[Federal Register Volume 69, Number 81 (Tuesday, April 27, 2004)]
[Notices]
[Pages 22877-22890]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-9225]


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NUCLEAR REGULATORY COMMISSION


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

Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and 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 April 2, 2004, through April 15, 2004. The 
last biweekly notice was published on April 13, 2004 (69 FR 19561).

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. Within 60 days after the date of publication of this 
notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, 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 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor

[[Page 22878]]

intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
pdr@nrc.gov.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: March 19, 2004.
    Description of amendment request: The licensee proposed to revise 
Section 4.2, ``Reactivity Control,'' of the Technical Specifications. 
Specifically, the amendment would revise Subsection 4.2.C, regarding 
surveillance requirements associated with control rod scram time 
testing (STT) by: (1) Eliminating unnecessary depressurized STT of non-
maintenance-affected control rods, (2) providing the required STT data 
necessary to apply actual scram times to implement improved minimum 
critical power ratio operating limits, and (3) eliminating the 
resulting redundant requirement to test ``eight control rods'' after a 
reactor scram or other outage. The amendment will also include 
editorial and pagination changes to accommodate the proposed technical 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change adds new surveillance requirements (SR) to 
the Minimum Critical Power Ratio (MCPR) Technical Specification (TS) 
which requires determination of the MCPR operating limit following 
the completion of scram time testing (STT) of the control rods. Use 
of the scram speed in determining the MCPR operating limit (i.e., 
Option B) is an alternative to the current method for determining 
the operating limit (i.e., Option A). The probability of an accident 
previously evaluated is unrelated to the MCPR operating limit that 
is provided to ensure no fuel damage results during anticipated 
operational occurrences. This is an operational limit to ensure 
conditions following an assumed accident do not result in fuel 
failure and therefore do not contribute to the occurrence of an 
accident. The proposed change eliminates unnecessary depressurized 
STT of non-maintenance[-]affected control rods and the requirement 
to test ``eight selected rods'' after a reactor scram or other 
outage. The requirement to test ``eight selected rods'' is replaced 
by a new SR to perform periodic STT. No active or passive failure 
mechanisms that could lead to an accident are affected by this 
proposed change. Therefore, the proposed change in STT requirements 
does not significantly increase the probability of an accident 
previously evaluated.
    The proposed change ensures that the appropriate MCPR operating 
limit is in place. By implementing the correct MCPR operating limit 
the MCPR safety limit will continue to be ensured. Ensuring the MCPR 
safety limit is not exceeded will result in prevention of fuel 
failure. Therefore, since there is no increase in the potential for 
fuel failure there is no increase in the consequences of any 
accidents previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change adds a new SR to the MCPR TS which requires 
determination of the MCPR operating limit following the completion 
of scram time testing of the control rods. The proposed change 
eliminates unnecessary depressurized STT of non-maintenance[-
]affected rods and the requirement to test ``eight selected rods'' 
after a reactor scram or other outage. The

[[Page 22879]]

requirement to test ``eight selected rods'' is replaced by a new SR 
to perform periodic STT. The proposed change does not involve the 
use or installation of new equipment. Installed equipment is not 
operated in a new or different manner. No new or different system 
interactions are created, and no new processes are introduced. No 
new failures have been created by the addition of the proposed SR 
and the use of the alternate method for determining the MCPR 
operating limit.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Use of Option B for determining the MCPR operating limit will 
result in a reduced operating limit in comparison to the use of 
Option A. However, a reduction in the operating limit margin does 
not result in a reduction in the safety margin. The MCPR safety 
limit remains the same regardless of the method used for determining 
the operating limit. The proposed change eliminates unnecessary 
depressurized STT of non-maintenance[-]affected control rods and the 
requirement to test ``eight selected rods'' after a reactor scram or 
other outage. The requirement to test ``eight selected rods'' is 
replaced by a new SR to perform periodic STT. No active or passive 
failure mechanisms that could adversely impact the consequences of 
an accident are affected by this proposed change. All analyzed 
transient results remain well within the design values for 
structures, systems and components.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: March 23, 2004.
    Brief description of amendments: The licensee proposed to revise 
the Technical Specifications (TSs) by eliminating the requirements for 
hydrogen/oxygen monitors. The proposed amendment supports 
implementation of the revision to 10 CFR 50.44, ``Standards for 
Combustible Gas Control System in Light-Water-Cooled Power Reactors,'' 
that became effective on October 16, 2003.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-447, ``Elimination of Hydrogen 
Recombiners and Change to Hydrogen and Oxygen Monitors.'' The 
availability of this TS improvement was published in the Federal 
Register on September 25, 2003 (68 FR 55416), on possible amendments 
concerning TSTF-447, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. In its application for 
amendment, the licensee affirmed the applicability of the following 
NSHC determination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee presented 
an analysis of NSHC by endorsing the model NSHC determination published 
in 68 FR 55416 (reproduced below):

    Criterion 1.--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen and oxygen 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
classification of the oxygen monitors as Category 2 and removal of 
the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs [Severe 
Accident Management Guidelines], the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.
    Criterion 2.--The Proposed Change Does Not Create the 
Possibility of a New or Different Kind of Accident From Any 
Previously Evaluated.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen and oxygen monitor equipment was intended to mitigate a 
design-basis hydrogen release. The hydrogen recombiner and hydrogen 
and oxygen monitor equipment are not considered accident precursors, 
nor does their existence or elimination have any adverse impact on 
the pre-accident state of the reactor core or post[-]accident 
confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3.--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-

[[Page 22880]]

basis LOCA. The Commission has found that this hydrogen release is 
not risk-significant because the design-basis LOCA hydrogen release 
does not contribute to the conditional probability of a large 
release up to approximately 24 hours after the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI [Three Mile Island], 
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
    Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.
    Therefore, this change does not involve a significant reduction 
in [a] margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    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: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: November 12, 2002, as supplemented 
March 5, 2004. This notice supersedes the notice that was published on 
February 18, 2003 (68 FR 7813).
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications to support an expansion of the core 
flow operating range, including the new automated backup stability 
protection function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will implement DSS-CD [Detect and Suppress 
Solution--Confirmation Density] as the long-term stability solution. 
The DSS-CD solution is designed to identify the power oscillation 
upon inception and initiate control rod insertion to terminate the 
oscillations prior to any significant amplitude growth. The DSS-CD 
provides protection against violation of the Safety Limit Minimum 
Critical Power Ratio (SLMCPR) for anticipated oscillations. 
Compliance with General Design Criteria (GDC) 10 and 12 of 10 CFR 
part 50, Appendix A is accomplished via an automatic action. The 
DSS-CD introduces an enhanced detection algorithm that detects the 
inception of power oscillations and generates an earlier power 
suppression trip signal exclusively based on successive period 
confirmation recognition. The existing Option III algorithms are 
retained, with generic setpoints, to provide defense-in-depth 
protection for unanticipated reactor instability events.
    A developing instability event is suppressed by the DSS-CD 
system with substantial margin to the SLMCPR and no clad damage, 
with the event terminating in a scram and never developing into an 
accident. In addition, the DSS-CD solution defense-in-depth features 
incorporate all the backup scram algorithms plus the licensed scram 
feature of the existing Option III system. The DSS-CD system does 
not interact with equipment whose failure could cause an accident. 
Scram setpoints in the DSS-CD will be established so that analytical 
limits are met. The reliability of the DSS-CD will meet or exceed 
that of the existing system. No new challenges to safety-related 
equipment will result from the DSS-CD solution. Because an 
instability event would reliably terminate in an early scram without 
impact on other safety systems, there is no significant increase in 
the probability of an accident.
    The existing requirement to initiate an alternate (i.e., manual) 
method to detect and suppress thermal hydraulic instability 
oscillations is expanded to include a requirement to either 
implement an Automated Backup Stability Protection (ABSP) (i.e., 
Required Action I.2.1) or exit the operating region most susceptible 
to rapid onset of Thermal Hydraulic Instability (THI) (i.e., 
Required Action I.2.2). The ABSP is an automatic reactor scram 
region, implemented by the Average Power Range Monitor (APRM) flow-
biased scram setpoint. It may be used if the Oscillation Power Range 
Monitoring (OPRM) system is inoperable to allow continued operation 
within the MELLLA+ [Maximum Extended Load Line Limit Analysis Plus] 
operating domain. Additionally, a new Required Action I.3 is 
included. Required Action I.3 ensures that a report is made to the 
NRC, if DSS-CD is inoperable for 120 days.
    To maintain the existing margin between equipment operability 
requirements and the region of power-flow operation where 
anticipated events could lead to thermal-hydraulic instability, (1) 
TS 3.3.1.1, Required Action J.1 is revised to require the plant to 
be < 18% RTP [rated thermal power] versus < 20% RTP in the event 
that the OPRM Upscale Function is inoperable and the Required 
Actions associated with Action I are not completed, and (2) the 
operability requirement for the OPRM Upscale Function (i.e., TS 
3.3.1.1, Table Function 2.f) is changed from = 20% RTP to 
= 18% RTP. This 5% margin is consistent with and 
maintains the existing 5% margin operability requirements for the 
Option III OPRM Upscale operability requirements.
    Overall, these changes result in more conservative plant 
operation. Other changes proposed in this supplement are either in 
direct support of ABSP or are administrative in nature.
    Proper operation of the DSS-CD system does not affect any 
fission product barrier or Engineered Safety Feature. Thus, the 
proposed change cannot change the consequences of any accident 
previously evaluated. As stated above, the DSS-CD solution meets the 
requirements of GDC 10 and 12 by automatically detecting and 
suppressing design basis thermal-hydraulic oscillations prior to 
exceeding the fuel SLMCPR.
    Based on the above, the operation of the DSS-CD solution within 
the framework of the Option III OPRM hardware will not increase the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The DSS-CD solution operates within the existing Option III OPRM 
hardware. No new operating mode, safety-related equipment lineup, 
accident scenario, system interaction, or equipment failure mode was 
identified. The ABSP automatic reactor scram region is implemented 
by adjusting the existing APRM flow-biased scram setpoint. 
Therefore, the DSS-CD solution will not adversely affect plant 
equipment.
    Because there are no hardware changes, there is no change in the 
possibility or consequences of a failure. The worst case failure of 
the equipment is a failure to initiate mitigating action (i.e., 
scram), but no failure can cause an accident of a new or different 
kind than any previously evaluated.
    Based on the above, the proposed change to the DSS-CD solution 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The DSS-CD solution is designed to identify the power 
oscillation upon inception and initiate control rod insertion to 
terminate the oscillations prior to any significant amplitude 
growth. The DSS-CD solution algorithm will maintain or increase the 
margin to the SLMCPR for anticipated instability events. The safety 
analyses in ``Detect And Suppress Solution--Confirmation Density 
Licensing Topical Report,'' Revision 3 demonstrate the margin to the 
SLMCPR for postulated bounding stability events. Existing margin 
between equipment operability requirements and the region of power-
flow operation where

[[Page 22881]]

anticipated events could lead to thermal-hydraulic instability are 
maintained. As a result, there is no impact on the SLMCPR identified 
for an instability event.
    The existing requirement to initiate an alternate method to 
detect and suppress thermal hydraulic instability oscillations is 
expanded to include a requirement to either implement an ABSP (i.e., 
Required Action I.2.1) or exit the operating region most susceptible 
to rapid onset of THI (i.e., Required Action I.2.2). Additionally, a 
new Required Action I.3 is included. Required Action I.3 ensures 
that a report is made to the NRC, if DSS-CD is inoperable for 120 
days. These change results in more conservative plant operation. 
Other changes proposed in this supplement are either in direct 
support of ABSP or are administrative in nature.
    The current Option III algorithms (i.e., Period Based Detection, 
Amplitude Based, and Growth Rate) are retained (with generic 
setpoints) to provide defense-in-depth protection for unanticipated 
reactor instability events.
    Based on the above, the proposed change will not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: William F. Burton, Acting.

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

    Date of amendment request: March 25, 2004.
    Description of amendment request: The proposed amendments would 
allow the use of the methodology described in Framatome-ANP (FRA-ANP) 
Topical BAW-10169-A ``RSG Plant Safety Analysis--B&W Safety Analysis 
Methodology for Recirculating Steam Generator Plants'', dated October 
1989 for the generation of mass and energy release rates during a Main 
Steam Line Break accident for Prairie Island Nuclear Generating Plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing the use of the 
methodology described in Framatome-ANP Topical BAW-10169-A ``RSG 
Plant Safety Analysis--B&W Safety Analysis Methodology for 
Recirculating Steam Generator Plants'' that utilizes the RELAP5/
MOD2-B&W code described in Topical BAW-10164-A ``RELAP5/MOD2-B&W--An 
Advanced Computer Program for Light-Water Reactor LOCA [loss-of-
coolant accident] and Non-LOCA Transient Analysis'' for the 
generation of predicted mass and energy releases during a Main Steam 
Line Break accident.
    The methodology used to perform an analysis of a main steam line 
break is not an accident initiator, thus changing the methodology 
does not increase the probability of an accident.
    The mass and energy releases generated by the proposed 
methodology will be utilized to demonstrate that the design basis 
limits for fission product barriers are not exceeded. The proposed 
methodology does not alter the nuclear reactor core, reactor coolant 
system, or equipment used directly in mitigation of a main steam 
line break, thus radioactive releases due to a main steam line break 
accident are not affected by the proposed change in analysis 
methodology. Therefore, this change does not increase the 
consequences of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing the use of the 
methodology described in Framatome-ANP Topical BAW-10169-A ``RSG 
Plant Safety Analysis--B&W Safety Analysis Methodology for 
Recirculating Steam Generator Plants'' that utilizes the RELAP5/
MOD2-B&W code described in Topical BAW-10164-A ``RELAP5/MOD2-B&W--An 
Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA 
Transient Analysis'' for the generation of predicted mass and energy 
releases during a Main Steam Line Break accident.
    The analysis of a main steam line break using the proposed 
methodology does not alter the nuclear reactor core, reactor coolant 
system, or equipment used directly in mitigation of a main steam 
line break.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing the use of the 
methodology described in Framatome-ANP Topical BAW-10169-A ``RSG 
Plant Safety Analysis--B&W Safety Analysis Methodology for 
Recirculating Steam Generator Plants'' that utilizes the RELAP5/
MOD2-B&W code described in Topical BAW-10164-A ``RELAP5/MOD2-B&W--An 
Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA 
Transient Analysis'' for the generation of predicted mass and energy 
releases during a Main Steam Line Break accident.
    The proposed licensing basis change will result in a 
conservative calculation of the mass and energy releases during a 
Main Steam Line Break accident. This will ensure that there is no 
reduction in the margin of safety for analyses that utilize the 
generated mass and energy releases as inputs. 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: March 4, 2004.
    Description of amendment request: The proposed amendment would 
revise the SSES 1 and 2 Technical Specification Table 3.3.5.1-1 to 
clarify that four low pressure coolant injection pump discharge 
pressure-high channels are required for each automatic depressurization 
system trip function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The Technical Specification required number of protection 
channels is not an initiator to any accident sequence analyzed in 
the Final Safety Analysis Report (FSAR). As discussed in this 
request, the change is editorial and involves no change in the 
number of ADS [Automatic Depressurization System] supporting 
protection channels

[[Page 22882]]

required by the Susquehanna Steam Electric Station (SSES) Technical 
Specifications (TS). The change does not have any effect on the 
initiator of any accident sequence analyzed in the Final Safety 
Analysis Report (FSAR) and does not affect any assumptions 
associated with the mitigation of accident or transient events. The 
change does not involve any physical change to structures, systems, 
or components (SSCs) and does not involve any physical change to 
structures, systems, or components (SSCs) and does not alter the 
method of operation or control of SSCs. The current assumptions in 
the SSES FSAR safety analysis regarding accident initiators and 
mitigation of accidents are unaffected by these changes. No 
additional failure modes or mechanisms are being introduced and the 
likelihood of previously analyzed failures remains unchanged.
    Operation in accordance with the proposed Technical 
Specification (TS) continues to ensure that the plant response to 
analyzed accidents remains capable of performing as described in the 
FSAR. Therefore, the mitigative functions supported by the system 
continue to provide the protection assumed by the analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints, at which protective or mitigative actions are 
initiated, affected by this change. This change does not alter the 
manner in which equipment operation is initiated, nor are the 
function demands on credited equipment be[ing] changed. No 
alterations in the procedures that ensure the plant remains within 
analyzed limits are being proposed, and no changes are being made to 
the procedures relied upon to respond to an off-normal event as 
described in the FSAR. As such, no new failure modes are being 
introduced. The change does not alter the assumptions made in the 
safety analysis and licensing basis.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The change is editorial and involves no technical 
changes to the Susquehanna Steam Electric Station (SSES) Technical 
Specifications (TS). Therefore the plant response to analyzed events 
continues to provide the margin of safety assumed by the 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: March 5, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement (SR) 3.6.4.1.3 
to require that only one secondary containment access door in each 
access opening be verified closed. In addition, this SR allows entry 
and exit access between required secondary containment zones that have 
a single door.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The Technical Specification Surveillance being revised, which 
verifies the status of the secondary containment access doors, is 
not an initiator to any accident sequence analyzed in the Final 
Safety Analysis Report (FSAR). The proposed change relaxes the 
acceptance criteria of this Surveillance such that maintenance on 
one of two airlock access doors can be performed. However, requiring 
that at least one door is closed, in conjunction with the continued 
requirement to maintain the building at a negative pressure, 
continues to assure that the secondary containment barrier is 
maintained operable. This provides adequate assurance that the 
secondary containment is capable of performing the accident 
mitigation function assumed in the accident analyses. As a result, 
the consequences of any accident previously evaluated are not 
significantly affected.
    The Note, which was added to the Technical Specifications, 
provides clarification and precludes a conflict with the explicit 
wording of SR 3.6.4.1.3. Since this Note is consistent with the 
intent as reflected in the Bases and with the prior SSES Technical 
Specifications, the change is considered editorial and reflects an 
administrative presentation preference and not a technical change.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints, at which protective or mitigative actions are 
initiated, affected by this change. This change does not alter the 
manner in which equipment operation is initiated, nor are the 
function demands on credited equipment changed. No alterations in 
the procedures that ensure the plant remains within analyzed limits 
are being proposed, and no changes are being made to the procedures 
relied upon to respond to an off-normal event as described in the 
FSAR. As such, no new failure modes are being introduced.
    The Note, which was added to the Technical Specifications, 
provides clarification and precludes a conflict with the explicit 
wording of SR 3.6.4.1.3. Since this Note is consistent with the 
intent as reflected in the Bases and with the prior SSES Technical 
Specifications, the change is considered editorial and reflects an 
administrative presentation preference and not a technical change.
    The change does not alter the assumptions made in the safety 
analysis and licensing basis.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The change could allow additional time for one of two 
airlock doors to be open for maintenance. However, the margin of 
safety is maintained by the continued closure of the remaining 
airlock door (as is currently allowed for normal entry and exit) and 
the continued requirement to be able to maintain the building at a 
negative pressure.
    The Note, which was added to the Technical Specifications, 
provides clarification and precludes a conflict with the explicit 
wording of SR 3.6.4.1.3. Since this Note is consistent with the 
intent as reflected in the Bases and with the prior SSES Technical 
Specifications, the change is considered editorial and reflects an 
administrative presentation preference and not a technical change.
    Therefore, the plant response to analyzed events continues to 
provide the margin of safety assumed by the 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

[[Page 22883]]

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant (BFN), Units 2 and 3, Limestone County, 
Alabama

    Date of amendment request: July 31, 2002, as supplemented by 
letters dated December 9, 2002, February 12, 2003, March 26, 2003, July 
11, 2003, and July 17, 2003.
    Description of amendment request: The proposed amendments request 
full implementation of an alternative source term (AST) for the Units 
1, 2, and 3 operating licenses. The amendments adopt the AST 
methodology by revising the current accident source term and replacing 
it with an accident source term as prescribed in 10 CFR 50.67. The 
submittals also propose to revise/delete the Technical Specification 
(TS) Sections associated with control emergency ventilation (CREV), 
standby gas treatment (SGT), standby liquid control (SLC), and 
secondary containment systems. Additionally, the submittals request 
modification of the licensing and design basis to reflect the 
application of the AST methodology and the function of the SLC system, 
and deletion of a license condition for Units 2 and 3, which all the 
actions have been completed.
    The supplements to the original application include the withdrawal 
of the request to delete one of the TS Sections described above, 
associated with the absorption of elemental iodine by the SGT and CREV 
systems charcoal filters. Also the supplements add a new TS Section to 
require verification that the minimum fuel decay period has passed 
prior to moving fuel after the reactor is shut down. The licensee 
indicated that these modifications/deletions do not affect the 
originally published no significant hazards consideration. The original 
no hazards consideration is reproduced below.
    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 AST and those plant systems affected by implementing AST do 
not initiate DBAs [design-basis accidents]. The AST does not affect 
the design or operation of the facility; rather, once the occurrence 
of an accident has been postulated, the new source term is an input 
to evaluate the consequences. The implementation of the AST has been 
evaluated in the analyses for the limiting DBAs at BFN. The 
equipment affected by the proposed change is mitigative in nature 
and relied upon following an accident. The proposed changes to the 
TS do revise certain performance requirements. However, these 
changes will not involve a revision to the parameters or conditions 
that could contribute to the initiation of a design basis accident 
discussed in Chapter 14 of the BFN Updated Final Safety Analysis 
Report.
    Plant specific radiological analyses have been performed and, 
based on the results of these analyses, it has been demonstrated 
that the dose consequences of the limiting events considered in the 
analyses are within the regulatory guidance provided by the NRC for 
use with the AST. This guidance is presented in 10 CFR 50.67, 
Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1. 
Therefore, the proposed amendment does not result in a significant 
increase in the consequences or a significant increase in the 
probability of any previously evaluated accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Implementation of AST does not alter any design basis accident 
initiators. These changes do not affect the design function or mode 
of operations of systems, structures, or components in the facility 
prior to a postulated accident. Since systems, structures, and 
components are operated essentially no differently after the AST 
implementation, no new failure modes are created by this proposed 
change. Therefore, the proposed license amendments will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The changes proposed are associated with a revision to the 
licensing basis for BFN. The results of accident analyses revised in 
support of the proposed change are subject to the acceptance 
criteria in 10 CFR 50.67. The analyzed events have been carefully 
selected, and the analyses supporting this submittal have been 
performed using approved methodologies. The dose consequences of 
these limiting events are within the acceptance criteria provided by 
the regulatory guidance as presented in 10 CFR 50.67, Regulatory 
Guide 1.183, and SRP 15.0.1.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, the 
changes are considered to not result in a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and based on 
this review, it appears that the three standards 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 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton (Acting).

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

    Date of amendment request: March 3, 2004 (TSC 03-10).
    Description of amendment request: The proposed amendment would 
revise the Updated Final Safety Analysis Report (UFSAR) and the 
Technical Specification Bases description of the seismic qualification 
of round flexible ducting, triangular ducting, and associated air bars 
installed as part of the suspended ceiling air delivery system in the 
main control room.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The design function of the MCR [main control room] ducting 
system is to support pressurization and cooling of the control room 
during normal and accident conditions. The design function of the 
MCR suspended ceiling is to remain in place during and subsequent to 
an accident, support the triangular and flexible ducts, and not 
damage safety-related equipment. The MCR ducting, including the 
classification and methodology changes, is a passive feature and 
does not act as an accident initiator, i.e., failure of the ducting 
would not initiate a design basis accident. The MCR suspended 
ceiling has been qualified such that it will remain in place and 
perform its safety function during and after an accident. 
Consequently, the changes associated with the MCR ducting and 
suspended ceiling do not affect the frequency of occurrence for 
accidents previously evaluated in the UFSAR.
    For the principal design basis accidents, loss of coolant 
accident (LOCA), internal flood, steam generator tube rupture 
(SGTR), main steam line break (MSLB), etc., the integrity of the MCR 
HVAC [heating, ventilation and air conditioning] system, including 
the suspended ceiling, will not be compromised. These accidents do 
not have a structural effect on the MCR. This means that for 
radiological or toxic chemical accidents, the ability to both 
pressurize and maintain MCR temperatures within the design limits is 
unaffected by the limited quality and seismic requirements for the 
flexible and triangular ducting.
    An accident that involves a fire that affects the MCR or the 
habitability of the MCR was not a consideration for the 
qualification of the air distribution components. A fire of this 
nature will result in plant operation from the Auxiliary Control 
Room (ACR) which is supported by a separate HVAC system.

[[Page 22884]]

    The physical effects of an earthquake (including the design 
basis SSE) is the only event in which the design basis for the MCR 
HVAC is potentially challenged. An evaluation by an industry seismic 
expert shows that the ducting and suspended ceiling will remain in 
place, will retain their structural integrity such that flow will 
not be impeded, and the ducting pressure boundary will not be lost. 
Thus, reducing the QA [quality assurance] and seismic qualification 
requirements for the MCR ducting and changing the method of seismic 
qualification will not result in loss of safety function for any 
design basis accident or event. Thus, the accident dose as 
previously evaluated in the UFSAR is not affected by the proposed 
license amendment.
    Based on the above discussion, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The MCR ducting addressed by the proposed amendment is not 
an accident initiator; i.e., failure of the ducting will not 
initiate a design basis accident. In addition, the subject ducting 
and suspended ceiling have been evaluated and a determination has 
been made that they will continue to perform their safety functions 
during normal and accident conditions. Consequently, this activity 
does not create a possibility of a new or different type of accident 
than any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The changes addressed in TVA' s proposed amendment are 
associated with changes in QA requirements and seismic qualification 
methodology for safety related air delivery components and for the 
suspended ceiling. The change does not affect specific HVAC 
equipment safety limits, design limits, set points, or other 
critical parameters. In addition, the new seismic analysis 
methodology and limited QA requirements ensure that these components 
will continue to perform their safety functions during normal and 
accident conditions. The previously implied margin of safety against 
structural or functional failure of the air delivery components or 
suspended ceiling during and after a design basis SSE [safe-shutdown 
earthquake] has not been reduced. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

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

    Date of amendment request: April 7, 2004.
    Description of amendment request: The proposed amendment would 
revise the maximum ultimate heat sink (UHS) temperature by revising the 
Technical Specification (TS) maximum essential raw cooling water (ERCW) 
temperature limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change to increase the UHS maximum temperature 
will not adversely alter the function, design, or operating 
practices for plant systems or components. The UHS is utilized to 
remove heat loads from plant systems during normal and accident 
conditions. This function is not expected or postulated to result in 
the generation of any accident and continues to adequately satisfy 
the associated safety functions with the proposed changes. 
Therefore, the probability of an accident presently evaluated in the 
safety analyses will not be increased. With the exception of re-
gearing the shutdown board room chiller compressors, no other plant 
equipment must be altered as a result of this change. Re-gearing of 
the shutdown board room chillers will ensure their continued 
performance in accordance with design concurrent with the increased 
UHS temperature. The heat loads that the UHS is designed to 
accommodate have been evaluated for functionality with the higher 
temperature limits. The result of these evaluations is that there is 
existing margin associated with the systems that utilize the UHS for 
normal and accident conditions. These margins are sufficient to 
accommodate the postulated normal and accident heat loads with the 
proposed changes to the UHS. Since the safety functions of the UHS 
are maintained, the systems that ensure acceptable offsite dose 
consequences will continue to operate as designed. The change in the 
maximum calculated containment pressure associated with the design 
basis loss of coolant accident remains below the ASME [American 
Society of Mechanical Engineers] Code design internal pressure. The 
change to clarify the maximum allowable internal containment 
pressure is administrative consistent with present wording in the TS 
Bases. Therefore, the consequence of any accident will be the same 
as those previously analyzed.
    Therefore, since the UHS safety function will continue to meet 
accident mitigation requirements and limit dose consequences to 
acceptable levels, TVA has concluded that the proposed TS change 
does not involve a significant increase [in] the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The UHS function provides accident mitigation capabilities 
and serves as a heat sink for normal and upset plant conditions; the 
UHS is not an initiator of any accident. By allowing the proposed 
change in the UHS temperature requirements, only the parameters for 
UHS operation are changed while the safety functions of the UHS and 
systems that transfer the heat sink capability continue to be 
maintained. The proposed change does not impact the response of the 
systems and components assumed in the safety analysis. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change has been evaluated for systems that are 
needed to support accident mitigation functions as well as normal 
operational evolutions. Operational margins were found to exist in 
the systems that utilize the UHS capabilities such that these 
proposed changes will not result in the loss of any safety function 
necessary for normal or accident conditions. The ERCW system has 
excess flow margins that will accommodate the increased flows 
necessary for the proposed temperature increase. While operating 
margins have been reduced by the proposed changes, safety margins 
have been maintained as assumed in the accident analyses for 
postulated events. The proposed change results in an increase in the 
maximum calculated containment peak pressure. However, the change in 
the maximum calculated containment peak pressure associated with the 
design basis LOCA [loss-of-coolant accident] is a small percentage 
of the margin between the current maximum calculated containment 
peak pressure and the ASME Code design internal pressure. The change 
to clarify the maximum allowable internal containment pressure is 
administrative. This aspect of the proposed change does not involve 
a significant reduction in a margin of safety. Additionally, the 
proposed changes do not require any further modification (the 
shutdown board room chiller will be re-geared) of component 
setpoints or operating provisions that are necessary to maintain 
margins of safety established by the WBN design. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

[[Page 22885]]

400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

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.

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagordo County, Texas

    Date of amendment request: March 4, 2004.
    Brief description of amendment request: The proposed amendment 
would allow South Texas Project (STP) Unit 2 to change modes with 
standby diesel generator 22 inoperable. This is a one-time change that 
would expire 14 days after entering Mode 4 on restart from the STP Unit 
2 Spring 2002 refueling outage.
    Date of publication of individual notice in Federal Register: March 
23, 2004.
    Expiration date of individual notice: April 22, 2004 (public 
comments), and May 24, 2004 (hearing requests).

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagordo County, Texas

    Date of amendment request: March 18, 2004.
    Brief description of amendment request: These amendments revise 
Technical Specification (TS) Surveillance Requirement 4.7.7.e.3 to add 
a footnote that allows an evaluation for points that do not meet the 1/
8 inch Water Gauge criterion of the current TS. These amendments close 
out Notice of Enforcement Discretion No. 04-6-001, which the Commission 
granted on March 23, 2004.
    Date of publication of individual notice in Federal Register: April 
5, 2004.
    Expiration date of individual notice: April 19, 2004 (public 
comments), and June 4, 2004 (hearing requests).

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: March 23, 2004.
    Description of amendments request: To allow both trains of control 
room air-conditioning system to be inoperable for up to 7 days, 
provided control room temperatures are verified every 4 hours to be 
less than or equal to 90 degrees Fahrenheit.
    Date of publication of individual notice in the Federal Register: 
April 14, 2004 (69 FR 19880).
    Expiration date of individual notice: May 14, 2004.

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, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

Calvert Cliffs Nuclear Power Plant, Inc., 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: July 14, 2003, as supplemented 
December 5, 2003, and February 12, 2004.
    Brief description of amendments: These amendments change the 
Surveillance Requirement 3.6.6.8 to verify each containment spray 
nozzle is unobstructed only following maintenance that could result in 
nozzle blockage.
    Date of issuance: April 8, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 264 and 241.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49814). The supplements dated December 5, 2003, and February 12, 2004, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 8, 2004.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., 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: February 25, 2004.
    Brief description of amendments: These amendments changes the 
implementation date for the new cooldown rates for pressure temperature

[[Page 22886]]

limits established by Amendment Nos. 261 and 238 for Calvert Cliffs 
Nuclear Power Plant, Unit Nos. 1 and 2, respectively, from 120 days 
after issuance, to July 1, 2004.
    Date of issuance: April 5, 2004.
    Effective date: As of the date of issuance, immediately changing 
the implementation date of Amendment Nos. 261 and 238 to July 1, 2004.
    Amendment Nos.: 263 and 240.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2004 (69 FR 
10487). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 5, 2004.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: March 20, 2003; as supplemented 
on March 31, April 17, June 11, July 21, and December 11, 2003; and 
January 20, February 10, and March 11, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to reflect an expanded operating domain resulting 
from the implementation of the Average Power Range Monitor, Rod Block 
Monitor TSs/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA).
    Date of Issuance: April 14, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented at the start of operating cycle 24.
    Amendment No.: 219.
    Facility Operating License No. DPR-28: Amendment revised the TS.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18276). The licensee's March 31, April 17, June 11, July 21, and 
December 11, 2003; and January 20, February 10, and March 11, 2004, 
letters provided clarifying information that did not change the scope 
of the proposed amendment as described in the original notice of 
proposed action published in the Federal Register, and did not change 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 14, 2004.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi; Entergy Gulf States, Inc., and Entergy Operations, 
Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana 
Parish, Louisiana; and Entergy Operations, Inc., Docket No. 50-382, 
Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of application for amendment: November 6, 2002, as 
supplemented by letters dated November 18, 2003, and January 30, 2004.
    Brief description of amendment: The amendment would revise the 
Facility Operating Licenses, Appendix B, Environmental Protection Plan 
(EPP) (Non-Radiological) for the respective plants.
    Date of issuance: April 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 165, Docket No. 50-416, NPF-29; 138, Docket No. 50-
458, NPF-47; 193, Docket No. 50-382, NPF-38.
    Facility Operating License Nos. NPF-29, NPF-47, and NPF-38: The 
amendments revise the EPPs for the respective plants.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75872).
    The licensee enclosed a revised no significant hazards 
consideration (NSHC) determination with the supplemental letter dated 
November 18, 2003. This revised NSHC determination contained minor 
wording changes as compared with the NSHC determination included in the 
original application dated November 6, 2002, changes made to reflect 
the new EPP changes, and did not expand the scope of the application as 
originally noticed, and did not change the conclusions of the NSHC 
determination as published in the Federal Register on December 10, 2002 
(67 FR 75872). The January 30, 2004, supplemental letter provided 
further clarification to the November 18, 2003, supplemental letter 
that did not change the conclusion of the NSHC determination published 
on December 10, 2002.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 12, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: June 20, 2003, as supplemented 
by letter dated December 12, 2003.
    Brief description of amendment: The amendment authorizes changes to 
the surveillance requirements for containment integrated leak rate 
testing in TS 4.4.a, ``Integrated Leak Rate Tests (Type A).''
    Date of issuance: April 6, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 173.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43391) . The supplemental letter contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, 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: March 27, 2003, as supplemented 
by letters dated October 30, and December 19, 2003.
    Brief description of amendments: The proposed amendment would 
approve a selective scope application of an alternative source term for 
fuel-handling accidents. Specifically, the amendments would revise 
Technical Specification 3.9.3, ``Containment Penetrations,'' to (1) 
change the Applicability statement to ``During movement of recently 
irradiated fuel assemblies within containment,'' and (2) modify the 
Required Action for Condition A to eliminate the requirement to suspend 
core alterations and add the requirement to suspend movement of 
recently irradiated fuel assemblies within containment if one or more 
containment penetrations are not in the required status.
    Date of issuance: April 2, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 213 and 218.

[[Page 22887]]

    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25656). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 2, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 25, 2003, as 
supplemented September 9, 2003.
    Brief description of amendment: The amendment added an allowed-
outage time for Engineered Safety Features Actuation System 
Instrumentation channels to be out of service in a bypassed state.
    Date of issuance: April 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. NPF-12: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15762). The September 9, 2003, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 18, 2004, as supplemented by 
letters dated April 7 and 13, 2004.
    Brief description of amendments: The amendments revise TS 
Surveillance Requirement (SR) 4.7.7.e.3 to add a footnote that allows 
use of alternate criteria for those measured points at positive 
pressure but that do not meet the \1/8\ inch Water Gauge criterion of 
the current TS. In addition the word ``that'' in the second line of the 
original text of SR 4.7.7.e.3 is changed to ``than'' to correct an 
existing typographical error. These amendments supersede Notice of 
Enforcement Discretion (NOED) No. 04-6-001, which the Commission staff 
granted to STPNOC on March 23, 2004.
    Date of issuance: April 15, 2004.
    Effective date: As of the date of issuance.
    Amendment Nos.: Unit 1-161; Unit 2-151.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC):
    Yes. A notice was published in the Federal Register on April 5, 
2004 (69 FR 17718). The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received. The notice also provided an opportunity to request 
a hearing within 60 days from the date of publication, but indicated 
that if the Commission makes a final NSHC determination, any such 
hearing would take place after issuance of the amendment. The 
supplements dated April 7 and 13, 2004, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2004.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: April 11, 2003, as supplemented 
by the October 2, 2003, meeting, and a letter dated February 20, 2004.
    Description of amendment request: The amendments revised Technical 
Specification (TS) Table 3.3.5.1-1 which will result in a change to the 
Updated Final Safety Analysis Report (UFSAR), Table 6.5-3.
    Date of issuance: April 1, 2004.
    Effective date: Date of issuance, to be implemented within 60 days 
for Unit 1, during Cycle 13 Refueling Outage for Unit 2 , and during 
Cycle 12 Refueling Outage for Unit 3.
    Amendment Nos.: 250, 289 & 248.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the TSs which will result in a change the UFSAR, 
Table 6.5-3.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28857). The October 2, 2003, meeting, and the February 20, 2004, 
letter, provided clarifying information that did not change the scope 
of the original request or the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 27, 2003, as supplemented 
by letters dated December 9, 2003, January 14 and April 5, 2004.
    Brief description of amendment: The amendment approves the 
application of leak-before-break methodology for the accumulator and 
residual heat removal lines and installation of an opening in the 
secondary shield wall in terms of the effect of the opening on 
occupational exposure. The shield wall opening is related to plant 
modifications that would facilitate maintenance on the replacement 
steam generators to be installed in Refueling Outage 14 (Fall 2005).
    Date of issuance: April 12, 2004.
    Effective date: April 12, 2004, and shall be implemented prior to 
entering Mode 4 during the startup from Refueling Outage 13 which is 
scheduled for the Spring of 2004.
    Amendment No.: 161.
    Facility Operating License No. NPF-30: The amendment revised the 
Final Safety Analysis Report.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43397).
    The December 9, 2003, January 14 and April 5, 2004, supplemental 
letters provided additional clarifying information, did not expand the 
scope of the application as originally noticed, and did not change the 
staff's original proposed no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 2004.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit 2, Louisa County, Virginia

    Date of application for amendment: March 28, 2002, as supplemented 
by letters dated May 13, June 19, July 9, July 25, August 2, August 16, 
and November 15, 2002, May 6, May 9, May 27, June 11 (2 letters), July 
18, August 20, August 26, September 4, September 5, September 22, 
September 26 (2

[[Page 22888]]

letters), November 10, December 8, and December 17, 2003, and January 
6, January 22 (2 letters), February 12, February 13, and March 1, 2004. 
The November 15, 2002, submittal replaced the submittals dated July 9, 
July 25, and August 16, 2002.
    Brief description of amendment: This amendment revises Improved 
Technical Specification Sections 2.1, 4.2, and 5.6.5 in order to allow 
Virginia Electric and Power Company to implement Framatome ANP Advanced 
Mark-BW fuel at North Anna Power Station, Unit 2.
    Date of issuance: April 1, 2004.
    Effective date: As of the date of issuance and shall be implemented 
prior to the initiation of core onload during Refueling Outage 16 
(Spring 2004).
    Amendment No.: 216.
    Renewed Facility Operating License No. NPF-7: Amendment changes the 
Improved Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43397). The supplements dated July 18, August 20, August 26, September 
4, September 5, September 22, September 26 (2 letters), November 10, 
December 8, and December 17, 2003, and January 6, January 22 (2 
letters), February 12, February 13, and March 1, 2004, contained 
clarifying information only and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 2004.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22.
    Therefore, pursuant to 10 CFR 51.22(b), no environmental impact 
statement or environmental assessment need be prepared for these 
amendments. If the Commission has prepared an environmental assessment 
under the special circumstances provision in 10 CFR 51.12(b) and has 
made a determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management System's (ADAMS) Public Electronic Reading Room on the 
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. 
If you do not have access to ADAMS or if there are problems in 
accessing the documents located in ADAMS, contact the NRC Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to pdr@nrc.gov.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, 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 2.309, which 
is available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1-

[[Page 22889]]

800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
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. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ 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/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by email to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon 
Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of application for amendment: April 2, 2004, as superseded by 
application dated April 8, 2004.
    Description of amendment request: The amendment revises the 
Technical Specification 3.3.5, ``Loss of Power (LOP) Diesel Generator 
(DG) Start Instrumentation,'' to allow performance of Surveillance 
Requirement (SR) 3.3.5.2 for the trip actuation device operational 
test, prior to first entry into MODE 4, by adding a note to the 
FREQUENCY column of SR 3.3.5.2 on a one-time basis.
    Date of issuance: April 15, 2004.
    Effective date: April 15, 2004, and shall be implemented within 10 
days from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. DPR-80: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. A public notice was published in the San 
Luis Obispo Tribune on April 13 and 14, 2004. The notice provided an 
opportunity to submit comments on the Commission's proposed NSHC 
determination. No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and

[[Page 22890]]

final NSHC determination are contained in a safety evaluation dated 
April 15, 2004.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

    Dated in Rockville, Maryland, this 19th day of April, 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-9225 Filed 4-26-04; 8:45 am]
BILLING CODE 7590-01-P