[Federal Register Volume 62, Number 29 (Wednesday, February 12, 1997)]
[Notices]
[Pages 6566-6587]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-20212]


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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

[[Page 6567]]

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 January 17, 1997, through January 31, 1997. 
The last biweekly notice was published on January 29, 1997 (62 FR 
4341).

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 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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 March 14, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no

[[Page 6568]]

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: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: November 26, 1996
    Description of amendment request: The proposed amendment would 
change the definition of ``Primary Containment Integrity,'' Note 6 on 
Table 3.2.A, correct a typographical error on Table 3.2 D, correct 
Table 3.2.F to reflect modifications to the plant and changes to Bases 
sections 3/4.6G and 3/4.7.A. These changes are considered 
administrative and have no effect on plant design, safety limit 
settings or plant system operation.
    Basis for proposed no significant hazards 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 amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed administrative changes involving typographical 
errors, additions for clarity and consistency, and updating the 
Bases do not affect plant design, safety limit settings, or plant 
system operation and, therefore, do not modify or add any initiating 
parameters that would significantly increase the probability or 
consequences of any previously analyzed accident.
    The changes to instrument numbers and type do not change the 
parameters being surveyed or the number of operable channels for 
these parameters. These changes do not modify or add any initiating 
parameters and do not affect plant design, safety limit settings, or 
plant system operation. Therefore, these instrument changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    These proposed changes do not involve any potential initiating 
events that would create any new or different kind of accident. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    These changes do not affect any safety analysis assumptions, 
system operation, structures, potential initiating events or safety 
limits. Therefore, it is concluded that the proposed amendment 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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Patrick D. Milano, Acting

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: January 24, 1997
    Description of amendment request: The proposed amendment will 
update the Safety Limit Minimum Critical Power Ratio (SLMCPR) in 
Technical Specification (TS) 2.1.2 and the associated Bases section to 
reflect the results of the latest cycle-specific calculation performed 
for the Pilgrim Nuclear Power Station Operating Cycle 12. In addition, 
the values provided in Note 5 of Table 3.2.C.1, which are based on the 
SLMCPR values, have been revised as a result of the changes to the 
SLMCPR value.
    Basis for proposed no significant hazards 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: 11.The proposed technical specification changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The derivation of the revised SLMCPR for Pilgrim for 
incorporation into the TS, and its use to determine cycle-specific 
thermal limits, have been performed using NRC approved methods. 
Additionally, interim implementing procedures that incorporate 
cycle-specific parameters have been used which result in a more 
restrictive value for SLMCPR. These calculations do not change the 
method of operating the plant and have no effect on the probability 
of an accident initiating event or transient.
    The basis of the MCPR [minimum critical power ratio] Safety 
Limit is to ensure no mechanistic fuel damage is calculated to occur 
if the limit is not violated. The new SLMCPR preserves the existing 
margin to transition boiling, and the probability of fuel damage is 
not increased.
    The basis of the MCPR criteria that define a limiting rod 
pattern is to ensure the SLMCPR is not violated in the event a 
control rod is fully withdrawn from the core. The new MCPR criteria 
that define a limiting rod pattern continue to ensure the SLMCPR is 
not violated in the event a control rod is fully withdrawn from the 
core. These new criteria do not change the method of operating the 
plant and have no effect on the probability of an accident 
initiating event or a transient.

[[Page 6569]]

    Therefore, the proposed changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes result only from a revised method of 
analysis for the Cycle 12 core reload. These changes do not involve 
any new method for operating the facility and do not involve any 
facility modifications. No new initiating events or transients 
result from these changes. Therefore, the proposed TS changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPR is calculated using NRC approved methods which 
are in accordance with the current fuel design and licensing 
criteria. Additionally, interim implementing procedures, which 
incorporate cycle-specific parameters, have been used. The SLMCPR 
remains high enough to ensure that greater than 99.9% of all fuel 
rods in the core will avoid transition boiling if the limit is not 
violated, thereby preserving the fuel cladding integrity.
    The new MCPR criteria that define a limiting rod pattern 
continue to ensure the SLMCPR is not violated in the event a control 
rod is fully withdrawn from the core.
    Therefore, the proposed TS changes do not involve a reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Patrick D. Milano, Acting

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of amendment request: January 29, 1997
    Description of amendment request: The proposed change adds a new 
entry 3.0.5 to the plant Technical Specifications (TS) to provide 
specific guidance for returning equipment to service under 
administrative control for the sole purpose of performing testing to 
demonstrate operability.
    Basis for proposed no significant hazards 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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change does not affect the operation or design of 
the plant in any way. Operation of plant equipment under this change 
will not differ in any way from its normal operational mode. The 
normal operation of plant equipment is not a precursor to any 
accident. The purpose of tests performed using this change are to 
demonstrate that required automatic actions are carried out. 
Equipment will be operated under administrative control for only a 
short period of time. Personnel will be immediately available to 
take appropriate manual action if it should be required. Therefore 
operation of equipment under this change is not expected to increase 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed testing allowance does not involve any physical 
alterations or additions to plant equipment or alter the manner in 
which any safety-related system performs its function. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3.
    The proposed amendment does not involve a signifcant reduction in 
the margin of safety.
    Equipment will be operated under administrative control for only 
a short period of time. Personnel will be immediately available to 
take appropriate manual action if it should be required. The purpose 
of the testing is to restore required equipment to an OPERABLE state 
which increases the automatic protection available and reduces the 
reliance on the compensatory measures provided by ACTION statements. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Mark Reinhart, Acting

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: October 24, 1996
    Description of amendment request: The proposed amendment to the 
Perry Nuclear Power Plant Technical Specifications revises those 
specifications associated with the Minimum Critical Power Ratio (MCPR) 
Reactor Core Safety Limit. The revision would increase the MCPR Safety 
Limit values to make them more conservative.
    Basis for proposed no significant hazards determination: The NRC 
staff provides its analysis of the issue of no significant hazards 
consideration below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    There is no change to any plant equipment, and increasing the 
MCPR Safety Limit is more conservative. Therefore, the proposed 
change does not significantly increase the probability or 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no physical changes to the plant, and increasing the MCPR 
Safety Limit is more conservative. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed MCPR Safety Limit values are more conservative, and 
were calculated using NRC approved methods. Therefore, the proposed 
change does not involve a significant reduction in a margin of safety.
    The staff has reviewed the amendment request and the licensee's no 
significant hazards consideration determination. Based on the review 
and the above discussions, the staff proposes to determine that the 
proposed changes do not involve a significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

[[Page 6570]]

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

    Date of amendment request: August 19, 1996
    Description of amendment request: The proposed amendments revise 
the steam generator (SG) repair criteria in the Byron, Unit 1, and 
Braidwood, Unit 1, Technical Specifications (TS). These revisions, if 
approved, would continue the use of the voltage-based SG tube repair 
criteria added by Amendment No. 77, dated November 9, 1995, to the 
Byron 1 TSs and by Amendment No. 69, dated November 9, 1995, to the 
Braidwood 1 TSs. The subject voltage-based repair criteria are 
applicable only for a specific form of SG tube degradation identified 
as outer diameter stress corrosion cracking (ODSCC), which is confined 
entirely within the thickness of the SG tube support plates (TSPs). 
Specifically, the pending amendments for both units would continue for 
one more operating cycle, the present use of a lower voltage repair 
limit of 3.0 volts on the hot leg side of the SGs using the Locked-Tube 
model. The cold leg side of the SGs and certain hot leg side tube/TSP 
intersections (e.g., dented SG tube intersections) would continue to be 
repaired using the Free-Span model. The proposed amendments are needed 
because the applicability of the revised voltage-based SG tube repair 
criteria for ODSCC which were added in the prior amendments cited 
above, was limited to only one full operating cycle for Braidwood 1 
ending in spring 1997 and for the operating cycle ending in late 1997 
for Byron 1.
    Additionally, the inspection and reporting requirements added to 
the Byron 1 and Braidwood 1 TSs by the prior amendments cited above, 
would also be continued for one more operating cycle for both units. 
The maximum permissible value of the iodine-131 concentration in the 
primary coolant in both the Byron 1 and Braidwood 1 TSs remains 
unchanged at 0.35 microcuries per gram of coolant. Finally, the Bases 
sections in the Byron 1 and Braidwood 1 TSs are proposed to be revised 
to introduce the terminology associated with the Locked-Tube SG tube 
model and that of the Free-Span model.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This amendment request proposes to renew the SG tube plugging/
repair criteria previously approved by the NRC in Amendments 69 and 
77 to Braidwood and Byron Technical Specifications, respectively.
    The previously evaluated applicable accidents are steam 
generator tube burst and Main Steam Line Break (MSLB). The 
postulated MSLB outside of containment but upstream of the Main 
Steam Isolation Valve (MSIV) represents the most limiting 
radiological condition relative to the IPC. The potential impact on 
public health and safety as a result of renewing the SG tube interim 
plugging criteria contained in the current Braidwood and Byron 
Technical Specifications is very low as discussed below. Tube burst 
due to predominantly axially oriented ODSCC at the TSP intersections 
is precluded during normal operating plant conditions since the tube 
support plates are adjacent to the degraded regions of the tube in 
the tube-to-tube support plate crevices.
    During accident conditions, i.e., MSLB, the tubes and TSP may 
move relative to each other. This can expose the crack length 
portion to free-span conditions. Testing has shown that the burst 
pressure correlates to the crack length that is exposed to the free-
span, regardless of the length that is still contained within the 
TSP bounds.
    Therefore, a more appropriate methodology has been established 
for addressing leakage and burst considerations. This methodology is 
based on limiting potential TSP displacements (Locked-Tube Model 
Intersections) during postulated MSLB events, thus reducing the 
free-span exposed crack length to minimal levels. The tube expansion 
process employed in conjunction with this tube plugging criteria is 
designed to provide postulated TSP displacements that result in 
negligible tube burst probabilities due to the minimal free-span 
exposed crack lengths. The tube expansions were performed during the 
first outage that the 3.0 volt IPC was applied (Braidwood refuel 
outage A1R05 -Fall 1995, and Byron midcycle outage B1P02 - Fall 
1995). These expansions will be inspected in accordance with an eddy 
current inspection probe that is sensitive to axial and 
circumferential indications. This program will ensure the integrity 
of the expansions for the additional cycle of operation. It has been 
demonstrated that axial indications in the expansion region will not 
result in a reduction of the load carrying capability of the 
expanded tubes.
    Thermal hydraulic modeling was used to determine TSP loading 
during MSLB conditions. A safety factor was conservatively applied 
to these loads to envelope the collective uncertainties in the 
analyses. Various operating conditions were evaluated and the most 
limiting operating condition was used in the analyses. Additional 
models were used to verify the thermal hydraulic results.
    Assessment of the tube burst probability for the Locked-Tube 
Model Intersections was based on a conservative assumption that all 
hot-leg TSP intersections (32,046) contained through wall cracks 
equal to the postulated TSP displacement and that the crack lengths 
were located within the boundaries of the TSP. Alternatively, it was 
assumed that all hot-leg TSP intersections contained through wall 
cracks with lengths equal to the thickness of the TSP. The 
postulated TSP motion was conservatively assumed to be uniform and 
equal to the maximum displacement calculated.
    The total burst probability for all 32,046 through wall 
indications, given a uniform MSLB TSP displacement of 0.31 inches, 
was calculated to be 1x10<SUP>-5. This is a factor of 1000 less than 
the GL 95-05 burst probability limit of 1x10<SUP>-2. Therefore, the 
functional design criteria for tube expansion was to limit the TSP 
motion to 0.31'' or less. However, the design goal for tube 
expansion limits the TSP MSLB motion to less than 0.1''. This design 
goal results in a total tube burst probability of 1x10<SUP>-10 for 
all 32,046 postulated through wall indications. Additional tubes 
were expanded to provide redundancy for the required expansions.
    The structural limit for the Locked-Tube Model Intersection SG 
tube repair criteria was based on axial tensile loading requirements 
to preclude axial tensile severing of the tube. Axially oriented 
ODSCC does not significantly impact the axial tensile loading of the 
tube. Based on the current voltage distributions and growth rates, 
Monte Carlo projections were performed for Braidwood Unit 1 and 
Byron Unit 1 for the additional cycle of operation that this 
proposed amendment is requesting. The End of Cycle (EOC) voltage 
projections for Braidwood Unit 1 Cycle 7 predict that the maximum 
voltage to be seen will be less than 10.5 volts. The number of 
indications predicted greater than ten volts at the end of Cycle 7 
for Braidwood Unit 1 is 0.3. The EOC voltage projections for Byron 
Unit 1 Cycle 9 predict that the maximum voltage to be seen will be 
less than 13.5 volts. The number of indications predicted greater 
than ten volts at the end of Cycle 9 for Byron Unit 1 is 4.59.
    Using a tensile rupture probability for a ten volt indication of 
3x10<SUP>-6, the probability of tensile rupture from the predicted 
0.3 indications at Braidwood is 1-(1-3x10<SUP>-6)<SUP>0.3 = 
9.0x10<SUP>-7. The probability of tensile rupture from the predicted 
4.59 indications at Byron is 1-(1-3x10<SUP>-6)<SUP>4.59 = 
1.38x10<SUP>-5. Both of these probabilities result in a negligible 
contribution to the total burst probability when compared to the 
1x10<SUP>-2 GL 95-05 limit.
    Cellular corrosion is a more limiting mode of degradation at the 
TSPs with respect to affecting the tube structural limit. Tensile 
tests that measure the force required to sever a tube with cellular 
corrosion and uncorroded cross sectional areas are used to establish 
the lower bound structural limit. Based upon these tests, a lower 
bound 95% confidence level structural voltage limit of 37 volts was 
established for cellular corrosion. This limit meets the Regulatory 
Guide (RG) 1.121, ``Basis for Plugging Steam Generator Tubes,'' 
structural requirements based upon the normal operating pressure 
differential with a safety factor of 3.0 applied. Due to the limited 
database supporting this value, the

[[Page 6571]]

structural limit was conservatively reduced to 20 volts. Accounting 
for voltage growth and Non-Destructive Examination (NDE) 
uncertainty, the full IPC upper limit exceeds ten volts. However, 
for added conservatism a single voltage repair limit of 3.0 volts 
for the Locked-Tube Model Intersection indications is specified in 
the current plugging/repair criteria. All indications at the Locked 
Tube Model Intersections with bobbin coil probe voltages greater 
than 3.0 volts will be plugged or repaired.
    The free-span tube burst probability must be calculated for the 
indications at the Free-Span Model Intersections. The total burst 
probability must be within the requirements of GL 95-05. The free-
span structural voltage limit is calculated using correlations from 
the database described in GL 95-05, with the inclusion of the recent 
Byron, Braidwood, and South Texas tube pull results. The structural 
limit for the Free-Span Model Intersections is 4.745 volts. The 
lower voltage repair limit for the indications at the Free-Span 
Model Intersections continues to be 1.0 volt. The upper voltage 
repair limit for the indications at the Free-Span Model 
Intersections will be calculated in accordance with GL 95-05.
    Since IPC will not be applied to indications at the Flow 
Distribution Baffle (FDB), no leakage or burst analyses are required 
for these indications.
    Per GL 95-05, MSLB leak rate and tube burst probability analyses 
are required to be performed prior to returning the unit to power. 
The results of these analyses are to be included in a report to the 
NRC within 90 days of restart. If allowable limits on leak rates and 
burst probability are exceeded, the results are to be reported to 
the NRC and a safety assessment of the significance of the results 
is to be performed prior to returning the SGs to service.
    A site specific calculation has determined the site allowable 
leakage limit for Braidwood and Byron. These limits use the 
recommended Dose Equivalent Iodine-131 transient spiking values 
consistent with NUREG-0800, ``Standard Review Plan'' and ensure site 
boundary doses are within a small fraction of the 10 CFR 100 
requirements.
    The projected leakage rate calculation methodology described in 
WCAP-14046, ``Braidwood Unit 1 Technical Support for Cycle 5 Steam 
Generator Interim Plugging Criteria,'' and WCAP-14277, ``SLB Leak 
Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP 
Intersections,'' will be used to calculate the EOC leakage. This 
method includes a Probability of Detection (POD) value of 0.6 for 
all voltage amplitude ranges and uses the accepted leak rate versus 
bobbin voltage correlation methodology (full Monte Carlo) for 
calculating the leak rate, as described in GL 95-05. The database 
used for the leak and burst correlations is consistent with that 
described in GL 95-05 with the inclusion of the Byron Unit 1, 
Braidwood Unit 1, and South Texas tube pull results. The EOC voltage 
distribution is developed from the POD adjusted beginning-of-cycle 
(BOC) voltage distributions and uses Monte Carlo techniques to 
account for variances in growth and NDE uncertainty.
    The Electric Power Research Institute (EPRI) leak rate 
correlation has been used. This correlation is based on free-span 
indications that have burst pressures above the MSLB pressure 
differential. There is a low but finite probability that indications 
may burst at a pressure less than MSLB pressure. With limited TSP 
motion for the Locked-Tube Model Intersections, the tube is 
constrained by the TSP and tube burst is precluded. However, the 
flanks of the crack may open up to contact the Inside Diameter (ID) 
of the TSP hole and result in a primary-to-secondary leak rate 
potentially exceeding that obtained from the EPRI correlation. This 
phenomenon is known as an Indication Restricted from Burst (IRB) 
condition.
    ComEd has performed laboratory testing to determine the bounding 
leak rate obtainable in an IRB condition (6.0 gallons per minute). 
The bounding leak rate value was then applied to a leak rate 
calculation methodology that accounts for the MSLB leak rate 
contribution from IRB indications to the total leak rate calculated 
as described above. Results indicate that the IRB contribution to 
the total leak rate value is negligible. However, ComEd will 
conservatively add a leakage contribution due to IRBs in addition to 
the leakage calculated in accordance with GL 95-05. When this is 
done, the dose at the site boundary resulting from the predicted 
leakage will be a small fraction (less than 10%) of the 10 CFR 100 
limits.
    Modification of the Braidwood and Byron TS to clarify 
application of the proposed tube plugging/repair criteria is purely 
administrative and will not have any effect on the probability or 
consequences of an accident previously evaluated.
    Operating experience over the last cycle with this plugging 
criteria applied has not revealed any unpredicted or unusual 
effects.
    For these reasons, renewal of the current Braidwood and Byron 
tube plugging criteria does not adversely affect SG tube integrity 
and results in acceptable dose consequences. By effectively 
eliminating tube burst at the Locked-Tube Model TSP intersections, 
the likelihood of a tube rupture is substantially reduced and the 
probability of occurrence of an accident previously evaluated is 
reduced.
    This conclusion is not affected by foreign or domestic plant SG 
experiences (NRC Information Notice 96-09 and its supplement). As 
the following evaluation shows, these experiences are not relevant 
to Braidwood or Byron.
    A foreign unit detected eddy current signal distortions in one 
area of the top TSP during a 1995 inspection. The steam generators 
had been chemically cleaned in 1992. Visual inspection showed that a 
small section of the top TSP had broken free and was resting next to 
the steam generator tube bundle wrapper. The support plate showed 
indications of metal loss.
    The chemical cleaning process used by the foreign unit was 
developed by the utility and differs significantly from the modified 
EPRI/SGOG process performed at Byron Unit 1 in 1994. The foreign 
chemical cleaning process, coupled with the specific application of 
the process, resulted in TSP corrosion of up to 250 mils compared to 
a maximum of 2.16 mils (11 mils maximum allowed) measured at Byron. 
During the Byron eddy current inspection performed after the 
chemical cleaning, no distortion of the tube support plate signals 
was reported. Therefore, these differences in cleaning processes 
imply that this foreign experience is irrelevant to the effects of 
the chemical cleaning process on the TSPs at Byron. Chemical 
cleaning of the SGs has not occurred at Braidwood.
    A number of units have experienced TSP cracking associated with 
severe tube denting due to TSP corrosion at the tube-to-TSP crevice. 
WCAP-14273, Section 12.4, shows that a diametral reduction of a SG 
tube of 0.065 inches is required to develop stress levels above 
yield in the TSP ligaments at dented intersections. The bobbin 
voltage range associated with a one mil radial dent is twenty to 
twenty-five volts.
    Although Braidwood Unit 1 and Byron Unit 1 have not seen 
corrosion induced denting, a 0.610 inch diameter bobbin coil probe 
will be used as a go/no-go gauge to assess dents at the Locked-Tube 
Model Intersections, if they occur in the future. If a tube has a 
dent at a Locked-Tube Model TSP intersection that fails to pass the 
go/no-go test probe, IPC will not be applied to that intersection. 
In addition, if the dent is determined to be corrosion induced, the 
Free-Span Model repair criteria will be applied to the intersections 
adjacent to the dented intersection. IPC repair limits will not be 
applied to tubes with dents greater than 5.0 volts since dent 
signals of this magnitude could mask a 1.0 volt ODSCC signal. Tube 
intersections with corrosion induced dents greater than 5.0 volts 
and the intersections adjacent to such an intersection were not 
selected for tube expansion to preclude adverse effects of the 
failure of such a tube on limiting TSP displacement. If corrosion 
induced denting, either greater than 5.0 volts or such that the tube 
is unable to pass a 0.610 inch diameter bobbin coil probe, are 
detected at an intersection adjacent to an expanded intersection, 
the dented intersection will be inspected by an EPRI developed 
technique to determine if the TSP is cracked. If a crack-like 
indication is identified in a TSP, a plus point inspection will be 
conducted per the EPRI TSP program. If the plus point inspection 
verifies the existence of a crack-like indication, the effect of 
that indication on TSP displacement will be evaluated. If this 
evaluation shows that TSP displacement would be greater than 0.1 
inches during a MSLB event, the effected area will either be 
mechanically corrected or the Free-Span Model criteria will be 
applied to the affected area. Based on the information presented 
above, the SG tube denting experience at other plants is not 
relevant to Braidwood or Byron.
    A foreign utility's SGs have experienced cracking at the top 
TSP. The cause of the cracking appears to be the configuration of 
the single anti-rotation device, connected between the SG shell and 
wrapper, and the wrapper internals. The single anti-rotation device 
carries the full load associated with the wrapper to shell motion. 
This rotational load is believed to be transferred to the TSP via 
the wrapper internals. The Byron/Braidwood Unit 1 SG design (D-4) 
uses three anti-rotation devices to spread the rotational load. The 
D-4 wrapper internals are configured such that this load is not 
directly transmitted to the TSP.

[[Page 6572]]

    No top TSP cracking has been detected at Braidwood Unit 1 or 
Byron Unit 1 and very few (<1%) of the ODSCC indications in the SG 
tubes at Braidwood and Byron, to date, have been at the top TSP 
elevation. Nevertheless, an analysis was performed to assess the 
impact of cracking of the top TSP. The results show an increase in 
the deflection of the top TSP for a very limited number of tubes to 
greater than the 0.10'' limit used in the 3.0 volt IPC analysis. The 
deflections of the lower support plates also increased, but remain 
within the 0.10'' limit. Thus, a large majority of the Locked-Tube 
Model indications continue to be bounded by the existing analysis 
even with a cracked top TSP. The Locked-Tube Model repair criteria 
will not be applied to any SG tube ODSCC indication where the TSP 
has been shown to be displaced by more than 0.1 inches during 
accident conditions.
    In response to these experiences at foreign and domestic 
utilities, ComEd developed an inspection plan for the SG internals 
to identify if indications detrimental to the load path components 
existed. This inspection plan was carried out at Braidwood during 
refueling outage A1R05 (Fall 1995) and at Byron during the midcycle 
outage B1P02 (Fall 1995) and refuel outage B1R07 (Spring 1996). 
These inspections revealed no degradation of the SG load path 
components necessary to support implementation of the 3.0 volt IPC. 
Inspections will be performed during the upcoming refuel outages at 
Braidwood Unit 1 and Byron Unit 1 to further ensure the integrity of 
the SG load path components necessary to support implementation of 
the 3.0 volt IPC.
    A domestic utility reported several distorted TSP signals over 
the past three refueling outages' SG tube inspections. It was 
determined that these signals were associated with the TSP geometry 
in an area where an access cover is welded to the TSP. These signal 
distortions are not attributed to TSP cracking or degradation. Since 
the distorted signals were due to TSP geometry which did not 
indicate or result in a defect of the TSP, there is no increase in 
the probability or consequences of an accident previously evaluated 
due to Braidwood Unit 1 and Byron Unit 1 steam generator TSP 
geometries which may result in distorted eddy current signals.
    One foreign unit observed a dislocation of the tube bundle 
wrapper when they were unable to pass sludge lancing equipment 
through a hand hole in the wrapper. The dislocation appears to be a 
result of improper attachment of the wrapper to the support 
structure. SG sludge lance operations have been successfully 
performed at Braidwood Unit 1 and Byron Unit 1 which indicates that 
no problem with the wrapper attachment exists. The foreign unit's 
wrapper support design is significantly different than that used on 
Braidwood Unit 1 and Byron Unit 1. Therefore, a similar wrapper 
dislocation will not occur and the foreign experience is not 
applicable to Braidwood or Byron. An inspection was conducted during 
the last Braidwood Unit 1 and Byron Unit 1 refueling outages which 
verified this conclusion.
    ComEd will continue to apply a maximum primary-to-secondary 
leakage limit of 150 gallons per day (gpd) through any one SG at 
Braidwood and Byron to help preclude the potential for excessive 
leakage during all plant conditions. The RG 1.121 criterion for 
establishing operational leakage limits that require plant shutdown 
are based on detecting a free-span crack prior to it resulting in 
primary-to-secondary operational leakage which could potentially 
develop into a tube rupture during faulted plant conditions. The 150 
gpd limit provides for leakage detection and plant shutdown in the 
event of an unexpected single crack leak associated with the longest 
permissible free-span crack length.
    Therefore, the proposed amendment does not result in any 
significant increase in the probability or consequences of an 
accident previously evaluated within the Braidwood and Byron Updated 
Final Safety Analysis Report (UFSAR).
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This amendment request proposes to renew the SG tube plugging/
repair criteria previously approved by the NRC in Amendments 69 and 
77 to Braidwood and Byron Technical Specifications, respectively.
    Renewal of the proposed steam generator tube plugging criteria 
with tube expansion does not introduce any significant changes to 
the plant design basis. Use of the criteria does not provide a 
mechanism which could result in an accident outside of the region of 
the tube support plate elevations as ODSCC does not extend beyond 
the thickness of the tube support plates and IPC is not allowed to 
be applied to indications that extend beyond the thickness of the 
tube support plate. Neither a single nor multiple tube rupture event 
would be expected in a SG in which the plugging criteria has been 
applied.
    The tube burst assessment involves a Monte Carlo simulation of 
the site specific voltage distribution to generate a total burst 
probability that includes the summation of the probabilities of one 
tube bursting, two tubes bursting, etc. For the Locked-Tube Model 
TSP Intersections, the maximum total probability of burst, by 
design, is estimated to be 1x10<SUP>-10 with all tube expansions 
functional. The burst probability for the Free-Span Model TSP 
intersections will be dependent on the number and size of 
indications at these applicable intersections. The total burst 
probability will be within the limit specified in GL 95-05.
    Accounting for the unlikely event of a failure of the expanded 
tubes, a sufficient number of redundant expansions exist to ensure 
that the burst probability remains below 1x10<SUP>-5. This includes 
the conservative assumption that all 32,046 hot-leg TSP 
intersections contain through wall indications. This level of burst 
probability is considered to be negligible when compared to the GL 
95-05 limit of 1x10<SUP>-2.
    In addressing the combined effects of a Loss Of Coolant Accident 
(LOCA) during a Safe Shutdown Earthquake (SSE) on the SG as required 
by General Design Criteria (GDC) 2, it has been determined that tube 
collapse may occur in the steam generators at some plants. The tube 
support plates may become deformed as a result of lateral loads at 
the wedge supports located at the periphery of the plate due to the 
combined effects of the LOCA rarefaction wave and SSE loadings. The 
resulting pressure differential on the deformed tubes may cause some 
of the tubes to collapse. There are two issues associated with SG 
tube collapse. First, the collapse of SG tubing reduces the Reactor 
Coolant System (RCS) flow area through the tubes. The reduction in 
flow area increases the resistance to flow of steam from the core 
during a LOCA which, in turn, may potentially increase the Peak Clad 
Temperature (PCT). Second, there is a potential that partial through 
wall cracks in the SG tubes could progress to through wall cracks 
during tube deformation or collapse. The tubes subject to collapse 
have been identified via a plant specific analysis and are excluded 
from application of any voltage-based criteria. This analysis is 
included in revision 3 to WCAP-14046 which was submitted to the NRC 
June 19, 1995.
    Modification of the Braidwood and Byron Technical Specifications 
to clarify application of the proposed tube plugging/repair criteria 
is purely administrative and will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Operating experience over the last cycle with this plugging 
criteria applied has not revealed any unpredicted or unusual 
effects.
    SG tube integrity will continue to be maintained following 
renewal of the 3.0 volt IPC voltage repair limit through inservice 
inspection, tube repair and primary-to-secondary leakage monitoring. 
By effectively eliminating tube burst at the Locked-Tube Model TSP 
Intersections, the potential for multiple tube ruptures is 
essentially eliminated.
    ComEd has evaluated industry experiences with TSP degradation, 
eddy current signal distortions, and component misalignment. Eddy 
current signal distortions due to TSP geometry are not indicative of 
TSP degradation and do not result in any kind of new or different 
accident.
    The component misalignment experienced by one unit is not 
applicable to Braidwood Unit 1 or Byron Unit 1 and, thus, will not 
result in any kind of new or different accident. Specific 
limitations, as discussed in response to Question 1, will be applied 
to indications at the Locked-Tube Model Intersections which contain 
dents. These limitations ensure that the integrity of the SG tubes 
is maintained consistent with the current analyses should tube 
denting or TSP cracking occur.
    Therefore, renewal of the current tube plugging/repair criteria 
at Braidwood Unit 1 and Byron Unit 1 will 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 use of the voltage-based, bobbin coil, tube support plate 
plugging criteria with tube expansion at Braidwood Unit 1 and Byron 
Unit 1 is demonstrated to maintain SG tube integrity commensurate 
with the criteria of RG 1.121. RG 1.121 describes a method

[[Page 6573]]

acceptable to the NRC staff for meeting GDC 14, 15, 31, and 32 by 
reducing the probability or the consequences of steam generator tube 
rupture.
    Reducing the probability or the consequences of steam generator 
tube rupture is accomplished by determining an eddy current 
inspection voltage value which represents a limit for leaving an 
axial, crack-like indication at an in service SG tube TSP 
intersection. Tubes with ODSCC voltage indications beyond this 
limiting value must be removed from service by plugging or repaired 
by sleeving. Implementation of a 3.0 volt IPC voltage repair limit 
for the Locked-Tube Model Intersections has been evaluated and shown 
not to present a credible potential for a steam generator tube 
rupture event during normal or faulted plant conditions, even with 
worst case assumptions. The total tube burst probability will 
include a contribution from the indications at the Locked-Tube Model 
Intersections and from indications at the Free-Span Model 
Intersections. The projected EOC voltage distribution of crack-like 
indications at the TSP elevations will be confirmed to result in 
acceptable primary-to-secondary leakage during all plant conditions 
such that radiological consequences are not adversely impacted.
    Addressing RG 1.83 considerations, implementation of the 
increased Locked-Tube Model Intersection bobbin coil voltage-based 
repair criteria is supplemented by enhanced eddy current inspection 
guidelines to provide consistency in voltage normalization and a 
100% eddy current inspection sample size at the affected TSP 
elevations.
    For the leak and burst assessments, the population of 
indications in the EOC voltage distribution is dependent on the POD 
function. The purpose of the POD function is to account for new 
indications that may develop over the cycle, and to account for 
indications not identified by the data analyst. In implementing this 
proposed IPC renewal, ComEd will continue to use the conservative GL 
95-05 POD value of 0.6 for all voltage amplitude ranges.
    Modification of the Braidwood and Byron Technical Specifications 
to clarify application of the proposed tube plugging/repair criteria 
is purely administrative and will not reduce any safety margins.
    Operating experience over the last cycle with this plugging 
criteria applied has not revealed any unpredicted or unusual 
effects.
    Implementation of the TSP elevation repair limits will decrease 
the number of tubes which must be repaired. Installation of steam 
generator tube plugs or sleeves reduces the RCS flow margin. Thus, 
implementation of the IPC will maintain the margin of flow that 
would otherwise be reduced in the event of increased tube plugging 
or sleeving.
    As discussed previously, ComEd has evaluated industry 
experiences with TSP degradation, eddy current signal distortions, 
and component misalignment. Eddy current signal distortions at tube 
support plates will be evaluated to attempt to determine the cause 
of the distortion. A signal distortion alone will not result in 
reduction in the margin of safety. The foreign unit that experienced 
the component misalignment was of a significantly different design 
than the Braidwood Unit 1 and Byron Unit 1 steam generators. 
Analysis of the design differences shows that component misalignment 
of that type is not applicable to Braidwood Unit 1 or Byron Unit 1 
and, thus, will not result in a reduction in the margin of safety. 
An inspection was conducted during the last Braidwood Unit 1 and 
Byron Unit 1 refueling outages which verified this conclusion.
    Specific limitations, as discussed previously, will be applied 
to indications at the Locked-Tube Model Intersections which contain 
dents. These limitations conservatively treat indications as free-
span to ensure that the integrity of the SG tubes is maintained 
consistent with current analyses should tube denting or TSP cracking 
occur. Application of the 3.0 volt Locked-Tube Model Intersection 
IPC and the 1.0 volt Free-Span Model Intersection IPC at Braidwood 
Unit 1 and Byron Unit 1, with the limitations specified, will not 
result in a reduction in a margin of safety.
    Thus, the implementation of this amendment does not result in 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendment request: January 6, 1997
    Description of amendment request: The proposed amendment would 
clarify and maintain consistency between the operability requirements 
for protective instrumentation and associated automatic bypass 
features.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
1) Involve a significant increase in the probability or consequences of 
an accident previously evaluated because of the following:
    The proposed changes are administrative in nature and do not 
affect the probability or consequences of any previously evaluated 
accidents for Dresden or Quad Cities Stations. The proposed 
amendment is consistent with the current safety analyses and 
represents sufficient requirements for the continued assurance and 
reliability of the RPS and Rod Block Instrumentation equipment, 
which is assumed to operate in the safety analysis, or provides 
continued assurance that specified parameters associated with RPS 
and Rod Block Instrumentation remain within their acceptance limits. 
Therefore, these changes will not affect the probability or 
consequences of a previously evaluated accident.
    The RPS and Rod Block Instrumentation related to this proposed 
amendment is not assumed in any safety analysis to initiate any 
accident sequence for Dresden or Quad Cities Stations; therefore, 
the probability of any accident previously evaluated is not affected 
by the proposed amendment.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed changes are administrative in nature and serve to 
maintain consistent and clear requirements for operability as 
specified in the Technical Specifications for the Limiting 
Conditions for Operation and Surveillance Requirements for the RPS 
and Rod Block Instrumentation. No new modes of operation or changes 
to any plant equipment are proposed by the proposed amendment 
request. The associated systems related to this proposed amendment 
are not assumed in any safety analysis to initiate any accident 
sequence for Dresden or Quad Cities. The proposed changes maintain 
the present level of operability; and therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident than any previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    The proposed changes are administrative in nature and do not 
affect existing plant safety margins or the reliability of the 
equipment assumed to operate in the safety analysis. The proposed 
changes have been evaluated and found to be acceptable for use at 
Dresden and at Quad Cities based on RPS and Rod Block 
Instrumentation system design, safety analysis requirements and 
operational performance. Since the proposed changes are 
administrative in nature and maintain necessary levels of the RPS 
and Rod Block reliability, the proposed changes do not involve a 
reduction in the margin of safety.
    The proposed amendment for Dresden and Quad Cities Stations will 
not reduce the availability of the RPS and Rod Block Instrumentation 
System which is required to mitigate accident conditions; therefore, 
the proposed changes do not involve a reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 6574]]

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: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: October 22, 1996
    Description of amendment request: The proposed amendments would 
allow continued plant operation at elevated Containment Lower 
Compartment temperatures between 125 deg. and 135 deg. F for a period 
not to exceed 72 cumulative hours.
    Basis for proposed no significant hazards 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 amendment involve a significant increase in 
the probability or consequences of any accident previously evaluated 
in the UFSAR [Updated Final Safety Analysis Report]?
    The increase in maximum Containment Lower Compartment 
temperature will not change the operation of any equipment which is 
important to safety. All components and instruments will continue to 
perform as designed in the higher temperature environment for the 
period that the revised Technical Specification allows. This 
temperature increase will not impact the ability of any component or 
instrument to perform its function in the event of an accident. 
Therefore, the probability of an accident is not impacted. The 
increased temperature will cause a decrease in the air mass in lower 
containment. This change has been evaluated for impact on 
containment temperature and pressure in accident conditions. The air 
mass change is conservative for peak containment pressure since the 
air mass is decreased. Maximum containment temperatures during a 
postulated accident are slightly increased as a result of higher 
initial Containment Lower Compartment temperature. The increase in 
peak temperature remains within the allowable values and thus does 
not increase the probability or consequence of an accident. The 
minimum containment pressure as a result of steam condensation in 
containment is lowered as a result of the decreased air mass in 
containment. Due to the conservative assumptions made in modeling 
containment for minimum pressure response, this change has no impact 
on the accident analysis.
    Based on the analysis of the bounding accidents that may be 
impacted by increased Containment Lower Compartment temperature and 
the review of the effect of the increased temperature on components 
in lower containment, it is determined that the probability and 
consequence of any analyzed accident is unchanged as a result of 
this change.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident not previously evaluated?
    The revised maximum Containment Lower Compartment temperature 
will not change any systems or operations procedures except to 
procedurally respond should Containment Lower Compartment 
temperature remain elevated for a period near the revised limiting 
period. The response of the systems and components are unaffected by 
this change. All instruments are qualified for the revised service 
conditions and will perform in the same manner as before. Normal 
operation and transient response will remain unchanged. Review of 
previously analyzed accidents show that no new transients are 
created as a result of this change. Based on this review there are 
no new or different accidents made possible by this change.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The amendment could potentially affect the containment system. 
The operation and analysis of the reactor coolant system and fuel 
are unaffected by this change. The maximum containment temperature 
is slightly increased while the maximum containment pressure is 
decreased. The minimum containment pressure could be slightly 
decreased and minimum containment temperature is unaffected. All 
these parameters have been reviewed and determined to be within 
assumptions made in these analyses. The accident transient analyses 
are unaffected beyond these small changes and remains acceptable in 
all cases. Therefore, the margin of safety is unaffected by this 
amendment.
    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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 6, 1997
    Description of amendment request: The proposed amendments would 
allow a one-time revision to Technical Specifications 3.6.1.1, 3.6.1.2, 
3.6.1.8, and 3.6.1.9 to allow operation of the Containment Purge 
Ventilation System (VP) during Modes 3 and 4 following the steam 
generator (SG) replacement outage. This one-time revision would be 
necessary due to respiratory hazardous gases released during heatup 
after the replacement of the SGs. The VP system would be used to remove 
the hazardous gases.
    Basis for proposed no significant hazards 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 activity does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The VP [Containment Purge Ventilation] System has no interfaces 
with any primary system, secondary system, or power transmission 
system. It has no interfaces with any reservoir of radioactive gases 
or liquids. None of the systems listed above are modified by the 
activity. In summary, no ``accident initiator'' is affected with the 
proposed operation of the VP System in Modes 3 and 4. For this 
reason, the activity does not involve an increase in the probability 
of an accident previously evaluated.
    Analyses have been performed to determine upper bounds to the 
source term, the offsite doses, and the Control Room dose. The 
results of that analyses are reported above. Both the source term 
and the doses were found to be significantly lower than the results 
of the corresponding design basis analyses. In addition, it has been 
determined that with no credit taken for any heat transfer from the 
fuel and cladding to the moderator channels, that sufficient time 
would exist for the operators to initiate recovery of flow from the 
ECCS [Emergency Core Cooling System] to the reactor core. The flow 
required from the ECCS to maintain the core in a coolable geometry 
was found to be well within the capacity of any one ECCS pump. 
Furthermore, it was determined that convective heat transfer to 
steam would be sufficient to prevent release of significant source 
term or a significant degree of fuel damage.
    For the above reasons, it is determined that operation of the VP 
System in Mode 3 or 4 immediately following the steam generator 
replacement outage does not involve a significant increase in either 
the probability or the consequences of an accident previously 
evaluated.
    2. The activity does not create the possibility of a new or 
different type of

[[Page 6575]]

accident from any accident previously evaluated.
    As discussed above, no ``accident initiators'' are affected by 
the proposed activity. Operation of the VP System proposed for Modes 
3 and 4 will be the same as that routinely carried in other modes of 
operation. For these reasons, the activity will not create the 
possibility of a new or different type of accident from any 
previously evaluated.
    3. The activity does not involve a significant reduction in the 
margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (the fuel and fuel cladding, the 
Reactor Coolant System pressure boundary, and the containment) to 
limit the level of radiation doses to the public. The proposed 
operation of the VP System will occur at the end of an extended 
outage. The level of decay heat and activity in the reactor is very 
low compared to the level of decay heat and activity associated with 
full power operations. For this reason, the likelihood of damage to 
the fuel following a DBLOCA [design basis loss-of-coolant accident] 
occurring during the proposed purging is reduced, as determined 
above. Both offsite doses and doses to the Control Room were found 
to be small compared to the limits of 10 CFR [Part] 100 and GDC 
[General Design Criterion] 19. For these reasons, the activity does 
not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 13, 1997
    Description of amendment request: The proposed amendments would 
implement the performance-based containment leak rate testing 
requirements of 10 CFR Part 50, Appendix J, Option B, for Type A 
testing.
    Compliance with 10 CFR Part 50, Appendix J, provides assurance that 
the primary containment, including those systems and components that 
penetrate the primary containment, do not exceed the allowable leakage 
rate values specified in the Technical Specifications and Bases. The 
allowable leakage rate is determined so that the leakage assumed in the 
safety analyses is not exceeded.
    On February 4, 1992, the NRC published a notice in the Federal 
Register (57 FR 4166) discussing a planned initiative to begin 
eliminating requirements marginal to safety that impose a significant 
regulatory burden. Appendix J to 10 CFR Part 50, ``Primary Containment 
Leakage Testing for Water-Cooled Power Reactors,'' was considered for 
this initiative and the staff undertook a study of possible changes to 
this regulation. The study examined the previous performance history of 
domestic containments and examined the effect on risk of a revision to 
the requirements of Appendix J. The results of this study are reported 
in NUREG-1493, ``Performance-Based Leak-Test Program.''
    Based on the results of this study, the staff developed a 
performance based approach to containment leakage rate testing. On 
September 12, 1995, the NRC approved issuance of this revision to 10 
CFR Part 50, Appendix J, which was subsequently published in the 
Federal Register on September 26, 1995, and became effective on October 
26, 1995. The revision added Option B ``Performance-Based 
Requirements'' to Appendix J to allow licensees to voluntarily replace 
the prescriptive testing requirements of Appendix J with testing 
requirements based on both overall and individual component leakage 
rate performance.
    Regulatory Guide 1.163, ``Performance-Based Containment Leak Test 
Program,'' was developed as a method acceptable to the staff for 
implementing Option B. Accordingly, the licensee has submitted, in its 
application dated January 13, 1997, proposed changes to the TS to 
implement 10 CFR Part 50, Appendix J, Option B, by referring to 
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test 
Program.''
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Containment leak rate testing is not an initiator of any 
accident; the proposed change does not affect reactor operations or 
accident analysis, and has no significant radiological consequences. 
... Therefore, this proposed change will not involve an increase in 
the probability or consequences of any previously-evaluated 
accident.
    2. The proposed change will not create the possibility of any 
new accident not previously evaluated.
    The proposed change does not affect normal plant operations or 
configuration, nor does it affect leak rate test methods. The test 
history at McGuire (two consecutive successful tests) provides 
continued assurance of the leak tightness of the containment 
structure.
    3. There is no significant reduction in a margin of safety.
    The proposed changes are based on NRC-accepted provisions, and 
maintain necessary levels of reliability of containment integrity. 
The performance-based approach to leakage rate testing recognizes 
that historically good results of containment testing provide 
appropriate assurance of future containment integrity; this supports 
the conclusion that the impact on the health and safety of the 
public as a result of extended test intervals is negligible. In 
addition, local leak[]rate testing will continue to provide 
assurances of overall containment integrity.
    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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 16, 1996
    Description of amendment request: The amendment requests to change 
the Waterford 3 Technical Specifications Table 4.3-1 to expand the 
applicability for Core Protection Calculator operability and to allow 
for the application of a Cycle Independent Shape Annealing Matrix.
    Basis for proposed no significant hazards 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 change will reduce the amount of non-conservatism 
presently allowed for linear power level, the CPC delta T power, and 
CPC nuclear power signals.

[[Page 6576]]

Changing the tolerance range from plus or minus 2% to between -0.5% 
and 10% between 15% and 80% RATED THERMAL POWER, except during 
physics testing, will allow more conservative settings than 
currently allowed. The consequences of an accident will be reduced 
due to the proposed change because it is less likely to be non-
conservative in power.
    This proposed change will allow use of Cycle Independent Shape 
Annealing Matrix (CISAM) elements. These elements will be validated, 
during startup testing, by monitoring the same parameters used for 
cycle specific shape annealing matrix (SAM) elements. If the CISAM 
is determined to be no longer valid, a cycle specific SAM will be 
calculated and used in the CPC's. In addition, use of CISAM gives 
better agreement throughout the cycle.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change to TS power calibration tolerance limits is 
conservative relative to the current TS requirement. CPC's cannot 
cause an accident and this change will not create the possibility of 
a new or different type accident. The changes ensures that the 
reactor will trip prior to the current condition due to higher CPC 
power.
    As stated previously, CISAM modeling removes some of the 
uncertainty associated with axial shape and provides increased 
assurance that the CPC is appropriately modeling the core.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change to the TS reduces the amount of non-
conservatism in safety system power indications and maintains the 
margin of safety for design basis events which take credit for the 
linear power level, the CPC delta T power, and CPC nuclear power 
signals.
    CISAM will be validated each cycle during startup testing and 
must meet the same parameters as cycle specific SAM elements. Since 
CISAM has a better accuracy than the cycle dependent SAM, the margin 
of safety is improved.
    Therefore, the proposed changes will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 27, 1996
    Description of amendment request: The proposed change request would 
change the acceptance criteria for the individual cell voltage from 
2.0v to 2.09v, change the surveillance frequency for battery specific 
gravities to implement the recommendations of IEEE 450-1995, delete 
surveillance requirement 4.7.B.4.d, add a clarifying phrase ``while on 
a float charge....'' where appropriate, and update the Basis to reflect 
these changes.
    Basis for proposed no significant hazards 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:
    This request has been determined to involve No Significant 
Hazards in that it does not:
    1. Involve a significant increase in the probability or 
consequences
    of an accident previous[ly] evaluated; (or)
    The proposed change in ICVs [individual cell voltages] does not 
increase the probability of an accident previously evaluated, as it 
increases the required voltage for each ICV.
    The proposed change in frequency does not increase the 
probability or consequences of an accident previously evaluated, as 
the change in the frequency of specific gravity testing is the 
result of industry experience gained over the years. The weekly 
reading of pilot cell specific gravity and cell voltage, along with 
the quarterly reading of all ICVs and a 10% sample of specific 
gravities from designated cells provides an acceptable means of 
determining cell operability as specified in IEEE 450-1995.
    The proposed deletion of Technical Specification Surveillance 
Requirement 4.7.B.4.d only removes an unnecessary Technical 
Specification surveillance and is consistent with NUREG-1433, 
Standard Technical Specifications General Electric Plants, BWR/4, 
Revision 1, April 1995. No change to plant systems, components or 
operating conditions are associated with this change. Existing 
Technical Specification station and diesel generator battery 
inspection and testing requirements adequately verify battery 
operability and condition.
    2. Create the possibility of a new or different kind of accident 
from any accident previous[ly] evaluated; (or)
    The proposed change does not create the possibility of a new or 
different kind of accident than previously evaluated, as the change 
only involves raising a required voltage, performing an existing 
surveillance on a different frequency, and removing an unnecessary 
annunciator surveillance requirement. The station battery and diesel 
generator battery low voltage annunciator setpoints do not meet any 
of the criteria codified in 10 CFR 50.36 for determining content of 
Technical Specifications and removal of surveillance requirement is 
consistent with NUREG-1433, Standard Technical Specifications 
General Electric Plants, BWR/4, Revision 1, April 1995. There is no 
change to hardware or operating conditions.
    3. Involve a significant decrease in the margin of safety.
    The proposed change to the ICV does not decrease the margin of 
safety, as increasing the required voltage actually increases the 
margin of safety. The proposed change to the frequency does not 
decrease the margin of safety as it continues to require testing and 
evaluation of the requisite surveillance points and implements 
requirements which have been determined to provide an adequate level 
of safety by the IEEE. The removal of Technical Specification 
surveillance requirements for the battery low voltage annunciator 
setpoints does not affect any plant systems, components or operating 
conditions.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Patrick D. Milano, Acting

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: January 23, 1997, as revised by letter 
dated January 28, 1997.
    Description of amendment request: The proposed amendment would make 
changes to Section 3.5/4.5.C of the technical specification (TS) bases 
to clarify the minimum residual heat removal (RHR) and residual heat 
removal service water (RHRSW) pump requirements for post-accident 
containment heat removal. In conjunction with the proposed amendment, 
the licensee requested NRC staff review and approval of an update to 
the design basis accident containment temperature and pressure response 
for the limiting single failure (loss of diesel generator) which 
results in minimum RHR and RHRSW pump availability.
    Basis for proposed no significant hazards determination: As 
required by

[[Page 6577]]

10 CFR 50.91(a), the licensee has provided its analysis of the issue of 
no significant hazards consideration. The NRC staff has reviewed the 
licensee's analysis against the standards of 10 CFR 50.92(c). The NRC 
staff's review 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 amendment will change TS bases to clarify the 
minimum RHR and RHRSW pump requirements for post-accident 
containment heat removal. The proposed amendment will also correct 
an error in a previous analysis on containment temperature and 
pressure response following a design basis accident (DBA) that was 
submitted for the NRC staff review on May 1, 1986. The proposed 
amendment does not affect the physical configuration of the plant or 
how it is operated. The licensee's analysis, using a new decay heat 
model, determined that the calculated maximum suppression pool 
temperature will be 2 degrees Fahrenheit greater (184 degrees 
Fahrenheit vs. 182 degrees Fahrenheit) than that predicted in its 
previous analysis, based on an earlier decay heat model, that was 
submitted for the NRC staff review on May 1, 1986. The licensee 
evaluated the effects of this increase on emergency core cooling 
system (ECCS) pump net positive suction head, wetwell attached 
piping, and environmental conditions in the ECCS pump rooms, and 
concluded that the change is acceptable. The consequences or 
probability of a previously evaluated accident will, therefore, not 
be significantly increased.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not create the possibility of a new or 
different kind of accident from any previously evaluated since the 
proposed amendment does not affect the physical configuration of the 
plant or how it is operated. The proposed amendment revises the TS 
bases to clarify the minimum RHR and RHRSW pump requirements for post-
accident containment heat removal.
    (3) The proposed changes do not result in a significant 
reduction inthe margin of safety.
    The proposed amendment will change TS bases to clarify the 
minimum RHR and RHRSW pump requirements for post-accident 
containment heat removal. The proposed amendment will also correct 
an error in a previous analysis on containment temperature and 
pressure response following a design basis accident (DBA) that was 
submitted for the NRC staff review on May 1, 1986. The proposed 
amendment does not affect the physical configuration of the plant or 
how it is operated. The licensee's analysis, using a new decay heat 
model, determined that the calculated maximum suppression pool 
temperature will be 2 degrees Fahrenheit greater (184 degrees 
Fahrenheit vs. 182 degrees Fahrenheit) than that predicted in its 
previous analysis, based on an ealier decay heat model, that was 
submitted for the NRC staff review on May 1, 1986. The licensee 
evaluated the effects of this increase on emergency core cooling 
system (ECCS) pump net positive suction head, wetwell attached 
piping, and environmental conditions in the ECCS pump rooms, and 
concluded that the change is acceptable. Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW. Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment requests: December 9, 1996
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance 
frequencies from at least once every 18 months to at least once per 
refueling interval (nominally 24 months) for the reactor trip system 
(RTS) and engineering safety features actuation systems (ESFAS) 
instrumentation channels, and make certain changes in trip setpoints 
and allowance values due to a setpoint methodology change in support of 
the calibration extensions. Channel operational tests (COTs) and trip 
actuating device operational tests (TADOTs) associated with these 
channels are also being extended. Revisions to the appropriate TS Bases 
are being revised to support the TS revisions.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS channel calibration, COT, and TADOT interval 
increases from 18 to 24 months, the setpoint change, and the 
allowable value changes do not alter the intent or method by which 
the channel calibrations are conducted, do not alter the way any 
structure, system, or component functions, and do not change the 
manner in which the plant is operated. The calibration and 
maintenance histories indicate that the equipment will continue to 
perform satisfactorily with longer surveillance intervals. With the 
exception of the pressurizer water level - high instrument, no 
recurring surveillance or maintenance problems were identified for 
the RTS or ESFAS instrumentation channels.
    The pressurizer water level instruments do not have a safety 
limit and are not credited in the DCPP safety analysis. The 
recurring surveillance problems were mainly due to calibration zero 
shift which is reflected in the statistically determined drift and 
in the proposed pressurizer water level high setpoint. The zero 
shift problem of these transmitters was a recurring problem with the 
calibration procedure. The procedures for calibrating these 
instruments have been revised to improve the repeatability of the 
surveillance activity.
    The trip setpoint and allowable value changes for pressurizer 
water level - high are each in the more restrictive direction. The 
revised setpoint would tend to trip the reactor sooner than the 
present settings. These changes ensure that sufficient margin is 
maintained for the pressurizer water level to accommodate the 
channel statistical uncertainty resulting from a 30-month operating 
cycle.
    A statistical analysis of channel uncertainty for a bounding 30-
month operating cycle has been performed. There is sufficient margin 
between the existing TS limits and the licensing basis safety 
analysis limits to accommodate the channel statistical uncertainty 
resulting from a 30-month operating cycle. The existing margin 
between the TS limits and the safety analysis limits provides 
assurance that plant protective actions will occur as required. 
However, a change to the safety analysis limit is proposed in order 
to provide additional margin for the RCS loss of [f]low-low 
setpoint.
    Westinghouse has evaluated the safety analysis limit for the RCS 
loss of flow-low setpoint and has determined that the limit can be 
changed from 87 percent of MMF to 85 percent of MMF with no impact 
on the probability and insignificant impact on the consequences of 
accidents already analyzed. The existing conclusions of the DCPP 
FSAR Update remain valid with the safety analysis limit change. 
Using the new safety analysis limit, sufficient margin exists 
between the TS limit and the safety analysis limit to accommodate 
the channel statistical uncertainty resulting from a 30-month 
operating cycle.
    The proposed changes to the allowable values ensure that drift 
assumptions regarding the protection racks and direct input 
functions are met.
    There are no known mechanisms that would significantly degrade 
the performance of the evaluated instrument channels during normal 
plant operation. All potential time-

[[Page 6578]]

 related degradation mechanisms have insignificant effects in the 
time frame of interest (maximum of 30 months). PG&E will continue to 
perform the maintenance required to maintain the qualification of 
this safety related equipment.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed pressurizer water level trip setpoint, RCS flow 
safety analysis limit, and various allowable value changes provide 
adequate margin to accommodate instrument channel uncertainty over a 
30-month operating cycle. Plant equipment, which will be set at, or 
more conservative than, the trip setpoints, will provide protective 
functions to assure that the safety analysis limits are not 
exceeded. The change to the RCS loss of flow safety analysis limit 
does not create the possibility of a new or different kind of 
accident since the setpoint will remain as currently specified and 
only results in an insignificant delay in the plant response to the 
accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    For almost all the existing DCPP RTS/ESFAS setpoints, the 
existing difference between the safety analysis limit and the 
setpoints was sufficient to accommodate any changes in instrument 
uncertainty.
    The change in the pressurizer water level - high setpoint does 
not affect a safety analysis limit and, therefore, has no effect on 
a margin to safety. Since the normal pressurizer level is maintained 
at 60 percent span and the no-load T<INF>avg control level is 22 
percent span, a change in the setpoint from less than or equal to 92 
percent span to less than or equal to 90 percent span is not 
significant to either DCPP plant operation or safety.
    The change in the RCS loss of flow-low safety analysis limit 
from 87 percent MMF to 85 percent MMF does not affect the existing 
plant setpoint and was evaluated to have a negligible effect on the 
limiting conditions of a partial loss of flow accident, a single RCP 
locked rotor, or RCP shaft break accident. This safety limit change 
was also found to have no effect on the DCPP minimum DNBR since the 
minimum DNBR is associated with the complete loss of flow accident. 
The complete loss of flow accident was evaluated to the Condition II 
fault criteria applicable to the partial loss of flow accident 
evaluation and was acceptable.
    Therefore, the proposed change 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 requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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 amendment request: December 23, 1996
    Description of amendment request: The proposed amendment would 
allow the use of Vantage Plus fuel.
    Basis for proposed no significant hazards 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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    The probability of occurrence or the consequences of an accident 
previously evaluated is not significantly increased. The VANTAGE + 
fuel assemblies containing ZIRLO<SUP>TM clad fuel rods, thimble and 
instrument tubes, IFMs, [intermediate fuel mixing assemblies] and 
LPD [low-pressure-drop] mid-grids meet the same fuel assembly and 
fuel rod design bases as VANTAGE 5 (without IFMs) fuel assemblies in 
the other fuel regions. In addition, the 10 CFR 50.46 criteria will 
be applied to the ZIRLO<SUP>TM clad fuel rods, thimble and 
instrument tubes, IFM grids, and LPD mid-grids. The use of these 
fuel assemblies will not result in a change to the proposed Indian 
Point Unit 3 VANTAGE 5 (without IFMs) transition core design and 
safety analysis limits. The ZIRLO<SUP>TM clad material is similar in 
chemical composition and has similar physical and mechanical 
properties as that of Zircaloy-4. Thus the cladding integrity is 
maintained and the structural integrity of the fuel assembly is not 
affected. The ZIRLO<SUP>TM clad fuel rod improves corrosion 
resistance and dimensional stability. In addition, the incorporation 
of LPD mid-grids and IFMs improves dimensional stability. Since the 
dose predictions in the safety analyses are not sensitive to the 
fuel assemblies material changes as specified in this report, the 
radiological consequences of accidents previously evaluated in the 
safety analyses remain valid. Therefore, the probability or 
consequences of an accident previously evaluated is not 
significantly increased.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created, since the VANTAGE + 
fuel assemblies containing ZIRLO<SUP>TM clad fuel rods, thimble, and 
instrument tubes, IFMs, and LPD mid-grids will satisfy the same 
design bases as that used for VANTAGE 5 (w/o IFMs) fuel assemblies 
in the other fuel regions. Since the original design criteria is 
being met, the ZIRLO<SUP>TM clad fuel rods, thimble and instrument 
tubes, IFMs, and LPD mid-grids will not be an initiator for any new 
accident. All design and performance criteria will continue to be 
met and no new single failure mechanisms have been created. In 
addition, the use of these fuel assemblies does not involve any 
alterations to plant equipment or procedures which would introduce 
any new or unique operational modes or accident precursors. 
Therefore, the possibility for a new or different kind of accident 
from any accident previously evaluated is not created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The margin of safety is not significantly reduced, since the 
VANTAGE + fuel assemblies containing ZIRLO<SUP>TM clad fuel rods, 
thimble and instrument tubes; IFMs, and LPD mid-grids do not change 
the proposed Indian Point 3 VANTAGE 5 (w/o IFMs) transition core 
design and safety analysis limits. The use of these fuel assemblies 
containing fuel rods, thimble and instrument tubes with ZIRLO<SUP>TM 
cladding alloy; IFMs and LPD mid-grids will take into consideration 
the normal core operating conditions allowed in the Technical 
Specifications. For the transition core and each future cycle reload 
core, these fuel assemblies will be specifically evaluated using 
standard reload design methods and approved fuel rod design models 
and methods. This will include consideration of the core physics 
analysis, peaking factors and core average linear heat rate effects. 
In addition, the 10 CFR 50.46 criteria will be applied each cycle to 
the ZIRLO<SUP>TM clad fuel rods, thimble and instrument tubes, IFMs, 
and LPD mid-grids. Analyses or evaluations will be performed each 
cycle to confirm the 10 CFR 50.46 will be met. Therefore, the margin 
of safety as defined in the Bases to the Indian Point Unit 3 
Technical Specifications and VANTAGE 5 (w/o IFMs) ZIRLO<SUP>TM 
licensing amendment approval is not significantly reduced.
    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Acting

[[Page 6579]]

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: July 9, 1996 (TXX-96393)
    Brief description of amendments: The proposed changes would 
increase the minimum allowable value of the Unit 1 Steam Line Pressure-
-Low Safety Injection and Steam Line Isolation functions. These changes 
are needed to ensure that the instrumentation error is properly 
accounted for in the Technical Specifications.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The changes in the License Amendment Request proposes more 
restrictive setpoint Allowable Values for the Steam Line pressure--
Low channels of the Engineered Safety Features Actuation System 
(ESFAS). These more restrictive values assure that all applicable 
safety analysis limits are being met. Changing an Allowable Value in 
the Technical Specifications has no impact on the probability of 
occurrence of any accident previously evaluated. None of the 
accident analyses were affected, therefore, the consequences of all 
previously evaluated accidents remain unchanged.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve the use of a more conservative 
value for the Allowable Value for the Steam Line Pressure--Low 
Safety Injection and Steam Line Isolation functions. As such, none 
of the changes effect plant hardware or the operation of plant 
systems in a way that could initiate an accident. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    There were no changes made to any of the accident analyses or 
safety analysis limits as a result of this proposed change. Further, 
the proposed change does not affect the acceptance criteria for any 
analyzed event. ESFAS will remain capable of performing its safety 
function, and the new requirement will continue to provide adequate 
assurance of that capability. Making the Allowable Value more 
restrictive provides increased assurance that the channels will 
function within the safety analysis limits assumed in the safety 
analyses. The margin of safety established by the Limiting 
Conditions for Operation also remains unchanged. Thus there is no 
effect on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: July 10, 1996 (TXX-96405), as 
supplemented by letter dated October 1, 1996 (TXX-96475)
    Brief description of amendments: The proposed change would take 
credit for the addition of train oriented Fan Coil Units for each UPS & 
Distribution Room and would provide redundancy to the existing Air 
Conditioning Units.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The UPS HVAC System is a support system for other safety related 
equipment, primarily the Uninterruptible Power Supplies and some of 
their distribution equipment. The only impact that this system can 
have on the probability or consequences of an accident must result 
from the failure of the system to provide adequate support to the 
supported safety related equipment when that supported safety 
related equipment is required to operate.
    Allowing same train cooling to satisfy the LCO is considered 
equivalent to the existing Technical Specification. The proposed 
changes allow the use of the same train UPS Room Fan Coil Units or 
the same train UPS A/C Train to support a UPS & Distribution Room.
    Surveillance requirements are added or modified to ensure that 
the credited support equipment will be available when needed. 
Unnecessary starts of the UPS A/C Trains have been eliminated from 
the specifications. Overall, this is considered an enhancement that 
will increase the reliability of the UPS HVAC Systems. Because both 
the existing specification and the proposed revision to the 
specification continue to ensure normal support and the availability 
of at least one train of equipment in the event of a design basis 
accident, with the same or increased reliability, the consequences 
of an accident previously evaluated is not affected.
    Changing the specification from a ``common'' specification which 
impacts both units simultaneously to a specification which applies 
to both units separately is basically just an administrative change. 
Having ``common'' specifications is an aid to the operator to 
provide an alert that both units are affected. With the new LCO, 
both units may not be affected because rooms may now be cooled 
separately. Because both CPSES Units remain properly covered, 
however, this change will not significantly increase the probability 
of consequences of an accident.
    The revision to the existing ACTION is considered equivalent 
except for the change of the Allowed Outage Time (AOT) from seven 
days to 30 days. This change is based on the significance of the 
heating and cooling function but does represent an increase in AOT 
and thus an increase in the probability that the supported functions 
could be unavailable. This increase is not considered significant 
based on the following several factors:
    a)the systems design is based on a conservative assessment of 
the worst postulated conditions in the rooms;
    b) generally, less than design cooling is required and a short 
duration or partial failure may have little or no impact on the 
systems ability to perform its function;
    c) the multiple backups available (two UPS A/C Trains and only 
one UPS Room Fan Coil Unit per each room) increase the potential of 
restoring additional cooling if needed;
    d) the ability to perform alternate actions if normal cooling is 
lost such as circulating air via existing fans or portable fans 
thereby extending the time before cooling must be restored; and
    e) the extended AOT would allow more time and opportunity to 
perform corrective maintenance to ensure high equipment reliability.
    The new ACTION for loss of cooling reflects requirements that 
already exist in the Technical Specifications. The AOT for this 
ACTION statement is 72 hours which is based on the risk from an
    event occurring requiring the inoperable UPS A/C Train, and the 
remaining UPS Room Fan Coil Units and A/C Train fans providing the 
required protection.
    The new ACTION for loss of cooling and ventilation reflects a 
conservative response to the potential impact of such a condition. 
The proposed AOT is one hour. One hour is based on the time lag 
available from the operating temperature to the maximum Technical 
Specification limit of the UPS & Distribution Rooms. The addition of 
a specific ACTION in lieu of relying on Specification 3.0.3, 
although essentially equivalent, is consistent with the methodology 
of the improved Standard Technical Specifications and alerts the 
operator to the significance of the situation.
    The changes made to the surveillance ensure that the UPS Room 
Fan Coil Units will operate. The UPS Room Fan Coil Units are 
connected to the emergency busses and TS 4.8.1.1.2f. demonstrates 
the energization of emergency busses with permanently connected 
loads. The changes made to the 18

[[Page 6580]]

month surveillances on the UPS A/C trains were changed from the 
Safety Injection signal with the Blackout Test signal to ``... 
actual or simulated actuation signal''. This is consistent with 
NUREG-1431, ``Standard Technical Specifications Westinghouse 
Plants''.
    The changes to the BASES are descriptive in nature to reflect 
the other changes and by themselves have no impact on the 
probability or consequences of an accident.
    The ability to cope with station blackout and design basis fires 
is maintained or enhanced. For station blackout coping, the UPS A/C 
fans are considered to remain available while additional cooling is 
provided by a single available Fan Coil Unit.
    In summary, the proposed changes take advantage of the increased 
reliability offered by the revised system design. It also maintains 
the level of support provided by the system while at worst, allowing 
a slight decrease in availability (in certain situations) which is 
not considered significant. As a result, it is concluded that none 
of the changes made to the existing Technical Specification involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Revising this specification to take credit for the new UPS Room 
Fan Coil Units, to take credit for same train UPS A/C Train support 
for a UPS and Distribution Room, to make the specification unit 
specific instead of common, to make the surveillances appropriate 
for the credited equipment, and to make the action statements 
appropriate for the credited equipment and their significance, does 
not by itself alter plant hardware. Plant procedures are only 
altered to the extent that the revised specification will allow 
different configurations of equipment in the UPS HVAC System to be 
operated at different times. These changes ensure continued support 
of the safety related equipment in the affected areas and do not 
affect the equipments failure or failure modes. As a result, these 
changes to the Technical Specification do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    None of the changes being proposed alter the environmental 
conditions which are to be maintained in the areas supported by an 
OPERABLE UPS HVAC System during normal operations and following an 
accident. As a result, the margin of safety for these functions 
remains the same. The only potential adverse impact is the system's 
postulated availability, as discussed in the response to question 1 
above. This reduction in availability is to a great extent mitigated 
by the projected increase in system reliability. As noted in the 
response to question 1, there is no significant impact on the 
accident analyses. Thus, even if system availability issues were 
considered an aspect of margin of safety, the proposed changes do 
not involve a significant reduction in 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: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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: December 17, 1996
    Description of amendment request: The proposed changes will allow 
one of the two service water loops to be isolated from the component 
cooling water heat exchangers (CCHXs) during power operation in order 
to refurbish sections of the isolated service water headers. The 
proposed temporary changes will be valid for two periods of up to 35 
days each for implementation of the service water upgrades associated 
with the repair of the sections of the 24-inch service water supply and 
return piping to/from the CCHXs.
    Basis for proposed no significant hazards 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:
    Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The piping refurbishment project and the proposed temporary 
changes to the SW [service water] and CC [component cooling] 
Technical Specifications have been evaluated to assess their impact 
on the normal operation of the SW and CC systems and to ensure that 
the design basis safety functions of each system are preserved. The 
SW system is required to function during all normal and emergency 
operating conditions. During normal plant operation, the SW system 
provides cooling water to the CCHXs, charging pump coolers, 
instrument air compressor coolers, and control room chiller 
condensers of both units. Within the first 168 hour Section 3/
4.7.4.1.d TS AS [Action Statement] of isolation of the header which 
is to be repaired, temporary 10'' diameter SW lines (one supply and 
one return) will be installed to supply the SW to the charging pumps 
coolers, instrument air compressors coolers, Unit 2 CR chillers and 
spent fuel pool (SFP) coolers to satisfy design basis conditions. 
These temporary lines will be routed from the operating part of the 
36'' SW headers while the 24'' headers to CCHXs are being repaired. 
The temporary lines will be dismantled when the repaired header is 
returned to operation (second 168 hour AS). During the two 35-day 
periods, one header will operate with its 24-inch piping to/from the 
CCHXs temporarily blanked. To avoid operation of the SW pump at 
abnormal conditions (low flow) on this ``partially deadlocked'' 
header, a temporary cross-connect will be installed to by-pass the 
CCHXs.
    SW system operation with the cross-connect installed was 
evaluated for design basis accident (DBA) conditions. The DBA 
condition for the SW system is a loss-of-coolant accident on one 
unit with simultaneous loss-of-offsite-power to both units. [An] SW 
system hydraulic analysis has been performed to verify that adequate 
flow is provided to the containment recirculation spray heat 
exchangers (RSHXs) with the temporary cross-connect installed and 
throttled open, assuming the occurrence of the most limiting single 
failure. Therefore, there is no increase in probability or 
consequences of the DBA condition.
    Utilizing only one SW header to supply flow to the CCHXs has the 
potential to affect the reliability of the CC system and all of the 
equipment cooled by CC. A review of the equipment affected by this 
phase of the SW restoration project was performed to evaluate the 
impact on initiating event frequency. Since the SW system and CC 
system are support systems used to remove heat, a failure in either 
of these systems does not affect the initiating event frequency of 
any design basis event. Additionally, an estimate of the impact on 
core damage frequency is provided below. The impact on the North 
Anna Probabilistic Safety Assessment (PSA) during implementation of 
this DCP [design change package] is similar to impact of work 
performed under DCP-94-010 since the scope of work of both DCPs is 
repair/replacement of different portions of the same 24'' SW headers 
to CCHXs. The only difference from a PSA standpoint is that CDF 
[core damage frequency] for DCP-94-010 was calculated based on 140 
days supply of CCHXs from one SW header while per this DCP it is 
only 70 days. Therefore, results of PSA evaluation for DCP-94-010 
are conservatively applied to this DCP. The activities to be 
performed during the refurbishment project and the various system 
alignments required have been evaluated using the Individual Plant 
Examination (IPE) Probabilistic Safety Assessment (PSA) model for 
North Anna Power Station. This model is used in a manner that is 
generally consistent with the Electric Power Research Institute 
(EPRI) PSA Applications Guide TR-105396. The effect on the PSA model 
is a slight increase in the frequency of reactor trips and an 
increase in the probability of RHR [residual heat removal] failure.
    The increased frequency of reactor trips is due to the decreased 
reliability of the CC system to supply cooling to the RCP [reactor 
coolant pump] motor. When only one SW header is available to the CC 
heat exchangers

[[Page 6581]]

the frequency of losing this single header is dominated by the 
probability of both SW pumps failing. Also considered was the 
frequency of pipe rupture anywhere in the single available header. 
When the single SW header fails to supply cooling to the CC heat 
exchangers, the CC system will heatup causing inadequate cooling for 
sustained operation of the RCPs. Tripping these pumps results in a 
reactor trip. The second SW header can be expected to supply other 
equipment with cooling. This scenario is appropriately modeled as a 
reactor trip with main feedwater available initiating event. A 
sensitivity analysis shows the increase in CDF to be about 1E-8/
year. The total effect of this DCP includes a failure analysis of 
the reactor coolant pump and motor in case of loss of CCW.
    The CC system is also included in the PSA model as a support 
system for RHR cooling. The RHR system is used to reduce reactor 
coolant system temperatures from 350 deg.F (hot shutdown) to 
140 deg.F (cold shutdown). The only accident initiator that requires 
the unit to be cooled down and placed on RHR cooling are sequences 
which are initiated with a steam generator tube rupture. (Note that, 
for the North Anna plant design, RHR is separate from the safety 
injection system and the low head safety injection pumps.) The 
increased probability for the loss of RHR when only one SW header is 
available to the CCHXs is estimated using fault tree analysis and is 
dominated by the failure of both SW pumps. The probability for the 
loss of both SW pumps aligned to the CCHXs is estimated to be 1.5E-
4. The effect of this increase in RHR failure probability was 
determined by adding this probability to the top single event in the 
RHR function and recalculating the new CDF. The resulting increase 
in CDF as a result of RHR system failure following a steam generator 
tube rupture is less than 1E-8 per year.
    The CC system is further included in the PSA model as part of 
the loss of RCP seal cooling as an initiating event and as a loss of 
function during other initiating event scenarios. The effect on the 
probability for a loss of RCP seal cooling due to losing CC cooling 
to the RCP thermal barriers is negligible due to the high 
reliability of the charging system to provide seal injection.
    The total effect of this DCP on core damage frequency (CDF) was 
estimated by a sensitivity analysis combining both the change in the 
reactor trip initiating event frequency and the increased failure 
probability of RHR. It was evaluated that during implementation of 
this DCP, CCHXs will be supplied from one SW header for 70 days (35 
x 2=70), therefore, the increase in CDF previously evaluated in DCP-
94-010 based on 140 days is conservative. This DCP does not affect 
the containment systems and there would not be any significant 
change in off-site dose since the containment heat removal portion 
of the SW system is not affected and the increase in CDF is 
insignificant. The small increase in CDF calculated for the repair 
activities and the procedure developed to provide contingency 
actions result in the conclusion that this work does not represent a 
significant increase in core damage frequency.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the allowed outage time only provide 
operational flexibility needed to perform necessary repairs. During 
the project, there will be a significant time period when all the 
CCHXs are aligned to one SW loop. The possibility of an interruption 
of SW supply to the heat exchangers during a DBA is eliminated by 
defeating the closure of the 24-inch SW isolation MOVs [motor-
operated valves] to the CCHXs on [an] SI/CDA [safety injection/
containment depressurization actuation] signal. Both SW headers will 
be available for equipment required for safe shutdown of the units 
(i.e., RSHXs, charging pumps, and CR/ESGR [control room/emergency 
switchgear room] chillers). The SW pipe repair activities and the 
installation/removal of the SW cross-connect and temporary piping do 
not create the possibility for a malfunction of equipment different 
than previously evaluated. Results of the Johnston Pump NPSH [net 
positive suction head] test proved to be satisfactory for the 
anticipated SW pump flow rates under modes of station operation for 
this project, therefore, the possibility for an accident of a 
different type than was previously evaluated in the Safety Analysis 
Report will not be created. Based on the above, implementation of 
the restoration project and approval of the proposed Technical 
Specifications changes will not introduce any new accident 
initiators nor affect the performance of accident mitigation 
systems.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to the schedule only provide operational 
flexibility to perform the required SW pipe refurbishment. The 
Technical Specifications continue to require the SW and CC systems 
to remain functional during the period with a single SW supply to 
the CCHXs. As stated in item (1) above, the SW system is fully 
capable of performing its DBA function during the course of the pipe 
refurbishment project with the proposed Technical Specification 
changes in place. The effect of this pipe refurbishment project on 
CC system reliability was estimated by a sensitivity analysis 
combining both the change in the reactor trip initiating event 
frequency and the increased failure probability of RHR resulting in 
about a 1E-8 per year increase in CDF. Since this project will not 
affect the containment systems, there would not be any significant 
change in off-site dose, except that resulting directly from the 
slight increase in CDF.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: F. Mark Reinhart (Acting)

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.

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

    Date of application for amendments: September 10, 1996
    Brief description of amendments: The amendments extend the 
automatic actuation logic channel functional test interval of the 
Engineering Safety

[[Page 6582]]

Features Actuation System and the surveillances test interval of the 
containment sump isolation valves from monthly to quarterly.
    Date of issuance: January 23, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 218 and 195
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52963) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated January 23, 1997 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: October 31, 1996
    Brief description of amendments: The amendments relocate the 
requirements for seismic monitoring instrumentation from Technical 
Specification (TS) 3/4.3.7.2, ``Seismic Monitoring Instrumentation'' to 
licensee-controlled documents in accordance with Generic Letter 95-10, 
``Relocation of Selected Technical Specifications Requirements Related 
to Instrumentation.'' The amendments also add a condition to the 
operating licenses which approves the relocation of the TS requirements 
to the UFSAR.
    Date of issuance: January 29, 1997 Effective date: Immediately, to 
be implemented within 90 days.
    Amendment Nos.: 117, 102
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and the license.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66703) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 29, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of application for amendment: September 9, 1996, as 
supplemented December 20, 1996
    Brief description of amendment: The amendment revises the Minimum 
Channels Operable requirement of Item 4.c (Steam Line Isolation, 
Containment Pressure Intermediate--High-High) of Technical 
Specification (TS) Table 3.3-3 from 3 channels to 2 channels provided 
the provisions of Action Statement 14 are followed. This change makes 
this Beaver Valley Power Station, Unit No. 1 TS consistent with the 
comparable Beaver Valley Power Station, Unit No. 2 TS. The amendment 
also revises the minimum charging pump discharge pressure in TS 3/4.5.5 
and associated Bases from 2311 psig to 2397 psig. This change ensures 
that safety analysis assumptions for safety injection flow are met.
    Date of issuance: January 27, 1997
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No: 201
    Facility Operating License No. DPR-66. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55032) The supplemental letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination or the original notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
January 27, 1997. No significant hazards consideration comments 
received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

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

    Date of application for amendment: August 15, 1996, and as 
supplemented by letters dated October 28, November 15, 1996, and 
January 7, 1997.
    Brief description of amendment: The amendment changes the Clinton 
Power Station (CPS) Technical Specifications to incorporate the revised 
Safety Limit Minimum Critical Power Ratio (SLMCPR) as calculated by 
General Electric (GE) for CPS Cycle 7. The need to change the SLMCPR 
resulted from the 10 CFR Part 21 condition reported by GE in their 
letter to the NRC dated May 24, 1996.
    Date of issuance: January 22, 1997
    Effective date: January 22, 1997
    Amendment No.: 113
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47978). The licensee's letters of October 28, November 15, 1996, and 
January 7, 1997, provided clarifying information and did not make 
significant changes to the initial Federal Register notice. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated January 22, 1997. No significant hazards 
consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: August 29, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications to (1) modify the applicability requirements for certain 
radiation monitors so that the radiation monitors are required to be 
operable only when secondary containment integrity is required to be 
operable; (2) delineate when secondary containment integrity is 
required; (3) modify standby gas treatment operability requirements; 
(4) make editorial corrections to clarify the configuration of the 
radiation monitors; and (5) revise the associated Bases sections.
    Date of issuance: January 14, 1997
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 98
    Facility Operating License No. DPR-21: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 1996 (61 FR 
54242) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 14, 1997 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, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

[[Page 6583]]

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

    Date of application for amendments: June 28, 1996, as supplemented 
by letters dated November 4 and 5, and December 9, 1996
    Brief description of amendments: These amendments revise the 
technical specifications to incorporate performance-based testing, in 
accordance with 10 CFR Part 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing For Water-Cooled Power Reactors,'' Option 
B. This option allows utilities to extend the frequencies of the Type A 
Containment Leak Rate Test, and Type B and C Local Leak Rate Tests 
based on the performance and design of the containment and components.
    Date of issuance: January 24, 1997
    Effective date: Both units, as of date of issuance and shall be 
implemented within 30 days.
    Amendment Nos.: 118 and 81
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55038) The supplemental letters provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or the original notice. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated Janaury 24, 1997. No significant hazards consideration 
comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464

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

    Date of application for amendments: May 20, 1996
    Brief description of amendments: These amendments revise Technical 
Specifications (TS) Sections 3/4.4.9.2, 3/4.9.11.1, 3/4.9.11.2, and the 
associated TS Bases 3/4.4.9 and 3/4.9.11, to more clearly describe that 
the Residual Heat Removal (RHR) system Shutdown Cooling mode of 
operation consists of four ``subsystems.'' These TS sections pertain to 
plant operations during Operational Conditions (OPCONs) 4, ``Cold 
Shutdown'' and 5, ``Refueling.'' In addition, the proposed TS change 
would make administrative changes to TS Section 3/4.4.9.1 to ensure 
consistency in terminology regarding the description of Shutdown 
Cooling ``subsystems.'' The proposed TS changes are consistent with the 
guidance delineated in the Improved TS (i.e., NUREG-1433, Revision 1, 
``Standard Technical Specifications General Electric Plants, BWR/4,'' 
dated April 1995) which indicates that the RHR Shutdown Cooling mode of 
operation is comprised of two loops and four subsystems (i.e., two 
subsystems per loop).
    Date of issuance: January 28, 1997
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 119 and 82
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55036) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 28, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464

Philadelphia Electric Company, Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: August 5, 1996, as supplemented 
December 4, 1996
    Brief description of amendment: The amendment revised TS Section 
2.1 and its associated TS Bases to reflect the change in the Minimum 
Critical Power Ratio Safety Limit due to the plant specific evaluation 
performed by General Electric Company (GE), for Limerick Generating 
Station, Unit 2, Cycle 4.
    Date of issuance: January 29, 1997
    Effective date: As of date of issuance and shall be implemented 
within 30 days
    Amendment No.: 83
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57491) The December 4, 1996, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or the initial notice. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated January 29, 1997. No significant hazards consideration comments 
received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464

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

    Date of application for amendments: March 29, 1996, as supplemented 
by letters dated December 5, 1996, and January 15, 1997
    Brief description of amendments: These amendments modify Technical 
Specification (TS) Section 4.5.1.d.2.b to delete the requirement to 
perform in-situ functional testing of the Automatic Depressurization 
System (ADS) valves once every 24-months as part of start-up testing 
activities.
    Date of issuance: January 29, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 120 and 84
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57488) The December 5, 1996, and January 15, 1997, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination nor the initial notice. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated January 29, 1997. 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: October 1, 1996
    Brief description of amendment: The amendment allows for a one-time 
extension of the surveillance intervals for the containment isolation 
valve seat leakage test, the isolation valve seal water test, the boron 
injection tank leakage test, the containment spray nozzle test, and the 
city water backup to the auxiliary boiler feed pump test.
    Date of issuance: January 28, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 172
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR

[[Page 6584]]

64393) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 28, 1997 No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: October 1, 1996, supplemented 
October 31, 1996
    Brief description of amendments: The amendments change Technical 
Specifications 3/4.7.1.5, ``Main Steam Line Isolation Valves (MSIVs),'' 
and 3/4.3.2, ``Engineered Safety Feature Actuation System 
Instrumentation.'' The amendments accommodate entry into Modes 3 and 2 
prior to performing MSIV closure time testing in Mode 2, allow 
additional time for the repair and testing of inoperable MSIVs in 
certain operating Modes, delete footnotes that are no longer 
applicable, and change the low steam line pressure trip setpoint value 
for safety injection, turbine trip and feedwater isolation to make it 
consistent with the actual plant configuration.
    Date of issuance: January 17, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented prior to entry into Mode 3 from the current outage.
    Amendment Nos. 187 and 170
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55040) The supplemental letter changed the TSs to provide greater 
consistency with requirements of NUREG-1431 ``Standard Technical 
Specifications - Westinghouse Plants,'' Revision 1, and did not change 
the initial proposed no significant hazards consideration determination 
or the Federal Register notice. The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated January 17, 
1997. No significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

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

    Date of application for amendments: October 24, 1996, as 
supplemented December 23, 1996
    Brief description of amendments: The amendments changed Technical 
Specification 3/4.7.1.2, ``Auxiliary Feedwater System.'' The changes 
revised the 18-month surveillance performed on the system's pumps and 
valves because testing of the turbine driven Auxiliary Feedwater pump 
can only be performed in higher modes when there is sufficient 
secondary steam pressure.
    Date of issuance: January 23, 1997
    Effective date: As of date of issuance, to be implemented within 30 
days
    Amendment Nos. 188 and 171
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1996 (61 
FR 58905) The December 23, 1996, letter proposed changes to TS 3/4.3.2 
to provide consistency with those proposed in the October 24, 1996, 
letter and therefore did not change the initial proposed no significant 
hazards consideration determination and was within the scope of the 
initial notice. The Commission's related evaluation of the amendments 
is contained in a Safety Evaluation dated January 23, 1997. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

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

    Date of application for amendments: September 25, 1996
    Brief description of amendments: The amendments relocate the list 
of containment isolation valves from the Technical Specifications to 
the Salem Updated Final Safety Analysis Report and correct references. 
Date of issuance: January 30, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos. 189 and 172
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications and the License.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55039) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 30, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

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

    Date of application for amendments: January 31, 1996, as revised 
November 26, 1996. The November 26, 1996, submittal withdrew the 
proposed change to surveillance tests being performed at power.
    Brief description of amendments: These amendments will revise the 
minimum emergency diesel generator day tank fuel oil volume.
    Date of issuance: January 17, 1997
    Effective date: January 17, 1997
    Amendment Nos.: 203 and 184
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7559) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 17, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: : The Alderman Library, 
Special Collections Department, University of Virginia, 
Charlottesville, Virginia 22903-2498.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of application for amendments: February 8, 1996, as 
supplemented August 15, December 2 and December 19, 1996, and January 
6, 1997
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.3.10, ``Control Rod and Power 
Distribution Limits,'' to improve the clarity of this section and add 
surveillance requirements to Section 15.4.1, ``Operational Safety 
Review.''
    Date of issuance: January 16, 1997
    Effective date: January 16, 1997, with full implementation within 
45 days
    Amendment Nos.: Unit 1 - 171, Unit 2 - 175
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10398) The August 15, December 2 and

[[Page 6585]]

December 19, 1996, and January 6, 1997, letters provided clarifying 
information and updated TS pages that were within the scope of the 
original application and did not change the NRC staff's initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated January 16, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

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

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point, Units 3 and 4, Dade County, Florida

    Date of amendment request: December 17, 1996
    Description of amendment request: The proposed amendments would 
modify the Turkey Point Units 3 and 4 Technical Specifications to 
change the Reactor Coolant Pump (RCP) flywheel surveillance 
requirement. The proposed change will require RCP flywheel inspections 
once every ten years.
    Date of publication of individual notice in Federal Register: 
January 10, 1997 (62 FR 1476)
    Expiration date of individual notice: February 10, 1997
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

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

    Date of amendment request: December 13, 1996
    Description of amendment request: The proposed amendment would 
approve transfer of Soyland Power Cooperative's 13.21% minority 
ownership interest in the Clinton Power Station to Illinois Power 
Company. This action would result in Illinois Power Company becoming 
the sole owner of the Clinton Power Station.
    Date of publication of individual notice in Federal Register: 
January 29, 1997 (62 FR 4337).
    Expiration date of individual notice: February 28, 1997
    Local Public Document Room location: : Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental

[[Page 6586]]

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 room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By March 14, 1997, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555-001, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-001, 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendments: January 13, 1997, as 
resubmitted January 17, 1997, and supplemented January 22, 1997.
    Brief description of amendments: The proposed amendments would: 
evaluate the Unreviewed Safety Question (USQ) associated with the 
operation of Dresden, Units 2 and 3, with the recently discovered error 
in the head loss across the Emergency Core Cooling System (ECCS) 
suction strainers; change the Technical Specification (TS) values by 
lowering the allowable water temperature in the suppression chamber and 
ultimate heat sink; change the basis of the TS to allow credit for two 
psig of containment pressure to compensate for a slight increase in the 
amount of Net Positive Suction Head (NPSH) deficiency during the first 
10 minutes following a design basis accident (DBA); and add a license 
condition to allow the licensee to change the Updated Final Safety 
Analysis Report to reflect the use of two psig of containment pressure 
to compensate for the deficiency in NPSH during the first 10 minutes 
following a DBA.
    Date of Issuance: January 28, 1997 Effective date: Immediately, to 
be implemented within 30 days.
    Amendment Nos.: 152/147
    Facility Operating License Nos. DPR-19 and DPR-25. The amendments 
revised the Technical Specifications and the Operating Licenses. Press 
release issued requesting comments as to proposed no significant 
hazards

[[Page 6587]]

consideration: Yes January 25, 1997 Joliet Herald News Comments 
received: No. The Commission's related evaluation of the amendment, 
finding of exigent circumstances, consultation with the State of 
Illinois and determination of no significant hazards consideration are 
contained in a Safety Evaluation dated January 28, 1997.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    NRC Project Director: Robert A. Capra
    Dated at Rockville, Maryland, this 5th day of February, 1997.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-3324 Filed 2-11-97; 8:45 am]
BILLING CODE 7590-01-F