[Federal Register Volume 63, Number 37 (Wednesday, February 25, 1998)]
[Notices]
[Pages 9589-9618]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-4620]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 2, 1998, through February 12, 1998. 
The last biweekly notice was published on February 11, 1998 (63 FR 
6968).

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

[[Page 9590]]

within 30 days after the date of publication of this notice will be 
considered in making any final determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By March 27, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(I)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

[[Page 9591]]

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

    Date of amendment request: October 4, 1996, as supplemented by 
letters dated June 6, September 19, November 7, and December 16, 1997.
    Description of amendment request: The proposed amendment for each 
unit identified above would change the distance criterion in Action b 
to Limiting Condition for Operation (LCO) 3/4.1.3, ``Movable Control 
Assemblies,'' by which more than one full-length or part-length control 
element assembly (CEA) is misaligned from any other CEA in its group. 
Action b states, in part, that if the misalignment is greater than the 
specified distance criterion, the reactor core is to be placed in at 
least hot standby within 6 hours. The proposed amendment would reduce 
the distance criterion from 19 inches to 9.9 inches, and replace hot 
standby in 6 hours by ``open the reactor trip breakers.''
    This proposed amendment is included as a ``more restrictive'' 
change in the conversion of the current Technical Specifications (CTS) 
to the Improved Technical Specifications, which was noticed in the 
Federal Register (62 FR 18153) on April 14, 1997. The proposed 
amendment would be included in Action F to LCO 3.1.5, ``Movable Control 
Assemblies,'' of the Improved Technical Specifications. This proposed 
amendment is a change to the current Technical Specifications and is in 
addition to the six proposed changes to the CTS or proposed deviations 
to the Improved Standard Technical Specifications (NUREG-1432) which 
were identified in the notice of April 14, 1997.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes provide more stringent requirements than 
previously existed in the CTS. The more stringent requirements will not 
result in operation that will increase the probability of initiating an 
analyzed event. If anything, the new requirements may decrease the 
probability or consequences of an analyzed event by incorporating the 
more restrictive changes discussed in the specific Discussion of 
Changes [for specification 3.1.5]. These changes will not alter 
assumptions relative to mitigation of an accident or transient event. 
The more restrictive requirements will not alter the operation and will 
continue to ensure process variables, structures, systems, or 
components are maintained consistent with safety analyses and licensing 
basis [for the plant]. These changes have been reviewed to ensure that 
no previously evaluated accident has been adversely affected. 
Therefore, these changes will not involve a significant increase in the 
probability or consequences of an accident evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Making existing requirements more restrictive and adding more 
restrictive requirements to the CTS will not alter the plant 
configuration (no new or different type of equipment will be installed) 
or change the methods governing normal plant operation. These changes 
do impose different requirements. However, they are consistent with the 
assumptions made in the safety analyses, licensing basis, and NUREG-
1432 [for the plant]. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes provide more stringent requirements than 
previously existed in the CTS. An evaluation of these changes concluded 
that adding these more restrictive requirements either increases or has 
no impact on the margin of safety. The changes provide additional 
restrictions which may enhance plant safety. These changes maintain 
requirements of the safety analysis, licensing basis, and NUREG-1432 
[for the plant]. As such, no question of safety is involved. Therefore, 
these changes will not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: September 19, 1997.
    Description of amendment request: The proposed amendment would 
relocate the Radioactive Effluent Technical Specifications (RETS) and 
the Radiological Environmental Monitoring Program to the Offsite Dose 
Calculation Manual (ODCM), in accordance with the recommendations of 
Generic Letter 89-01 and NUREG-1433. In addition, changes to other 
sections of the TSs are being proposed to align the current TSs with 
NUREG-1433.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Operation of PNPS in accordance with the proposed change will not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated because of the following:

Definitions

    Definitions perform a supporting function for other sections of the 
TS. The proposed change to incorporate the definition for the Offsite 
Dose Calculations Manual (ODCM) into Section 5.0, ``Programs and 
Manuals'', subsection 5.5.1 of the proposed TS will carry forward the 
requirements contained in the DEFINITION, with minor editorial 
rewording to be consistent with NUREG 1433, and result in no technical 
changes. Since the requirements will remain, the impact on initiators 
of analyzed events or the assumptions assumed in the mitigation of 
accidents or transient events will not change. Editorial rewording 
(either adding or deleting) and reformatting is proposed to provide 
clarity and does not change any technical requirements.
    The definitions being proposed for relocation do not impact reactor 
operation, identify a parameter which is an initial condition 
assumption for a DBA or transient, identify a significant abnormal 
degradation of the reactor coolant pressure boundary, and do not

[[Page 9592]]

provide any mitigation of a design basis event.

RAD Effluents

    All editorial rewording (either adding or deleting) and renumbering 
is made to restructure the section accounting for the requirements 
relocated in accordance with Generic Letter 89-01. During the editorial 
rewording and renumbering of the Improved Technical Specifications, no 
technical changes (either actual or interpretational) to the TS were 
made unless they were identified and justified.
    Adding a note to clearly indicate that the first sample for noble 
gas activity is not required for 31 days after SJAE is placed in 
operation has always been considered the intent of this surveillance 
requirement. This allowance is consistent with the frequency for the 
required surveillance and allows time for concentrations of longer 
lived isotopes to reach equilibrium. In addition, other instrumentation 
continuously monitors the offgas to alert operators of significant 
increases in radioactivity.
    The proposed change provides more stringent requirements than 
previously existed in the Technical Specifications. The more stringent 
requirements will not result in operation that will increase the 
probability of initiating an analyzed event. If anything, the new 
requirements may decrease the probability or consequences of an 
analyzed event by incorporating the more restrictive changes discussed 
above. The change will not alter assumptions relative to mitigation of 
an accident or transient event. The more restrictive requirements will 
not alter the operation of process variables, structures, systems, or 
components as described in the safety analyses.
    These proposed changes relocate requirements from the Technical 
Specifications to the T. S. BASES, FSAR, or ODCM. The licensee 
controlled document containing the relocated requirements will be 
maintained using the provisions of 10 CFR 50.59 or a change control 
process in the Administrative Controls Section of the Technical 
Specifications. Since any changes to these licensee controlled 
documents will be evaluated per an NRC approved change control process, 
no increase in the probability or consequences of an accident 
previously evaluated will be allowed.
    Basing the potential fission product release rate on gross gamma 
activity rate is more representative of the whole body dose that would 
be received by an individual at the site boundary should a release 
occur. Therefore, reasonable assurance that the potential whole body 
accident dose to an individual at the exclusion area boundary will not 
exceed a small fraction of the limits specified in 10 CFR Part 100 is 
maintained.
    Allowing the sample to be taken from either pretreatment monitor 
station will have no effect on the objective of assuring that the 
potential whole body accident dose to an individual at the exclusion 
area boundary will not exceed a small fraction of the limits specified 
in 10 CFR Part 100, because both monitor stations are prior to 
treatment, adsorption, or delay of the noble gases.

RAD Material Source

    The requirements for miscellaneous radioactive materials do not 
impact reactor operation, identify a parameter which is an initial 
condition assumption for a DBA or transient, identify a significant 
abnormal degradation of the reactor coolant pressure boundary, and do 
not provide any mitigation of a design basis event.

Major Design Features

    The reformatting, renumbering, and rewording along with the other 
changes listed involve no technical changes to existing Technical 
Specifications. The proposed changes are administrative in nature and 
do not impact initiators or assumptions of analyzed accidents or 
transient events.
    The proposed change provides more stringent requirements than 
previously existed in the Technical Specifications. The more stringent 
requirements will not result in operation that will increase the 
probability of initiating an analyzed event. If anything, the new 
requirements may decrease the probability or consequences of an 
analyzed event by incorporating the more restrictive changes discussed 
above. The change will not alter assumptions relative to mitigation of 
an accident or transient event. The more restrictive requirements will 
not alter the operation of process variables, structures, systems, or 
components as described in the safety analyses.
    These proposed changes relocate requirements from the Technical 
Specifications to the FSAR. Since any changes to the FSAR must be 
evaluated per 10 CFR 50.59, no increase (significant or insignificant) 
in the probability or consequences of an accident previously evaluated 
will be allowed.

Administrative Controls

    The reformatting, renumbering, and rewording along with the other 
changes listed involves no technical changes to existing Technical 
Specifications. The change to the existing Technical Specifications was 
done in order to be consistent with the NUREG-1433. During development 
of NUREG-1433, certain wording preferences or English language 
conventions were adopted. The proposed change to this section is 
administrative in nature and does not impact initiators of analyzed 
events. It also does not impact the assumed mitigation of accidents or 
transient events.
    The proposed change provides more stringent requirements than 
previously existed in the Technical Specifications. These more 
stringent requirements are administrative in nature (e.g., specifying 
additional responsibilities for plant personnel, ensuring overtime 
control, incorporating program and manual requirements already in 
place, and adding details to reports). These additional requirements 
will not alter the plant configuration (no new or different type of 
equipment will be installed) or changes in methods governing normal 
plant operation, not alter assumptions relative to the mitigation of an 
accident or transient event, or alter the operation of process 
variables, structures, systems, or components as described in the 
safety analyses.
    This proposed change relocates requirements from the Technical 
Specifications to licensee controlled documents. The licensee 
controlled documents containing the relocated requirements are required 
to meet the applicable regulation and any change process invoked by the 
regulation. Since any changes to the licensee controlled document must 
continue to meet the regulation, no increase (significant or 
insignificant) in the probability or consequences of an accident 
previously evaluated will be allowed.
    This change proposes to provide flexibility in meeting the minimum 
shift staffing for up to two hours in order to provide for unexpected 
absence. The proposed change does not affect the probability of an 
accident. The actions of an individual are not assumed to be an 
initiator of any analyzed event. Also, the change does not negate the 
requirement to have licensed individuals in the control room. This 
proposed change does not impact the assumptions of any design basis 
accident. This change will not alter assumptions relative to the 
mitigation of an accident or transient event.
    This change proposes to relax the requirement to have an individual 
qualified in radiation protection procedures to be onsite when fuel is 
in the reactor. The proposed change will allow the position to be 
vacant for up

[[Page 9593]]

to two hours in order to provide for unexpected absence.
    The proposed change does not affect the probability of an accident. 
The actions of an individual qualified in radiation protection 
procedures are not assumed to be an initiator of any analyzed event. 
Also, the consequences of an accident are not affected by the presence 
of an individual qualified in radiation protection. This proposed 
change does not impact the assumptions of any design basis accident. 
This change will not alter assumptions relative to the mitigation of an 
accident or transient event. This change will not have any impact on 
the plant safety because the presence of a person qualified in 
radiation protection is not required for the mitigation of any 
accident.
    This change proposes to relax the requirement for submitting the 
Radioactive Effluent Release Report and to relocate the report details 
outside the TS. The current TS require the report to be submitted semi-
annually. This proposed change will allow the report to be submitted 
annually as required by 10 CFR 50.36a. The proposed change does not 
affect the probability of an accident. Neither the submittal 
requirements nor the contents of the Radioactive Effluent Release 
Report is assumed to be an initiator of any analyzed event. Also, the 
consequences of an accident are not affected by submittal requirements 
nor the contents of the Radioactive Effluent Release Report. This 
proposed change does not impact the assumptions of any design basis 
accident. This change will not alter assumptions relative to the 
mitigation of an accident or transient event. This change has no impact 
on the safe operation of the plant. The report will still be required 
to be submitted and does not affect any plant equipment or requirements 
for maintaining plant equipment. The submittal of this report is not 
required for the mitigation of any accident.
    The proposed alternatives for control of access to high radiation 
areas are consistent with the intent of 10 CFR 20.1601(a) and (b). The 
proposed changes do not affect the probability of an accident. The 
controls used for access to high radiation areas are not assumed in the 
initiation of any analyzed event. Also, the consequences of an accident 
are not affected by these changes. These changes are both consistent 
with good radiological safety practice and will provide an adequate 
level of radiation protection. These proposed changes do not impact the 
assumptions of any design basis accident. These changes will not alter 
assumptions relative to the mitigation of an accident or transient 
event. These changes have no impact on safe operation of the plant.

Radiological Environmental Monitoring

    The proposed changes only alter the format and location of 
procedural details and administrative controls of the radioactive 
effluents, radiological environmental monitoring, and solid radioactive 
waste programs. The changes are administrative in nature and do not 
involve any change to the configuration or operation of plant 
equipment. The Radiological Effluent Technical Specifications (RETS) 
procedural details are being moved to the Offsite Dose Calculation 
manual (ODCM). In addition, new administrative controls have been added 
to the Technical Specifications which will provide an equivalent level 
of assurance that activities involving radioactive effluents, solid 
radioactive waste, and radiological environmental monitoring are 
conducted in full compliance with regulatory requirements. Since any 
changes to these requirements will require NRC approval, no increase in 
the probability or consequences of an accident previously evaluated 
will be allowed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Operation of PNPS in accordance with the proposed change will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated because of the following:

Definitions

    These proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
change will not impose any new or different requirements or eliminate 
any existing requirements.
    Relocating these definitions will not alter the plant configuration 
(no new or different type of equipment will be installed) or change 
methods governing normal plant operation. Relocating requirements will 
not impose different requirements and adequate control of information 
will be maintained. Relocating these definitions will not alter 
assumptions made in the safety analysis and licensing basis.

RAD Effluents

    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
change will not impose any new or different requirements or eliminate 
any existing requirements.
    Making existing requirements more restrictive and adding more 
restrictive requirements to the Technical Specifications will not alter 
the plant configuration (no new or different type of equipment will be 
installed) or change methods governing normal plant operation. These 
changes are consistent with current design bases, licensing bases or 
assumptions made in the safety analysis.
    These changes do not alter the plant configuration (no new or 
different type of equipment will be installed) or methods governing 
normal plant operation. These changes will not impose different 
requirements and adequate control of information will be maintained. 
These changes do not alter assumptions made in the safety analysis and 
licensing basis.
    The proposed change will not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. Operation of the 
plant will not be altered by this change. This change will not place 
the plant in any new condition or introduce any mode of operation not 
previously analyzed.
    The proposed change will not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. Operation of the 
plant will not be altered by this change. This change will not place 
the plant in any new condition or introduce any mode of operation not 
previously analyzed.

RAD Material Source

    Relocating these requirements will not alter the plant 
configuration (no new or different type of equipment will be installed) 
or change methods governing normal plant operation. Relocating 
requirements will not impose different requirements and adequate 
control of information will be maintained. Relocating requirements does 
not alter assumptions made in the safety analysis and licensing basis.

Major Design Features

    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
change will not impose any new or different requirements or eliminate 
any existing requirements.

[[Page 9594]]

    Making existing requirements more restrictive and adding more 
restrictive requirements to the Technical Specifications will not alter 
the plant configuration (no new or different type of equipment will be 
installed) or changes in methods governing normal plant operation. The 
change does impose different requirements. However, the change is 
consistent with assumptions made in the safety analyses.
    These changes relocate requirements to the FSAR. These changes do 
not alter the plant configuration (no new or different type of 
equipment will be installed) or the methods governing normal plant 
operation. These changes do not impose different requirements and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis and licensing basis.

Administrative Controls

    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
change will not impose any new or different requirements or eliminate 
any existing requirements.
    Making existing requirements more restrictive and adding new 
requirements to the Technical Specifications will not alter the plant 
configuration (no new or different type of equipment will be installed) 
or changes in the methods governing normal plant operation.
    This change relocates requirements to a licensee controlled 
document. This change will not alter the plant configuration (no new or 
different type of equipment will be installed) or changes in methods 
governing normal plant operation. This change will not impose different 
requirements and adequate control of information will be maintained. 
This change will not alter assumptions made in the safety analysis and 
licensing basis.
    This change proposes to provide flexibility in meeting the minimum 
shift staffing for up to two hours in order to provide for an 
unexpected absence. The proposed change will not create the possibility 
of an accident. This change will not physically alter the plant (no new 
or different type of equipment will be installed).
    This change proposes to relax the requirement to have an individual 
qualified in radiation protection procedures to be onsite when fuel is 
in the reactor. The proposed change will allow the position to be 
vacant for up to two hours in order to provide for unexpected absence. 
The proposed change will not create the possibility of an accident. 
This change will not physically alter the plant (no new or different 
type of equipment will be installed) or the methods of operation.
    This change will not physically alter the plant (no new or 
different type of equipment will be installed). The changes in methods 
governing normal plant operation are consistent with the current safety 
analysis assumptions.
    The proposed change will not create the possibility of an accident. 
This change will not physically alter the plant (no new or different 
type of equipment will be installed). The changes in methods governing 
normal plant operation are consistent with the current safety analysis 
assumptions.

Radiological Environmental Monitoring

    The procedural requirements of the RETS will be maintained in the 
ODCM. Operation of the plant will not be altered by the changes 
proposed to the administration of the RETS. This change will not place 
the plant in any new condition or introduce any mode of operation not 
previously analyzed.
    3. Does this change involve a significant reduction in a margin of 
safety?
    Operation of PNPS in accordance with the proposed change will not 
involve a significant reduction in a margin of safety because of the 
following:

Definitions

    Definitions perform a supporting function for other sections of the 
TS and the proposed editing, omission or relocation of definitions 
associated with this change will not, by itself, reduce existing 
restrictions on plant operations.
    The definitions to be transposed from the Technical Specifications 
to the ODCM are the same as the existing Technical Specifications. 
Future changes to the ODCM will be controlled in accordance with 
proposed technical specification 5.5.1 ``Offsite Dose Calculation 
Manual (ODCM)''.

RAD Effluents

    The change is administrative in nature and does not involve any 
technical changes. The proposed change will not reduce a margin of 
safety because it has no impact on any safety analysis assumptions. 
Also, because the change is administrative in nature, no question of 
safety is involved.
    Adding these new requirements and making existing ones more 
restrictive does not affect any safety analysis assumptions. As such, 
no question of safety is involved.
    The requirements to be relocated from the Technical Specifications 
to the FSAR T.S. BASES, or ODCM are the same as the existing Technical 
Specifications and any future changes to this licensee controlled 
document will be evaluated per an NRC approved change control process.
    Specifying a release rate based only on gamma activity is more 
representative of the whole body dose that would be received by an 
individual at the site boundary should a release occur. The actual 
margin of safety could be increased because potential errors in 
converting beta activity to whole body exposures are eliminated
    The sample used to determine the gaseous activity rate will 
continue to be taken prior to treatment, adsorption, or delay of the 
noble gases.

RAD Material Source

    This change relocates requirements from the Technical 
Specifications to a licensee controlled document. This change will not 
reduce a margin of safety since it has no impact on any safety analysis 
assumptions. In addition, the requirements to be transposed from the 
Technical Specifications to the licensee controlled documents are the 
same as the existing Technical Specifications. Since any future changes 
to these licensee controlled documents must be evaluated per the cited 
regulations or requirements of 10 CFR 50.59, no reduction (significant 
or insignificant) in a margin of safety will be allowed.

Major Design Features

    The changes are administrative in nature and do not involve any 
technical changes. The proposed changes do not impact initiators or 
assumptions of analyzed accidents or transient events.
    These new or more restrictive requirements are consistent with the 
current design and licensing bases; therefore, a margin of safety is 
not affected.
    These changes relocate requirements from the Technical 
Specifications to the FSAR. The requirements to be are the same as the 
existing Technical Specifications. Since any future changes to the FSAR 
must be evaluated per the requirements of 10 CFR 50.59, no reduction 
(significant or insignificant) in a margin of safety will be allowed.

Administrative Controls

    The change is administrative in nature and will not involve any 
technical changes. The proposed change will not reduce a margin of 
safety because it has no impact on any safety analysis assumptions.

[[Page 9595]]

    Adding these new requirements and making existing ones more 
restrictive does not introduce any new tests or changes in methods 
governing normal plant operation. Therefore, the changes do not impact 
any safety analysis assumptions.
    This change relocates requirements from the Technical 
Specifications to a licensee controlled document. The licensee 
controlled documents containing the relocated requirements are required 
to meet the applicable regulation and any change process invoked by the 
regulation. Since any changes to a licensee controlled document must 
continue to meet the regulation, no increase (significant or 
insignificant) in the probability or consequences of an accident 
previously evaluated will be allowed.
    This change proposes to provide flexibility in meeting the minimum 
shift staffing for up to two hours in order to provide for unexpected 
absence. This proposed change has no effect on the assumptions of a 
design basis accident. The safety analysis assumptions will still be 
maintained; thus, no question of safety exists.
    This change proposes to relax the requirement to have an individual 
qualified in radiation protection procedures to be onsite when fuel is 
in the reactor. The proposed change will allow the position to be 
vacant for up to two hours in order to provide for unexpected absence. 
The margin of safety is not affected by the presence or absence on site 
of an individual qualified in radiation protection procedures. This 
proposed change has no effect on the assumptions of the design basis 
accident. This change will not have any impact on the plant safety 
because the presence of a person qualified in radiation protection is 
not required for the mitigation of any accident. The safety analysis 
assumptions will still be maintained; thus, no question of safety 
exists.
    This proposed change has no effect on the assumptions of the design 
basis accident. This change has no impact on the safe operation of the 
plant. The report will still be required to be submitted and does not 
affect any plant equipment or requirements for maintaining plant 
equipment. The safety analysis assumptions will still be maintained; 
thus, no question of safety exists.
    The proposed alternatives for control of access to high radiation 
areas are consistent with the intent of 10 CFR 20.1601(a) and (b). The 
margin of safety is not reduced due to these proposed changes. These 
changes are both consistent with good radiological safety practices and 
have been found to provide an adequate level of radiation protection. 
In addition, these changes provide the benefit of ensuring radiation 
dose to all workers is minimized by providing the flexibility to select 
the best means of providing a barrier and access control to a high 
radiation area given the plant location and radiological conditions. 
These proposed changes have no impact on the safe operation of the 
plant. The safety analysis assumptions will still be maintained; thus, 
no question of safety exists.

Radiological Environmental Monitoring

    The proposed changes relocate the procedural details and Bases for 
RETS from the Technical Specifications to the ODCM. The RETS procedural 
details and Bases will be maintained by these programs. In addition, 
new administrative controls have been added to the Technical 
Specifications which assure the proper control and maintenance of these 
documents and provide an equivalent level of assurance that activities 
involving radioactive effluents, solid radioactive waste, and 
radiological environmental monitoring are conducted in full compliance 
with regulatory requirements.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 0236.
    Attorney for licensee: W.S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: November 7, 1997.
    Description of amendment request: The proposed amendment would 
revise the technical specifications and associated bases to allow the 
licensee to perform 10 CFR Part 50, Appendix J, Type A testing on 
Byron, Unit 2, and Braidwood, Unit 2, containments at least once per 10 
years based on a single successful Type A test, rather than two 
successful Type A tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Performance of Type A tests at a different interval does not 
involve a change to any structures, systems, or components, does not 
affect reactor operations, is not an accident initiator, and does not 
change any existing safety analysis previously evaluated in the UFSAR 
[Updated Final Safety Analysis Report]. Therefore, there is no 
significant increase in the probability of an accident previously 
evaluated.
    Several tables of UFSAR Chapter 15, ``Accident Analyses,'' provide 
containment leak rate values used in assessing the consequences of 
accidents discussed in this chapter. Although decreasing the test 
frequency can increase the probability that an increase in containment 
leakage could go undetected for an extended period of time, the risk 
resulting from this proposed change is inconsequential as documented in 
NUREG-1493, ``Performance-Based Containment Leakage Test Program''. 
This document indicated that given the insensitivity of reactor risk to 
containment leakage rate and a small fraction of leakage paths are 
detected solely by Type A testing, increasing the interval between 
integrated leak rate tests is possible with minimal impact on public 
risk. Further, industry experience presented in this document indicated 
that Type A testing has had insignificant impact on uncertainties 
involved with containment leak rates.
    Based on risk information presented in NUREG-1493, the proposed 
change does not increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not alter the plant design, systems, 
components, or reactor operations, only the frequency of test 
performance. New conditions or parameters that contribute to the 
initiation of accidents would not be created as a result of this 
proposed change. The change does not involve new equipment and existing 
equipment does not have to be operated in a

[[Page 9596]]

different manner, therefore there are no new failure modes to consider.
    Changing test intervals as shown in NUREG-1493 has no impact on, 
nor contributes to the possibility of a new or different kind of 
accident as evaluated in the UFSAR. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    With the exception of the test frequency, the actual tests will not 
change. Quantitative risk studies documented in NUREG-1493 regarding 
extended testing intervals demonstrated that there was minimal impact 
on the public health and safety. Reducing the frequency, as stated in 
the NUREG resulted in an ``imperceptible'' increase in risk to public 
safety. Further, a table in this NUREG regarding risk impacts due to a 
reduction in testing frequency suggested that there was also minimal 
difference in risk to the public safety when the test frequency was 
relaxed.
    The proposed change will not reduce the availability of systems and 
components associated with containment integrity that would be required 
to mitigate accident conditions nor are any containment leakage rates, 
parameters or accident assumptions affected by the proposed change.
    The proposed change does not involve a significant reduction in a 
margin of safety, based on the above information.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: December 30, 1997.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 3.7.1.3, ``Condensate Storage 
Tank,'' (CST) and its associated Bases for Byron and Braidwood to raise 
the minimum allowable CST level to ensure that a sufficient volume of 
water is available to meet the design basis requirements for the 
auxiliary feedwater (AFW) system supply. The proposed amendment would 
also revise the AFW system transfer to essential service water (SX) 
trip setpoint and allowable value in Table 3.3-4 to ensure that the 
design basis requirements for the AFW system are accurately reflected 
in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The amount of water in the CST [Condensate Storage Tank] at the 
beginning of an accident and the setpoint for AF [auxiliary feedwater] 
pump suction pressure-low trip have no impact on the probability of 
occurrence of any accident analyzed in the UFSAR [Updated Final Safety 
Analysis Report]. This is due to the availability of the safety-related 
SX [essential service water] water supply as a backup system. 
Therefore, the probability of an accident previously evaluated is 
unchanged.
    The loss of the Safety Category II CST under accident conditions 
has already been evaluated in the UFSAR. The SX system is the emergency 
source of water supply to the AF system under accident conditions. The 
design basis analysis for the essential service water (SX) system and 
the Limiting Condition for Operation requirements for the ultimate heat 
sink ensure that a sufficient supply of water is available to plant 
operators to mitigate the consequences of all analyzed accidents. None 
of the proposed changes to the CST minimum level or the setpoints 
documented in TS Table 3.3-4, functional unit 6.g. has any negative 
impact on the assumptions or results of these analyzed accidents. To 
the contrary, the proposed changes will ensure that the CST remains 
available as the primary supply of water to the AF system and that 
automatic suction transfer will occur for circumstances where the 
Safety Category II CST becomes unavailable (e.g., seismic event or 
tornado).
    The level in the CST and the associated instrumentation and 
setpoints help ensure that sufficient water is available to plant 
operators to mitigate the consequences of accidents that are analyzed 
in the UFSAR. The SX system is the emergency source of water credited 
in the UFSAR. However, the proposed Technical Specification Bases 
require that sufficient water be maintained in the CST to respond to 
postulated events where the CST remains available (e.g., non-seismic 
related events and events with no tornado assumed). The proposed CST 
levels ensure that this requirement is met. The water level requirement 
for the CST provides additional assurance that plant operators remain 
capable of responding to postulated events as described in the UFSAR. 
Therefore, the proposed changes do not increase the consequences of an 
accident previously evaluated.
    Therefore this proposed amendment does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes are being implemented to account for 
instrument accuracy and AF system suction requirements that affect the 
volume of useable water in the CST. The amendment request incorporates 
the full design requirements of the AF System and components to ensure 
that sufficient water is maintained in the CST. The changes reduce the 
probability of an undesirable introduction of lower quality essential 
service (SX) system water into the steam generators unless required due 
to the unavailability of the CST during emergency conditions (e.g., 
seismic event or tornado). Although the SX system is the safety-related 
water supply to AF, the water contains high levels of impurities and 
sediment that could eventually degrade the steam generators. The CST 
contains demineralized water. Therefore, the long term reliability and 
availability of the steam generators is enhanced by precluding 
introduction of SX water into the steam generators unless required 
under emergency conditions. The proposed CST levels account for the 
incremental increase in CST water

[[Page 9597]]

volume required due to the larger metal mass and primary volume of the 
replacement steam generators for Byron Unit 1 and Braidwood Unit 1. 
Finally, the trip setpoint and allowable values in Table 3.3-4 of the 
TS are being updated to reflect the current design basis of the AF 
system. The required CST level changes when plant modifications are 
completed. Each configuration has been evaluated and the associated CST 
level maintains a sufficient water volume to perform its design 
function.
    The modification to the suction pressure circuitry involves the 
addition of an electronic ``lead-lag'' circuit card for the motor-
driven AF pump, which experiences the most severe startup suction 
pressure transients. This circuit card will be set up for ``lag'' only 
operation and will filter the suction pressure signal during transients 
associated with pump startup or other sudden changes in flow or 
pressure. This will prevent an inadvertent trip during transient 
conditions when the CST is available. In situations where the CST is 
unavailable, the suction pressure will decrease with no recovery until 
switchover. Under this condition, the output of the lead-lag card will 
continue to decrease as well until the switchover setpoint is reached. 
The time constant of the lead-lag card was selected such that the 
resulting time delays in actuating SX switchover and pump trip are 
consistent with pump protection requirements.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated. This 
conclusion is also valid when considering the planned modifications to 
the AF suction pressure transient circuitry.
    3. The change does not involve a significant reduction in a margin 
of safety.
    The proposed change is made in the conservative direction with 
respect to the current TS requirements for minimum CST level and AF 
pump CST to SX switchover setpoints. Increasing the volume of water 
contained in the CST level provides redundancy to the safety-related 
source of water to the AF supply, which is the SX system. In 
combination, the CST and the SX system ensure that sufficient water is 
available to feed the steam generators under all anticipated normal and 
emergency conditions to cool a unit from full power conditions down to 
350 degrees Fahrenheit, when the residual heat removal system can be 
placed into service. The proposed changes ensure the CST will have 
sufficient water to meet all normal operating conditions and mitigate 
the consequences of all analyzed accidents except those that result in 
CST unavailability. In addition, automatic switchover of the AF water 
supply from the CSTs to SX will occur as assumed in the current safety 
analyses for events where the CST becomes unavailable. The SX system 
remains capable of supplying the emergency source of water to the AF 
supply.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: January 28, 1998 (NRC-98-0002).
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) surveillance requirements 
4.8.2.1.a.2, 4.8.2.1.b, and 4.8.2.1.c.4 to accommodate differences in 
the monitored parameters between the existing batteries and the 
batteries that will be installed for Division II during the sixth 
refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a change in the manner in which 
the plant is operated. TS Section 4.8.2.1 is being revised to reflect 
the new Division II battery cell/system characteristics and associated 
requirements. The new battery will have an increased capacity over the 
present battery, while maintaining the existing battery system voltage 
requirements. This is possible because the present and new battery 
specific gravity (1.215) and type (lead calcium) are the same. Also, 
the end of battery system discharge voltage remains the same as 210 
VDC. The Division II batteries will continue to furnish power to 
redundant essential loads as required and as designed. The new 
surveillance requirement voltages are based on the same volts/cell 
criteria used for the existing batteries. Furthermore, failure or 
malfunction of the station batteries does not initiate any of the 
analyzed accidents previously evaluated in the UFSAR [updated final 
safety analysis report]. The changes described will therefore not 
involve an increase in the probability or consequences of an accident 
previously evaluated.
    2. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The new battery is Class 1E qualified equipment and is being 
maintained within the same overall design parameters as the existing 
battery. That is, the battery terminal voltage on float voltage 
conditions (2.167 volt[s]/cell), overvoltage conditions (2.5 volts/
cell) and charger capability (2.15 volts/cell) are the same as the 
original design. Furthermore, the end of system discharge voltage of 
the battery system is maintained the same; therefore, there is no 
negative impact to plant loads supplied by the batteries. Failures of 
the batteries and chargers have been considered in both the existing 
and modified configurations. The proposed changes will not change 
performance or reliability nor introduce any new or different failure 
modes or common mode failure and will therefore not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The changes do not involve a significant reduction in the margin 
of safety.
    The changes act to increase overall battery capacity from 560 
ampere-hours to 1200 ampere-hours with the minimum battery discharge 
voltage remaining at 210 VDC (or 105 VDC per battery). The battery 
terminal voltage on float voltage conditions (2.167 volt[s]/cell), 
overvoltage conditions (2.5 volts/cell) and charger capability (2.15 
volts/cell) are the same as the original design. The new surveillance 
requirement voltages are based on the same volts/cell criteria used for 
the existing batteries. The batteries' ability to satisfy the design 
requirements (battery duty cycle) of the dc system will not be reduced 
from original plant design and will therefore not have any negative 
impact to plant loads the battery supplies. The

[[Page 9598]]

proposed changes therefore do not involve a reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: January 28, 1998 (NRC-98-0003).
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) 3.4.10, TS Figure 3.4.10-1 and the 
associated bases by changing the prohibited and restricted operating 
regions associated with core thermal-hydraulic stability. TS 3.4.1.4, 
TS Figure 3.4.1.4-1, and the associated bases would also be revised to 
reflect stability-related improvements in operating restrictions for 
idle recirculation loop startup. Finally, in an unrelated change, TS 
Tables 3.3.7.5-1 and 4.3.7.5-1 would be revised to delete neutron flux 
from the parameters the licensee is required to monitor by TS 3.3.7.5, 
Accident Monitoring Instrumentation.
    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:

Thermal Hydraulic Stability and Idle Recirculation Loop Startup

    1. The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    These changes act to prohibit operations which have been found to 
carry a significant potential for the formation of core thermal-
hydraulic instabilities and eliminates inappropriate technical 
specifications for maintaining <50% recirculation loop flow before 
starting the idle recirculation pump. As such, operation in compliance 
with the proposed provisions does not affect any initiating mechanism 
for previously evaluated accidents or the response of the plant to a 
previously evaluated accident. The actions taken lead to placing the 
plant in a safe condition and are not themselves associated with an 
initiator for a previously evaluated accident. Therefore, the change 
does not represent a significant increase in the probability or 
consequences of any previously evaluated accident.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    As discussed above, the change acts to restrict operations 
previously allowed. The change also provides remedial actions that act 
to place the plant in a safe condition. The actions specified are 
within the analyzed domain of plant operations. Unless an instability 
event is in progress, the new allowance to use a core flow increase to 
leave the Exit Region is no different than normal plant maneuvering. If 
an instability event is in progress, the new ACTION 3.4.10.c to scram 
the reactor takes precedence. The allowance to start an idle loop with 
the active loop flow <50% of rated flow has been shown to have no 
adverse [e]ffect on scram avoidance or jet pump riser brace vibration. 
Therefore, the proposed changes do not create a new or different type 
of accident.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    Consistent with the latest BWROG [Boiling Water Reactor Owners 
Group] guidance, the changes act to expand the Exit region compared to 
the current TS for core thermal-hydraulic instability and provide 
improved remedial actions which promptly terminate the potential for 
instability. These changes therefore do not involve a significant 
reduction in a margin of safety.

Post-Accident Monitoring

    1. The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change does not involve a change in plant design or a 
change in the manner in which the plant is operated. The long term 
post-accident design requirements of the Neutron Monitoring System 
(NMS) are not based on operator use for transients with scram, 
accidents with scram, and other occurrences without scram (Reference 6 
[of January 28, 1998, application]). For lesser events such as 
transients without scram, the NMS enhances the operator actions, since 
successful verification that power is below approximately 3% power can 
avoid non-routine operator actions (Reference 6). These lesser events 
establish design requirements for the NMS. The failure of this 
instrumentation during post-accident conditions will not prevent the 
operator from determining reactor power levels. Alternate parameter 
status will be available from which reactor power may be inferred. 
Based on the multiple inputs available to the operator, sufficient 
information will be available upon which to base operational decisions 
and to conclude that reactivity control has been accomplished. This 
change will therefore not represent a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change does not introduce a new mode of plant 
operation and does not involve the installation of any new equipment or 
modifications to the plant. Therefore, it does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed change eliminates a TS listing of a function to 
reflect the actual safety significance. As such it has no effect on 
actual plant operation and thus no impact on any margin of safety.
    Based on the above, Detroit Edison has determined that the proposed 
amendment does not involve a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: January 28, 1998 (NRC-98-0006).

[[Page 9599]]

    Description of amendment request: The proposed amendment would 
revise technical specification (TS) surveillance requirement 
4.4.3.2.2.a to extend the interval for leak rate testing of pressure 
isolation valves from 18 months to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed change revises the periodicity of TS Surveillance 
Requirement (SR) 4.4.3.2.2.a from ``At least once per 18 months'' to 
``At least once per 24 months.'' This change revises the testing 
periodicity only; no other testing methodology is being affected. The 
testing periodicity is being revised to be consistent with other 
Category ``A'' valves since the Pressure Isolation Valves (PIVs) are 
classified as Category ``A'' valves. Both ASME [American Society of 
Mechanical Engineers] [Code] Section XI and NUREG-1482 require Category 
``A'' valves to be leak tested on a periodicity of at least once every 
2 years.
    The function of the PIVs is to protect the low pressure portions of 
safety systems from the RCS [reactor coolant system] pressure. Periodic 
valve leak rate testing is performed on the PIVs to assure system 
integrity is maintained and to prevent the design pressure of the low 
pressure systems from being exceeded. The frequency of the inservice 
test could increase the probability that an increase in PIV seat 
leakage may occur. If this were to occur and the leakage was 
significant (assuming leakage through both the inboard and outboard 
valves of the same penetration), the excess leakage would be detected 
by the system leakage detection instrumentation which would require 
corrective actions to be taken to assure that leakage remained within 
allowable limits. Considering that past test results show very minimal 
seat leakage changes over years of service, the consequences and 
probabilities resulting from the proposed change is considered minimal.
    The proposed change does not impose or eliminate any testing 
requirements. This change is only a change to the frequency (testing 
interval) for measuring the seat leakage through the PIVs. The PIVs 
will continue to be tested in accordance with ASME Code Section XI. 
This change does not affect any of the parameters or conditions that 
could contribute to the initiation of any accidents previously 
evaluated and therefore cannot increase the consequences or 
probabilities of any accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change does not involve a change to the plant design 
or operation. As a result, the proposed change does not affect any of 
the parameters or conditions that could contribute to the initiation of 
any accidents. This change only involves the lengthening of the PIVs' 
testing frequency from 18 months to 24 months. The method for 
performing the actual tests are not changed. No new accident scenarios 
are created by extending the testing intervals. No safety-related 
equipment or safety functions are altered as a result of this change. 
Therefore, extending the test frequency does not create the possibility 
of a new or different kind of accident or malfunction from those 
previously analyzed.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    The proposed change only affects the frequency of the PIVs' seat 
leakage tests. The frequency is proposed to be extended to reflect the 
ASME Section XI, 1980 Edition, Winter 1980 Addenda, Section IWV-3422 
seat leakage testing periodicity requirement of 24 months. No other 
testing methodology is being changed. The allowable leakage limits will 
not be affected by this change. The margin of safety as defined in the 
bases of any Technical Specification will, therefore, not be reduced by 
extending the testing periodicity of the subject valves.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: January 28, 1998 (NRC-98-0008).
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) by modifying the ``#'' 
footnote to Table 1.2 and the ``*'' footnote to surveillance 
requirements 4.9.1.2 and 4.9.1.3 to permit the Reactor Mode Switch to 
be placed in the Run or Startup/Hot Standby positions to test switch 
interlock functions provided that all control rods are verified to 
remain fully inserted in core cells containing one or more fuel 
assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change would permit the Reactor Mode Switch to be 
placed in the Run or Startup/Hot Standby positions to test the switch 
interlock functions provided that all control rods are verified to 
remain inserted in core cells containing one or more fuel assemblies. 
The existing TS requires that all control rods be verified to remain 
inserted regardless of whether core cells are defueled. The reactor 
mode switch refuel position interlocks restrict the operation of 
refueling equipment or withdrawal of control rods to reinforce unit 
procedures that prevent the reactor from achieving criticality during 
refueling operations. As such, the refueling equipment interlocks 
preserve the assumptions for the analyses of a control rod withdrawal 
event or loading of a fuel assembly into an uncontrolled cell during 
refueling operations. The reactor mode switch refuel position 
interlocks are not initiators of any previously evaluated accident. The 
revised footnote requires that all control rods remain fully inserted 
in core cells containing one or more fuel assemblies while the mode 
switch is moved to support interlock testing. Additionally, when the 
reactor mode switch is unlocked to support interlock testing, TS 3.9.1 
prohibits core alterations. With all control rods fully inserted in 
core cells containing one or more fuel assemblies and no core 
alterations in progress, there are no credible mechanisms to initiate a 
reactivity excursion during the interlock

[[Page 9600]]

testing. Therefore, the proposed change does not involve a significant 
increase in the probability of a previously evaluated accident.
    The proposed change accommodates reactor mode switch refuel 
position interlock testing with one or more control rods removed as 
permitted by TS 3.9.10.1 and 3.9.10.2. In addition to requiring all 
fuel assemblies to be removed from core cells associated with removed 
control rods, TS 3.9.10.1 and 3.9.10.2 require minimum shutdown margin 
to be maintained in accordance with TS 3/4.1.1. Under these conditions, 
it is not possible for criticality to occur in the event of a 
withdrawal of a single control rod or loading of fuel assemblies into a 
single core cell with no control rod inserted. Therefore, the proposed 
change does not involve a significant increase in the consequences of a 
previously evaluated accident.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Repositioning of the reactor mode switch to test refueling position 
interlocks is permitted by both the existing and proposed TS. The 
proposed change affects only the conditions under which the mode switch 
can be repositioned. The proposed changes do not change underlying 
principles affecting the way in which the plant is operated and no new 
or different failure modes are introduced by the proposed change for 
any plant system or component. No new limiting single failure has been 
identified as a result of the proposed changes. Therefore, no new or 
different types of failures or accident initiators are introduced by 
the proposed changes.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change described above affects the conditions under 
which the reactor mode switch can be repositioned to accommodate refuel 
position interlock testing. The proposed change in combination with 
existing restrictions within the TS provide assurance that there is no 
credible mechanism to initiate a reactivity excursion during interlock 
testing. 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.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: January 28, 1998 (NRC-98-0011).
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) 3.4.2.1 by changing the tolerance 
for the as-found setpoints of the safety/relief valves (SRVs) from 
[plus or minus] 1 percent to [plus or minus] 3 percent of the nominal 
setpoint. The revised tolerance would be used when evaluating whether 
setpoint test results were acceptable. However, after initial testing, 
the as-left setpoints of the SRVs would be adjusted to within [plus or 
minus] 1 percent of the nominal setpoint.
    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 this change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change allows an increase in the SRV setpoint 
tolerance, determined by test after the valves have been removed from 
service, from [plus or minus] 1% to [plus or minus] 3%. The proposed 
change does not alter the SRV lift setpoints, the SRV lift setpoint 
test frequency, or the number of SRVs required to be operable. This 
change does not involve physical changes to the SRVs, nor does it 
change the operating characteristics or safety function of the SRVs. 
This change requires that the SRVs be adjusted to within [plus or 
minus] 1% of their nominal lift setpoints following testing and prior 
to installation in the plant.
    The only change, other than the change in setpoint tolerance, will 
be to increase the maximum rated speed of the RCIC [reactor core 
isolation cooling] turbine and pump. The increased speed is within the 
design limits of the system and the overspeed trip function retains 
adequate margin; therefore, RCIC operability is not affected by this 
change. Additionally, SRV actuation is not a precursor to any design 
basis accident analyzed for the Fermi 2 plant. Therefore, this change 
will not significantly increase the probability of an accident 
previously evaluated.
    Generic considerations related to the change in setpoint tolerance 
were addressed in NEDC-31753P, ``BWROG In-Service Pressure Relief 
Technical Specification Revision Licensing Topical Report,'' and were 
reviewed and approved by the NRC. The plant specific evaluations 
identified in the NRC[']s Safety Evaluation for NEDC-31753P were 
performed in order to support the proposed change (Cycle 6 reload 
licensing report, Power Uprate Safety Analysis, and NEDC-32788P, 
``Safety Review for Enrico Fermi Energy Center Unit 2 Safety/Relief 
Valve Setpoint Tolerance Relaxation Analyses''). These evaluations 
included transient analysis of the anticipated operational occurrences 
(AOOs); analysis of the design basis overpressurization event; 
evaluation of the performance of high pressure systems, motor operated 
valves, and vessel instrumentation and associated piping; and 
evaluation of the containment response during LOCA [loss of coolant 
accident] and the hydrodynamic loads on the SRV discharge lines and 
containment. Although not specified in the generic topical report NEDC-
31753P, an analysis of the short term pressurization phase of an ATWS 
[anticipated transient without scram] event was also performed. These 
analyses show that there is adequate margin to the design core thermal 
limits and to the reactor vessel pressure limits using a [plus or 
minus] 3% SRV setpoint tolerance. They also show that operation of the 
high pressure injection systems will not be adversely affected; and the 
containment response during LOCA will be acceptable. Therefore, this 
change will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Does this change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change to allow an increase in the SRV setpoint 
tolerance from [plus or minus] 1% to [plus or minus] 3% does not alter 
the SRV lift setpoints, the minimum SRV lift setpoint test frequency, 
or the number of SRVs required to be operable. This change does not 
involve physical changes to the SRVs, nor does it change the operating 
characteristics or the safety function of the SRVs. The only change to 
plant equipment will be to increase the RCIC turbine/pump maximum rated 
speed from 4550 rpm to 4600 rpm. The RCIC pump and turbine

[[Page 9601]]

have been verified to be capable of operating at the increased speed, 
pressure and temperature associated with this increase in maximum rated 
speed. These changes do not result in any changed component 
interactions. The SRVs and the RCIC System will continue to function as 
designed. Therefore, this change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    While the calculated peak vessel pressures for the ASME [American 
Society of Mechanical Engineers] overpressure event and the ATWS MSIVC 
[main steam isolation valve closure] event are higher than those 
calculated without the setpoint tolerance relaxation, both are still 
within the respective licensing acceptance limits associated with these 
events. Similarly, although the loads associated with SRV blowdown 
could increase slightly, containment loadings have been determined to 
remain within acceptance limits. These licensing acceptance limits have 
been determined by the NRC to provide a sufficient margin of safety. 
Additionally, the increased setpoint tolerances have been determined to 
have a negligible effect on the other accidents and transients 
analyzed. Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: January 17, 1998.
    Description of amendment request: The proposed amendment would 
revise the waste gas system line break accident analysis. The proposed 
changes would affect Beaver Valley Power Station, Unit No. 1 Updated 
Final Safety Analysis Report (UFSAR) Tables 11.3-7, ``Postulated 
Control Room Accident Dose,'' and 14.2-8, ``Parameters Used In Control 
Room Habitability Analysis Of The Waste Gas System Failure Analysis.'' 
The analysis references on Tables 11.3-7 and 14.2-8 would be revised 
due to the reanalysis of the waste gas system line break accident. In 
Table 11.3-7, the waste gas system line break accident gamma dose value 
would be revised from 0.0031 Rem to less than 0.01 Rem and the beta 
dose value would be revised from 0.013 Rem to less than 1.0 Rem.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change has no effect on the probability of an accident 
previously evaluated. The proposed change results from the correction 
of values and change to assumptions utilized in the original 
calculation to address resultant dose to Control Room operators in the 
event of the postulated Waste Gas System line break.
    The proposed change also corrects an error in UFSAR Table 14.2-8 
whereby the fraction of fuel with defects was assumed to be one 
percent, not 0.0026. This correction reflects the value used in the 
calculation and does not alter the results.
    The proposed change does not significantly increase the 
consequences of an accident previously analyzed. Although the 
correction to the calculation and revision to the assumptions used 
result in an insignificant increase to the postulated dose to the 
Control Room operators, the results remain below the acceptance limit 
of other postulated accidents presented in the UFSAR (Table 11.3-7) and 
the acceptance approved by the NRC in the NRC Safety Evaluation Report, 
Section 15.1, dated October 1974. The proposed change does not alter 
the currently approved Technical Specification. The proposed change 
does not affect the dose to the public.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not alter the physical plant or modify the 
modes of operation. The proposed change does not involve modifications 
to plant equipment nor does it alter operation of plant systems. 
Therefore operation of the facility with the proposed change does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    The proposed change does not reduce the margin of safety. The 
proposed change does not affect any plant systems or equipment. 
Therefore, the response of the plant to any actual events will not be 
affected, and the change does not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: November 13, 1997.
    Description of amendment request: The proposed change will modify 
Technical Specification (TS) 6.8.4.a, ``Primary Coolant Sources Outside 
Containment,'' to add portions of the containment vacuum relief (CVR) 
system and the primary sampling system to the program at Waterford 3.
    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. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds the containment vacuum relief (CVR) system 
and the primary sampling system to the Primary Coolant Sources Outside 
Containment Program in the Technical Specifications. The program will 
require preventative maintenance

[[Page 9602]]

and periodic visual inspection, and leak rate testing on appropriate 
portions of these systems to ensure leakage of radioactive fluids are 
as low as practicable. The addition of these two systems to the program 
will not affect the probability of an accident. Neither the CVR system 
nor the primary sampling system are initiators of any analyzed event. 
The consequences of an accident are not affected by this change. The 
maximum allowed leakage limits are not being increased due to the 
addition of these two systems. Any leakage from the CVR system will be 
factored into the overall leakage limits and any leakage from the 
primary sampling system will be kept to a minimum by performing 
required maintenance. This change does not affect the mitigation 
capabilities of any component or system nor does it affect the 
assumptions relative to the mitigation of accidents or transients. The 
addition of these systems to the program also helps ensure that the 
systems will perform their intended function. Therefore, the proposed 
change will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different type of accident 
from any accident previously evaluated?
    Response: No.
    The proposed change adds the CVR system and the primary sampling 
system to the Primary Coolant Sources Outside Containment Program in 
the Technical Specifications. The program will require preventative 
maintenance and periodic visual inspection, and leak rate testing on 
appropriate portions of these systems to ensure leakage of radioactive 
fluids are as low as practical. Neither the design nor configuration of 
the plant is being changed due to the addition of the CVR system to the 
program. Also, as a result of the CVR system being added to the 
program, there has been no physical change to plant systems, structures 
or components nor will the addition of the CVR system reduce the 
ability of any of the safety-related equipment required to mitigate 
anticipated operational occurrences (AOOs) or accidents.
    Although the addition of the primary sampling system to the program 
was a result of a change to the configuration of the plant, it does not 
reduce the ability of any safety-related equipment required to mitigate 
AOOs or accidents. Any leakage from the primary sampling system will be 
kept to a minimum by performing required maintenance.
    Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    Response: No.
    The proposed change adds the CVR system and the primary sampling 
system to the Primary Coolant Sources Outside Containment Program in 
the Technical Specifications. The program will require preventative 
maintenance and periodic visual inspection, and leak rate testing on 
appropriate portions of these systems to ensure leakage of radioactive 
fluids are as low as practical. This change will not affect the maximum 
containment leakage allowed in the Technical Specifications. The 
leakage from the CVR system will be added to the overall containment 
leakage rate. Any leakage from the primary sampling system will be kept 
to a minimum by performing required maintenance. The overall 
containment leakage requirement is required to be met and therefore, 
this change will not result in an increase in the analyzed dose 
consequences assumed in the Waterford 3 safety analysis. Therefore, the 
proposed change will not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: December 31, 1997.
    Description of amendment request: The proposed amendment will 
revise Technical Specification 5.6.1 and associated Figure 5.6-1, and 
Specification 5.6.3, to permit an increase in the allowed Spent Fuel 
Pool (SFP) storage capacity. The analyses supporting this request, in 
part, assume credit for up to 1266 ppm boron concentration existing in 
the SFP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    Analyses to support the proposed fuel pool capacity increase have 
been developed using conservative methodology. The analysis of the 
potential accidents summarized below has shown that there is no 
significant increase in the consequences of any accident previously 
analyzed. A review of relevant plant operations has also demonstrated 
that there is no significant increase in the probability of occurrence 
of any accident previously analyzed. This conclusion is also discussed 
below.
    Previously evaluated accidents that were examined for this proposed 
license amendment include: Fuel Handling Accident, Spent Fuel Cask Drop 
Accident, and Loss of all Fuel Pool Cooling.
    There will be no change in the mode of plant operation or in the 
availability of plant systems as a result of this proposed change; the 
systems interfacing with the spent fuel pool have previously 
encountered borated pool water and are designed to interact with 
irradiated spent fuel and remove the residual heat load generated by 
isotopic decay. The proposed amendment does not require a change in the 
maintenance interval or maintenance scope for the fuel pool cooling 
system or for the spent fuel cask crane. The frequency of cask handling 
operations and the maximum weight carried by the crane is not increased 
as a result of the proposed license amendment. Thus, there will be no 
increase in the probability of a loss of fuel pool cooling or in the 
probability of a failure of the cask crane as a result of the proposed 
amendment.
    There will not be a significant increase in the frequency of 
handling discharged assemblies in the fuel pool as a result of this 
change; any handling of fuel in the spent fuel pool will continue to be 
performed in borated water. If the license amendment is approved, there 
will be a one-time repositioning of certain discharged assemblies 
stored in the fuel pool to comply with the revised positioning 
requirements, but the increased pool storage capacity will permit the 
deferral of spent fuel handling associated with cask loading 
operations. Fuel manipulation during the repositioning activity will be 
performed in the same

[[Page 9603]]

manner as for fuel placed in the spent fuel pool during refueling 
outages. There will be no changes in the manner of handling fuel 
discharged from the core as a result of refueling; administrative 
controls will continue to be used to specify fuel assembly placement 
requirements. The relative positions of Region I and Region II storage 
locations will remain the same within the fuel pool. Therefore, the 
probability of a fuel handling accident has not been significantly 
increased.
    The consequences of a fuel handling accident have been evaluated. 
The radioactive release consequences of a dropped fuel assembly are not 
affected by the proposed increase in fuel pool storage capacity. They 
remain bounded by the results of calculations performed to justify the 
existing St. Lucie Unit 2 fuel storage racks and burnup limits. At the 
limiting fuel assembly burnup, radioactive releases from a dropped 
assembly would be only a small fraction of NRC guidelines. The input 
parameters employed in analyzing this event are consistent with the 
current values of fuel enrichment, discharge burnup and uranium content 
used at St. Lucie Unit 2 and with future use of the ``value-added'' 
fuel pellet design. Thus, the consequences of the fuel assembly drop 
accident would not be significantly increased from those previously 
evaluated.
    The capability of the fuel pool cooling system to handle the 
increased number of discharged assemblies has been examined. The impact 
of a total loss of spent fuel pool cooling flow on available equipment 
recovery time and on fuel cladding integrity has also been evaluated. 
For the limiting full core discharge, sufficient time remains available 
to restore cooling flow or to provide an alternate makeup source before 
boiloff results in a fuel pool water level less than that needed to 
maintain acceptable radiation dose levels. Analysis has shown that in 
the event of a total loss of fuel pool cooling fuel cladding integrity 
is maintained. Therefore, the consequences of a loss of fuel pool 
cooling event, including the effect of the proposed increase in fuel 
pool storage capacity, have not been significantly increased from 
previously analyzed results for this type of accident.
    The analysis of record pertaining to the radiological consequences 
of the hypothetical drop of a loaded spent fuel cask just outside the 
Fuel Handling Building was examined to determine the impact of the 
increased fuel storage capacity on this accident's results. The results 
of the previously performed analysis were determined to bound the 
conditions described by the proposed license amendment, thus the 
consequences of the cask drop accident would not be significantly 
increased as a result of this change.
    It is concluded that the proposed amendment to increase the storage 
capacity of the St. Lucie Unit 2 spent fuel pool will not involve a 
significant increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a new 
or different type of accident from any accident previously evaluated.
    In this license amendment FPL proposes to credit the negative 
reactivity associated with a portion of the soluble boron present in 
the spent fuel pool. Soluble boron has always been present in the St. 
Lucie Unit 2 spent fuel pool; as such the possibility of an inadvertent 
fuel pool dilution has always existed. However, the spent fuel pool 
dilution analysis demonstrates that a dilution of the Unit 2 spent fuel 
pool which could increase the pool keff to greater than 0.95 
is not a credible event. Neither implementation of credit for the 
reactivity of fuel pool soluble boron nor the proposed increase in the 
fuel pool storage capacity will create the possibility of a new or 
different type of accident at St. Lucie Unit 2.
    An examination of the limiting fuel assembly misload has determined 
that this would not represent a new or different type of accident. None 
of the other accidents examined as a part of this license submittal 
represent a new or different type of accident; each of these situations 
has been previously analyzed and determined to produce acceptable 
results.
    The proposed license amendment will not result in any other changes 
in the mode of spent fuel pool operation at St. Lucie Unit 2 or in the 
method of handling irradiated nuclear fuel. The spatial relationship 
between the fuel storage racks and the cask crane range of motion is 
not affected by the proposed change.
    As a result of the evaluation and supporting analyses, FPL has 
determined that the proposed fuel pool capacity increase does not 
create the possibility of a new or different type of accident from any 
accident previously evaluated.
    3. The proposed amendment will not involve a significant reduction 
in the margin of safety.
    FPL has determined, based on the nature of the proposed license 
amendment that the issue of margin of safety, when applied to this fuel 
pool capacity increase, should address the following areas:

(1) Fuel Pool reactivity considerations
(2) Fuel Pool boron dilution considerations
(3) Thermal-Hydraulic considerations
(4) Structural loading and seismic considerations

    The Technical Specification changes proposed by this license 
amendment, the proposed spent fuel pool storage configuration and the 
existing Technical Specification limits on fuel pool soluble boron 
concentration provide sufficient safety margin to ensure that the array 
of fuel assemblies stored in the spent fuel pool will always remain 
subcritical. The revised spent fuel storage configuration is based on a 
Unit 2 specific criticality analysis performed using methodology 
consistent with that approved by the NRC. Additionally, the soluble 
boron concentration required by current Technical Specifications 
ensures that the fuel pool keff will always be maintained 
substantially less than 0.95.
    The Unit 2 criticality analysis established that the 
keff of the spent fuel pool storage racks will be less than 
1.0 with no soluble boron in the fuel pool water, including the effect 
of all uncertainties and tolerances. Credit for the soluble boron 
actually present is used to offset uncertainties, tolerances, off-
normal conditions and to provide margin such that the spent fuel pool 
keff is maintained less than or equal to 0.95. FPL has also 
demonstrated that a decrease in the fuel pool boron concentration such 
that keff exceeds 0.95 is not a credible event.
    Current Technical Specifications require that the fuel pool boron 
concentration be maintained greater than or equal to 1720 ppm. This 
boron value is substantially in excess of the 520 ppm required by the 
uncertainty and reactivity equivalencing analyses discussed in this 
evaluation and the 1266 ppm value required to maintain keff 
less than or equal to 0.95 in the presence of the most adverse 
mispositioned fuel assembly.
    The St. Lucie Unit 2 fuel pool boron concentration will continue to 
be maintained significantly in excess of 1266 ppm; the proposed license 
amendment will not result in changes in the mode of operation of the 
refueling water tank (RWT) or in its use for makeup to the fuel pool. 
Thus, operation of the spent fuel pool following the proposed change, 
combined with the existing fuel pool boron concentration Technical 
Specification limit of 1720 ppm, will continue to ensure that 
keff of the fuel pool will be substantially less than 0.95.

[[Page 9604]]

    Even if this not-credible dilution event was to occur, no radiation 
would be released; the only consequence would be a reduction of 
shutdown margin in the fuel pool. The volume of unborated water 
required to dilute the fuel pool to a keff of 0.95 is so 
large (in excess of 358,900 gallons to dilute the fuel pool to 520 ppm 
boron) that only a limited number of water sources could be considered 
potential dilution sources. The likelihood that this level of water use 
could remain undetected by plant personnel is extremely remote.
    In meeting the acceptance criteria for fuel pool reactivity, the 
proposed amendment to increase the storage capacity of the existing 
fuel pool racks does not involve a significant reduction in the margin 
of safety for nuclear criticality.
    Calculations of the spent fuel pool heat load with an increased 
fuel pool inventory were performed using ANSI/ANS-5.1-1979 methodology. 
This method was demonstrated to produce conservative results through 
benchmarking to actual St. Lucie Unit 2 fuel pool conditions and by 
comparison of its results to those generated by a calculation using 
Auxiliary Systems Branch Technical Position 9-2 methodology. 
Conservative methods were also used to demonstrate fuel cladding 
integrity is maintained in the absence of cooling system forced flow. 
The results of these calculations demonstrate that, for the limiting 
case, the existing fuel pool cooling system can maintain fuel pool 
conditions within acceptable limits with the increased inventory of 
discharged assemblies. Therefore, the proposed change does not result 
in a significant reduction in the margin of safety with respect to 
thermal-hydraulic or spent fuel cooling considerations.
    The primary safety function of the spent fuel pool and the fuel 
storage racks is to maintain discharged fuel assemblies in a safe 
configuration for all environments and abnormal loadings, such as an 
earthquake, a loss of pool cooling or a drop of a spent fuel assembly 
during routine spent fuel handling. The proposed increase in spent fuel 
inventory on the fuel pool and the existing storage racks have been 
evaluated and show that relevant criteria for fuel rack stresses and 
floor loadings have been met and that there has been no significant 
reduction in the margin of safety for these criteria.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: November 22, 1996, as revised and 
replaced on February 2, 1998.
    Description of amendment request: The licensee proposed to change 
the Technical Specifications (TS) to allow the use of a temporary fuel 
oil storage system for up to 10 days in order to perform a surveillance 
requirement on the Unit 3 fuel oil storage tank with Unit 3 in Modes 5, 
6, or defueled.
    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:
    Question 1  Does the proposed license amendment involve a 
significant increase in the probability or consequences of an accident 
previously evaluated?
    The proposed amendment will allow the installation of a temporary 
fuel oil storage and transfer system for up to 10 days, once every 10 
years. EDGs [emergency diesel generators] are designed as backup AC 
power sources for essential safety systems in the event of a loss of 
offsite power. Since the EDGs are not accident initiators, the 
probability of occurrence of accidents previously analyzed has not been 
increased.
    The temporary fuel oil storage tanks will be located greater than 
fifty (50) feet from safety related or safe shutdown components or 
circuits. This does not produce any threat to fire protection or safe 
shutdown capability and therefore represents a configuration that is 
bounded by existing fire hazards analysis.
    The proposed amendment will not change the condition or minimum 
amount of operating equipment assumed in the plant safety analyses for 
accident mitigation. The temporary fuel storage and transfer system 
provides a reliable means of performing the required delivery support 
function for the Unit 3 EDGs.
    An insignificant increase in the consequences of an accident 
previously evaluated is possible since the temporary storage and 
transfer system will not meet requirements for Seismic Category I or 
Class 1E. However, the probability of a seismic event will be very low 
due to the limited time that the temporary storage system will be in 
use.
    The increase in the consequences of an accident previously 
evaluated is insignificant due to the following:
    Manual actions required to provide a 7 day supply of fuel to the 
EDGs can easily be accomplished in the 17 hours of EDG operation 
provided by the 3880 gallon capacity of a single EDG day and skid tank. 
The location of the temporary fuel oil supply inside the protected area 
security fence by the Central Receiving Facility provides multiple 
access routes to transfer fuel to the Unit 3 EDGs and is in close 
proximity to a severe weather shelter for the mobile tanker.
    Additionally, more than 17 hours will be available to manually 
transfer fuel from the temporary fuel storage tanks located inside the 
protected area, by filling the Unit 4 EDG storage tanks with 
approximately 8600 gallons of fuel oil above that required for Unit 4 
EDG operability. This extra capacity will be available to the Unit 3 
EDGs prior to taking the permanent Unit 3 storage tank out of service. 
This will be done by filling the Unit 4 fuel tanks to 39,000 gallons, 
which is just below the high level alarm. This gives a capacity of 4300 
gallons in each tank above the Unit 4 Technical Specification minimum 
required volume of 34,700 gallons. The Unit 4 tanks are contained 
within a Seismic Class 1 structure and protected by installed fire 
protection equipment.
    Combining the excess available fuel from the Unit 4 storage tanks 
and the nominal volume of the Unit 3 day and skid tanks gives a total 
of 12,480 gallons (4300 x 2+3880) of available fuel to either of the 
Unit 3 EDGs. This allows a run time for a Unit 3 EDG of 55 hours 
(assuming fuel oil transfer from Unit 4) prior to reaching the 
Technical Specification minimum volume for the Unit 4 fuel oil storage 
tanks. Manual actions to replenish the Unit 4 or Unit 3 fuel oil 
storage tanks from the temporary storage tanks, via the mobile tanker, 
can easily be accomplished within the 55 hours. Procedures currently 
exist for the transfer of fuel from (1) the mobile tanker to the 
auxiliary fill station at the Unit 3 EDGs, and (2) from the Unit 4 EDG 
storage tanks to the Unit 3 day tanks by using either of the Unit 4 
transfer pumps. The

[[Page 9605]]

Unit 4 transfer pumps are powered from redundant Class 1E power 
supplies.
    The temporary storage tanks will be located inside the protected 
area in the vicinity of the Nuclear Plant Central Receiving Facility. 
The temporary tanks will be located greater than fifty (50) feet from 
safety related or safe shutdown components or circuits. This does not 
produce any threat to fire protection or safe shutdown capability and 
therefore represents a configuration that is bounded by existing fire 
hazards analysis.
    A dedicated mobile tanker staged inside the protected area to 
transfer fuel from the temporary storage tanks to the permanent day/
skid tank system. The mobile tanker will have an integral transfer pump 
to facilitate movement of fuel to either of the two truck fills at the 
Unit 4 EDG building or day tank truck fills (auxiliary fill station) at 
the Unit 3 EDGs. One truck fill at the Unit 4 EDG building supplies 
fuel to the 4A and 4B storage tanks, the other truck fill at the Unit 4 
EDG building can provide fuel directly to the Unit 3 day tanks. This 
fuel supply will provide continued operation for 7 days. The temporary 
storage and transfer system will not meet requirements for Seismic 
Category I or Class 1E.
    The capability to operate an Unit 3 EDG for 7 days during the tank 
cleaning evolution will be assured by an approved plant procedure that 
controls the following:

 A minimum fuel supply of 3880 gallons from the Unit 3 day and 
skid tank. This provides 17 hours of operation.
 The extra fuel supply of 8600 gallons in the Unit 4 EDG tanks 
which will be transferred by using one of the installed Unit 4 transfer 
pumps. This provides an additional 38 hours of operation.
 Three temporary tanks containing a minimum fuel supply of 
38,000 gallons. This fuel supply will provide continued operation for 7 
days.

Consequently, operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Question 2  Does the proposed license amendment create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    The proposed amendment will not change the physical plant or modes 
of plant operation defined in the Turkey Point Units 3 and 4 operating 
license. The change will not involve addition or modification of 
equipment for Unit 3 EDG fuel storage and transfer. The temporary fuel 
supply system provides a reliable means of performing the required fuel 
delivery support function for the Unit 3 EDGs.
    Consequently, operation of either unit in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Question 3  Does the proposed amendment involve a significant 
reduction in the margin of safety?
    The proposed amendment is designed to provide flexibility to 
schedule and perform required surveillance activities. Surveillance 
intervals or operating requirements are not changed by the proposal; 
only the method of fuel oil storage on a temporary basis for a single 
operable EDG is addressed. The proposed change will not alter the basis 
for any Technical Specification that is related to the establishment 
of, or maintenance of, a nuclear safety margin.
    Consequently, operation of Turkey Point Units 3 and 4 in accordance 
with this proposed amendment would not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: January 9, 1998.
    Description of amendment request: The licensee proposed to change 
the Technical Specifications (TS) to allow the use of 
ZIRLOtm fuel rod clad material.
    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:
    Question 1  Does the proposed license amendment involve a 
significant increase in the probability or consequences of an accident 
previously evaluated?
    Implementation of ZIRLOtm fuel rod cladding will have no 
impact on the probability or consequences of any Design Basis Event 
occurrences which were previously evaluated. The determination that 
fuel design limits are met will continue to be performed using NRC 
approved fuel performance analysis methodology. Changing to 
ZIRLOtm fuel rod cladding poses no significant increase in 
the probability or consequences of any accident previously evaluated.
    No new performance requirements are being imposed on any system or 
component in order to support implementation of ZIRLOtm fuel 
rod cladding. Since the LOCA and Non-LOCA analysis results will remain 
within design limits, the inputs to the radiation dose analysis do not 
change. Therefore, the consequences to the public resulting from any 
accident previously evaluated in the Updated Final Safety Analysis 
Report (UFSAR) is not increased.
    Fuel rod design criteria will be evaluated every cycle to ensure 
proper compliance with fuel rod design limits and therefore the UFSAR. 
The evaluation of the fuel design against fuel design limits will be 
performed in accordance with 10 CFR 50.59, which ensures that the 
reload will not involve an increase in the probability or consequence 
of an accident previously evaluated.
    Question 2  Does the proposed license amendment create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    Implementation of ZIRLOtm fuel rod cladding will have no 
impact, nor does it contribute in any way to the probability or 
consequences of an accident.
    No new accident scenarios, failure mechanisms or limiting single 
failures are introduced as a result of using ZIRLOtm fuel 
rod cladding. The institution of ZIRLOtm fuel rod cladding 
will have no adverse effect on, and does not challenge the performance 
of, any safety related system.
    The determination that the fuel rod design limits are met will be 
performed using NRC approved methodology. Therefore, the proposed 
amendment does not in any way create the possibility of a new or 
different kind of accident from any accident previously evaluated.

[[Page 9606]]

    Question 3  Does the proposed amendment involve a significant 
reduction in the margin of safety?
    The margin of safety is not affected by the implementation of 
ZIRLOtm fuel rod cladding. Use of ZIRLOtm fuel 
rod cladding has been approved by the NRC and does not constitute a 
significant reduction in the margin of safety.
    The margin of safety provided in the fuel design limits is 
acceptable and will be maintained and not reduced.
    In addition, each future reload will involve a 10 CFR 50.59 review 
to assure that operation of the units within the cycle specific limits 
will not involve a reduction in the margin of safety. Therefore, the 
proposed amendment does not significantly reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: January 22, 1998.
    Description of amendment request: The amendment would incorporate 
the proposed revision into Chapter 9 of the Millstone Unit 3 Final 
Safety Analysis Report. The proposed revision to the Millstone Unit 3 
licensing basis would accept the existing use of epoxy coatings on 
safety-related components.
    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:
    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10CFR50.92(c) are not satisfied. The 
proposed revision does not involve [an] SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    Past experience indicates that failure of previous ARCOR 
applications may have degraded the performance of SWS [service water 
system] heat exchangers within one train, but there is no indication 
that failure of multiple heat exchangers on both trains is feasible. 
Furthermore, the likelihood of ARCOR material being released has been 
reduced by improving the application procedure and performing 
destructive testing to detect disbondment. In addition, the completion 
of normal heat exchanger performance surveillance's and periodic visual 
inspections minimizes the potential for disbonded ARCOR to degrade SWS 
components.
    Therefore, the presence of ARCOR coating material within the SWS 
does not involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The application of ARCOR material may lead to the degradation of 
SWS heat exchangers. However, multiple ARCOR application failures 
occurring simultaneously either instantaneously or gradually resulting 
in failure of all SWS heat exchangers in both trains is not considered 
feasible. An instantaneous failure is discounted by analysis which 
concludes that normal system operations are more likely to cause the 
release of degraded ARCOR than what might be expected following a 
seismic event. Gradual degradation is not expected since normal SWS 
heat exchanger performance surveillance's will identify heat exchanger 
tubesheet fouling and thus, provide early detection of coating 
failures. Therefore, the use of ARCOR coating material within the SWS 
does not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Although the gradual release of ARCOR material creates the 
potential to simultaneously degrade the performance of mitigating 
equipment in both trains of safety systems, it is determined to be 
unrealistic due to normal heat exchanger performance surveillance's. 
These surveillance's are expected to identify heat exchanger tubesheet 
fouling and provide early detection and mitigation of a problem with 
the pipe coatings. Therefore, the application of ARCOR coating within 
the SWS does not involve a significant reduction in the margin of 
safety.
    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: July 26, 1996, as supplemented September 
5 and December 4, 1997.
    Description of amendment request: The proposed amendment would, as 
part of the licensee's power rerate program, increase the maximum power 
level to 1775 megawatts thermal (MWt). This change is approximately 6.3 
percent above the current maximum power level of 1670 MWt.
    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 will not involve a significant increase 
In the probability or consequences of an accident previously evaluated.
    The probability of occurrence and consequences of an [accident] 
previously evaluated have been evaluated for MNGP [Monticello Nuclear 
Generating Plant] Power Rerate. This evaluation has concluded that MNGP 
Power Rerate will not involve a significant increase in the probability 
of occurrence or consequences of previously evaluated accidents.

1. Evaluation of Accident Consequences

(a) ECCS-LOCA Analysis
    The Emergency Core Cooling System Loss of Coolant Accident (ECCS-
LOCA)

[[Page 9607]]

performance analysis has been evaluated for MNGP Power Rerate using 
methodology which has been approved by the NRC for LOCA 10CFR50.46 
analyses [requirements]. The current ECCS performance requirements were 
used in the power rerate analysis; no further parameter relaxations 
were included in the analysis. The ECCS-LOCA analysis was performed for 
MNGP Power Rerate for the existing licensed rated thermal power and at 
a bounding thermal power level of 1880 MWt that is approximately 6% 
greater than the proposed power rerate to 1775 MWt [megawatts thermal]. 
In addition, the bounding thermal power level was increased by an 
additional 2% in accordance with regulatory guidance. The licensing 
peak clad temperature for the bounding analyzed thermal power level 
remains below the 10CFR50.46 required limit of 2,200'F. Therefore the 
analysis demonstrates that MNGP will continue to comply with 10CFR50.46 
and 10CFR50, Appendix K at rerated conditions thus the consequences of 
a LOCA is not significantly increased for the proposed power rerate.
(b) Abnormal Operating Transient Analysis
    An evaluation of the Updated Safety Analysis Report (USAR) and 
reload transients has been performed for MNGP Power Rerate to 
demonstrate that the proposed power rerate has no adverse effect on 
plant safety. This evaluation was performed for a power level of 1775 
MWt, with the exception that certain event evaluations were performed 
at 102% of the rerate power level. The transient analysis performed to 
demonstrate the acceptability of MNGP Power Rerate used the NRC 
approved methods identified in the MNGP Technical Specifications.
    The limiting transient events at the power rerate conditions have 
been analyzed. This includes all events that establish the core thermal 
operating limits and the events that bound other transient acceptance 
criteria. These limiting transients were benchmarked against the 
existing rated thermal power level by performance of the event analysis 
at both the proposed rerate power level and the existing rated power 
level. In addition, an expanded group of transient events was evaluated 
to confirm that these events were less severe with the power rerate 
than the most limiting transients. The events included in the expanded 
group of transient events were chosen based on those events which have 
been demonstrated to be sensitive to initial power level. This 
evaluation confirmed that the existing set of limiting transient events 
remains valid for MNGP Power Rerate. The evaluation was performed for a 
representative core and demonstrated the overall capability to meet all 
transient safety criteria for the power rerate. Cycle specific analysis 
will continue to be performed for each fuel reload to demonstrate 
compliance with the applicable transient criteria and to establish 
cycle specific operating limits.
    The results of the evaluation of transients demonstrate that the 
power rerate can be accomplished without a significant increase in the 
consequences of the transients evaluated. The fuel thermal-mechanical 
limits at the power rerate conditions are within the specific design 
criteria for the GE [General Electric] fuels currently loaded in the 
MNGP core. Also, the power-dependent and flow-dependent MCPR [minimum 
critical power ratio] and Maximum Average Planar Linear Heat Generation 
Rate (MAPLHGR) methods developed as part of the core performance 
improvement program remain applicable to rerate conditions. The 
transient event evaluation confirmed that MNGP Power Rerate has no 
significant effect on the power-dependent and flow-dependent MCPR and 
MAPLHGR limits. The peak reactor pressure vessel bottom head pressure 
remains within the ASME [American Society of Mechanical Engineers] 
requirement for reactor pressure vessel overpressure protection.
    The effects of plant transients were evaluated by assessing a 
number of disturbances of process variables and malfunctions or 
failures of equipment consistent with USAR. The transient events were 
evaluated against the Safety Limit Minimum Critical Power Ratio, 
(SLMCPR). The SLMCPR is determined using NRC-approved methods. The 
limiting transient events are slightly more severe when initiated from 
the rerate power level. The power rerate transient evaluation results 
show a slightly more limiting event initial CPR [critical power ratio] 
(less than or equal to 0.02) than that initiated from the present rated 
power level for the near limiting transients. However, for the most 
limiting transient, the evaluation of a representative core showed that 
no change is required to the Operating Limit MCPR for the power rerate 
and that the integrity of the SLMCPR is maintained. The margin of 
safety established by the SLMCPR is not affected and the event 
consequences are not significantly affected by the proposed power 
rerate to 1775 MWt. Cycle specific analysis will continue to be 
performed for each fuel reload to demonstrate compliance with the 
applicable transient criteria and to establish cycle specific operating 
limits.
    The results demonstrate that the MNGP core thermal power output can 
be safely increased to the power rerate level without significant 
effect on the consequences of previously evaluated postulated transient 
events. The results of the rerate transient analysis are summarized as 
follows.
(1) Events Resulting in a Nuclear System Pressure Increase
(a) Main Generator Load Rejection with No Steam Bypass
    At rerated conditions, the fuel transient thermal and mechanical 
overpower results remain below the NRC accepted design criteria.
(b) Main Turbine Trip with No Steam Bypass
    At rerate conditions, the fuel transient thermal and mechanical 
overpower results remain below the NRC accepted design criteria.
(c) Main Steam Isolation Valve Closure, Flux Scram
    The peak reactor pressure vessel bottom head pressure for rerate 
conditions is slightly higher than the reactor pressure vessel bottom 
head pressure at current conditions. However, the resultant pressure is 
still below the ASME overpressure limit of 1,375 psig [pounds per 
square inch].
(d) Slow Closure of a Single Turbine Control Valve
    The results of this transient for the power rerate remain non-
limiting as compared with other more severe pressurization events.
(2) Event Resulting in a Reactor Vessel Water Temperature Decrease
(a) Feedwater Controller Failure-Maximum Demand
    The delta CPR calculated for this event at rerate conditions is 
about 0.01 higher than the corresponding value for the current rated 
power when the impact of the new condensate pumps is factored in. The 
trend for the Feedwater Controller Failure-Maximum Demand event is 
consistent with the analysis for the current rated power. The fuel 
thermal margin results are within the acceptable limits for the fuel 
types analyzed.
(b) Loss of Feedwater Heating
    This event at the rerate conditions remains significantly less than 
the cycle operating MCPR limit. The results at low core flow conditions 
are actually slightly higher than for the high core

[[Page 9608]]

flow condition because of increased inlet coolant subcooling into the 
reactor core. The calculated thermal and mechanical overpower limits at 
the power rerate conditions for this event also meet the fuel design 
criteria.
(c) Inadvertent HPCI [high-pressure coolant injection] Actuation
    For the limiting condition analyzed, both the high water level 
setpoint and the high reactor pressure vessel steam dome pressure scram 
setpoints are not reached. Based on the peak average fuel surface heat 
flux results, the HPCI actuation event will be bounded by the limiting 
pressurization event with respect to delta Critical Power Ratio 
([delta] CPR) considerations. In addition, the fuel transient thermal 
and mechanical overpower limits remain within the NRC accepted design 
values.
(3) Event Resulting in a Positive Reactivity Insertion
(a) Rod Withdrawal Error (RWE)
    The current Rod Block Monitor (RBM) system for MNGP with power 
dependent setpoints was analyzed for the rod withdrawal error event at 
the power rerate conditions using a statistical approach consistent 
with NRC approved methods. The analysis concluded that the transient is 
slightly more severe with a greater delta Critical Power Ratio ([delta] 
CPR) from the initial most limiting CPR. However, the fuel and 
mechanical overpower results remain within the NRC accepted design 
criteria.
(4) Event Resulting in a Reactor Vessel Coolant Inventory Decrease
(a) Pressure Regulator Failure to Full Open
    The results of this transient for the power rerate remain non-
limiting as compared with other more severe pressurization events.
(b) Loss of Feedwater Flow
    This transient event does not pose any direct threat to the fuel in 
terms of a power increase from the initial conditions. Water level 
declines rapidly and a low level causes a reactor scram. The closure of 
the main steam isolation valves and the actuation of High Pressure 
Coolant Injection and Reactor Core Isolation Cooling terminate the 
event. This event was included in the power rerate evaluation to 
provide assurance that sufficient water makeup capability is available 
to keep the core covered when all normal feedwater is lost. The generic 
analysis performed in support of the extended power uprate program 
shows that at the power rerate conditions a large amount of water 
remains above the top of the active fuel. These sequences of events do 
not require any new operator actions or shorter operator response 
times. Therefore, the operator actions for the event do not 
significantly change for the power rerate.
(5) Event Resulting in a Core Coolant Flow Decrease
(a) Recirculation Pump Seizure
    The recirculation pump seizure assumes instantaneous stoppage of 
the pump motor shaft of one recirculation pump. As a result, the core 
flow decreases rapidly. The heat flux decline lags core power and flow 
and could result in a degradation of core heat transfer. At the power 
rerate conditions, the transient results confirmed that the 
consequences of the pump seizure event remain non-limiting.
(6) Event Resulting in a Core Coolant Flow Increase
(a) Recirculation Flow Controller Failure Increasing Flow
    The results of this transient for the power rerate remain non-
limiting as compared with other more severe pressurization events.
(c) Design Basis Accident Challenges to the Containment
    The primary containment response to the limiting design basis 
accident was evaluated for a bounding reactor power level approximately 
6% greater than the proposed power rerate to 1775 MWt. In addition, the 
bounding reactor power level was increased by an additional 2% in 
accordance with regulatory guidance. The effect of the power rerate on 
the short term containment response (peak values) as well as the long 
term containment response for containment pressure and temperature 
confirms the suitability of the plant for operation at the bounding 
power level, thus the proposed power rerate to 1775 MWt is acceptable. 
Factors of safety provided in the ASME Code are maintained and safety 
margin is not affected for the power rerate to 1775 MWt.
    Short-term containment response analyses were performed for the 
limiting design basis LOCA consisting of a double-ended guillotine 
break of a recirculation suction line, to demonstrate that operation at 
a bounding reactor power will not result in exceeding the containment 
design limits. This limiting design basis LOCA event results in the 
highest short-term containment pressures and dynamic loads. The 
analysis determined that for a bounding reactor power the maximum 
drywell pressure values are bounded by the current USAR analysis value 
and by the containment design pressure. The power rerate to 1775 MWt 
has no adverse effect on the containment structural design pressure.
    Because there will be more residual heat with increased thermal 
power, the containment long term response will have slightly higher 
temperatures. Long term suppression chamber temperatures remain within 
the design temperature of the structure, thus factors of safety 
provided in the ASME code are maintained and safety margin is not 
affected. Analysis confirmed that ECCS pump NPSH is adequate for this 
temperature response. It was confirmed that the long term response does 
not adversely affect the containment structure or the environmental 
qualification (EQ) of equipment located in the drywell or suppression 
chamber room. The drywell long term temperature response is not 
adversely affected for a bounding reactor power. An analytical power 
level of 1880 MWt bounds the decay heat associated with the 1775 MWt 
power level with a one sided confidence interval of 95%. The 
containment long term response is therefore acceptable for the power 
rerate to 1775 MWt.
    The impact of a reactor power increase on the containment dynamic 
loads have been determined, evaluated and found to have no adverse 
effects for conditions which well bound the proposed power rerate. Thus 
the containment dynamic loads were found to be acceptable for the power 
rerate to 1775 MWt.
    The MNGP Power Rerate evaluation of the primary containment 
response to the design basis accident confirmed that the power rerate 
does not result in a significant increase in consequences for a 
bounding reactor power approximately 6% greater than the proposed power 
rerate to 1775 MWt.
(d) Radiological Consequences of Design Basis Accidents
    For MNGP Power Rerate, the radiological consequences of the 
limiting design basis accidents were re-evaluated. These evaluations 
included the effect of the power rerate on the radiological 
consequences of accidents presented in USAR Section 14.7.
    This evaluation was performed using inputs and evaluation 
techniques consistent with the current regulatory guidance, the current 
GE analysis methods, and the appropriate plant design basis. The inputs 
and analysis methods used for MNGP Power Rerate differ from those 
utilized in the current licensing basis evaluation presented in

[[Page 9609]]

the USAR and the AEC [Atomic Energy Commission] safety evaluation 
supporting plant initial licensing. The MNGP Power Rerate evaluations 
used the more contemporary staff approved methods. The inputs used in 
the MNGP Power Rerate evaluation provide a conservative assessment of 
the potential radiological consequences. The conclusions of these 
evaluations are consistent with the original licensing basis 
evaluations. The radiological consequences of the limiting design basis 
accidents remain well within 10CFR100 guidelines for a bounding thermal 
power approximately 6% greater than the proposed power rerate of 1775 
MWt. In addition the bounding thermal power level was increased by an 
additional 2% in accordance with regulatory guidance.
    To conservatively analyze the change in consequences, the 
evaluation of radiological consequences using the analysis inputs and 
methods was performed for the existing licensed rated thermal power and 
a thermal power bounding the proposed power rerate. This provides a 
conservative bounding change in consequences for the requested power 
rerate to 1775 MWt.
    The MNGP Power Rerate evaluation of the radiological consequences 
of design basis accidents confirmed that the power rerate does not 
result in a significant increase in consequences for a bounding power 
level approximately 6% greater than the proposed power rerate. The 
results remain below the 10CFR100 guideline values as well as the 
licensing basis established in the March 18, 1970 AEC safety 
evaluation. Therefore, the postulated radiological consequences do not 
represent a significant change in accident consequences and are clearly 
within the regulatory guidelines for the proposed power rerate to 1775 
MWt.
(e) Other Evaluations
(1) Performance Improvements
    The MNGP Power Rerate safety analysis has been performed taking 
into account the implementation of the following previously approved 
special operational features.
(a) Maximum Extended Load Line Limit/Increase Core Flow (MELLL/ICF)
    The safety analysis for rerate conditions shows that the extended 
operating domain as analyzed by MELLL/ICF remains valid for the power 
rerate conditions.
(b) Average Power Range Monitor/Rod Block Monitor Technical 
Specification (ARTS) Improvements
    The safety analysis for rerate conditions shows that the ARTS 
improvements remain valid for the power rerate conditions.
(c) Single Loop Operation (SLO)
    The safety analysis for rerate conditions shows that the single 
loop operating mode remains valid for the power rerate conditions. The 
MELLLA trip setpoints determined for two-loop operation were confirmed 
to be acceptable for single loop operation with a correction applied to 
account for the actual effective drive flow applied when operating in 
single loop. The single loop settings have been conservatively 
established to be consistent with the two loop settings while ensuring 
the appropriate corrections are applied to the MAPLHGR and the 
operating limit MCPR to account for single loop operation.
(2) Effect of Power Rerate on Support Systems
    An evaluation was performed to address the effect of MNGP Power 
Rerate on accident mitigation features, structures, systems, and 
components within the balance of plant. The results are as follows:
    Auxiliary systems such as, building heating, Ventilation and Air 
Conditioning (HVAC) systems, reactor building closed cooling water, 
service water and emergency service water, spent fuel pool cooling, 
process auxiliaries such as instrument air and makeup water and the 
post-accident sampling system were confirmed to operate acceptably 
under normal and accident conditions at rerate conditions.
    The secondary containment and standby gas treatment system were 
confirmed to be able to adequately contain, process, and control the 
release of normal and post-accident levels of radioactivity at rerate 
conditions.
    Instrumentation was reviewed and confirmed to be capable of 
performing its control and monitoring functions under rerate 
conditions. As required, analyses were performed to determine the need 
for setpoint changes for various functions (e.g., APRM [average power 
range monitor] neutron flux scram setpoints). In general, setpoints are 
to be changed only to maintain adequate difference between plant 
operating parameters and trip setpoints, while ensuring safety 
performance is demonstrated. The revised setpoints have been 
established using the NRC reviewed methodology as guidance.
    Electric power systems including the turbine generator and 
switchgear components were verified as being capable of providing the 
electrical load as a result of the rerate power levels. An evaluation 
of the auxiliary power system for the power rerate conditions confirmed 
that the system has sufficient capacity with the changes identified in 
Exhibit I [of the 12/4/97 submittal] to support all required loads for 
safe shutdown, to maintain a safe shutdown condition, and to operate 
the required engineered safeguards equipment following postulated 
accidents. No safety-related electrical loads were affected which would 
adversely impact the emergency diesel generators.
    Piping systems were evaluated for the effect of operation at higher 
power levels, including transient loading. The evaluation confirmed 
that, with few exceptions, piping and supports are adequate to 
accommodate the increased loading resulting from operation at rerate 
power conditions. In a few cases, piping supports will be modified to 
accept higher forces due to rerate conditions.
    The effect of rerate conditions on high energy line break (HELB) 
was evaluated. The evaluation confirmed structures, systems, and 
components important to safety are capable of accommodating the effects 
of jet impingement and blowdown forces and the environmental effects 
resulting from HELB events at rerate conditions.
    Control room habitability was evaluated. With the implementation of 
minor hardware and non-hardware changes to the control room ventilation 
system, Post-accident Control Room and Technical Support Center doses 
at rerate conditions were confirmed to be within the guidelines of 
General Design Criterion 19 of 10CFR50, Appendix A.
    The environmental qualification of equipment important to safety 
was evaluated for the effect on normal and accident operating 
conditions at rerate power levels. The equipment remains qualified for 
the new conditions. Minor adjustments will reflect some changes to 
maintenance frequencies. The preventative maintenance program will 
continue to provide for equipment maintenance or replacement to ensure 
equipment environmental qualification at rerate power conditions.
(3) Effect on Special Events
    The consequences of special events (i.e., ATWS [anticipated 
transient without scram], 10CFR50, Appendix R, and Station Blackout) 
remain within NRC accepted criteria for rerate conditions. Concurrent 
malfunctions assumed to occur during accidents have

[[Page 9610]]

been accounted for in the safety analyses for rerate conditions. The 
consequences of these equipment malfunctions does not change with 
implementation of the MNGP Power Rerate program. The generic ATWS 
analysis for operation at rerate conditions is being revised. The 
revision is not expected to affect MNGP compliance with NRC acceptance 
criteria.
(f) Conclusion
    The evaluation of the Emergency Core Cooling System performance has 
demonstrated the criteria of 10CFR50.46 are satisfied, thus the margin 
of safety established by the criteria is maintained. The analysis 
demonstrated that the ECCS will function with the most limiting single 
failure to mitigate the consequences of the accidents and maintain fuel 
integrity. The system will continue to perform as required under rerate 
conditions to mitigate the consequences of accidents and thus the power 
rerate does not adversely affect ECCS performance in a manner to 
increase the severity of consequences. Challenges to the containment 
have been evaluated and the integrity of the fission product barrier 
has been confirmed. The radiological consequences of design basis 
accidents have been evaluated and it was found that the effect of the 
proposed power rerate on postulated radiological consequences does not 
result in a significant increase in accident consequences. These 
evaluations have been performed for a bounding reactor power 
approximately 6% greater than the proposed power rerate. In addition 
the bounding reactor power level was increased by an additional 2% in 
accordance with regulatory guidance. Thus the evaluations provide 
conservative bounding results for the proposed power rerate to 1775 MWt 
and demonstrate that the proposed power rerate does not result in 
significant increase in accident consequences.
    The abnormal transients have been analyzed under the power rerate 
conditions, and the analysis has confirmed that the power rerate to 
1775 MWt has only a minor effect on the minimum critical power ratio 
and that no change to the safety limit critical power ratio results, 
thus the margin of safety as assured by the safety limit critical power 
ratio is maintained. The effect of the power rerate on the consequences 
of abnormal transients which result from potential component 
malfunctions has been shown to be acceptable, thus the power rerate 
does not result in a significant increase in transient event 
consequences.
    The spectrum of analyzed postulated accidents and transients has 
been investigated, and has been determined to meet the current 
regulatory criteria for the MNGP at rerate conditions. In the area of 
core design, the fuel operating limits will still be met at the rerate 
power level, and fuel reload analyses will show plant transients meet 
the criteria accepted by the NRC as specified in the plant Technical 
Specifications. The evaluation of transient and accident consequences 
was performed consistent with the proposed changes to the plant 
Technical Specifications. Therefore, the proposed Operating License and 
Technical Specification changes will not cause a significant increase 
in the consequences of an accident previously evaluated for the 
Monticello plant.

2. Evaluation of the Probability of Previously Evaluated Accidents

    The proposed power rerate imposes only minor increases in the plant 
operating conditions. No changes are required to the rated core flow, 
rated reactor pressure, or turbine throttle pressure. The power rerate 
will result in moderate flow increases in those system[s] associated 
with the turbine cycle (i.e., condensate, feedwater, main steam, etc.). 
For MNGP Power Rerate, the small increase in operating temperatures for 
balance of plant support systems has no significant effect on LOCA or 
other accident probabilities.
     The increase in flow rates in balance of plant systems is 
addressed by compliance with NRC Generic Letter 89-08, ``Erosion/
Corrosion in Piping.'' The MNGP Power Rerate evaluations have confirmed 
that the power rerate has no significant effect on flow induced 
erosion/corrosion. The worst case limiting feedwater and main steam 
piping flow increases were evaluated to be approximately proportional 
to the power increase. The affected systems are currently monitored by 
the MNGP Erosion/Corrosion program. Continued monitoring of the systems 
provides a high level of confidence in the integrity of potentially 
susceptible high energy piping systems.
    The occurrence frequency of accident precursors and transients 
[has] been addressed when required by applying the guidance of NRC 
reviewed setpoint methodology to insure that acceptable trip avoidance 
is provided during operational transients subsequent to implementation 
of rerate. The setpoint evaluation has confirmed that MNGP Power Rerate 
does not result in any increase in challenges to the plant protective 
instrumentation.
    Plant systems, components, and structures have been verified to be 
capable of performing their intended functions under rerate conditions 
with a few minor exceptions. Where necessary, some components will be 
modified prior to implementation of the MNGP Power Rerate Program to 
accommodate the revised operating conditions (e.g., a limited number of 
pipe supports changes, instrumentation setpoint changes, control room 
habitability improvements). MNGP Power Rerate does not significantly 
affect the reliability of plant equipment. Where reliability effects 
have been identified, modifications and administrative controls will be 
implemented prior to the power rerate to adequately compensate. No new 
components or system interactions that could lead to an increase in 
accident probability are created due to the power rerate.
    The probability (i.e., frequency of occurrence) of design basis 
accidents occurring is not affected by the increased power level, as 
the applicable criteria established for plant equipment (e.g., ANSI 
Standard B31.1, ASME Code,) will still be followed as the plant is 
operated at the rerate power level. The MNGP Power Rerate analysis 
basis assures that the power dependent margin prescribed by the Code of 
Federal Regulations (CFR) will be maintained by meeting the appropriate 
regulatory criteria. Similarly, factors of safety specified by 
application of the Code design rules have been demonstrated to be 
maintained, as have other margin-assuring acceptance criteria used to 
judge the acceptability of the plant. Reactor scram setpoints as 
established are such that there is no significant increase in scram 
frequency due to rerate conditions. No new challenges to safety-related 
equipment will result from the power rerate. Therefore, the proposed 
Operating License and Technical Specifications changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    B. The proposed Operating License changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The basic Boiling Water Reactor configuration, operation and event 
response is unchanged by the power rerate. Analysis of transient events 
has confirmed that the same transients remain limiting and that no 
transient events result in a new sequence of events which could lead to 
a new accident scenario. The MNGP Power Rerate analyses confirmed that 
the accident progression is basically unchanged by the power rerate.

[[Page 9611]]

    An increase in power level will not create a new fission product 
release path, or result in a new fission product barrier failure mode. 
The same fission product barriers such as the fuel cladding, the 
reactor coolant pressure boundary and the reactor containment, remain 
in place. Fuel rod cladding integrity is ensured by operating within 
thermal, mechanical, and exposure design limits and is demonstrated by 
the MNGP Power Rerate transient analysis and accident analysis. 
Similarly, analysis of the reactor coolant pressure boundary and 
primary containment have demonstrated that the power rerate has no 
adverse effect on these fission product barriers. The proposed changes 
to the plant Technical Specifications to support the power rerate 
implementation are consistent with the MNGP Power Rerate analyses and 
assure transient and accident mitigation capability in compliance with 
regulatory requirements.
    The effect of MNGP Power Rerate on plant equipment has been 
evaluated. No new operating mode, safety-related equipment lineup, 
accident scenario, or equipment failure mode resulting from the power 
rerate was identified. The full spectrum of accident considerations 
defined in the USAR have been evaluated and no new or different kind of 
accident resulting from the power rerate has been identified. MNGP 
Power Rerate uses already developed technology and applies it within 
the capabilities of already existing plant equipment in accordance with 
presently existing regulatory criteria which includes accepted codes, 
standards, and methods. GE has designed BWRs of higher power levels 
than the rerate power of any of the currently operating BWR fleet and 
no new power dependent accidents have been identified. In addition, 
MNGP Power Rerate does not create any new sequence of events or failure 
modes that lead to a new type of accident.
    All actions to ensure that safety-related structures, systems, and 
components will remain within their design allowable values and ensure 
they can perform their intended functions under rerate conditions will 
be taken prior to implementation of the power rerate. MNGP Power Rerate 
does not increase challenges to or create any new challenge to safety-
related equipment or other equipment whose failure could cause an 
accident. Plant modifications required to support implementation of 
MNGP Power Rerate will be made to existing systems (e.g., a limited 
number of pipe supports, instrumentation setpoints, control room 
habitability improvements), rather than by adding new systems of a 
different design which might introduce new failure modes or accident 
sequences. The Technical Specification changes required to implement 
the power rerate require little change to the plant's configuration, 
and all changes have been evaluated and are acceptable.
    Therefore, the proposed Operating License and Technical 
Specification changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    C. The proposed Operating License changes do not involve a 
significant reduction in a margin of safety.
    The accident analysis, as well as a majority of the plant specific 
evaluations performed in support of MNGP Power Rerate have been 
performed assuming a bounding steady state power level 112.6% of the 
existing licensed limit of 1670 MWt, and approximately 6% above the 
licensed maximum thermal power level of 1775 MWt proposed by MNGP Power 
Rerate. In addition, the bounding reactor power level was increased by 
an additional 2% in accordance with regulatory guidance when applicable 
for the evaluation of accidents and transients. For plant conditions 
associated with a bounding analysis power level, the analyses 
demonstrated operating margin to criteria establishing margins of 
safety, thus additional operating margin is demonstrated and assured 
for the proposed power rerate to 1775 MWt and added confidence is 
established in the integrity of criteria establishing margin to safety.
    The cycle specific transient analysis, as well as the analysis to 
establish plant instrumentation set points have been performed assuming 
a plant steady state power level of 1775 MWt. This analysis approach 
was taken in order to demonstrate safety and equipment margins while 
ensuring appropriate cycle specific operating limits. The evaluation of 
transient events and instrument setpoints demonstrated operating margin 
to criteria establishing margins of safety for the proposed power 
rerate conditions.
    The MNGP Power Rerate analysis basis assures that the power 
dependent safety margin assuring criteria prescribed by the Code of 
Federal Regulations (CFR) will be maintained by meeting the appropriate 
regulatory criteria. Similarly, factors of safety specified by 
application of the code design rules have been maintained, as have 
other margin-assuring acceptance criteria used to judge the 
acceptability of the plant.

1. Fuel Thermal Limits

    No change is required in the basic fuel design to achieve the 
rerate power levels or to maintain the margins as discussed above. No 
increase in the allowable peak bundle power is requested for the power 
rerate. The abnormal transients have been evaluated under the power 
rerate conditions for a representative core configuration. The analysis 
has confirmed that the power rerate has no adverse effect on the 
operating limit Minimum Critical Power Ratio (MCPR) and that no change 
to the safety limit MCPR results, thus the margin of safety as assured 
by the safety limit MCPR is maintained. The fuel operating limits such 
as Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and the 
operating limit MCPR will still be met at the rerate power level. The 
MNGP Power Rerate analyses have confirmed the acceptability of these 
operating limits for the power rerate without an adverse effect on 
margins to safety. Cycle specific analysis will continue to be 
performed for each fuel reload to demonstrate compliance with the 
applicable transient criteria and to establish cycle specific operating 
limits.

2. Design Basis Accidents Challenges to Fuel

    The evaluation of the Emergency Core Cooling System performance has 
demonstrated the criteria of 10CFR50.46 are satisfied, thus the margin 
of safety established by the criteria is maintained. This evaluation 
was performed for a bounding reactor power level approximately 6% 
greater than the proposed power rerate. In addition the bounding 
reactor power level was increased by an additional 2% in accordance 
with regulatory guidance. The analysis demonstrates that MNGP will 
continue to comply [with] the 10 CFR 50.46 at the rerate conditions and 
that the margin of safety established by the regulation is maintained 
for the proposed power rerate.

3. Design Basis Accident Challenges to Containment

    The primary containment response to the limiting design basis 
accident was evaluated for a bounding reactor power level approximately 
6% greater than the proposed power rerate to 1775 MWt. In addition, the 
bounding reactor power level was increased by an additional 2% in 
accordance with regulatory guidance. The effect of the power rerate on 
the short term containment response (peak values) as well as the long 
term containment response for containment pressure and temperature 
confirms the

[[Page 9612]]

suitability of the plant for operation at the bounding power level, 
thus the proposed power rerate to 1775 MWt is acceptable. Factors of 
safety provided in the ASME Code are maintained and safety margin is 
not affected for the power rerate to 1775 MWt.
    Short-term containment response analyses were performed for the 
limiting design basis LOCA consisting of a double-ended guillotine 
break of a recirculation suction line, to demonstrate that operation at 
a bounding reactor power will not result in exceeding the containment 
design limits. The analysis determined that for a bounding reactor 
power the maximum drywell pressure values are bounded by the current 
USAR analysis value and by the containment design pressure. The power 
rerate to 1775 MWt has no adverse effect on the containment structural 
design pressure.
    Long term suppression chamber temperatures remain within the design 
temperature of the structure, thus factors of safety provided in the 
ASME code are maintained and safety margin is not affected. An 
analytical power level of 1880 MWt bounds the decay heat associated 
with the 1775 MWt power level with a one sided confidence interval of 
95%. Analysis confirmed that ECCS pump NPSH is not adversely affected 
with this temperature response. It was confirmed that the long term 
response does not significantly affect the containment structure or the 
environmental qualification (EQ) of equipment located in the drywell or 
suppression chamber room.
    The impact of a reactor power increase on the containment dynamic 
loads [has] been determined, evaluated and found to have no adverse 
effects for conditions which well bound the proposed power rerate. Thus 
the containment dynamic loads were found to be acceptable for the power 
rerate to 1775 MWt.
    The MNGP Power Rerate evaluation of the primary containment 
response to the design basis accident confirmed that the power rerate 
does not result in a reduction in margins of safety for a bounding 
reactor power approximately 6% greater than the proposed power rerate 
to 1775 MWt.

4. Design Basis Accident Radiological Consequences

    The Updated Safety Analysis Report (USAR) provides the radiological 
consequences for each of the design basis accidents. The magnitude of 
the potential consequences is dependent upon the quantity of fission 
products released to the environment, the atmospheric dispersion 
factors and the dose exposure pathways. For power rerate, the 
atmospheric dispersion factors and the dose exposure pathways do not 
change. Therefore, the only factor which will influence the magnitude 
of the consequences is the quantity of activity released to the 
environment. This quantity is a product of the activity released from 
the core and the transport mechanisms between the core and the effluent 
release point.
    The radiological consequences of design basis accidents have been 
evaluated, and it was found that the consequences did not result in a 
significant increase in consequences for a bounding reactor power level 
approximately 6% greater than the proposed power rerate. In addition, 
the bounding reactor power level was increased by an additional 2% in 
accordance with regulatory guidance. The results remain below the 
10CFR100 guideline values as well as the licensing basis established in 
the March 18, 1970 AEC safety evaluation. Therefore, the postulated 
radiological consequences are clearly within the regulatory guidelines 
and all radiological safety margins are maintained for the power rerate 
to 1775 MWt.

5. Transient Evaluations

    The effects of plant transients were evaluated by assessing a 
number of disturbances of process variables and malfunctions or 
failures of equipment consistent with USAR. The transient events were 
evaluated against the Safety Limit Minimum Critical Power Ratio, 
(SLMCPR). The SLMCPR is determined using NRC-approved methods. The 
Power Rerate transient analyses were performed using the approved 
methodology specified in the plant Technical Specifications. The 
limiting transient events are slightly more severe when initiated from 
the rerate power level. The power rerate transient evaluation results 
show a slightly more limiting transient initial CPR (less than or equal 
to 0.02) than that initiated from the present rated power level for the 
near limiting transients. However, for the most limiting transient, the 
evaluation of a representative core showed that no change is required 
to the Operating Limit MCPR for the power rerate and that the integrity 
of the SLMCPR is maintained. Cycle specific analysis will continue to 
be performed for each fuel reload to demonstrate compliance with the 
applicable transient criteria and to establish cycle specific operating 
limits.
    The fuel thermal-mechanical limits at the power rerate conditions 
are within the specific design criteria for the GE fuels currently 
loaded in the MNGP core. Also, the power-dependent and flow-dependent 
MCPR and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) 
methods developed as part of the core performance improvement program 
remain applicable to rerate conditions. The transient event evaluation 
confirmed that MNGP Power Rerate has no significant effect on the 
power-dependent and flow-dependent MCPR and MAPLHGR limits. The peak 
reactor pressure vessel bottom head pressure remains within the ASME 
requirement for reactor pressure vessel over pressure protection.
    The margin of safety established by the SLMCPR is not affected by 
the proposed power rerate to 1775 MWt.

6. Technical Specification Changes

    The Technical Specifications ensure that the plant and system 
performance parameters are maintained at the values assumed in the 
safety analysis. The Technical Specification (setpoints, trip settings, 
etc.) are selected such that the actual equipment is maintained equal 
to or conservative with respect to the inputs used in the safety 
analysis. Proper account is taken of inaccuracies introduced by 
instrument drift, instrument accuracy, and calibration accuracy. The 
Technical Specifications address equipment availability and limit 
equipment out-of-service to assure that the plant can be expected to 
have at least the complement of equipment available to deal with plant 
transients as that assumed in the safety analysis. The evaluations and 
analyses performed to demonstrate the acceptability of MNGP Power 
Rerate were performed using inputs consistent with the proposed changes 
to the plant Technical Specifications.
    The events that form the Technical Specification Bases were 
evaluated for the power rerate conditions using inputs and initial 
conditions consistent with the proposed Technical Specification 
changes. Although some changes to the Technical Specifications are 
required for the power rerate, no NRC acceptance limit will be 
exceeded. Therefore, the margins of safety assured by safety limits and 
other Technical Specification limits will be maintained. The changes to 
the Technical Specification Bases proposed by this submittal are 
consistent with the evaluations which demonstrated acceptability of the 
power rerate.

7. Conclusion

    The spectrum of postulated accidents, transients, and special 
events has been investigated and [has] been determined to meet the 
current regulatory criteria

[[Page 9613]]

for the MNGP at the power rerate conditions. In the area of core 
design, the fuel operating limits will still be met at the rerate power 
level, and fuel reload analyses will show plant transients meet the 
criteria accepted by the NRC as specified in the plant Technical 
Specifications. Challenges to fuel or ECCS performance were evaluated 
and shown to meet the criteria of 10 CFR 50.46 and 10 CFR 50, Appendix 
K. Challenges to the containment have been evaluated and the integrity 
of the fission product barrier has been confirmed. Radiological release 
events have been evaluated and shown to meet the guidelines of 10 CFR 
100. The proposed Operating License and Technical Specification changes 
are consistent with the MNGP Power Rerate evaluation performed. The 
evaluations demonstrated compliance with the margin assuring acceptance 
criteria contained in applicable codes and regulations. Therefore, the 
proposed Operating License and Technical Specifications changes 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.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Cynthia A. Carpenter

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating 
Station (LGS), Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: February 9, 1998.
    Description of amendment request: The amendment request proposes to 
revise the LGS, Unit 1 Technical Specifications (TS) Section 2.1 and 
its associated TS Basis to reflect the change in the minimum critical 
power ratio (MCPR) safety limit due to the plant-specific evaluation 
performed by General Electric Company (GE) for LGS, Unit 1, Cycle 8.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The revised MCPR Safety Limits for LGS Unit 1 Technical 
Specifications, and their use to determine cycle-specific thermal 
limits, have been calculated using NRC-approved methods (i.e., GESTAR-
II, Rev. 13) and are based on LGS Unit 1 Cycle 8 specific inputs. The 
use of these methods assures that the [safety limit for minimum 
critical power ratio] SLMCPR value is within the existing design and 
licensing basis, and cannot increase the probability or severity of an 
accident.
    The basis of the MCPR Safety Limit calculation is to ensure that 
greater than 99.9% of all fuel rods in the core avoid transition 
boiling if the limit is not violated. The MCPR Safety limit preserves 
the existing margin to transition boiling and fuel damage in the event 
of a postulated accident. The probability of fuel damage is not 
increased.
    Therefore, the proposed TS change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The MCPR Safety Limit is a Technical Specification numerical value 
designed to ensure that fuel damage from transition boiling does not 
occur as a result of the limiting postulated accident. The MCPR Safety 
Limit is not an accident initiator; therefore, it cannot create the 
possibility of any new type of accident. The new MCPR Safety Limits are 
calculated using NRC-approved methods (i.e., GESTAR-II, Rev. 13) and 
are based on LGS Unit 1, Cycle 8 specific inputs.
    Therefore, the proposed TS change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS change does not involve a significant reduction 
in the margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new MCPR Safety Limits are calculated using NRC-approved 
methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the 
current fuel design and licensing criteria, and are based on LGS Unit 1 
Cycle 8 specific inputs. The MCPR Safety Limit remains high enough to 
ensure that greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated, thereby preserving the 
fuel cladding integrity.
    Therefore, the proposed TS change does not involve a reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: John F. Stolz.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: October 14, 1997.
    Description of amendment request: The proposed changes would 
correct the maximum exposure dependent, infinite lattice multiplication 
factor for fuel bundles and provide for installation of additional 
storage racks to increase spent fuel capacity.
    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:
    Operation of the FitzPatrick plant in accordance with the proposed 
Amendment would not involve a significant hazards consideration as 
defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    A change in the infinite lattice neutron multiplication factor for 
a fuel bundle in the reactor core geometry which ensures the 
criticality limit for fuel in the spent fuel pool [SFP] geometry is met 
does not affect initiation of any accident.
    Operation in accordance with the revised limit ensures the 
consequences of previously analyzed accidents are not changed. Storage 
of additional fuel assemblies in the pool does not affect the 
probability or consequences of dropping a fuel assembly, since this 
accident is localized to a small area of the storage array. Likewise, 
addition of

[[Page 9614]]

specifications containing details presently in plant design documents 
and editorial changes do not change the probability or consequences of 
a previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
for any accident previously evaluated because:
    A change in the infinite lattice neutron multiplication factor for 
a fuel bundle in the reactor core geometry which ensures the 
criticality limit for fuel in the spent fuel pool geometry is met does 
not affect the types of reactivity accidents which may occur. Therefore 
changing the limit will not [create the possibility of] a new or 
different type of accident. Maintenance of available decay heat removal 
systems ensures that no new type of loss of cooling accident associated 
with the SFP will occur as a result of storing additional irradiated 
fuel assemblies. Likewise, addition of specifications containing 
details presently in plant design documents and editorial changes do 
not create the possibility of a new or different type of accident.
    3. Involve a significant reduction in a margin of safety because:
    The revised limit on infinite lattice neutron multiplication factor 
for a fuel bundle in the reactor core geometry ensures maintenance of 
the same margin of safety with respect to criticality as presently 
exists for storage of fuel in the SFP. Storing additional irradiated 
fuel assemblies in the pool does not affect the margin of safety with 
regard to pool cooling since sufficient heat removal systems will be 
maintained available to ensure maintenance of acceptable pool 
temperatures. Addition of specifications containing details presently 
in other design documents and editorial changes have no effect on the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: February 9, 1998.
    Description of amendment request: The proposed amendment would 
revise the Virgil C. Summer Nuclear Station Technical Specifications 
(TS) to remove emergency diesel generator (1) accelerated testing 
requirements (TS 3/4.8.1, Table 4.8-1), and (2) special reporting 
requirements (TS Surveillance Requirement 4.8.1.1.3) in accordance with 
NRC Generic Letter (GL) 94-01, ``Removal of Accelerated Testing and 
Special Reporting Requirements for Emergency Diesel Generators.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    This change will provide flexibility to structure the emergency 
diesel generator maintenance program based on the risk significance of 
the structures, systems, and components that are within the scope of 
the maintenance rule. The removal of the diesel generator accelerated 
testing is acceptable as the maintenance rule applies system and train 
specific performance criteria to monitor diesel generator performance. 
These criteria include a running availability and reliability measure. 
The performance criteria for the diesel generator reliability and 
unavailability established by the maintenance rule, and the causal 
determinations and corrective actions required for functional failures 
and/or exceeding performance criteria, is considered to be an 
acceptable method for monitoring diesel generator performance.
    As the diesel generator performance will [continue] to be assured 
by the maintenance rule, the proposed changes do not affect any of the 
initiators for an accident previously evaluated. The changes do not 
impact the diesel's design sources, operating characteristics, system 
functions, or system interrelationships. The failure mechanisms for the 
accidents previously analyzed are not affected, and no additional 
failure modes are created that could cause an accident previously 
evaluated. Since the changes are administrative in nature, and the 
diesel generator performance and reliability will continue to be 
assured by the maintenance rule, the proposed changes cannot involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. This request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change does not involve a change to the plant design 
or operation. As a result, the proposed change does not affect any of 
the parameters or conditions that could contribute to the initiation of 
any accidents. The proposed changes only affect the methods used to 
monitor and assure diesel generator performance. The performance 
criteria for both the diesel generator reliability and unavailability 
established by the maintenance rule, and the causal determinations and 
corrective actions required for functional failures and/or exceeding 
performance criteria, is considered by GL 94-01 to be an acceptable 
method for monitoring diesel generator performance.
    No SSC [structure, system, or component], method of operating, or 
system interface is altered by this change. The changes do not impact 
the diesel's design sources, operating characteristics, system 
functions, or system interrelationships. The failure mechanisms for the 
accidents are not affected, and no additional failure modes are 
created. Because the proposed changes are administrative in nature, and 
the diesel generator performance and reliability will continue to be 
assured by the maintenance rule, the proposed changes cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. This request does not involve a significant reduction in a 
margin [of] safety.
    The proposed changes only affect the methods used to monitor and 
assure diesel generator performance. The performance criteria for both 
the diesel generator reliability and unavailability established by the 
maintenance rule, and the causal determinations and corrective actions 
required for functional failures and/or exceeding performance criteria, 
is considered by GL 94-01 to be an acceptable method for monitoring 
diesel generator performance. No margin [of] safety as defined in the 
basis for any technical specification is impacted by these changes. 
This change does not impact any uncertainty in the design, 
construction, or operation of any SSC.

[[Page 9615]]

Diesel generator response to accident initiators is unchanged. No SSC, 
method of operating, or system interface is altered by this change. The 
changes do not impact the diesel's design sources, operating 
characteristics, system functions, or system interrelationships. 
Because the proposed changes are administrative in nature, and the 
diesel generator performance and reliability will continue to be 
assured by the maintenance rule, the proposed changes cannot involve a 
significant reduction in the margin [of] safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: William M. Dean.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: January 28, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Sections 6.3 and 6.12 to reflect 
the merger of the positions of Superintendent Radiation Protection and 
Superintendent Chemistry into one new position, Manager Chemistry/
Radiation Protection.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change does not involve a significant increase in the 
probability of consequences of an accident previously evaluated. These 
changes involve administrative changes to the WCNOC organization.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. This 
change is administrative in nature and does not involve a change to the 
installed plant systems or the overall operating philosophy of Wolf 
Creek Generating Station.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change does not involve a significant reduction in a 
margin of safety. This change does not involve any changes in overall 
organizational commitments and will not affect qualification 
requirements of any unit staff personnel. A position and title change 
alone does not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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.

Northern States Power Company, Docket No. 50-282, Prairie Island 
Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota

    Date of amendment request: January 15, 1998.
    Description of amendment request: The proposed amendment would 
initiate a one-time only change for Prairie Island Unit 1 Cycle 19 that 
would allow the use of the moveable incore detector system for 
measurement of the core peaking factors with less than 75% and greater 
than or equal to 50% of the detector thimbles available.
    Date of individual notice in the Federal Register: January 30, 1998 
(63 FR 4676).
    Expiration date of individual notice: March 2, 1998.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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

[[Page 9616]]

provision in 10 CFR 51.12(b) and has made a determination based on that 
assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: May 16, 1997, as supplemented 
November 14, 1997.
    Brief description of amendment: The amendment involves replacing 
the service water (SRW) heater exchangers with new plate and frame heat 
exchangers (PHEs), having increased thermal performance capability. The 
Saltwater (SW) and SRW piping configuration will be modified as 
necessary to allow proper fit-up to the new components. A flow control 
scheme to throttle saltwater flow to the heat exchangers and the 
associated bypass lines will be added. Saltwater strainers with an 
automatic flushing arrangement will be added upstream of each heat 
exchanger. The majority of the physical work associated with this 
modification is restricted to the SRW pump room. The amendment is 
partially denied to the extent that the licensee is not authorized to 
operate with one PHE secured, and removing one containment air cooler 
from service to enable the affected subsystem to remain operable while 
the one PHE is secured.
    Date of issuance: February 10, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 225.
    Facility Operating License No. DPR-53: Amendment revised the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33118).
    The November 14, 1997, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: November 6, 1997, as 
supplemented by letter dated January 28, 1998.
    Brief Description of amendments: The amendments to Technical 
Specification (TS) Limiting Conditions for Operation (LCO) 3.3.5.5, 
Instrumentation for Control Room Emergency Ventilation System (CREVS) 
and 3.7.2, Control Room Emergency Ventilation System, and associated 
Bases for the Brunswick Steam Electric Plant (BSEP) Units 1 and 2 will 
be limited in duration (approximately 3 months) and will allow 
operation of both BSEP units to continue while upgrades to the control 
building ventilation system, including new air conditioning (AC) units 
and improved ductwork supports, are being installed. Part of the 
planned work requires opening the ductwork at the evaporative (i.e. 
cooling) coils. Temporary barriers will be constructed to preserve the 
leakage integrity of the control room pressure boundary; however, the 
temporary barriers will not be seismically qualified. While the 
permanent AC units are out of service, temporary AC units will be 
utilized. During the upgrade installation, the AC for the control room 
will not be protected from certain external events (e.g., seismic 
events, environmental hazards such as tornadoes and hurricanes, 
radiological sabotage, and missile hazards), as required by the system 
design and licensing basis, and will not fully meet single failure 
criteria.
    Date of issuance: February 6, 1998.
    Effective date: February 6, 1998.
    Amendment Nos.: 191 and 222.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
authorize changes to the facility's Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63973).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendments: February 28, 1997. Information 
related to the proposed restoration of the primary coolant dose 
equivalent iodine-131 (DEI) to their original licensing basis had been 
previously submitted in Commonwealth Edison Company's (ComEd) letter 
dated November 13, 1996, which was supplemented in subsequent letters 
dated March 20, June 24, August 19 and November 3, 1997.
    Brief description of amendments: The amendments revise the 
technical specifications (TS) to reflect the forthcoming replacement of 
the original steam generators (OSG) in Byron, Unit 1, and Braidwood, 
Unit 1, which are Westinghouse Model D4 steam generators (SG), with the 
replacement steam generators (RSG) which are Babcock and Wilcox, 
International (BWI) SG. The present revisions to the TS remove the 
interim plugging criteria (IPC) related to outer diameter stress 
corrosion cracking (ODSCC) in the OSG as well as the F* alternative 
repair criteria and two separate SG tube sleeving methodologies which 
are not needed for the RSG.
    Date of issuance: February 3, 1998
    Effective date: This license amendment is effective as of the date 
of its issuance and shall be implemented in the first operating cycle 
after installation of the BWI replacement steam generators
    Amendment Nos.: 101, 101, 92 and 92.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66134). The November 13, 1996, and March 20, June 24, August 19 and 
November 3, 1997, submittals provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 3, 1998.
    No significant hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for

[[Page 9617]]

Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

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

    Date of application for amendments: November 6, 1995, and March 11, 
1996, as supplemented June 5, 1997. The June 5, 1997, letter provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
request beyond the scope of the December 20, 1995, and April 10, 1996, 
Federal Register notices.
    Brief description of amendments: These amendments revise the alarm 
setpoints for the effluent radiation and in-containment area radiation 
monitors listed in Technical Specification (TS) Table 3.3-6. These 
revisions make these alarm setpoints consistent with criteria for the 
Emergency Action Levels (EALs) approved by the Nuclear Regulatory 
Commission in August 1994. The EALs use these monitors as an indication 
of fission product barrier challenges or failures. These amendments 
also revise Action Statement 36 of TS Table 3.3-6 to reflect a 
previously approved change (License Amendment Nos. 188 and 70) in 
reporting frequency (change from semi-annual to annual) for effluent 
releases. The revision to Action Statement 36 makes it consistent with 
the previously approved change. These amendments include several 
editorial changes to the TSs which do not change the intent of the TSs.
    Date of issuance: February 9, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 211 and 89.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Dates of initial notice in Federal Register: December 20, 1995 (60 
FR 65677) and April 10, 1996 (61 FR 15988). The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
February 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendment: November 4, 1997.
    Brief description of amendment: The amendment revises Item 6.a.2, 
``4.16kV Emergency Bus (Start Diesel),'' of Table 3.3-4 of Technical 
Specification 3.3.2.1. The change reduces the trip setpoint for 
starting the emergency diesel generators on emergency bus undervoltage 
from a trip setpoint of greater than or equal to 83 percent with a 12-
cycle delay time to a setpoint of greater than or equal to 75 percent 
of nominal bus voltage with a time delay of less than 0.9 seconds 
including auxiliary relay times. The amendment also revises the 
allowable value from greater than or equal to 81 percent of nominal bus 
voltage to greater than or equal to 74 percent of nominal bus voltage 
with a time delay of less than 0.9 seconds including auxiliary relay 
times.
    Date of issuance: February 11, 1998.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No: 212.
    Facility Operating License No. DPR-66. Amendment revised the 
Technical Specifications.

    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63976).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 11, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendment: August 6, 1997.
    Brief description of amendment: The amendment eliminated the 
provisions in Technical Specification 3.8.1, ``AC Sources--Operating,'' 
for accelerated testing of the emergency diesel generators (DG). The 
changes are the following: (1) the frequency of verifying DG starts and 
operation in Surveillance Requirements (SRs) 3.8.1.2 and 3.8.1.3, 
respectively, are changed to 31 days, from the present reference to 
Table 3.8.1-1, and (2) Table 3.8.1-1, ``Diesel Generator Test 
Schedule,'' is deleted. The emergency diesel generators provide 
emergency AC power to the site with the loss of offsite AC power.
    Date of issuance: February 9, 1998.
    Effective date: February 9, 1998.
    Amendment No: 134.
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50003).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

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

    Date of application for amendment: September 3, 1997.
    Brief description of amendment: The amendment authorizes Northeast 
Nuclear Energy Company, through a license condition, to incorporate 
changes to the description of the facility in the Updated Final Safety 
Analysis Report (UFSAR). This change revises the UFSAR by modifying the 
operation of the onsite emergency diesel generators and their 
associated fuel oil supplies.
    Date of issuance: January 23, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 212.
    Facility Operating License No. DPR-65: Amendment revised the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50009).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

[[Page 9618]]

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

    Date of application for amendments: October 24, 1997.
    Brief description of amendments: The amendments revise the 
containment hydrogen analyzer Technical Specifications (TSs) 
surveillance requirements of TS 4.6.4.1 to increase the calibration 
frequency from once per refueling outage to quarterly.
    Date of issuance: January 29, 1998.
    Effective date: As of the date of issuance.
    Amendment Nos. 204 and 186.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66140).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of amendments request: June 13, 1997, as supplemented by 
letter dated January 7, 1998.
    Brief Description of amendments: The amendments change Technical 
Specification (TS) 3.9.13 by adding a footnote to clarify the required 
electrical power sources for the penetration room filtration system 
when it is aligned to the spent fuel pool room during refueling 
operations. In addition, the associated Bases section of the TS will be 
modified to provide additional details concerning the proposed TS 
change.
    Date of issuance: February 5, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-134; Unit 2-126.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38138).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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

    Date of amendments request: October 16, 1997.
    Brief Description of amendments: The amendments change the Farley 
Units 1 and 2 TS by revising the number of allowable charging pumps 
capable of injecting into the reactor coolant system (RCS) when the 
temperature of one or more of the RCS cold legs is equal to or less 
than 180 deg. F.
    Date of issuance: February 5, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-135; Unit 2-127.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63983).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: October 28, 1996, as 
supplemented by letters dated August 19, 1997, and October 16, 1997.
    Brief description of amendment: This amendment revises TS Section 
3/4.8.1, ``A.C. Sources,'' TS Section 3/4.8.2, ``Onsite Power 
Distribution Systems,'' TS Table 4.8.1, ``Battery Surveillance 
Requirements,'' and the associated bases. Surveillance requirements 
have been modified to account for the increase in the fuel cycle. 
Administrative changes were also made.
    Date of issuance: February 3, 1998.
    Effective date: February 3, 1998.
    Amendment No.: 219.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
132).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 1998.
    No significant hazards consideration comments received: No. The 
supplemental information provided by the Licensees did not affect the 
proposed no significant hazards consideration determination.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

    Dated at Rockville, Maryland, this 18th day of February 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-4620 Filed 2-24-98; 8:45 am]
BILLING CODE 7590-01-P