[Federal Register Volume 65, Number 8 (Wednesday, January 12, 2000)]
[Notices]
[Pages 1918-1934]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-611]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 13, 1999, through December 31, 
1999. The last biweekly notice was published on December 29, 1999 (64 
FR 73083).

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

[[Page 1919]]

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 February 11, 2000, 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 electronically from 
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: 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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of amendment request: October 25, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification allowable values for the reactor 
protection system electric power monitoring assembly overvoltage and 
undervoltage trip setpoints.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or

[[Page 1920]]

consequences of any accident previously evaluated.
    The proposed Technical Specification (TS) change revises the 
Reactor Protection System (RPS) Electric Power Monitoring Assembly 
overvoltage and undervoltage Allowable Values. The new Allowable 
Values and setpoints will continue to provide adequate margin to the 
normal operating voltage range for the RPS and MSIV [main steam 
isolation valve] solenoids, thus minimizing the potential for 
inadvertent trips. The proposed change does not have a detrimental 
impact on the condition or performance of any plant structure, 
system, or component that may initiate an analyzed event. The 
proposed change does not physically impact the plant nor does it 
impact any design or functional requirements of the associated 
system. That is, the proposed change does not degrade the 
performance or increase the challenges of any safety systems assumed 
to function in the accident analysis. Further, the proposed change 
does not impact the Surveillance Requirements themselves nor the way 
in which the Surveillances are performed. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased.
    Additionally, the proposed change does not effect the affect the 
availability of equipment or systems required for mitigating the 
consequences of an accident. The revision of the overvoltage and 
undervoltage setpoints will ensure that the associated trip 
functions continue to protect the RPS scram solenoids and main steam 
isolation valve (MSIV) solenoids so that these devices will perform 
their intended safety function. Thus, the affected equipment is 
still required to be maintained Operable and capable of performing 
the accident mitigation functions assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly affected.
    Therefore, based on the above, this change does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS change revises the Reactor Protection System 
(RPS) Electric Power Monitoring Assembly overvoltage and 
undervoltage Allowable Values. The proposed change does not involve 
a physical alteration of the plant (no new or different type of 
equipment will be installed) or a change in the methods governing 
normal plant operation. The revised setpoints will continue to 
ensure that the RPS bus would be disconnected from its power supply 
under specified conditions that could damage the RPS bus powered 
equipment. Thus, this change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.

    The proposed TS change revises the Reactor Protection System 
(RPS) Electric Power Monitoring Assembly overvoltage and 
undervoltage Allowable Values. The proposed change provides 
necessary conservatism in the Allowable Values in the RPS 
Surveillance Requirement to ensure that the equipment used to meet 
the Limiting Condition for Operation (i.e., each of the two electric 
power monitoring assemblies) can continue to perform its required 
functions. At the same time, the revised setpoint/Allowable Values 
continue to provide adequate margin to the expected operating 
voltage range to prevent inadvertent or unnecessary tripping of the 
electric power monitoring assemblies (thus preventing unnecessary or 
excessive transfer to the alternate power source). The affected 
equipment will thus continue to be tested (calibrated and 
functionally tested) in a manner that gives confidence that the 
equipment can perform its assumed safety function. Therefore, this 
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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius 
LLP, 1800 M Street, NW, Washington, DC 20036.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 25, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification definitions for channel 
calibrations, channel functional tests, and logic system functional 
tests in accordance with Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-205, Revision 3, 
``Revision of Channel Calibration, Channel Functional Test, and Related 
Definitions.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The proposed change clarifies the Technical Specification 
requirements for performance of channel calibrations, channel 
functional test, and logic system functional tests. Specifically, 
the proposed change incorporates the NRC-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification 
Change Traveler, TSTF-205, Revision 3, ``Revision of Channel 
Calibration, Channel Functional Test, and Related Definitions.'' The 
change approved per this TSTF is not expected to adversely affect 
the performance and effectiveness of required testing as testing 
appropriate to the associated Surveillance Requirements will 
continue to be performed. The proposed change does not have a 
detrimental impact on the condition or performance of any plant 
structure, system, or component that initiates an analyzed event. 
Consequently, the probability of an accident previously evaluated is 
not significantly increased. The equipment being tested is still 
required to be operable and capable of performing the accident 
mitigation functions assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly affected. Therefore, this change does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The scope of the proposed change is limited to the clarification 
of existing test requirements. As such, the proposed change does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or a change in the methods governing 
normal plant operation. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    As noted above, the proposed change clarifies requirements for 
the performance of channel calibrations, channel functional tests, 
and logic system functional tests. Specifically, the proposed change 
incorporates the NRC-approved Technical Specification Task Force 
(TSTF) Standard Technical Specification Change Traveler, TSTF-205, 
Revision 3, ``Revision of Channel Calibration, Channel Functional 
Test, and Related Definitions.'' No changes or setpoints to plant 
process limits are involved. The surveillance requirements as 
revised will continue to ensure that affected equipment is tested in 
a manner that gives confidence that the equipment can perform its 
appropriate safety function. Therefore, this 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius 
LLP, 1800 M Street, NW, Washington, DC 20036.
    NRC Section Chief: Anthony J. Mendiola.

[[Page 1921]]

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

    Date of amendment request: December 16, 1999.
    Description of amendment request: The proposed amendment would 
allow a one-time extension of some Technical Specification surveillance 
intervals to support elimination of a planned spring 2000 mid-cycle 
outage (PO-8). For the applicable surveillances, the licensee proposes 
to extend their current surveillance intervals to November 30, 2000, 
the scheduled startup date from refueling outage 7 (RF-7).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The proposed Technical Specification (TS) changes involve a one-
time only change in the surveillance test intervals of selected 
Surveillance Requirements (SRs). The proposed TS changes do not 
impact the TS surveillance performance requirements themselves nor 
the way in which the surveillances are performed. The proposed TS 
changes do not physically involve any changes to the plant, nor do 
they impact any design or functional requirements of the associated 
systems. Thus, the proposed TS changes do not increase the 
challenges of any safety systems assumed to function in the accident 
analysis.
    In addition, the proposed TS changes do not significantly affect 
the availability of equipment or systems required to mitigate the 
consequences of an accident because (1) extension of the test 
intervals to the extent requested is not expected to have a 
significant impact on availability (i.e., no extended test interval 
would exceed 30 months), and (2) other or more frequent testing 
performed for the affected systems or components, as well as for 
redundant systems or components, supports continued availability of 
the affected function. The equipment subject to testing per the 
affected SRs is still required to be operable and capable of 
performing any accident mitigation functions assumed in the accident 
analysis. Furthermore, a historical review of surveillance test 
results identified no failures that would invalidate these 
conclusions.
    Based on the above, the proposed TS changes do not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes involve a one-time only change in the 
surveillance testing intervals of selected SRs. Such changes do not 
introduce any failure mechanisms of a different type than those 
previously evaluated since there are no physical changes being made 
to the facility. In addition, the surveillance test requirements 
themselves, and the way surveillance tests are performed, will 
remain unchanged. Therefore, the proposed TS changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The one-time extended surveillance frequencies do not result in 
a significant reduction in the margin of safety. Although the 
proposed TS changes will result in an increase in the interval 
between surveillance tests, the impact, if any, on system 
availability is small. This is because, as noted previously, 
extension of the test intervals to the limited extent proposed would 
not be expected to have a significant impact on availability. Other 
or more frequent testing performed for the affected systems or 
components, as well as the testing performed for redundant systems 
or components, supports continued availability of the affected 
functions.
    In addition, the proposed changes do not involve any physical 
changes to the affected systems or components, nor do they involve 
any changes to setpoints, operating limits, or safety limits.
    Based on the above, the assumptions in the licensing basis are 
not impacted, and the proposed TS changes do not significantly 
reduce a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius 
LLP, 1800 M Street, NW, Washington, DC 20036.
    NRC Section Chief: Anthony J. Mendiola.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: September 17, 1999.
    Description of amendment request: The proposed change to the 
Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Technical 
Specifications would lower the maximum limit for contents of the 
gaseous radioactive system from 300,000 curies (Ci) to 78,782 Ci and 
82,400 Ci for ANO-1 and ANO-2, respectively. This limit would ensure 
that, upon an uncontrolled release of the tank's contents over a 2-hour 
period, the resulting total whole body exposure to a member of the 
public at the nearest exclusion area boundary would not exceed 0.5 
roentgen equivalent man (rem).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    An evaluation of the proposed change has been performed in 
accordance with 10CFR50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10CFR50.92(c). A discussion of 
these standards as they relate to this amendment request follows:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The proposed change to lower the current technical specification 
(TS) gas storage tank activity limits does not require new hardware 
or physical equipment modifications to the plant design. By lowering 
the setpoint, the resultant exposure at the exclusion area boundary 
upon an inadvertent release of a gas storage tank's content will be 
limited to 0.5 rem. Therefore the consequences of such an 
uncontrolled release of activity are effectively reduced. 
Additionally, no new accident is introduced by the proposed 
reduction in activity limits associated with the gas storage tanks.
    Therefore, reducing the gas storage tank limits from 300,000 
Curies (Ci) to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2, 
respectively) does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From any Previously Evaluated

    The proposed change affects the consequences of an event 
associated with the loss of gas storage tank radioactive contents on 
either ANO-1 or ANO-2. Since this event has been previously 
evaluated, no new or different accident can be associated with the 
proposed change. Decreasing the present TS activity limits results 
in an exposure at the exclusion area boundary to be limited to 0.5 
rem in the event of an inadvertent release of a gas storage tank's 
content.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The proposed change conservatively lowers the existing TS GRW 
[Gaseous Radwaste] System gas storage tank activity limits from 
300,000 Ci to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2 
respectively). In doing so, the resultant exposure to a member of 
the public at the exclusion area boundary during an inadvertent 
release of gas storage tank contents over a two-hour period is 
reduced to 0.5 rem or less. The proposed change, therefore, retains 
the margin of safety for both ANO-1 and ANO-2.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based on the reasoning presented above and the 
previous discussion

[[Page 1922]]

of the amendment request, Entergy Operations, Inc. has determined 
that the requested change does not involve a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: December 16, 1999.
    Description of amendment request: The proposed license amendment 
request would revise Fuel Handling Accident (FHA) dose calculations for 
3 scenarios documented in the River Bend Station, Unit 1 (RBS), Updated 
Safety Analysis Report (USAR). The first is a FHA in the fuel building, 
assumed to occur 24 hours post-shutdown. A second FHA analysis was 
prepared to support Amendment 35 to RBS Technical Specifications (TS) 
which assumed a FHA occurs in the primary containment 80 hours post-
shutdown during Local Leakage Rate Testing (LLRT). A third analysis was 
prepared in support of Amendment 85 to the RBS TS which assumed the 
containment is open at 11 days.
    These analyses are being updated to account for several changes. 
The primary reason for the revisions, as stated by the licensee, was to 
update the analyses to reflect current RBS operating strategies and 
make the analyses consistent with each other. Specifically, Cases 1 and 
2 of the three analyses assumed a Radial Peaking Factor (RPF) of 1.5 
consistent with Regulatory Guide (RG) 1.25. However, current core 
design strategies could lead to an RPF as high as 1.65. In addition, to 
account for the potential impact of extended burnup fuel in future 
operating cycles, an increased iodine-131 gap fraction of 0.12 was more 
conservatively assumed in lieu of the 0.10 recommended by RG 1.25. The 
revised analysis also includes a change to the control room atmospheric 
dispersion factors (/Q) for the Main Control Room (MCR) 
ventilation system. Credit is taken for Standard Review Plan (SRP) 
Section 6.4 guidance for manual dual control room air intakes in that 
the /Q's are divided by 4. The revised FHA analyses also 
credit this action at a 20 minute delay to be consistent with the Loss 
of Coolant Accident (LOCA) analysis.
    Furthermore, an error was discovered in one of the FHA 
calculations. The release rate assumed in the analysis did not ensure 
that the RG 1.25 assumption of a 2-hour release was preserved. The 
error is the result of an inherent bias in the secondary mixing effects 
in the dose calculation. The results continue to be bounded by the 
guidance contained in SRP 15.7.4 and RG 1.25.
    Reanalysis showed that the release rate error, compounded with the 
other changes discussed above, resulted in calculated doses greater 
than those currently found in the RBS USAR. In addition, some of the 
doses were also greater than those presented in the Amendment 85 
submittal. However, the licensee has stated that the results of the 
revised analyses remain ``well within'' 10 CFR 100, the guidance 
contained in SRP 15.7.4, and RG 1.25. Since the analyses results are 
above those reported in the RBS USAR, the criterion of 10 CFR 
50.59(a)(2)(i) is, therefore, satisfied. Accordingly, the licensee has 
concluded that these changes involve an unreviewed safety question.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The analyses changes described by this proposed change to the 
USAR are not initiators to events, and, therefore, do not involve 
the probability of an accident. The changes to the FHA calculations 
for radiological doses following a FHA reflect the current operating 
strategies and make the analyses consistent. These changes included:
     Accounting for the impact of extended burnup fuel,
     Addressing a change to the control room atmospheric 
dispersion factors assumed in the analysis, and
     Revising the Radial Peaking Factor (RPF) used in the 
analysis. Current core design strategies could lead to a RPF higher 
[than] that assumed in Regulatory Guide 1.25.
    The TRANSACT code is used for offsite dose and control room dose 
calculations. The TRANSACT code is derived from the TACT V code 
documented in
    NUREG/CR-5109. RBS has benchmarked the TRANSACT code as 
discussed in the request dated August 17, 1995, (RBG-41728) which 
resulted in the NRC granting Amendment 85.
    The revisions to the FHA are used to establish operational 
conditions where specific activities represent situations where 
significant radioactive releases can be postulated. These 
operational conditions include:
     Initial fuel movement in the Fuel Building 24 hours 
after shutdown,
     Fuel movement in Primary Containment after 80 hours 
with leakrate testing being conducted, and
     Fuel movement in Primary Containment with the Primary 
Containment open.
    Because the analyses affected by the changes are not considered 
an initiator to any previously analyzed accident, these changes 
cannot increase the probability of any previously evaluated 
accident. Therefore, this change does not increase the probability 
of occurrence of an accident evaluated previously in the safety 
analysis report (SAR).
    This proposed change to the USAR does increase the consequences 
of an accident, but the increase is within all regulatory limits and 
guidance. While the calculated off-site and control room doses of a 
FHA did increase, the dose consequences remain below the regulatory 
limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC [General Design 
Criterion]-19 as approved per NUREG-0989, and the guidance contained 
in SRP 15.7.4 of less than 25% of the 10 CFR 100 limits. The cause 
of these events remains the failure of the fuel assembly lifting 
mechanism. These analyses demonstrated that for the worst case 
bundle drop, the regulatory dose guidelines of SRP 15.7.4 continue 
to be satisfied for the required decay periods.
    This change accounts for the potential effects of current fuel 
design and operating strategies including increased burnup of fuel, 
increased iodine-131 fraction released, Main Control Room 
ventilation system operation, and release rate timing assumptions. 
Reanalysis of the off-site dose calculation demonstrates that the 
revised doses are increased but remain less than the regulatory 
limits of 10 CFR 100 and within the guidance of SRP 15.7.4. 
Therefore, this change does not significantly increase the 
consequences of an accident previously evaluated in the SAR.
    The proposed changes, in conjunction with existing 
administrative controls, bound the conditions of the current design 
basis fuel handling accident analysis. The analysis also concludes 
the limiting offsite radiological consequences are well within the 
acceptance criteria of NUREG[-]0800, Section 15.7.4 and 10 CFR 50, 
Appendix A, GDC[-]19. The analysis is also conducted in a 
conservative manner containing margins in the calculation of 
mechanical analysis, iodine inventory, and iodine decontamination 
factor. Each of these conservatisms will further decrease the 
consequences. Therefore, the proposed changes do not significantly 
increase the probability or consequences of any previously evaluated 
accident.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous[ly] analyzed.
    This change does not involve initiators to any events in the 
SAR, nor does the activity create the possibility for any new 
accidents. Rather, this change is a result of the

[[Page 1923]]

evaluation of the most limiting FHA, which can occur at River Bend.
    The proposed changes to the dose analyses are consistent with 
previous limits, only revising previous evaluations to account for 
current operating strategies and assumptions. These changes 
included:
     Accounting for the impact of extended burnup fuel,
     addressing a change to the control room atmospheric 
dispersion factors assumed in the analysis, and
     Revising the Radial Peaking Factor (RPF) used in the 
analysis. Current core design strategies could lead to a RPF higher 
[than] that assumed in Regulatory Guide 1.25.
    The radiological consequences remain within accepted limits of 
10 CFR 100 and guidance of the Standard Review Plan (NUREG-0800) 
Section 15.7.4. Therefore, these changes are consistent with the 
design basis analysis. The proposed changes do not introduce any new 
modes of plant operation and do not involve physical modifications 
to the plant. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previous[ly] analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The dose consequences are calculated in accordance with 
regulatory guidance found in Regulatory Guide 1.25 and the SRP 
[S]ection 15.7.4. The RBS analyses conservatively assumed that 
failures are consistent with those in the standard General Electric 
GESTAR II. These analyses result in a bounding number of fuel 
failures. The RBS analyses are also consistent with those approved 
by the NRC [Nuclear Regulatory Commission] in support of Technical 
Specification Amendments 35 and 85 to the River Bend Station license 
(NPF-47). The radiological dose consequences resulting from these 
failures are therefore analyzed using accepted methods and criteria. 
In addition, the analyses contain known conservatisms and margins to 
ensure the results will remain bounding.
    The revised limits are used to establish operational conditions 
where specific activities represent situations where significant 
radioactive releases can be postulated. These operational conditions 
are consistent with the design basis analysis and are established 
such that the radiological consequences are at or below the current 
regulatory limits and guidance. Safety margins and analytical 
conservatisms have been evaluated and are well understood. 
Conservative methods of analysis are maintained through the use of 
accepted methodology and benchmarking the proposed methods to 
previous analysis. Margins are retained to ensure that the analysis 
adequately bounds all postulated event scenarios. The proposed 
change only eliminates some excess conservatism from the analysis.
    In addition, EOI [Entergy Operations, Inc.] has implemented 
NUMARC [Nuclear Management and Resources Council (now NEI)] 91-06 
guidelines for shutdown operations at RBS. Shutdown Operations 
Protection Plan and Primary-Secondary Containment Integrity 
procedures presently include guidance for closure of the containment 
hatch and other significant openings in containment, in addition to 
the requirements contained in the license and design basis. This 
additional protection will enhance the ability to limit offsite 
effects.
    Acceptance limits for the fuel handling accident are provided in 
10 CFR 100 with additional guidance provided in NUREG[-]
0800, Section 15.7.4. The proposed changes continue to ensure that the 
whole-body and thyroid doses at the exclusion area and low population 
zone boundaries, as well as control room doses, are below the 
corresponding regulatory limits. These margins are unchanged, 
therefore, the proposed changes do not involve a significant reduction 
in a margin of safety.
    The commission has provided guidance concerning the application 
of the standards of 10 CFR 50.92 by providing certain examples (51 
FR 7751, March 6, 1986) of amendments that are not considered likely 
to involve a significant hazards consideration. This proposed 
amendment is very similar to example (vi):
    (vi) A change which either may result in some increase to the 
probability or consequences of a previously-analyzed accident or may 
reduce in some way a safety margin, but where the results of the 
change are clearly within all acceptable criteria with respect to 
the system or component specified in the Standard Review Plan: for 
example, a change resulting from the application of a small 
refinement of a previously used calculational model or design 
method.
    As we have shown in the preceding discussion, this refinement to 
the FHA dose calculation results in a small increase to the 
consequences of a previously analyzed accident, but the results of 
the change remain clearly within the guidelines of 10 CFR 100, 
Appendix A, GDC[-]19, and the guidance of SRP [S]ection 15.7.4, 
without reducing 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.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 17, 1999 (L-99-241)
    Description of amendment request: The proposed amendment would 
revise the St. Lucie Unit 1 and Unit 2 Technical Specifications (TS) to 
require laboratory testing of activated charcoal samples for applicable 
engineered safety feature ventilation systems using the American 
Society for Testing and Materials (ASTM) D3803-1989 protocol. The 
affected TS are Units 1 and 2 shield building ventilation system, TS 
4.6.6.1; Unit 1 emergency core cooling system area ventilation system, 
TS 4.7.8.1; Unit 1 control room emergency ventilation system, TS 
4.7.7.1; Unit 2 control room emergency air cleanup system, TS 4.7.7; 
and Unit 1 fuel pool ventilation system--fuel storage, TS 4.9.12.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated. The new charcoal testing protocol is performed offsite on 
samples extracted from the safety-related ventilation systems. 
Therefore, there is no impact on any accident initiator and 
therefore, no changes on the probability. The proposed testing 
protocol is more conservative than previous tests; therefore, the 
efficiency of charcoal for the affected safety-related systems would 
not be overestimated. With the new testing protocol, more 
conservative testing results are expected since the temperature at 
which testing is performed is lower and the charcoal retention 
capability is more consistent with actual accident conditions. The 
proposed change thus ensures that the charcoal in service will 
comply with the penetration requirements to meet the design basis 
accident conditions.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed new charcoal testing protocol only affects 
surveillance testing requirements for ventilation systems. The 
functions of these systems remain unchanged and unaffected. No new 
system interactions have been introduced by the proposed amendment, 
which would create a new or different type of accident than 
previously analyzed. No physical changes are being made to any 
structure, system or component. The operation of the facility has 
not been altered by the proposed amendment. The systems involved are 
not considered to initiate any accidents as previously evaluated.

[[Page 1924]]

    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment, nor do they alter the design of St. Lucie plant 
systems. Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.

    The proposed amendment does not involve a reduction in the 
margin of safety. The margin of safety of the Technical 
Specifications, its bases, the Final Safety Analysis Report, the 
Safety Evaluation Report or in any other design document has not 
been affected by the proposed amendment. The change provided in this 
proposed amendment is related to introducing an improved testing 
protocol for the activated charcoal in safety related ventilation 
systems. The change consists of testing the charcoal with a new 
testing protocol and with lower test temperatures to resemble more 
closely accident conditions and to eliminate potential 
overestimation of charcoal efficiency.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Section Chief: Richard P. Correia

GPU Nuclear, Inc., et al., Three Mile Island Nuclear Station, Unit 2 
(TMI-2), Docket No. 50-320, Dauphin County, Pennsylvania

    Date of amendment request: November 5, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Sections 6.8.1.4, 6.5.4.6, 6.13, 
6.14, and 6.8.3. Specifically, Sections 6.13, 6.14 and 6.8.3 would be 
revised to eliminate the requirement to notify the Nuclear Regulatory 
Commission (NRC) of exceeding environmental limits and changes to 
environmental permits such as National Pollution Discharge Elimination 
System (NPDES). The requirements contained in the individual 
environmental permits and program regulations administered by the U.S. 
Environment Protection Agency (EPA), Pennsylvania Department of 
Environmental Protection (PADEP), and other regulatory agencies with 
program jurisdiction for reporting are included in plant procedures and 
data base tracking systems. Sections 6.8.1.4 and 6.5.4.6 are changes to 
the amendment that are administrative in nature and reflect a 
streamlining of the GPU Nuclear, Inc. management structure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes to the TMI-2 [Three Mile Island, Unit 2] 
Technical Specifications do not involve a significant increase in 
the probability of occurrence or consequences of an accident or 
malfunction of equipment important to safety previously analyzed in 
the safety analysis report. The changes have no impact on plant 
operations or the release of radioactive materials.
    2. The proposed changes to the TMI-2 Technical Specifications 
will not create the possibility for an accident or malfunction of a 
different type than any previously evaluated in the safety analysis 
report because no plant configuration or operational changes are 
involved.
    3. The changes will not involve a significant reduction in the 
margin of safety as defined in the basis for any technical 
specification for TMI-2 because no change to operational limits will 
be made.
    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.
    Attorney for Licensee: Ernest L. Blake, Jr Esq., Shaw, Pittman, 
Potts & Trowbridge, 2300 N. Street, N.W., Washington, DC 20037.
    NRC Section Chief: Mike Masnik.

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

    Date of amendment request: November 10, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.7.c, to commit to the American 
Society for Testing and Materials (ASTM) D3803-1989 test protocol for 
the ventilation filter testing program. The proposed changes are 
consistent with Attachment 2, Sample Technical Specifications, in 
Generic Letter 99-02. Because the current TS penetration limits do not 
reflect a safety factor in excess of that assumed in the dose 
calculations of the accident analysis, the TS change request would also 
revise the allowable penetration values to correspond to a safety 
factor of 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below for the administrative changes:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The ESF [engineered safety feature] ventilation systems are not 
initiators of any accident previously evaluated and the change in 
testing protocol to ASTM D3803-1989 as requested by the NRC will be 
more accurate and realistic and provide greater assurance of 
consistency. The acceptance criteria will be more conservative than 
those currently used in TS 5.5.7.c.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No new types of accidents are being introduced because no 
modifications or changes in operations are being proposed for the 
ESF [engineered safety feature] ventilation systems. The proposed 
changes to TS 5.5.7.c impact acceptance criteria and test protocols 
only.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The margin of safety is not reduced. The proposed change in ESF 
ventilation testing protocol includes a safety factor of two (2) for 
the penetration limit in excess of that assumed in the dose 
calculations of the DAEC [Duane Arnold Energy Center] accident 
analysis.

    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.
    Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: November 22, 1999.
    Description of amendment request: The proposed amendment would 
adopt selected NRC-approved generic changes to the Improved Technical 
Specifications (ITS) NUREGs. The 16 changes come from the Technical 
Specification Task Force (TSTF) process

[[Page 1925]]

developed by the industry and the NRC. Three of these changes are 
Bases-only changes but are included for completeness relative to the 
TSTF process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below for the administrative changes:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves reformatting, renumbering, and 
rewording the existing Technical Specifications. The reformatting, 
renumbering, and rewording process involves no technical changes to 
the existing Technical Specifications. As such, this change is 
administrative in nature and does not affect initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose any new or eliminate any old requirements. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analyses assumptions. This change is 
administrative in nature. Therefore, the change does not involve a 
significant reduction in a margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below for more restrictive changes:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change provides more stringent requirements for 
operation of the facility. These more stringent requirements do not 
result in operation that will increase the probability of initiating 
an analyzed event and do not alter assumptions relative to 
mitigation of an accident or transient event. The more restrictive 
requirements continue to ensure process variables, structures, 
systems, and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change does impose different requirements. However, these changes 
are consistent with the assumptions in the safety analyses and 
licensing basis. Thus, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The imposition of more restrictive requirements either has no 
effect on or increases the margin of plant safety. As provided in 
the justification, each change in this category is, by definition, 
providing additional restrictions to enhance plant safety. The 
change maintains requirements within the safety analyses and 
licensing basis. Therefore, the change does not involve a 
significant reduction in a margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below for less restrictive changes--removed detail:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates certain details from the Technical 
Specifications to other documents under regulatory control. The 
Bases, UFSAR [updated final safety analysis report], and Technical 
Requirements Manual will be maintained in accordance with 10 CFR 
50.59. In addition to 10 CFR 50.59 provisions, the Technical 
Specification Bases are subject to the change control provisions in 
the Administrative Controls Chapter of the Technical Specification. 
The UFSAR is subject to the change control provisions of 10 CFR 
50.71(e). Other documents are subject to controls imposed by 
Technical Specifications or regulations. Since any changes to these 
documents will be evaluated, no significant increase in the 
probability or consequences of an accident previously evaluated will 
be allowed. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not impose or eliminate any requirements and 
adequate control of the information will be maintained. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analyses assumptions. In addition, 
the details to be moved from the Technical Specifications to other 
documents are the same as the existing Technical Specifications. 
Since any future changes to these details will be evaluated, no 
significant reduction in a margin of safety will be allowed. A 
significant reduction in the margin of safety is not associated with 
the elimination of the 10 CFR 50.92 requirement for NRC review and 
approval of future changes to the relocated details. The proposed 
change is consistent with the BWR [Boiling Water Reactor]/4 Standard 
Technical Specifications, NUREG-1433, issued by the NRC Staff, 
revising the Technical Specifications to reflect the approved level 
of detail, which indicates that there is no significant reduction in 
the margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below for less restrictive changes--category 3, relaxation of 
completion time:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relaxes the Completion Time for a Required 
Action. Required Actions and their associated Completion Times are 
not initiating conditions for any accident previously evaluated and 
the accident analyses do not assume that required equipment is out 
of service prior to the analyzed event. Consequently, the relaxed 
Completion Time does not significantly increase the probability of 
any accident previously evaluated. The consequences of an analyzed 
accident during the relaxed Completion Time are the same as the 
consequences during the existing Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
Required Actions and associated Completion Times have been evaluated 
to ensure that no new accident initiators are introduced. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The relaxed Completion Time for a Required Action does not 
involve a significant reduction in the margin of safety. As provided 
in the justification, the change has been evaluated to ensure that 
the allowed Completion Time is consistent with the safe operation 
under the specified Condition, considering the operability status of 
the

[[Page 1926]]

redundant systems of required features, the capacity and capability 
of remaining features, a reasonable time for repairs or replacement 
of required features, and the low probability of a DBA [design basis 
accident] occurring during the repair period. Therefore, this change 
does not involve a significant reduction in a margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below for less restrictive changes--category 4, relaxation of 
required action.

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relaxes Required Actions. Required Actions 
and their associated Completion Times are not initiating conditions 
for any accident previously evaluated and the accident analyses do 
not assume that required equipment is out of service prior to the 
analyzed event. Consequently, the relaxed Required Actions do not 
significantly increase the probability of any accident previously 
evaluated. The Required Actions in the change have been developed to 
provide assurance that appropriate remedial actions are taken in 
response to the degraded condition, considering the operability 
status of the redundant systems of required features, and the 
capacity and capability of remaining features while minimizing the 
risk associated with continued operation. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
Required Actions and associated Completion Times in the change have 
been evaluated to ensure that no new accident initiators are 
introduced. Thus, this change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The relaxed Required Actions do not involve a significant 
reduction in the margin of safety. As provided in the justification, 
the change has been evaluated to minimize the risk of continued 
operation under the specified Condition, considering the operability 
status of the redundant systems of required features, the capacity 
and capability of remaining features, a reasonable time for repairs 
or replacement of required features, and the low probability of a 
DBA [design basis accident] occurring during the repair period. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below for less restrictive changes--category 6, relaxation of 
surveillance requirement acceptance criteria:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relaxes the acceptance criteria of 
Surveillance Requirements. Surveillances are not initiators to any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be Operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The relaxed acceptance criteria for Surveillance Requirements do 
not result in a significant reduction in the margin of safety. As 
provided in the justification, the relaxed Surveillance Requirement 
acceptance criteria have been evaluated to ensure that they are 
sufficient to verify that the equipment used to meet the LCO 
[limiting condition for operation] can perform its required 
functions. Thus, appropriate equipment continues to be tested in a 
manner that gives confidence that the equipment can perform its 
assumed safety function. Therefore, this 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

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: September 3, 1998, as supplemented by 
letters dated January 22, February 5, March 17, and November 24, 1999. 
The September 3, 1998, amendment application was previously noticed in 
the Federal Register on December 16, 1998 (63 FR 69345).
    Description of amendment requests: The amendment would revise 
Section 5.6.6, ``Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE 
LIMITS REPORT (PTLR),'' of the improved Technical Specifications (TSs), 
that were issued in Amendment Nos. 135 and 135 on May 28, 1999. The 
amendment would add the phrase ``and LTOP'' (low-temperature 
overpressure protection) to the first sentence of item 5.6.6.b that 
identifies the limits that can be determined by the licensee in the 
PTLR, and (2) replace the current list of documents listed in item 
5.6.6.b by the NRC letter that would approve this amendment and 
Westinghouse WCAP-14040-NP-A, ``Methodology Used to Develop Cold 
Overpressure Mitigation System Setpoints and RCS Heatup and Cooldown 
Limit Curves,'' dated January 1996. WCAP-14040-NP-A is the NRC-approved 
topical report which provides a methodology for developing the LTOP 
setpoints and RCS heatup and cooldown limit curves for Westinghouse 
plants, such as Diablo Canyon Nuclear Power Plant, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to Figures 3.4-2 and 3.4-3 of Technical 
Specification (TS) 3.4.9.1 and the associated Bases adjust the 
reactor coolant system (RCS) heatup and cooldown pressure/
temperature (P/T) limits to permit operation through 16 effective 
full power years (EFPY). The 16 EFPY P/T limits are more restrictive 
than the current limits; this accounts for an expected incremental 
increase in reactor vessel embrittlement, and assures the reactors 
will continue to be operated within acceptable stresses and at 
temperatures for which the reactor vessel metal exhibits ductile 
properties. The P/T limits developed for 16 EFPY were determined in 
accordance with 10 CFR 50, Appendix G, and maintain the same margins 
of safety as the current limits. The proposed changes will not 
impact the probability of overpressurization or brittle fracture of 
the vessel, and therefore will not impact the consequences of an 
accident.
    The present low temperature overpressure protection (LTOP) 
pressure and enable temperature setpoints were reviewed and found to 
be acceptable and conservative for

[[Page 1927]]

use through 16 EFPY, based on use of ASME [American Society of 
Mechanical Engineers] Code Case N-514, which provides acceptable 
margins to the prevention of vessel overpressurization and brittle 
fracture. Therefore, there is no change to the consequences of 
accidents previously analyzed. Since no changes are proposed in the 
actual LTOP setpoints, nor any physical alteration of the LTOP 
system, nor a change to the method by which the LTOP system performs 
its function, there would be no change to the probability of an 
accident previously evaluated. The proposed change to the Bases 
incorporates use of ASME Code Case N-514, which will benefit DCPP 
[Diablo Canyon Power Plant] by not resulting in a reduced RCS P/T 
window and reduced power-operated relief valve (PORV) pressure 
setpoint for LTOP. This maintains the current level of operator 
flexibility during heatup and cooldown, and prevents an increase in 
the probability of an accident associated with an inadvertent PORV 
actuation.
    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 changes to TS 3.4.9.1, ``Reactor Coolant System--
Pressure/Temperature Limits,'' do not involve any physical 
alteration to any plant system or change the method by which any 
safety-related system performs its function. The changes to TS 
3.4.9.1 account for the effects of an incremental increase in 
reactor vessel embrittlement and are requested in order to restrict 
future reactor operation to within acceptable stress levels and 
temperature regimes in accordance with 10 CFR 50, Appendix G, 
requirements. These changes are needed to maintain the current P/T 
limit margins of safety as defined by 10 CFR 50, Appendix G, and 
ASME XI, Appendix G, for operation through 16 EFPY. The possibility 
of a new kind of accident such as catastrophic failure of the 
reactor vessel is prevented by maintaining acceptable margins of 
safety.
    The present LTOP pressure setpoint was reviewed and found to be 
acceptable and conservative for the extension of the P/T curves to 
16 EFPY.
    Additionally, the proposed changes will not affect the ability 
of the LTOP system to provide pressure relief at low temperatures, 
thereby maintaining the LTOP design basis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes to TS 3.4.9.1 adjust the RCS heatup and 
cooldown P/T limits to permit operation through 16 EFPY. The P/T 
limits have been determined in accordance with 10 CFR 50, Appendix 
G, and include the safety margins with regard to brittle fracture 
required by the ASME Section XI, Appendix G, which maintain the same 
margins of safety as the current limits.
    The LTOP setpoints were reevaluated using the requirements of 
ASME Code Case N-514. This code case was developed to provide the 
necessary margins of safety for the prevention of reactor vessel 
overpressurization and brittle fracture. The LTOP evaluation results 
conclude the current LTOP setpoints are conservative for operation 
through 16 EFPY. In addition, avoiding an unnecessary reduction in 
the LTOP, the PORV pressure setpoint prevents an increase in the 
likelihood of an inadvertent PORV actuation
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: December 1, 1999.
    Description of amendment request: The proposed amendment would 
revise the Humboldt Bay Power Plant (HBPP) Unit 3 Technical 
Specifications (TS) related to fire protection, administrative 
controls, and quality assurance audits. The fire protection 
requirements would be relocated verbatim from the TS to the HBPP 
Defueled Safety Analysis Report (DSAR). The administrative controls 
requirements would be revised to (1) refer to the DSAR for a 
description of the plant organization, (2) modify information 
pertaining to plant staff titles and qualifications to reflect the 
current organization, and (3) replace a reference to the Final Hazards 
Summary Report with a reference to the DSAR. Quality assurance audit 
requirements would be relocated from the TS to the Quality Assurance 
Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration, which are presented below:
    For the proposed changes to the fire protection requirement, the 
licensee's analysis states:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The FPP [Fire Protection Program] and FPS [Fire Protection 
System] are not being changed. Operability requirements and 
procedural controls of the FPP and FPS are not being changed. The 
proposed changes involve only where the FPP and FPS description is 
located and how changes can be made. Consequently, the changes will 
not affect the probability or consequences of an accident occurring.
    Future changes to the FPP and FPS as described in the Defueled 
Safety Analysis Report would be made in accordance with 10 CFR 
50.59. This ensures that adequate controls will remain in place so 
that the public health and safety will be protected.
    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 FPP and FPS are not being changed. Operability requirements 
and procedural controls of the FPP and FPS are not being changed. 
The proposed changes involve only where the FPP and FPS description 
is located and how changes can be made. Consequently, the changes 
will not affect the probability or consequences of an accident 
occurring.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The FPP and FPS are not being changed. Operability requirements 
and procedural controls of the FPP and FPS are not being changed. 
The proposed changes involve only where the FPP and FPS description 
is located and how changes can be made. Consequently, the changes 
will not affect the probability or consequences of an accident 
occurring.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    For the proposed changes to the administrative controls 
requirements, the licensee's analysis states:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The organization title and responsibility changes update the 
Technical Specification (TS) to reflect the current organization and 
have no impact on the function or operability of plant systems, 
structures, or components, or the ability of the plant to safely 
maintain SAFSTOR status. Consequently, the changes will not affect 
the probability or consequences of an accident occurring.
    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

[[Page 1928]]

accident from any accident previously evaluated.
    The organization title and responsibility changes update the TS 
to reflect the current organization and have no impact on the 
function or operability of plant systems, structures, or components, 
or the ability of the plant to safely maintain SAFSTOR status. 
Consequently, the changes will not affect the probability or 
consequences of an accident occurring.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The organization title and responsibility changes update the TS 
to reflect the current organization and have no impact on the 
function or operability of plant systems, structures, or components, 
or the ability of the plant to safely maintain SAFSTOR status. 
Consequently, the changes will not affect the probability or 
consequences of an accident occurring.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    For the proposed changes to the quality assurance audit 
requirements, the licensee's analysis states:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes simplify the Technical Specifications (TS), 
meet regulatory requirements for relocated TS, and implement: (1) 
The recommendations of NRC's letter dated October 25, 1993, from 
William T. Russell to the chairpersons of the industry owners 
groups; (2) the Commission's Final Policy Statement on TS 
Improvements; and (3) the current revision of 10 CFR 50.36. Future 
changes to these requirements will be controlled by 10 CFR 50.54. 
This ensures that adequate controls will remain in place so that the 
public health and safety will be protected. The proposed changes are 
administrative in nature and do not involve any modifications to any 
plant equipment or affect plant operation.
    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 changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by 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 change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not alter implementation of the basic 
regulatory requirements and do not affect any safety analyses. 
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 requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Michael Masnik.

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

    Date of application for amendments: October 12, 1999 (TS 99-15).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79 
(Unit 2) by revising the Technical Specification (TS) to provide for 
unisolation of containment penetrations under administrative controls. 
This revision will add a footnote to Specification 3.9.4.c indicating 
this allowance and the necessary Bases addition for this section to 
clarify the use of this allowance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision will allow the opening of specific 
containment penetrations during the movement of irradiated fuel or 
core alterations provided administrative controls are implemented. 
These controls will establish the proper awareness of the unisolated 
penetration condition, designate individuals to isolate the 
penetration in the event of an FHA [fuel handling accident], and 
[to] ensure the auxiliary building gas treatment system (ABGTS) is 
available. The status of containment penetrations does not impact 
the generation of an accident nor does the ability to unisolate 
penetrations affect this potential. The proposed revision does not 
alter any plant equipment or operating practices other than 
penetration isolation such that the probability of an accident is 
increased.
    The administrative controls provide adequate requirements to 
provide timely identification and closure of penetrations opened 
under this allowance should a fuel handling event occur. Designated 
individuals ensure that adequate resources are available to isolate 
the penetration such that the offsite dose consequences are not 
significantly impacted. The lack of motive force in containment 
during fuel movement to expel the radioactive material allows a more 
flexible isolation interval. The exception for the containment 
ventilation isolation valves is based on being exposed to a motive 
force and the flow paths outside the auxiliary building secondary 
containment enclosure (ABSCE) is based on being exposed to an 
unfiltered atmosphere. Timely isolation of the specified flow paths 
is required to ensure that the unlikely transmission of radioactive 
material does not occur. Interactions that may occur during the 
period of time before isolation will be controlled by operation of 
the ABGTS and will not significantly increase the consequences of an 
accident as previously evaluated. Completion of penetration 
isolation and operation of the ABGTS, as required by the 
administrative controls, will maintain the offsite dose consequences 
well within the 10 CFR 100 limits.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed allowance to open penetrations in Mode 6 will not 
alter plant functions or equipment operating practices other than 
penetration isolation. Containment penetration status is not 
considered to be the source of an accident. Therefore, since the 
plant functions and equipment are not altered and the isolation 
status of containment penetrations do not contribute to the 
initiation of postulated accidents, the proposed revision will not 
create a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The isolation requirements for containment penetrations ensure 
that the release of radioactivity is minimized to maintain the 10 
CFR 100 limits for offsite dose consequences in the event of an FHA. 
The proposed change to allow penetrations to be unisolated does not 
significantly affect the expected dose consequence because of the 
absence of containment pressurization potential during fuel movement 
or core alterations. The most significant offsite dose contributor 
to the fuel handling event is the containment purge system that 
generates a motive force for the radioactive material. This flow 
path is excluded from the proposed allowance because of this motive 
force potential along with flow paths outside the ABSCE. Without 
this motive force, as is the case with other penetrations during 
fuel movement or core alterations, the potential for additional 
offsite dose consequence is unlikely. As an additional measure, this 
allowance applies to flow paths that can be filtered by the ABGTS. 
Therefore, the margin of safety provided by the containment building 
penetration requirements is not significantly impacted by the 
proposed allowance to open penetrations under administrative 
controls. With the timely provision to identify and isolate affected 
penetrations and the provision for ABGTS operability, the margin of 
safety is maintained without a significant reduction.


[[Page 1929]]


    The NRC 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.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of application for amendments: November 24, 1999 (TS 99-16).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79 
(Unit 2) by updating the Technical Specification (TS) surveillance 
requirements for penetration efficiency tests of charcoal adsorbers to 
comply with American Society for Testing and Materials (ASTM) test 
standard ASTM D3803-1989 as directed by NRC Generic Letter (GL) 99-02.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision will require laboratory tests of safety-
related charcoal filter adsorbers to tighter specifications. NRC 
research indicates that the new test protocols yield more accurate 
measures of filter efficiency and better reproducibility of test 
results. No physical change is made to the filter by these expanded 
timeframes of testing and tighter controls; therefore, no change to 
the filter behavior is expected. Current methods for selecting and 
obtaining charcoal samples for testing will be retained without 
change. The proposed revision does not alter any plant equipment or 
operating practices other than filter tests that are conducted away 
from the plant site, and as such the probability of an accident is 
not increased.
    Laboratory test acceptance criteria contain a safety factor to 
ensure that the efficiency assumed in the accident analysis is still 
valid at the end of the operating cycle. Because ASTM D3803-1989 is 
a more accurate and demanding test than older tests, upgrading TSs 
to the ASTM D3803-1989 protocol allows use of a safety factor of 2 
for determining the acceptance criteria for charcoal filter 
efficiency. This safety factor can be used for systems with or 
without humidity control because the lack of humidity control is 
already accounted for in the test conditions.
    Applying the ASTM D3803-1989 test methodology and using the new 
safety factor is expected to yield a net improvement in safety. The 
ASTM D3803-1989 test protocol is expected to improve the 
identification of degraded charcoal filters and lead to their timely 
replacement without any adverse effects on filter performance. 
Therefore, the change in testing does not significantly increase the 
consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change in laboratory tests performed on charcoal 
filters will not alter plant functions or equipment operating 
practices other than possibly resulting in more frequent replacement 
of charcoal filters. As stated previously, current methods for 
selecting and obtaining charcoal samples for testing will be 
retained without change. The ASTM D3803-1989 test methodology is not 
expected to alter the filters; therefore, it will not adversely 
alter the resulting filter performance. Since the plant functions 
and equipment are not altered, the proposed revision will not create 
a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Analyses of design-basis accidents assume a particular ESF 
[Engineered Safety Feature] charcoal filter adsorption efficiency 
when calculating offsite and control room operator doses. Charcoal 
filter samples are tested to determine whether the filter adsorber 
efficiency is greater than that assumed in the design-basis accident 
analysis. The laboratory test acceptance criteria contains a safety 
factor to ensure that the efficiency assumed in the accident 
analysis is still valid at the end of the operating cycle. Because 
ASTM D3803-1989 is a more accurate and demanding test than older 
tests, NRC indicated in GL 99-02 that licensees upgrading their TS 
to this new protocol will be able to use a safety factor as low as 2 
for determining the acceptance criteria for charcoal filter 
efficiency. This safety factor can be used for systems with or 
without humidity control because the lack of humidity control is 
already accounted for in the test conditions. As stated in the GL, 
the new test protocol and associated safety factors have been 
reviewed and found to not significantly decrease the margin of 
safety.

    The NRC 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.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: December 16, 1999.
    Brief description of amendment: The proposed change would amend 
Technical Specification 4.18.5.b to allow tube 110/60 to remain in 
service through the current operating cycle (cycle 16) with two axial 
indications that have potential through-wall depths greater than the 
plugging limit. The axial indications are located in the roll 
transition region and are contained within the upper tubesheet.

    Date of publication of individual notice in Federal Register: 
December 29, 1999 (64 FR 73080).
    Expiration date of individual notice: Comments on no significant 
hazards considerations by January 12, 2000; requests for hearing by 
January 28, 2000. Clarification: The December 29, 1999, notice 
indicated that requests for a hearing with respect to issuance of this 
amendment must be filed by January 12, 2000. The correct deadline for 
this action is January 28, 2000.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: December 16, 1999.
    Description of amendment request: The proposed changes would revise 
the River Bend Station (RBS) Technical Requirements Manual, Section TR

[[Page 1930]]

3.9.14, and add an exception to the current prohibition for travel of 
loads in excess of 1200 pounds over fuel assemblies in the spent fuel 
storage pool. The exception would allow the licensee to move the spent 
fuel pool (SFP) watertight gates, which separate the SFP from the cask 
and lower transfer pools, to perform maintenance and repairs on the 
gates and watertight seals. Related sections of the RBS Updated Safety 
Analysis Report would also be revised to be consistent with the 
exception. The licensee determined that movement of the gate, with its 
associated rigging, over spent fuel would involve an unreviewed safety 
question in accordance with Title 10 of the Code of Federal 
Regulations, Section 50.59.
    Date of publication of individual notice in Federal Register: 
December 21, 1999 (64 FR 71511).
    Expiration date of individual notice: January 20, 2000. Correction: 
The December 21, 1999, notice indicated that requests for a hearing 
with respect to issuance of this amendment must be filed by January 28, 
2000. The correct deadline for this action is January 20, 2000.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.

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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of application for amendments: December 16, 1998, as 
supplemented July 16, September 29, and December 21, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications 3.8.1 and 3.37 to ensure that the appropriate actions 
are taken to prevent double sequencing of safety-related loads and that 
the setpoint allowable values for the degraded voltage relays reflect 
the required function of the relays.
    Date of issuance: December 29, 1999.
    Effective Date: December 29, 1999, to be implemented within 90 
days.
    Amendment Nos: Unit 1-123, Unit 2-123, Unit 3-123.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 24, 1999 (64 FR 
14279) The July 16, September 29, and December 21, 1999, letters 
provided additional clarifying information that was written within the 
scope of the original application and Federal Register notice and did 
not change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated December 29, 1999.
    No significant hazards consideration comments received: No.

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 1, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications as follows:
    1. Technical Specification 1.1 is changed to replace the definition 
of Azimuthal Power Tilt with a new definition.
    2. Technical Specification 2.1.1.2 is changed by replacing the peak 
linear heat rate safety limit with less than or equal to 22 kW/ft.
    3. Technical Specification Surveillance Requirement (SR) 3.3.6.2 is 
changed by replacing the degraded voltage function with transient 
degraded voltage and steady-state degraded voltage functions.
    4. Technical Specification SRs 3.8.1.9 and 3.8.1.15 are changed by 
replacing the steady-state voltage range with the range of greater than 
or equal to 4060 volts and less than or equal to 4400 volts.
    5. Technical Specification 5.6.5.a is changed by adding Technical 
Specifications 3.1.4 and 3.3.1 to the list.
    6. Technical Specification Figure 2.1.1-1 is changed by removing 
the reference to Figure B2.1-1.
    7. Various Technical Specifications and Figures 2.1.1-1a are 
changed by removing references to Unit 2, Cycle 12, and deleting Figure 
2.1.1-1a.
    8. Technical Specification 5.6.5.b, Item 41.ii is changed by 
correcting CEN-99(B)-P to CEN-119(B)-P.
    Date of issuance: December 15, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 232 and 208.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54372).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated December 15, 1999.
    No significant hazards consideration comments received: No.

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: August 6, 1999, as supplemented 
on November 15, 1999.
    Brief description of amendments: The amendments revised Technical

[[Page 1931]]

Specification 3/4.4.6, ``Vacuum Relief'' to remove specific operability 
requirements related to position indication for the suppression 
chamber-drywell vacuum breakers. The amendments also reformat the 
action statement for inoperable vacuum breakers, increase the 
surveillance interval for verifying that the vacuum breakers are 
closed, and delete the requirement to verify that the manual isolation 
valves are closed for an inoperable and open vacuum breaker.
    Date of issuance: December 21, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 138 and 122.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46428).
    The November 15, 1999, submittal provided additional clarifying 
information that did not change the staff's initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 16, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification 4.7.D.6 by replacing the leakage limit of 11.5 standard 
cubic feet per hour (scfh) for each main steam isolation valve (MSIV) 
with a limit of 46 scfh on the total combined leakage for the MSIVs of 
all four main steam lines.
    Date of issuance: December 21, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 192 and 188.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46429).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: October 15, 1999.
    Brief description of amendments: The amendments revise Section 
5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the 
Technical Specifications by adding a new paragraph. The existing single 
paragraph of Section 5.5.7 requires that inspection of each reactor 
coolant pump flywheel be done per the recommendations of Regulatory 
Position C.4.b of Regulatory Guide 1.14. The amendments add a new 
paragraph which specifies that in lieu of Regulatory Positions C.4.b(1) 
and C.4.b(2), alternative inspection techniques may be used. Date of 
issuance: December 21, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: 182 (Unit 1); 174 (Unit 2).
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62705).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 15, 1999.
    Brief description of amendments: The amendments revise Section 
5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the 
Technical Specifications by adding a new paragraph. The existing single 
paragraph of Section 5.5.7 requires that inspection of each reactor 
coolant pump flywheel be done per the recommendations of Regulatory 
Position C.4.b of Regulatory Guide 1.14. The amendments add a new 
paragraph which specifies that in lieu of Regulatory Positions C.4.b(1) 
and C.4.b(2), alternative inspection techniques may be used.
    Date of issuance: December 21, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: 190 (Unit 1); 171 (Unit 2).
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62706).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: December 18, 1998, as 
supplemented September 13, 1999.
    Brief description of amendment: This amendment revises the St. 
Lucie, Unit 2 (SL-2), Technical Specifications (TS) Index Page III, TS 
1.10, Dose Equivalent iodine-131; TS 2.1.1.2, Linear Heat Rate; TS 
3.1.1.1/4.1.1.1.1, Shutdown Margin--Tavg Greater than 200 
deg.F; TS 3/4.1.1.2, Shutdown Margin--Tavg Less Than or 
Equal to 200 deg.F; TS 3.1.2.2, Boration Systems Flow Paths--Operating; 
TS 3.1.2.4, Charging Pumps--Operating; TS 3.1.2.6, Boric Acid Makeup 
Pumps--Operating; TS 3.1.2.8, Borated Water Sources--Operating; and TS 
6.9.1.11, Core Operating Limits Report (COLR). The amendment also 
relocates the core operating limits for shutdown margin to the SL-2 
COLR. The following Bases have also been changed in connection with 
this amendment: TS Bases 2.1.1, Reactor Core; Bases Figure B2.1-1, 
Axial Power Distributions for Thermal Margin Safety Limits; TS Bases 
2.2.1, Reactor Trip Setpoints (Variable Power Level-High); TS Bases 3/
4.1.1.1 and 3/4.1.1.2, Shutdown Margin; and TS Bases 3/4.1.2, Boration 
Systems.
    Date of Issuance: December 21, 1999.
    Effective Date: As of date of issuance, to be implemented prior to 
fuel reload for Cycle 12.
    Amendment No.: 105.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6697). The supplemental letter dated September 13, 1999, provided 
additional information that did not expand the scope of the amendment 
request as noticed or change the original proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No.

[[Page 1932]]

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

    Date of application for amendments: April 26, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) for Turkey Point Units 3 and 4 to correct 
the Technical Specification Index and to remove inconsistencies, and 
make administrative changes. A portion of the request, related to the 
proposed deletion of dates for the approved security plans, was denied.
    Date of issuance: December 20, 1999.
    Effective date: December 20, 1999.
    Amendment Nos.: 203 and 197.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TS.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29711).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 1999.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit 1 Dauphin County, Pennsylvania

    Date of application for amendment: December 3, 1998, as 
supplemented January 11, February 4, March 4, March 10, and March 15, 
1999.
    Brief description of amendment: This amendment conforms the license 
to reflect the transfer of Facility Operating License No. DPR-50 for 
the Three Mile Island Nuclear Station, Unit 1, from GPU Nuclear, Inc., 
et al., to AmerGen Energy Company, LLC, as previously approved by Order 
dated April 12, 1999.
    Date of issuance: December 20, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 218.
    Facility Operating License No. DPR-50: Amendment revised the 
license and the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1998 (63 
FR 70436).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 1999.
    Comments received: Yes. See safety evaluation dated April 12, 1999.

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

    Date of application for amendment: April 12, 1999, as supplemented 
October 5 and 8, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Surveillance Requirement (SR) 3.6.1.3.7 to allow a 
``representative sample'' of reactor instrumentation line excess flow 
check valves (EFCVs) to be tested every 24 months, instead of testing 
each EFCV every 24 months.
    Date of issuance: December 29, 1999.
    Effective date: December 29, 1999.
    Amendment No.: 230.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38028).
    The October 5 and 8, 1999, letters provided clarifying information 
that was within the scope of the original Federal Register notice and 
did not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 29, 1999.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: December 3, 1998.
    Brief description of amendment: This amendment revised the 
Technical Specifications for sealed source leakage testing to 
specifically address testing requirements for fission detectors.
    Date of issuance: December 20, 1999.
    Effective date: December 20, 1999, with full implementation within 
45 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-58: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43773).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 20, 1999.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: September 17, 1999, as 
supplemented November 10 and 19, 1999.
    Brief description of amendments: The amendments would approve the 
licensee's revision of the Updated Final Safety Analysis Report and 
Emergency Operating Procedures to use methodology to credit the 
negative reactivity provided by insertion of the rod cluster control 
assemblies (RCCAs) into the reactor core following any design basis 
loss-of-coolant accident, during realignment from a cold leg 
recirculation to a hot leg recirculation configuration. This change to 
the licensing basis, when evaluated by the licensee in accordance with 
10 CFR 59.59, resulted in an unreviewed safety question that requires 
prior approval by the NRC staff in accordance with the provisions of 10 
CFR 50.90 prior to implementation. The amendments also change the Bases 
for Technical Specifications Section 3/4.5.5, Refueling Water Storage 
Tank.
    Date of issuance: December 28, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 236 and 218.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56531).
    The licensee's letters of November 10 and 19, 1999, provided 
additional information that did not change scope of the application or 
the staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 1999.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: November 5, 1999.
    Brief description of amendment: The amendments would revise Unit 1 
and 2 Technical Specification (TS) 3.5.1, Action ``a'' and ``b,'' to 
reflect the monitoring of pressure from the Reactor Coolant System 
instead of the pressurizer. The amendment would also revise Unit 1 and 
2 TS Surveillance Requirement 4.5.1.c to require verification that 
power is removed from each emergency core cooling system accumulator 
isolation valve operator instead of verification that each accumulator 
isolation valve breaker is physically removed from the circuit. 
Furthermore, the amendment would make administrative changes to Unit 1 
and 2 TS Bases 3/4.5.1.
    Date of issuance: December 23, 1999.

[[Page 1933]]

    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 237 and 219.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1999 (64 
FR 65735).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 23, 1999.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: August 17, 1999.
    Brief description of amendment: The amendment removes the steam 
generator voltage-based repair criteria, F* repair criteria, and 
sleeving methodologies from the Unit 1 Technical Specifications and 
clarifies the Bases sections accordingly.
    Date of issuance: December 22, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 238.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54375).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 8, 1999.
    Brief description of amendment: The amendment changed action 
statements, definitions, and footnotes pertaining to the Technical 
Specifications for primary containment leakage and primary containment 
purge system to allow an alternative approach for isolating a bypass 
leakage path and/or purge system line.
    Date of issuance: December 16, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 87.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 16, 1999 (64 
FR 62228).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 1999.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: March 31, 1999.
    Brief description of amendment: Amendment changes Technical 
Specification Table 3.6.1.2-1 by adding two relief valves, and 
associated leak rate criteria, to be installed on the drywell equipment 
drain line and drywell floor drain line.
    Date of issuance: December 16, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 88.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24197).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 1999.
    No significant hazards consideration comments received: No

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 5, 1998.
    Description of amendment request: This amendment revises the 
Technical Specifications (TSs) by relocating the procedural details of 
the Radiological Effluent Technical Specifications (RETS) to the 
Offsite Dose Calculation Manual. The TSs were also revised to relocate 
procedural details associated with solid radioactive wastes to the 
Process Control Program. In addition, the Administrative Controls 
section of the TSs was revised to incorporate programmatic controls for 
radioactive effluents and environmental monitoring. These changes are 
consistent with the guidance provided in Generic Letter 89-01, 
``Implementation of Programmatic Controls for Radiological Effluent 
Technical Specifications in the Administrative Controls Section of the 
Technical Specifications and the Relocation of Procedural Details of 
RETS to the Offsite Dose Calculation Manual or to the Process Control 
Program.''
    Date of issuance: December 15, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 120 days.
    Amendment No.: 66.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19972). The Commission received comments which were addressed in the 
staff's Safety Evaluation dated December 15, 1999.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 15, 1999.
    No significant hazards consideration comments received: Yes.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: October 20, 1999.
    Brief description of amendment: The amendment changes from December 
31, 1999, to June 30, 2001, the date specified in TS 4.3.1.1.b Note 
associated with maintaining spent fuel pool boron concentration >2300 
ppm at all times until a permanent resolution to the current 
criticality concerns is implemented.
    Date of issuance: December 21, 1999.
    Effective date: December 21, 1999.
    Amendment No.: 75.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1999 (64 
FR 63345).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: September 4, 1998, as 
supplemented on February 8, April 16, August 26, September 16, and 
November 17, 1999.
    Brief description of amendment: The amendment increases the spent 
fuel pool storage capacity from 2,870 to 3,353 fuel assemblies.
    Date of Issuance: December 21, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.

[[Page 1934]]

    Amendment No.: 182
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 1, 1998 (64 FR 
52774). The supplemental information did not affect the staff's 
proposed no significant hazards consideration determination, and was 
within the scope of the original amendment application as published.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated December 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 18, 1998, as 
supplemented by letter dated October 22, 1999.
    Brief description of amendments: The amendments change the North 
Anna Power Station Technical Specifications (TS) to increase the 
allowable groundwater elevation at the southeast section of the service 
water reservoir dike from 277 to 280 feet at the toe and from 280 to 
295 feet at the crest. In addition, TS Table 3.7-6 has been reorganized 
to clarify zones of interest in the Service Water Reservoir, the 
location of piezometer devices, and piezometer device numbers. The 
proposal to eliminate device numbers from the TS was denied because the 
device number helps to indicate the location of the piezometer within 
the zone as well as the piezometer itself. Finally the column heading 
for Allowable Drain Flow Rate was clarified to be the total flow rate.
    Date of issuance: As of the date of issuance and shall be 
implemented within 30 days.
    Effective date: December 29, 1999.
    Amendment Nos.: 220 and 201.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69349). The supplemental letter dated October 22, 1999, contained 
clarifying information only, and did not change the initial no 
significant hazards determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 12, 1999.
    Brief description of amendments: These amendments update references 
in the Technical Specifications to information in the updated Final 
Safety Analysis Report (FSAR). The update is necessary to reflect 
relocation of the referenced information in the updated FSAR.
    Date of issuance: December 23, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-192; Unit 2-197.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24204).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 23, 1999.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 5th day of January 2000.

    For the Nuclear Regulatory Commission.
Suzanne Black,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 00-611 Filed 1-11-00; 8:45 am]
BILLING CODE 7590-01-P