on the design of pipe 
whip restraints, and in Table 3.6-6 acceptance criteria are 
provided. This table shows that the energy absorbing portions of the 
pipe whip restraint are allowed to go plastic, thereby absorbing 
energy. While Table 3.6-6 of the UFSAR deals with energy absorbing 
portions of the pipe whip restraints, wide-flange shapes are not 
addressed. Wide-flange shapes absorb energy through flexural 
deformations.
    Guidance on appropriate acceptance criteria for flexural members 
is provided in Appendix A to SRP 3.5.3, ``Barrier Design 
Procedures.'' This appendix indicates that for tension due to 
flexure in structural steel members, a ductility ratio value not to 
exceed 10.0 is acceptable. SRP 3.8.4, paragraph III.5 also notes 
that some localized points on the structure, the allowable stresses 
specified for ``structural steel'' may be exceeded, provided that 
integrity of the structure is not affected.
    Note that only one of the Unit 2 walls affected by these HELB 
loads required the use of the elasto-plastic acceptance criteria for 
two structural steel members.
    In summary, these alternate criteria for reassessment of the 
integrity of the LaSalle Reactor Building Ventilation Exhaust Plenum 
masonry walls in conjunction with the design changes adding a 
pressure relief damper, an excess flow check damper and masonry wall 
support steel changes, assures that the walls will maintain their 
integrity during a MSLB. The safety factor is reduced; however, the 
walls have sufficient strength and safety margin to maintain 
structural integrity and thus perform their intended safety function 
during the pressurization transient due to a MSLB accident.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois

[[Page 32288]]

Valley Community College, Oglesby, Illinois 61348-9692.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix, County, Michigan

    Date of amendment request: May 11, 1999 (Accession No. 9905170189).
    Description of amendment request: The proposed amendment would 
delete from the Defueled Technical Specifications (DTS) the definition 
for site boundary and Figure 5.1-1, Big Rock Point Site Map, and revise 
the description of the Big Rock Point site under subsection 5.1. The 
amendment also proposes editorial changes associated with the above 
proposed revisions.
    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:

    In accordance with 10 CFR 50.91, Consumers Energy Company has 
made a determination that the proposed amendment does not involve 
significant hazards considerations. Consumers Energy Company has 
concluded that the proposed amendment will 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 proposed change is administrative in nature and has no 
[e]ffect on the health and safety of the public. There is no 
reduction or elimination of federal regulatory requirements 
associated with the proposed amendment. The information being 
removed from the Defueled Technical Specifications is unnecessary 
since Site Boundary is already defined in 10 CFR Part 20, and the 
site map [Defueled Technical Specification Figure 5.1-1] is already 
provided in the Updated Final Hazards [Summary] Report. Furthermore, 
the proposed changes are consistent with the guidance provide in 
NUREG-1625 ['Proposed Standard Technical Specifications for 
Permanently Defueled Westinghouse Plants''].
    The proposed change does not:
    (1) Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed amendment does not change the site boundary as it 
currently exists. Deleting the Site Boundary definition and changing 
the upper case characters to lower case throughout the DTS and the 
Bases where it appears, and deleting the site figure from the DTS 
and related references will not increase the probability or 
consequences of a new or different kind of accident previously 
evaluated. This proposed change is administrative in nature and does 
not involve fuel handling or affect or modify any system, structure 
or component.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed amendment does not change the site boundary as it 
currently exists. Deleting the Site Boundary definition and changing 
the upper case characters to lower case throughout the DTS and the 
Bases where it appears, and deleting the site figure from the DTS 
and related references will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This proposed change is administrative in nature and does not 
involve fuel handling or affect or modify any system, structure or 
component.
    (3) Involve a significant reduction in the margin of safety.
    The proposed changes do not involve any physical changes to the 
plant or plant procedures. There will be no 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: North Central Michigan 
College, 1515 Howard Street, Petosky, MI 49770.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Dr. Michael T. Masnik.

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 amendment request: October 2, 1998, supplemented May 13, 
1999.
    Description of amendment request: The proposed amendments would 
resolve an unreviewed safety question involving use of credit for 
reactor building overpressure in the licensing basis for the available 
net positive suction head for the reactor building spray pumps and the 
low pressure injection pumps. If approved, the appropriate changes 
would be incorporated in the Oconee Updated Final Safety Analysis 
Report.
    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.

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The reactor building spray (RBS) and low pressure injection 
(LPI) systems are not considered as initiators of any analyzed 
event, therefore, this change has no impact on the probability of an 
event previously analyzed.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed for the analysis, the availability 
and successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the set points at which these 
actions are initiated. The proposed change permits limited reactor 
building overpressure to be credited in the calculation of available 
net positive suction head (NPSH) for the RBS and LPI pumps for a 
limited period of time during the sump recirculation phase. It is 
supported by calculations which demonstrate that adequate reactor 
building overpressure will be available to ensure the RBS and LPI 
systems will be capable of performing their safety functions. Thus, 
the proposed change does not significantly increase the consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from the accidents previously evaluated?
    The proposed change permits limited reactor building 
overpressure to be credited in the calculation of available NPSH for 
the RBS and LPI pumps for a limited period of time during the sump 
recirculation phase. It does not involve a physical alteration of 
the plant. The proposed change is supported by calculations which 
demonstrate that adequate reactor building overpressure will be 
available to ensure the RBS and LPI systems will be capable of 
performing their safety functions. This change will not alter the 
manner in which the RBS or LPI system is initiated, nor will the 
function demands on the RBS or LPI system be changed. Thus, the 
proposed change does not create the possibility of a new or 
different kind of accident.
    3. Involve a significant reduction in a margin of safety?
    The proposed change permits limited reactor building 
overpressure to be credited in the calculation of available NPSH for 
the RBS and LPI pumps for a limited period of time during the sump 
recirculation phase. Crediting a slight amount of overpressure does 
not result in a significant reduction in the margin of safety, 
because conservative analyses demonstrate that adequate reactor 
building overpressure will be available to ensure the RBS and LPI 
systems will be capable of performing their safety functions. Thus, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    Duke has concluded based on the above information that there are 
no significant hazards involved in this LAR.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 32289]]

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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

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 amendment request: May 11, 1999.
    Description of amendment request: The proposed amendments would: 
(a) revise the pressure-temperature (P-T) limits of Technical 
Specification (TS) 3.4.3 for heatup, cooldown, and inservice test 
limitations for the Reactor Coolant System to a maximum of 33 Effective 
Full Power Years; (b) revise TS 3.4.12, Low Pressure Overpressure 
Protection System (LTOP), to reflect the revised P-T limits of the Unit 
1, 2, and 3 reactor vessels; (c) permit operation during LTOP 
conditions with two reactor coolant pumps in operation in a single 
loop; and (d) relax the LTOP operating envelope, thereby reducing 
potential challenges to the reactor coolant system power operated 
relief valves.
    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.

    A. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No.
    These proposed Technical Specification (TS) changes were 
developed utilizing the procedures of ASME XI, Appendix G, in 
conjunction with Code Cases N-514, N-588 and N-626, as described in 
the Technical Justification. Usage of these procedures provides 
compliance with the underlying intent of 10 CFR 50 Appendix G and 
provide safety limits and margins of safety that ensure failure of a 
reactor vessel will not occur.
    The proposed changes do not impact the capability of the reactor 
coolant pressure boundary (i.e., no change in operating pressure, 
materials, seismic loading, etc.) and therefore do not increase the 
potential for the occurrence of a loss of coolant accident (LOCA). 
The changes do not modify the reactor coolant system pressure 
boundary, nor make any physical changes to the facility design, 
material, or construction standards. The probability of any design 
basis accident (DBA) is not affected by this change, nor are the 
consequences of any DBA affected by this change. The proposed 
Pressure-Temperature (P-T) limits, Low Temperature Overpressure 
(LTOP) limits and setpoints, and allowable operating reactor coolant 
pump combinations are not considered to be an initiator or 
contributor to any accident analysis addressed in the Oconee UFSAR.
    The proposed changes do not adversely affect the integrity of 
the RCS such that its function in the control of radiological 
consequences is affected. Radiological off-site exposures from 
normal operation and operational transients, and faults of moderate 
frequency do not exceed the guidelines of 10 CFR 100. In addition, 
the proposed changes do not affect any fission product barrier. The 
revised PORV LTOP setpoint is established to protect reactor coolant 
pressure boundary. The changes do not degrade or prevent the 
response of the PORV or safety-related systems to previously 
evaluated accidents. In addition, the changes do not alter any 
assumption previously made in the mitigation of the radiological 
consequences of an accident previously evaluated.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be increased by approval of the 
requested changes.
    B. Create the possibility of a new or different kind of accident 
from the accident previously evaluated?
    No.
    The proposed license amendment revises the Oconee reactor vessel 
P-T limits, LTOP limits and setpoints, and allowable operating 
reactor coolant pumps combinations. Compliance with 10 CFR 50 
Appendix G, includes utilization of ASME XI, Appendix G, as modified 
by Code Cases N-514, N-588 and N-626 to meet the underlying intent 
of the regulations.
    Operation of Oconee in accordance with these proposed Technical 
Specifications changes will not create any failure modes not bounded 
by previously evaluated accidents. Consequently, approval of these 
changes will not create the possibility of a new or different 
accident from any accident previously evaluated.
    C. Involve a significant reduction in a margin of safety?
    No.
    The proposed Technical Specification (TS) changes were developed 
utilizing the procedures of ASME XI, Appendix G, in conjunction with 
Code Cases N-514, N-588 and N-626, as described in the Technical 
Justification. Usage of these procedures provides compliance with 
the underlying intent of 10 CFR 50 Appendix G and provides safety 
limits and margins of safety which ensure failure of a reactor 
vessel will not occur.
    No plant safety limits, set points, or design parameters are 
adversely affected. The fuel, fuel cladding, and Reactor Coolant 
System are not impacted. Therefore, there will be no significant 
reduction in any margin of safety as a result of approval of the 
requested changes.
    Duke has concluded based on this information there are no 
significant hazards considerations involved in this amendment 
request.

    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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of amendment request: May 17, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications associated with the enabling of the 
Oscillation Power Range Monitor (OPRM) instrumentation reactor 
protection system (RPS) trip function. The OPRM is designed to detect 
the onset of reactor core power oscillations resulting from thermal-
hydraulic instability and suppresses them by initiating a reactor scram 
via the RPS trip logic.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change specifies limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRM system and allows operation in regions of the power to flow map 
currently restricted by the requirements of Interim Corrective 
Actions (ICAs) and certain limiting conditions of operation of 
Technical Specifications (TS) 3.4.1. The OPRM system can 
automatically detect and suppress conditions necessary for thermal-
hydraulic (T-H) instability. A T-H instability event has the 
potential to challenge the Minimum Critical Power (MCPR) safety 
limit. The restrictions of the ICAs and TS 3.4.1 were imposed to 
ensure adequate capability to detect and suppress conditions 
consistent with the onset of T-H oscillations that may develop into 
a T-H instability event. With the installation of the OPRM System, 
these restrictions are no longer required.

[[Page 32290]]

    The probability of a T-H instability event is most significantly 
impacted by power to flow conditions such that only during operation 
inside specific regions of the power to flow map, in combination 
with power shape and inlet enthalpy conditions, can the occurrence 
of an instability event be postulated to occur. Operation in these 
regions may increase the probability that operation with conditions 
necessary for a T-H instability can occur.
    However, when the OPRM is operable with operating limits as 
specified in the COLR [Core Operating Limits Report], the OPRM can 
automatically detect the imminent onset of local power oscillations 
and generate a trip signal. Actuation of an RPS trip will suppress 
conditions necessary for T-H instability and decrease the 
probability of a T-H instability event. In the event the trip 
capability of the OPRM is not maintained, the proposed change 
includes actions which limit the period of time before the effected 
OPRM channel (or RPS system) must be placed in the trip condition. 
If these actions would result in a trip function, an alternate 
method to detect and suppress thermal hydraulic oscillations is 
required. In either case the duration of this period of time is 
limited such that the increase in the probability of a T-H 
instability event is not significant. Therefore the proposed change 
does not result in a significant increase in the probability of an 
accident previously evaluated.
    An unmitigated T-H instability event is postulated to cause a 
violation of the MCPR safety limit. The proposed change ensures 
mitigation of T-H instability events prior to challenging the MCPR 
safety limit if initiated from anticipated conditions by detection 
of the onset of oscillations and actuation of an RPS trip signal. 
The OPRM also provides the capability of an RPS trip being generated 
for T-H instability events initiated from unanticipated but 
postulated conditions. These mitigating capabilities of the OPRM 
system would become available as a result of the proposed change and 
have the potential to reduce the consequences of anticipated and 
postulated T-H instability events. Therefore, the proposed change 
does not significantly increase the consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change specifies limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRM system and allows operation in regions of the power to flow map 
currently restricted by the requirements of ICAs and TS 3.4.1. The 
OPRM system uses input signals shared with APRM [Average Power Range 
Monitor] and rod block functions to monitor core conditions and 
generate an RPS trip when required. Quality requirements for 
software design, testing, implementation and module self-testing of 
the OPRM system provide assurance that no new equipment malfunctions 
due to software errors are created. The design of the OPRM system 
also ensures that neither operation nor malfunction of the OPRM 
system will adversely impact the operation of other systems and no 
accident or equipment malfunction of these other systems could cause 
the OPRM system to malfunction or cause a different kind of 
accident. Therefore, operation with the OPRM system does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Operation in regions currently restricted by the requirements of 
ICAs and TS 3.4.1 is within the nominal operating domain and ranges 
of plant systems and components for which postulated equipment and 
accidents have been evaluated. Therefore operation within these 
regions does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed change which specifies limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRM system and allows operation in certain regions of the power to 
flow [map] does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change specifies limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRM system and allows operation in regions of the power to flow map 
currently restricted by the requirements of ICAs and TS 3.4.1.
    The OPRM system monitors small groups of LPRM signals for 
indication of local variations of core power consistent with T-H 
oscillations and generates an RPS trip when conditions consistent 
with the onset of oscillations are detected. An unmitigated T-H 
instability event has the potential to result in a challenge to the 
MCPR safety limit. The OPRM system provides the capability to 
automatically detect and suppress conditions which might result in a 
T-H instability event and thereby maintains the margin of safety by 
providing automatic protection for the MCPR safety limit while 
significantly reducing the burden on the control room operators. In 
the event the trip capability of the OPRM is not maintained, the 
proposed change includes actions which limit the period of time 
before the effected OPRM channel (or RPS system) must be placed in 
the trip condition. If these actions would result in a trip 
function, an alternate method to detect and suppress thermal 
hydraulic oscillations is required. Since, in either case, the 
duration of this period of time is limited so that the increase in 
the probability of a T-H instability event is not significant. 
Operation with the OPRM system does not involve a significant 
reduction in a margin of safety.
    Operation in regions currently restricted by the requirements of 
ICAs and TS 3.4.1 is within the nominal operating domain assumed for 
identifying the range of initial conditions considered in the 
analysis of anticipated operational occurrences and postulated 
accidents. Therefore, operation in these regions does not involve a 
significant reduction in the margin of safety.
    The proposed change, which specifies limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRM system and allows operation in certain regions of the power to 
flow map, does 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station 
(VCSNS), Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: May 17, 1999.
    Description of amendment request: The proposed amendment would 
change VCSNS Technical Specification 3.7.1.3 ``Condensate Storage 
Tank--Limiting Conditions for Operation'' to revise the tank minimum 
contained water volume from 172,000 gallons to 179,850 gallons.
    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. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    FSAR [Final Safety Analysis Report] 10.4.9.1 states that minimum 
required usable volume for the Condensate Storage Tank (CST) is 
158,570 gallons based on maintaining the plant at HOT STANDBY 
conditions for eleven hours. This volume has already been adjusted 
for both plant uprate conditions and replacement steam generator 
requirements. This change to LCO [Limiting Condition for Operation] 
3.7.1.3 will ensure that 160,054 gallons is maintained in the CST, 
being available and dedicated to the Emergency Feedwater (EFW) 
System. Thus, this change will ensure that the EFW System has an 
adequate water supply to perform its design basis function in regard 
to maintaining the plant in HOT STANDBY condition.
    2. This request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This change increases the minimum required volume of water in 
the CST, thus

[[Page 32291]]

ensuring that the EFW System can perform its required safety 
function. The maximum and normal water levels in the CST are not 
being changed. Therefore, no new failure modes of the CST, or 
flooding concerns are created.
    3. This request does not involve a significant reduction in a 
margin to safety[.]
    This change does not reduce any margin associated with the CST 
inventory available to the EFW. In fact, a small gain in margin 
(less than 1%) is realized by specifying the minimum required volume 
based on the maximum volume available due to nozzle locations and 
other physical characteristics of the tank instead of the minimum 
required to maintain HOT STANDBY for 11 hours. Additionally, the 
requirement for sufficient CST volume to maintain HOT STANDBY for 11 
hours is still met and the Service Water System still provides the 
long term supply of safety grade cooling water to the EFW System. 
The Service Water supply is not affected by this change, and thus 
the margin for safety grade cooling water to the EFW System (or 
safety grade cooling of the RCS) is not affected.

    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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

Southern Nuclear Operating Company, Inc, Docket No. 50-348 Joseph M. 
Farley Nuclear Plant Unit 1, Houston County, Alabama

    Date of amendment request: April 30, 1999.
    Description of amendment request: The proposed amendment would add 
an additional condition to the Farley Nuclear Plant (FNP), Unit 1 
license. This condition would allow cycle 16 operation based on a risk-
informed approach to evaluate steam generator tube structural 
integrity.
    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 do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the FSAR [Final Safety Analysis Report]. The probability of tube 
burst is slightly increased as a result of this proposed amendment 
but is within current industry guidance. Therefore, the probability 
of a previously evaluated accident are not significantly increased. 
There is no change in the FNP design basis as a result of this 
change and, as a result, this change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    The proposed changes to the TSs [technical specifications] do 
not increase the possibility of a new or different kind of accident 
than any accident already evaluated in the FSAR. No new limiting 
single failure or accident scenario has been created or identified 
due to the proposed changes. Safety-related systems will continue to 
perform as designed. The proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed changes do not involve a significant reduction in 
the margin of safety. There is no impact in the accident analyses. 
These proposed changes are technically consistent with the 
requirements of NEI [Nuclear Energy Institute] 97-06, ``Steam 
Generator Program Guidelines,'' Draft Regulatory Guide DG 1074, 
``Steam Generator Tube Integrity,'' and Regulatory Guide (RG) 1.174, 
``An Approach for Using Probabilistic Risk Assessment In Risk-
Informed Decisions on Plant-Specific Changes to the Licensing 
Basis.'' Thus the proposed changes do not involve a significant 
reduction in the margin of safety.
    Accordingly, SNC [Southern Nuclear Operating Company] has 
determined that the proposed amendment to the Facility Operating 
License NPF-2 does not involve a significant hazards consideration.

    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama.
    NRC Section Chief: Richard L. Emch, Jr.

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: May 3, 1999.
    Description of amendment request: The proposed changes will modify 
the Technical Specifications to ensure the emergency ventilation system 
is maintained operable consistent with the assumptions in the 
radiological dose consequence reanalysis from a Large Break Loss-of-
Coolant Accident and to clearly identify that the ventilation system is 
a shared system between the two units.
    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. There is no significant change in the probability or 
consequences of an accident previously evaluated. There are no 
system changes which would increase the probability of occurrence of 
an accident. The dose consequences of the accidents have been 
reviewed, and in some cases the doses at the EAB [exclusion area 
boundary] * * * and the doses to the control room personnel were 
found to increase. However, this increase is not significant because 
the revised doses remain below the limits of 10 CFR 100 and below 
the limits of GDC [General Design Criterion]--19 of Appendix A of 10 
CFR 50.
    2. No new accident types or equipment malfunction scenarios have 
been introduced. Therefore, the possibility of an accident of a 
different type than any evaluated previously in the UFSAR [Updated 
Final Safety Analysis Report] is not created.
    3. There is no significant reduction in the margin of safety, as 
the revised dose calculations for all accidents continue to meet the 
appropriate GDC-19 limits.

    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: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard L. Emch, Jr.

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: May 6, 1999.
    Description of amendment request: The proposed changes will modify 
the Technical Specifications, revising the surveillance frequency for 
the Reactor Trip System (RTS) and Engineered Safety Features Actuation 
System (ESFAS) analog instrumentation

[[Page 32292]]

channels and also revising the allowed outage time and action times for 
the RTS and ESFAS analog instrumentation channels and the actuation 
logic.
    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 Reactor 
Trip System (RTS) and Engineered Safety Features Actuation System 
(ESFAS) Technical Specification changes for the North Anna Units 1 
and 2 and determined that a significant hazards consideration is not 
involved. In support of this conclusion, the following evaluation is 
provided.
    Criterion 1--Operation of North Anna Units 1 and 2 in accordance 
with the proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The determination that the results of the 
proposed changes remain within acceptable criteria was established 
in the SER(s) [Safety Evaluation Reports] prepared for WCAP-10271, 
WCAP-10271 Supplement 1, WCAP-10271 Supplement 2, WCAP-10271 
Supplement 2, Revision 1 and WCAP-14333 issued by letters dated 
February 21, 1985, February 22, 1989, April 30, 1998, and July 15, 
1998.
    Implementation of the proposed changes is expected to result in 
an increase in total RTS and ESFAS yearly unavailability. The 
proposed changes have been shown to result in a small increase in 
the core damage frequency (CDF) due to the combined effects of 
increased RTS and ESFAS unavailability and reduced inadvertent 
reactor trips.
    The values determined by the WOG [Westinghouse Owners Group] and 
presented in the WCAP for the increase in CDF were verified by 
Brookhaven National Laboratory (BNL) as part of an audit and 
sensitivity analyses for the NRC [Nuclear Regulatory Commission] 
Staff. Based on the small value of the increase compared to the 
range of uncertainty in the CDF, the increase is considered 
acceptable. The analysis performed by the WOG and presented in the 
WCAP included changes to the surveillance frequencies for the 
automatic actuation logic and actuation relays and the reactor trip 
and bypass breakers. The overall increase in the CDF, including the 
changes to the surveillance frequencies for the automatic actuation 
logic and actuation relays and the reactor trip and bypass breakers, 
was approximately 6 percent. However, even with this increase, the 
overall CDF remains lower than the NRC safety goal of 
10-4/reactor year.
    Changes to surveillance test frequencies for the RTS and ESFAS 
interlocks do not represent a significant reduction in testing. The 
currently specified test interval for interlock channels allows the 
surveillance requirement to be satisfied by verifying that the 
permissive logic is in its required state using the annunciator 
status light. The surveillance as currently required only verifies 
the status of the permissive logic and does not address verification 
of channel setpoint or operability. The setpoint verification and 
channel operability is verified after a refueling shutdown. The 
definition of the channel check includes comparison of the channel 
status with other channels for the same parameter. The requirement 
to routinely verify permissive status is a different consideration 
than the availability of trip or actuation channels which are 
required to change state on the occurrence of an event and for which 
the function availability is more dependent on the surveillance 
interval. Therefore, the change in the interlock surveillance 
requirement to at least once every 18 months does not represent a 
significant change in channel surveillance and does not involve a 
significant increase in unavailability of the RTS and ESFAS.
    For the additional relaxations in WCAP-14333, the WOG evaluated 
the impact of the additional relaxation of allowed outage times and 
completion times, and action statements on core damage frequency. 
The change in core damage frequency is 3.1 percent for those plants 
with two out of three logic schemes that have not implemented the 
proposed surveillance test interval, allowed outage times, and 
completion times evaluated in WCAP-10271 and its supplements. This 
analysis calculates a significantly lower increase in core damage 
frequency than the WCAP-10271 analysis calculated. This can be 
attributed to more realistic maintenance intervals used in the 
current analysis and crediting the AMSAC [ATWS (anticipated 
transient without scram) mitigating system actuation circuitry] 
system as an alternative method of initiating the auxiliary 
feedwater pumps. Therefore, the overall increase in CDF is estimated 
to be 3.1% for the proposed changes per the generic Westinghouse 
analysis.
    The NRC performed an independent evaluation of the impact on 
core damage frequency (CDF) and large early release fraction (LERF). 
The results of the staff's review indicate that the increase in core 
damage frequency is small (approximately 3.2%) and the large early 
release fraction would increase by only 4 percent for 2 out of 3 
logic schemes that have not implemented the proposed surveillance 
test interval, allowed outage times, and completion times evaluated 
in WCAP-10271 and its supplements. Further, the absolute values for 
CDF still remain within NRC safety goals.
    Therefore, the proposed changes do not result in a significant 
increase in the severity or consequences of an accident previously 
evaluated. Implementation of the proposed changes affects the 
probability of failure of the RTS and ESFAS but does not alter the 
manner in which protection is afforded or the manner in which 
limiting criteria are established.
    Criterion 2--The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes do not result in a change in the manner in 
which the RTS or ESFAS provide plant protection. No change is being 
made which alters the functioning of the RTS or ESFAS (other than in 
a test mode). Rather the likelihood or probability of the RTS or 
ESFAS functioning properly is affected as described above. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident as defined in the Safety Analysis 
Report.
    The proposed changes do not involve hardware changes. Some 
existing instrumentation is designed to be tested in bypass and 
current Technical Specifications allow testing in bypass. Testing in 
bypass is also recognized by IEEE [Institute of Electrical and 
Electronics Engineers] Standards. Therefore, testing in bypass has 
been previously approved and implementation of the proposed changes 
for testing in bypass does not create the possibility of a new or 
different kind of accident from any previously evaluated. 
Furthermore since the other proposed changes do not alter the 
physical operation or functioning of the RTS or ESFAS the 
possibility of a new or different kind of accident from any 
previously evaluated has not been created.
    Criterion 3--The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed changes do not alter the safety limits, limiting 
safety system setpoints or limiting conditions for operation. The 
RTS and ESFAS analog instrumentation remain operable to mitigate as 
assumed in the accident analysis. The impact of reduced testing 
other than as addressed above is to allow a longer time interval 
over which instrument uncertainties (e.g., drift) may act.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety by less frequent testing of the RTS 
and ESFAS analog instruments will result in less inadvertent reactor 
trips and actuation of Engineered Safety Features components.
    This analysis demonstrates that the proposed amendment to The 
North Anna Unit 1 and 2 Technical Specifications does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident and does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.

[[Page 32293]]

    NRC Section Chief: Richard L. Emch Jr.

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. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois and Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: March 22, 1999.
    Brief description of amendments: The amendments modify the 
technical specifications to permit the use of the Gamma-Metrics Post 
Accident Neutron Monitors source range neutron flux detectors in 
addition to the Westinghouse source range neutron flux monitors to 
satisfy the requirement that two source range neutron flux monitors be 
operable during Mode 6 operations (refueling).
    Date of issuance: June 2, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 109 & 109, 102 & 102.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1999 (64 FR 
14944). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 2, 1999.
    No significant hazards consideration comments received: No.
    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.

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: December 2, 1996, as 
supplemented on May 27, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification 3/4.4.2 to reduce the number of required Safety/Relief 
valves (SRVs). This change supports a modification to remove five of 
the currently installed SRVs due to excess capacity and to reduce the 
amount of valve maintenance and associated worker radiation dose. The 
revised TS requires that 12 of the remaining installed 13 SRVs be 
operable.
    Date of issuance: June 3, 1999.
    Effective date: Immediately, to be implemented prior to startup of 
L1C10 for Unit 1 and prior to startup of L2C9 for Unit 2.
    Amendment Nos.: 133 & 118.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4343). The May 27, 1999, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.

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

    Date of application for amendment: March 23, 1999 (NRC-99-0025).
    Brief description of amendment: The amendment revises Technical 
Specification Surveillance Requirement (SR) 4.4.1.1.1 to require each 
recirculation pump discharge valve be demonstrated operable at least 
once every 18 months, deletes the ``*'' footnote from the SR, and 
revises the footnote itself to read ``Not used.''
    Date of issuance: May 25, 1999.
    Effective date: May 25, 1999, with full implementation within 90 
days.
    Amendment No.: 133.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19555)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 25, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

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 application for amendments: July 9, 1998, as supplemented 
March 31, 1999.
    Brief description of amendments: These amendments revised Technical 
Specification (TS) 3/4.7.1.1 and associated Bases for both units. This 
amendment specifies maximum allowable reactor power level based on the 
number of operable main steam safety valves (MSSVs) rather than 
requiring reduction in reactor trip setpoint. This change is consistent 
with the Nuclear Regulatory Commission's improved Standard Technical 
Specifications for Westinghouse plants (NUREG-1431, Revision 1). The 
maximum allowable reactor power level with inoperable MSSVs will be 
calculated based on the recommendations of Westinghouse Nuclear Safety 
Advisory Letter 94-01. The change to the Unit 1 TS 3.7.1.1 also deletes 
reference to 2 loop operation since 2 loop operation is not a licensed

[[Page 32294]]

condition for either unit. Unit 1 TS Table 3.7-3 is then renumbered to 
be Table 3.7-2.
    The March, 31, 1999 letter withdrew a portion of the amendment 
which would have removed the values of the orifice diameter of each 
MSSV from the TSs. This information will be maintained in the TSs.
    Date of issuance: June 3, 1999.
    Effective date: Units 1 and 2 as of date of issuance and shall be 
implemented within 60 days.
    Amendment Nos.: 223 and 99.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43203). The March 31, 1999 letter did not change the initial proposed 
no significant hazards consideration determination or expand the 
amendment beyond the scope of the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendment: August 31, 1998.
    Brief description of amendment: Changes the Crystal River Unit 3 
Technical Specifications to add additional instrumentation variables to 
Improved Technical Specification Table 3.3.17-1, Post-Accident 
Monitoring Instrumentation.
    Date of issuance: June 3, 1999.
    Effective date: As of date of issuance, to be implemented prior to 
commencing cycle 12 operation.
    Amendment No.: 177.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56250).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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

    Date of application for amendment: November 23, 1998, as 
supplemented January 29 and May 7, 1999.
    Brief description of amendment: The amendment changes the Improved 
Technical Specifications for several reactor protection system and 
engineered safeguards actuation system setpoint values, and changes the 
surveillance requirement to verify valve position for valves in the 
high pressure injection system flowpath.
    Date of issuance: May 21, 1999.
    Effective date: As of date of issuance, to be implemented prior to 
commencing Cycle 12 operation.
    Amendment No.: 178.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71966). The supplemental letters dated January 29 and May 7, 1999, 
did not change the original proposed no significant hazards 
consideration determination, or expand the scope of the amendment 
request as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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

    Date of amendment request: December 16, 1998.
    Description of amendment request: These amendments consist of 
changes to the Technical Specifications (TS) in response to Florida 
Power & Light's (FPL) application dated December 16, 1998, regarding 
facility staff qualifications for multi-discipline supervisor (MDS) 
positions at Lucie Units 1 and 2. The amendments revise the 
administrative controls in TS Section 6.3, ``Unit Staff 
Qualifications,'' by modifying FPL's commitment to ANSI/ANS 3.1-1978, 
``Selection and Training of Nuclear Power Plant Personnel,'' to 
incorporate specific staff qualifications for the position of MDS.
    Date of Issuance: May 25, 1999.
    Effective Date: May 25, 1999.
    Amendment Nos.: 161 and 102.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the TS.
    Date of Initial Notice in Federal Register: February 10, 1999 (64 
FR 6698).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 25, 1999.
    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.

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

    Date of application for amendment: November 5, 1998, as 
supplemented February 18, 1999.
    Brief description of amendment: The amendment modifies the safety 
limits and surveillances of the LPRM and APRM systems and related Bases 
pages to ensure the APRM channels respond within the necessary range 
and accuracy and to verify channel operability. In addition, an 
unrelated change to the Bases of Specification 2.3 is included to 
clarify some ambiguous language.
    Date of Issuance: June 2, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 208.
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69342). The February 18, 1999, supplemental letter provided 
clarifying information, was within the scope of the original 
application, and did not change the staff's original no significant 
hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated June 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336, 
and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New 
London County, Connecticut

    Date of application for amendment: December 22, 1998, as 
supplemented March 19, 1999.
    Brief description of amendment: The amendment replaces specific 
titles in Section 6.0 of the Technical

[[Page 32295]]

Specifications of all three Millstone units with generic titles.
    Date of issuance: June 3, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 105, 235, and 171.
    Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: 
Amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4158). The March 19, 1999 letter provided clarifying information that 
did not change the scope of the December 22, 1998, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendments: April 20, 1999.
    Brief description of amendments: The amendments revised the 
implementation date for the relocation of the requirements specified in 
Technical Specification Sections 3.1.E and 5.1 to the Updated Final 
Safety Analyis Report. On December 7, 1998, the NRC had previously 
issued license amendments 141 and 132 for Units 1 and 2, respectively, 
approving the relocation of aforementioned requirements by June 1, 
1999. The proposed amendments would postpone the implementation date to 
September 1, 1999.
    Date of issuance: June 2, 1999.
    Effective date: June 2, 1999, with full implementation within 30 
days .
    Amendment Nos.: 145 and 136.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 29, 1999 (64 FR 
23131) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating 
Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: January 4, 1999.
    Brief description of amendments: These amendments revise the 
administrative section of the Technical Specification pertaining to 
controlled access to high radiation areas, and the reporting dates for 
the annual occupational radiation exposure report and the annual 
radioactive effluent release report.
    Date of issuance: May 24, 1999.
    Effective date: Units 1 and 2, as of date of issuance and shall be 
implemented within 30 days.
    Amendment Nos.: 135 and 100.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6706) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

Power Authority of the State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County , New York

    Date of application for amendment: January 25, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by relocating certain requirements from the TSs to 
the Final Safety Analysis Report.
    Date of issuance: May 24, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 189.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19562).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit 
1, Luzerne County, Pennsylvania

    Date of application for amendment: March 12, 1999.
    Brief description of amendment: This amendment would change the 
allowable values for both the core spray system and the low pressure 
coolant injection system reactor steam dome pressure-low functions.
    Date of issuance: May 25, 1999.
    Effective date: As of date of issuance, and shall be implemented 
within 30 days after startup from the Unit 1 eleventh refueling and 
inspection outage currently scheduled for spring 2000.
    Amendment No.: 181.
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17028).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 25, 1999.
    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.

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

    Date of application for amendment: October 27, 1998, as 
supplemented by letters in 1999 dated January 11, January 29, February 
25, and April 7 (two letters), and May 17.
    Brief description of amendment: The amendment revised Technical 
Specification 4.4.5.4, Table 4.4-3 and the associated Bases to allow 
the repair of the steam generator tubes with the Electrosleeve tube 
repair method.
    Date of issuance: May 21, 1999.
    Effective date: May 21, 1999, to be implemented within 30 days from 
the date of issuance. The amendment includes a two cycle operating 
limit that requires all steam generator tubes repaired with 
Electrosleeves to be removed from service at the end of two operating 
cycles following installation of the first Electrosleeve in the steam 
generators.
    Amendment No.: 132.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66604). The supplemental letters in 1999 dated January 11, January 29, 
February 25, and April 7 (two letters)

[[Page 32296]]

provided additional clarifying information that did not expand the 
staff's original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia Missouri 65201.

    Dated at Rockville, Maryland, this 9th day of June 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-15098 Filed 6-15-99; 8:45 am]
BILLING CODE 7590-01-P