# Biweekly Notice

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3A94-31196

## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** December 21, 1994

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving
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 November 28, 1994, through December 9, 1994.
The last biweekly notice was published on December 7, 1994.

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

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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 January 13, 1995, 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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

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

Date of amendment requests: November 2, 1994
Description of amendment requests: The proposed amendment would
delete the Condenser Vacuum Exhaust release point reference on Figure
5.1-3 and combine it with the Plant Vent Exhaust release point on the
revised Figure 5.1-3. In addition to the figure change, Bases Section
3/4.3.3.6 is amended to note the deletion of radiation monitor RU-142
and the relocation of RU-144 and RU-146 from Table 3.3-13 (previously
deleted) to the Offsite Dose Calculation Manual (ODCM).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Combining the condenser vacuum and the plant vent exhausts has
no affect
[sic] on the operation of the radiation monitoring system or its
intended functions. Routing of the condenser vacuum exhaust to the
plant vent exhaust is in the same area as the old system and does
not affect accident initiation or consequences. The change has no
affect
[sic] on the operation of the plant. The radiation monitors
affected by this change do not provide engineered safety features or
protection system actuation signals. Therefore, the change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Standard 2 -- Does the proposed change create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The overall system is designed to assist the operators in
evaluating and controlling the radiological consequences of normal
plant operations, anticipated operational occurrences, and
postulated accidents. The change does not affect the way the system
is operated. Therefore, the change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Standard 3 -- Does the proposed change involve a significant
reduction in a margin of safety?
Combining of the condenser vacuum and plant vent exhaust into a
single release path does not involve a significant reduction in a
margin of safety. The change involves the removal of one high range
monitor in the condenser vent, however, its function is provided by
the high range monitor in the plant vent. The ranges of the monitors
are the same. The existing plant effluent radiation monitors will
serve to monitor both the plant and condenser air removal system
effluent. The normal range monitors have the ability to adequately
detect radiation over five decades and these monitors will stay in
place and they have the ability to perform the anticipated radiation
release detection. 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
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Basis for proposed no significant hazards consideration
determination: I11Attorney for licensees: 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: Theodore R. Quay

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 amendments request: November 16, 1994 Description of
amendments request: The proposed Technical Specification (TS) change
would (1) revise TS 4.6.1.2 by removing the schedular requirements for
Type A overall integrated leakage rate tests to be performed at 40 plus
or minus 10 month intervals and replacing the acceptance criteria for
these Type A integrated leakage rate tests with a reference to the
containment integrated leakage testing requirements of Appendix J to 10
CFR Part 50, (2) delete TS 4.6.1.2.a through TS 4.6.1.2.c because they
are no longer needed, (3) revise TS 4.6.1.2.h to remove the prohibition
against applying TS 4.0.2 to the 40 plus or minus 10 month integrated
leakage rate test frequency, (4) delete Unit 1 one-time footnote *
located on TS page 3/4 6-3A and on Table 4.6.1.2-1 listed on TS page 3/
4 6-3B since the exception provision has expired, (5) delete Unit 1
one-time footnote ** located on TS page 3/4 6-3A since the exception
has expired, (6) delete Unit 2 footnote * located on TS page 3/4 6-3
because the exception constitutes an approved exemption.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed amendments remove the detailed technical and
schedular information pertaining to primary containment integrated
leakage rate testing from the Technical Specifications and
references the corresponding requirements that are located in the
Appendix J to 10 CFR Part 50. As such, the proposed amendments are
an administrative change since the actual requirements for the
performance of primary containment integrated leakage rate testing
are not being changed. No safety-related equipment, safety function,
or plant operations will be altered as a result of the proposed
amendments. The change does not affect the design, materials, or
construction standards of the primary containment nor the test
methods, test acceptance criteria, or testing frequencies applicable
to primary containment integrated leakage rate testing. Based on the
above, the proposed license amendments do not create a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, no safety-related equipment, safety
function, or plant operations will be altered as a result of the
proposed change. The proposed amendments do not change the primary
containment design or the test methods, test acceptance criteria, or
testing frequencies for primary containment integrated leakage rate
testing. As such, the proposed license amendments cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendments do not involve a significant
reduction in the margin of safety. The proposed amendments do not
involve any changes to the test methods, acceptance criteria, or
testing frequency for primary containment integrated leakage rate
testing. Thus, the proposed amendments will not affect the ability
of the primary containment to perform its intended safety function
and no margins of safety, as defined by the plant's accident
analyses, are impacted. Primary containment integrated leakage rate
testing will continue to be performed in accordance with the
regulatory requirements of Appendix J to 10 CFR Part 50. Based on
the above reasoning, the proposed license amendments 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman

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

Date of amendment request: September 28, 1994
Description of amendment request: Amendments will update the ``Loss
of Power'' functional unit of the Engineered Safety Features Actuation
System (ESFAS) Instrumentation tables within the Technical
Specifications for McGuire Nuclear Station.
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:
As required by 10 CFR 50.91, this analysis is provided
concerning whether the requested amendments involve significant
hazards considerations, as defined by 10 CFR 50.92. Standards for
determination that an amendment request involves no significant
hazards considerations are if operation of the facility in
accordance with the requested 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 requested amendments update the existing one-level
undervoltage protection to be exclusively for loss of voltage, and
add a second level of undervoltage protection to be exclusively for
degraded voltage.
In 48 FR 14870, the Commission has set forth examples of
amendments that are considered not likely to involve significant
hazards considerations. Example vi describes a change which either
may result in some increase to the probability or consequences of a
previously-analyzed accident or may reduce in some way a safety
margin, but where the results of the change are clearly within all
acceptable criteria with respect to the system or component
specified in the Standard Review Plan. The requested amendments are
similar to example vi in that they result in some increase to the
probability of a previously-analyzed accident, the Loss of Offsite
Power accident, but where the changes are clearly based on the
recommendations of Branch Technical Position PSB-1.
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The requested amendments will involve some increase in
the probability of an accident previously evaluated. Automatic
separation from offsite power (a LOOP accident) will be more
probable because the voltage setpoints for the new relaying will be
higher than the settings for the existing relaying. The closer relay
settings are to 100% bus voltage, the more frequently actual bus
voltage can be expected to occur at or below the setpoint. The
occurrence of a LOOP presents a challenge to safety systems. More
probable (e.g., more frequent) LOOPs increase the frequency of
safety system challenges, which increases the probability of
malfunction of equipment important to safety. However, offsetting
this probability increase is a probability decrease due to the
protection of safety equipment from degraded voltage conditions,
given by the added protective relaying. The EPC system is required
to provide power for equipment used for accident mitigation and safe
shutdown. The ability of the EPC system to perform its required
safety functions will not be degraded by the implementation of this
TS change. No common failure modes are created between redundant EPC
system power trains. Therefore, the consequences of an accident or
malfunction of equipment important to safety evaluated in the SAR
are not increased.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No new failure modes are created by the implementation of
this TS change. No accidents previously considered incredible are
made credible. The added protective relaying is expected to be as
reliable as the existing relaying. The added equipment is QA
Condition 1, and qualifications of equipment enclosures have been
maintained. Thus, the possibility of an accident or malfunction of
equipment of a different type than evaluated in the SAR will not be
created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. The setpoints for the existing Loss
of Power protective relays are lowered by this TS change. The new
setpoints have been evaluated and will not prevent the protective
relaying from performing its required safety function. The fission
product barriers (RCS pressure boundary, containment, fuel pellets,
and cladding) are not degraded. No assumptions made in any accident
analysis are affected by the implementation of this TS change,
except as previously discussed for probability of a Loss of Offsite
Power. Therefore, the margin of safety as defined in the basis for
any Technical Specification is not decreased.
Based on the preceding analyses, Duke Power concludes that the
requested amendments do 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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow

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 amendment request: August 31, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs), Section 6, Administrative
Controls, and includes line-item improvements suggested by Generic
Letter 93-07. The proposed changes include the following:
1. Elimination of the references to specific frequencies for
each of the Technical Specification required audits.
2. Elimination of the references to reviews and audits of the
Emergency Plan and Security Plan.
3. Separation of the Inservice Inspection (ISI) and Inservice
Testing (IST) Programs surveillance requirements and removal of the
requirement that relief requests be granted before they are
implemented for both IST and ISI.
4. Editorial changes which were necessitated by a
reorganization.
5. Elimination of the reference to Appendix A of 10 CFR Part 55.
6. Elimination of the requirement to perform an independent fire
protection and loss prevention program inspection annually.
7. Inclusion of the Offsite Dose Calculation Manual and Process
Control Program and associated implementing procedures into the list
of required audits.
8. Updates of the Beaver Valley Power Station (BVPS) Unit 2
License Conditions to reflect completion of activities.
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 likelihood that an accident will occur is neither increased
or decreased by this proposed Technical Specification change which
only affects review and audit frequencies, removes redundancies in
the audit program, corrects editorial information, and updates the
Unit 2 license conditions. This Technical Specification change will
not impact the function or method of operation of plant equipment.
Thus, there is not a significant increase in the probability of a
previously analyzed accident due to this change. No systems,
equipment, or components are affected by the proposed change. Thus,
the consequences of a malfunction of equipment important to safety
previously evaluated in the Updated Final Safety Analysis Report are
not increased by this change.
The proposed change affects audit frequencies, types of audits
listed in the technical specifications, references for some
technical specification sections, the time frame for Inservice
Testing (IST) and Inservice Inspection (ISI) relief request
submittals, and editorial changes necessitated by an internal
reorganization. As such, the proposed change has no impact on
accident initiators or plant equipment, and therefore, does not
affect the probabilities or consequences of an accident.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed technical specification revisions do not involve
changes to the physical plant or operations. Since program audits,
organizational titles, and technical specification references do not
contribute to accident initiation, a change related to the areas
listed in the description section [***] cannot produce a new
accident scenario or produce a new type of equipment malfunction.
Therefore, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change concerns the conduct of audits, technical
specification references, ISI and IST relief request submittals,
completed License conditions, and organizational title changes and
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are not affected by this
proposed change.
Therefore, use of the proposed Technical Specification would 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 involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Alquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler

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

Date of amendment request: August 11, 1994, as supplemented
December 2, 1994.
Description of amendment request: The requested change would remove
cycle-specific variables from the Waterford 3 Technical Specifications
(TSs) and control them under a new document called the Core Operating
Limits Report (COLR). All cycle-specific limits that are to be included
in the COLR must be calculated using NRC approved methodologies. The
proposed change is consistent with the TS line-item improvement
guidelines provided by the NRC in Generic Letter (GL) 88-16, ``Removal
of Cycle-Specific Parameter Limits From Technical Specifications,''
dated October 3, 1988.
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. Removing cycle-specific variables from the TS and placing
them into a COLR, is consistent with the NRC guidance provided in GL
88-16. These changes are administrative in nature and have no impact
on plant operation or accident analyses. The TS will continue to
require operation within the core operational limits for each cycle
reload calculated by the approved reload methodologies. If these
limits are violated, Technical Specifications will continue to
ensure that the appropriate actions are taken.
The cycle-specific evaluation demonstrates that changes in the
fuel cycle design and the corresponding COLR do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Listing the NRC approved methodologies in the COLR as opposed to
the TS Administrative Controls section is purely an administrative
change in contrast to NUREG 1432. The proposed change requires the
use of NRC approved methodologies. Listing the approved
methodologies in the TS provides the potential for an increased
licensee and NRC administrative burden without a commensurate
increase in safety or control.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes, to relocate the cycle-specific
variables from TS to the COLR, are administrative in nature. No
change in the design, configuration, or method of operation of the
plant is made by this amendment. The cycle-specific variables will
continue to be calculated using NRC approved methods. TS will
continue to require operation within the required core operating
limits and appropriate actions will be taken if the limits are
exceeded.
Listing the NRC approved methodologies in the COLR as opposed to
the TS Administrative Controls section is purely an administrative
change in contrast to NUREG 1432. The proposed change requires the
use of NRC approved methodologies. Listing the approved
methodologies in the TS provides the potential for an increased
licensee and NRC administrative burden without a commensurate
increase in safety or control.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The margin of safety presently provided is not affected by
removing cycle-specific core operating limits from TS. The core
limits contained in the COLR are obtained through analyses using NRC
approved methodologies. The TS still: (1) require that the core be
operated within these limits and (2) specify appropriate actions to
be taken if the limits are violated. The cycle-
specific COLR limits for future reload will also be developed
based on NRC-approved methodologies. In addition, each reload will
involve a 10CFR 50.59 safety review to assure that operation of the
unit within the cycle-specific limits will not involve a reduction
in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Safety and Significant Hazard Determination
Based on the above safety analysis, it is concluded that: (1)
the proposed change does not constitute a significant hazards
consideration as defined by 10CFR50.92; and (2) there is a
reasonable assurance that the health and safety of the public will
not be endangered by the proposed change; and (3) this action will
not result in a condition which significantly alters the impact of
the station on the environment as described in the NRC Final
Environmental Statement.
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, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia

Date of amendment request: October 13, 1994
Description of amendment request: The proposed amendments would
revise the Hatch Technical Specifications (TS) as follows:
1. Lower the anticipated transient without scram-recirculation pump
trip (ATWS-RPT) setpoint by approximately 2 feet 2 inches to minimize
the potential for recirculation pump trips following reactor scrams.
2. Allow restarting the recirculation pump following an RPT when
the temperature differential between the coolant at the reactor bottom
head and the reactor steam dome cannot be obtained, provided certain
conditions are met.
The licensee believes the above changes will aid in preventing
thermal stratification and unnecessary thermal cycles resulting from
the rapid cooldown of the bottom head region and the reduction in
reactor pressure to atmospheric conditions.
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:
Proposed Change 1
Proposed Change 1 does not involve a significant hazards
consideration, because it does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Lowering the ATWS-RPT trip will not increase the probability of
occurrence of any design basis accident or transient, since this
change does not physically affect any component of the reactor
coolant pressure boundary (RCPB). Therefore, the probability of a
LOCA event is not increased. Lowering the ATWS-RPT water level
setpoint does not increase the probability of an ATWS event, since
no component of the CRD system or the reactor protection system is
being physically altered by this change. Also, the operation of
these two systems is not affected.
Reducing the setpoint may require installation of new slave trip
units; however, this addition does not increase the probability of
occurrence of accidents or transients. The new trip units will be
functionally identical to other slave trip units already in use at
Plant Hatch and are within the design capabilities of ATTS. In
conclusion, no safety-related plant system or component is being
affected in a manner that would render it more susceptible to
failure.
Lowering the setpoint does not result in an increase of the
consequences of a previously evaluated accident. GE reviewed the
proposed reduction and determined the results of the ATWS event with
the lowered setpoint remain acceptable. An approved analytical
method (REDY) was used to evaluate a bounding ATWS event -- LOFW
[loss of feedwater]. The results indicate that reactor power with
the new ATWS-RPT setpoint remains stable, with no unacceptable power
spikes. Hot and cold reactor shutdowns can still be ultimately
attained.
The consequences of non-ATWS events are not increased. For LOCA
events, reducing the recirculation pump low water level trip
setpoint allows the recirculation pumps to run longer. The forced
circulation provided by the recirculation pumps keeps the fuel
cooler for a longer period of time. The ECCS-LOCA analysis assumes
the pump trip and coastdown early in the event. Therefore, lowering
the ATWS-RPT makes the ECCS-LOCA analysis more conservative and, as
a result, it does not need to change.
Based on the above discussion, Proposed Change 1 does not
constitute an increase in the probability or consequences of a
previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
Lowering the ATWS-RPT trip will not alter the design or
operation of any safety-related system. The change may require
adding new slave trip units to ATTS; however, the new trip units
will be functionally identical to the equipment already in use at
Plant Hatch. Furthermore, the addition of this slave trip unit is
within the design capabilities of ATTS.
Since no new operation modes, accident scenarios, or failure
modes are introduced, Proposed Change 1 does not create the
possibility of a new type of accident.
3. Involve a significant reduction in the margin of safety.
As stated previously, reducing the ATWS-RPT low water level
setpoint will not cause unacceptable results for ATWS events.
Specifically, the LOFW event is bounding for all the ATWS events. An
evaluation using approved analytical methods indicates that reducing
the ATWS-RPT setpoint will not result in power instabilities or
unacceptable power spikes, or prevent the mitigation of an ATWS
event. (Reference Enclosure 1 [of the licensee's submittal],
Proposed Change 1).
The ATWS-RPT aids in maintaining the level above the top of the
active fuel. The reduction of core flow reduces the neutron flux and
thermal power and, therefore, the rate of coolant boil-off. However,
the setpoint reduction does not significantly reduce the margin of
safety since a substantial margin remains to the top of the active
fuel.
For non-ATWS events, delaying the RPT will provide a slight
improvement in the current ECCS-LOCA analysis, thereby improving the
margin of safety.
The margin of safety for transients is not reduced because plant
transient (MCPR) analyses do not take credit for the ATWS-RPT trip.
Proposed Change 2
Proposed Change 2 does not involve a significant hazards
consideration, because it does not:
1. Involve a significant increase in the probability of
occurrence or the consequences of a previously analyzed accident.
Allowing a recirculation pump restart within 30 minutes of a
trip, when the temperature differential is unknown, will not
increase the probability of occurrence of a previously analyzed
accident because this change does not physically alter the RCPB.
Additionally, the proposed change does not alter the design or
function of any safety-related systems.
Furthermore, no recirculation system equipment is being changed
as a result of this amendment. The start circuitry and trip
circuitry remain[s] unaffected. Operation of the recirculation
system with the reactor at power is also unaffected. As a result,
the probability of the chapter 14 and 15 events dealing with the
recirculation system are not increased; i.e., trip of one or both
recirculation pumps, recirculation pump seizure, recirculation flow
controller failure, etc.
The purpose of the 145 deg.F temperature differential
requirement is to avoid thermal shock caused by hot water on the
cold CRD stub tubes during recirculation pump restart. If the
temperature differential is unable to be determined, restart within
30 minutes of the trip will not increase the probability or severity
of thermal fatigue on the stub tubes. As discussed in Enclosure 1,
Basis for Proposed Change 2, stratification will not develop within
a 30-minute period following pump trip, thus, the temperature
differential will not exceed 145 deg.F. Additional caveats are
provided to insure the required temperature differential is met.
These involve certain conditions of ECCS injection, feedwater
temperature, and drive flow.
General Electric verified that this provision for recirculation
pump restart will not affect any plant safety analysis, including
radiological analysis. Therefore, the consequences of previously
analyzed events are not increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed recirculation pump restart provisions do not
introduce any new plant operating modes, accident scenarios, or
equipment failure modes. All other requirements for recirculation
pump restart; e.g., those addressing equipment protection and power
oscillations, will continue to apply.
3. Involve a significant reduction in the margin of safety.
The 145 deg.F differential temperature requirement is in place
to avoid thermal fatigue on the CRD stub tubes and the in-core
housing welds. Allowing the early restart with the listed caveats,
when temperature indication is not available, is acceptable because
the conditions for re-start insure that a stratified condition has
not yet developed. Thus, the cooler vessel structures at the vessel
bottom will not experience a severe thermal shock resulting from
exposure to hot water following the pump restart.
This change will actually aid in preventing the development of a
stratified condition, since the recirculation pumps will be
restarted before a stratified condition can develop, thereby helping
to maintain RCPB integrity. In the past, it has often been necessary
to depressurize the RPV [reactor pressure vessel] to atmospheric
pressure before the required temperature differential was met.
Proposed Change 2 should reduce the number of times depressurization
is required, thus avoiding unnecessary thermal cycles on the RPV.
Therefore, the margin of safety regarding the protection of RPV
components from severe thermal stresses, and the integrity of the
RCPB has not been reduced, and may actually increase.
The margin of safety in existing plant analyses is not reduced,
because none of the analyses are adversely affected as a result of
allowing the pump restart within 30 minutes of the RPT, as indicated
in GE's review of plant transient and accident analyses.
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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow

Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1
and 2, Burke County, Georgia

Date of amendment request: October 3, 1994
Description of amendment request: This amendment would replace the
reactor coolant system heatup and cooldown limitations for VEGP Units 1
and 2, contained in Technical Specification figures 3.4-2a through 3.4-
3b, and the maximum allowable nominal power-operated relief valve
(PORV) setpoint for the cold overpressure protection system. These
changes are the results of new analyses that account for the
nonconservatisms identified in NRC Information Notice 93-58, the
results of reactor pressure vessel surveillance capsule examinations,
and recently issued ASME Code Case N-514.
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:
Conformance of the proposed amendment with the standards for a
determination of no significant hazards, as defined in the three
factor test of 10 CFR 50.92, is shown in that the proposed
amendment:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The revised heatup and cooldown limits and PORV setpoints ensure
that the Appendix G pressure/temperature limits are not exceeded and
therefore, help ensure that RCS integrity is maintained. The changes
do not result in a condition where the design, material, and
construction standards of the RCS are altered. In addition, the
safety function of the COMS (cold over-pressure mitigation system),
which is related to accident mitigation, has not been degraded.
Therefore, the probability of an accident is not increased by the
PORV setpoint change.
The changes do not adversely affect the integrity of the RCS
such that its function in the control of radiological consequences
is affected. In addition, the changes do not affect any fission
barrier. The changes do not degrade or prevent the response of the
COMS or other safety-related system to accident scenarios, as
described in FSAR chapter 15. In addition, the changes do not alter
any assumption previously made in the radiological consequence
evaluations nor affect the mitigation of the radiological
consequences of an accident described in the FSAR. Therefore, the
consequences of an accident previously evaluated in the FSAR will
not be increased.
Thus, operation of VEGP Units 1 and 2 in accordance with the
proposed license amendment, does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The changes do not cause the initiation of any accident nor
create any new credible limiting single failure for safety-related
systems and components. The changes do not result in any event
previously deemed incredible being made credible. As such, it does
not create the possibility of an accident different than any
evaluated in the FSAR.
The changes do not have any effect on the ability of the safety-
related systems to perform their intended safety functions. The
changes do not create failure modes that could adversely impact
safety-related equipment. Therefore, it will not create the
possibility of a malfunction of equipment important to safety
different than previously evaluated in the FSAR. Thus, the proposed
license amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does not involve a significant reduction in a margin of
safety.
The evaluation has shown that the PORV setpoints ensure that the
Appendix G pressure/temperature limits are not exceeded. The
analysis to support the proposed PORV setpoint change demonstrates
that the appropriate criteria, including that of ASME Code Case N-
514, are met for the postulated RCS pressures and temperatures. An
adequate margin of safety against vessel failure is assured, in
part, by the safety factors identified in Appendix G to Section III
of the ASME Boiler and Pressure Vessel Code, and [in] the basis for
ASME Code Case N-514 as well as [in the] added margin to prevent
lifting of the PORVs. The heatup and cooldown limits are designed to
prevent nonductile failure of the reactor vessel and take into
account the results of surveillance capsule Y on the reactor vessel
materials for VEGP Unit 1. The actuation of the safety-related
components and responses of the safety-related systems will remain
as modeled in the safety analyses. The changes will have no adverse
[effect] on the availability, operability, or performance of the
COMS. Therefore, the changes will not reduce the margin of safety,
as described in the bases to any Technical Specification.
Thus, [this] proposed license amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow

Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 14, 1994, as supplemented by
letter dated November 10, 1994.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) by removing component lists
from the TSs in accordance with NRC Generic Letter (GL) 91-08 and by
removing the schedule for withdrawal of reactor vessel material
specimen capsules from the TSs in accordance with GL 91-01. This
proposed amendment was originally noticed in the Federal Register on
May 23, 1994, (59 FR 26675). The licensee's letter dated November 10,
1994, provides clarification of the wording in the proposed TSs and
does not change the proposed determination that the amendment request
involves no significant hazards consideration. However, the notice is
being repeated here.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will not result in any hardware or operating
changes. The proposed change is based upon Generic Letters 91-01 and
91-08 and merely removes component lists, removes details relating
to the component lists, provides clarifying information supporting
the removal of the component listings, or removes details (which are
considered administrative) that are no longer applicable to the
Technical Specifications. The components listed in the affected
Technical Specifications are assumed in the mitigation of accident
and transient events. The removal of tabular component listings from
the Technical Specifications does not impact affected component
OPERABILITY requirements. Technical Specifications will continue to
require the components to be OPERABLE. Action statements and
surveillance requirements for the components will also remain in the
Technical Specifications. The tabular component lists are relocated
to the Technical Requirements Manual which will be in accordance
with the change control provisions specified in the Administrative
Controls Section of the Technical Specifications (Specification
6.5.2). Therefore, this change is administrative in nature and does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes to parameters governing normal plant
operation. The proposed change will not impose any different
requirements and adequate control of information will be maintained.
No new failure modes are introduced. Therefore, this proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumption. The proposed
changes do not alter the scope of equipment currently required to be
OPERABLE or subject to surveillance testing, nor do the proposed
changes affect any instrument setpoints or equipment safety
functions. Therefore the 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: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner

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

Date of amendment request: November 10, 1994
Description of amendment request: The proposed amendment revises
the Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Section 3.2.A to refer to the Offsite Dose Assessment Manual (ODAM) for
the setpoint of the Offgas Stack Radiation Monitor and makes the
``Applicable Operating Mode'' and the ``Action'' statements for these
instruments consistent with the required function. The Action statement
for the other instruments which initiate Secondary Containment
isolation is also revised to be consistent with the current practice
and with the function of those instruments. The Basis is also revised
to add further description of the function and requirements.
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 provided below:
1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated because the instruments will still be required
to be operable to initiate an isolation at a setpoint which will
assure that the offsite dose limits are preserved, as designed, or
else administrative controls will be established for the venting of
primary containment. Through either means, offsite releases will be
maintained within the limits established in the ODAM. The change to
the applicable operating mode simply will require that the
instruments be operable when they are assumed to be operable in
previously analyzed accidents. The change to the required action
when the TS requirement cannot be met will assure that the flow path
from containment is isolated or that positive control is established
so that any offsite radioactive gaseous release is within the limits
analyzed in the ODAM.
2) The proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated
because the affected instruments are inputs to the secondary
containment isolation and the revised specification will assure that
they are operable or adequately compensated when they are assumed to
perform their function. The instruments initiate a secondary
containment isolation in the event that high radiation levels are
detected in the monitored effluent.
3) The proposed amendment will not involve a significant
reduction in a margin of safety because the revised applicability
statement will assure that the instruments are operable when they
are required to perform their function. The proposed compensatory
action allows administrative control of the isolation valves when
the instruments are inoperable and it is necessary to continue
venting. This allowance recognizes that venting is a controlled
evolution and that operator action would be adequate to prevent
excessive releases in the event of high radioactivity in the offgas
piping. The revision to the setpoint will not affect system
operation, but will continue to assure that the gaseous effluents
released are within the limits specified in the ODAM.
In summary, the proposed changes do not change the probability
or consequences of an accident previously evaluated, do not create
the possibility for a new or different kind of accident and do not
involve a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, 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: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Newman,
Bouknight & Edgar, PC, 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Leif J. Norrholm

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

Date of amendment requests: October 7, 1994
Description of amendment requests: The proposed amendments would
remove the requirements for the Nuclear Safety and Design Review
Committee (NSDRC) to audit, and for the Plant Nuclear Safety Review
Committee (PNSRC) to review, the Emergency and Security plans and
implementing procedures. The composition of the PNSRC and the NSDRC
would also be revised to reflect organizational changes. Changes would
be made to the delegation of responsibility by the Site Vice President/
Plant Manager, and title corrections would be made on all pages
affected by the above changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:We [the licensee] have
evaluated the proposed T/S changes and have determined that the changes
should involve no significant hazards consideration. The proposed
amendment involves changes to the administrative controls section of
the T/Ss only. Because all changes reflect organizational/title changes
only or guidance from GL 93-07, they do not:
1) 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; or
3) involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske
MemorialLibrary, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon

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

Date of amendment requests: November 16, 1994
Description of amendment requests: The proposed amendments would
allow core offload 100 hours after core subcriticality instead of the
168 hours currently required. Also included in this submittal are minor
typographical corrections to Figure 5.6-1, ``Normal Storage Pattern
(Mixed Three Zone), and Figure 5.6-2, ``Interim Storage Pattern
(Checkerboard).
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:
We [the licensee] have evaluated the proposed T/S, editorial and
clarification changes and have determined that they do not represent
a significant hazards consideration based on the criteria
established in 10 CFR 50.92(c). Operation of Cook Nuclear Plant in
accordance with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Although one of the proposed changes results in initiation of
core offload earlier after subcriticality than is currently allowed,
it does not increase the probability or consequences of an accident
previously evaluated. The bulk pool water temperatures, fuel rod
clad temperatures, and pool wall concrete temperatures will be
within acceptable limits as shown in Attachment 2 [of the November
16, 1994, submittal]. In addition, the subject change will not
result in an uncontrolled release of radiation to the environment
and will not initiate an accident. The remaining changes are
editorial in nature and have no [e]ffect on probability or
consequences of a postulated accident.
(2) Create the possibility of a new or different kind of
accident from an accident previously evaluated.
As previously stated, the earlier fuel movement change will not
result in bulk pool water, fuel rod clad, or concrete temperatures
which would initiate bulk pool boiling, challenge fuel rod integrity
or jeopardize the structural integrity of the pool. This change will
also have no impact on the criticality, structural, seismic, or
dropped assembly accident analysis previously performed and accepted
by the NRC. Consequently, the proposed T/S change does not create
the possibility of a new or different kind of accident from any
previously analyzed. The remaining changes have no [e]ffect on [the]
nature or probability of a postulated accident.
(3) Involve a significant reduction in a margin of safety.
The proposed change for earlier fuel movement will not result in
bulk pool water temperatures, fuel rod clad temperatures or concrete
temperatures which would initiate bulk pool boiling, challenge fuel
rod integrity or jeopardize the structural integrity of the pool.
This proposed change will not affect the results of any other
analysis associated with the spent fuel pool. It is, therefore,
concluded that this change poses no significant reduction in a
margin of safety. The remaining changes have no [e]ffect on the
nature or probability of a postulated accident.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon

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

Date of amendment requests: November 18, 1994
Description of amendment requests: The license amendment requests
propose a change to Technical Specification (T/S) 4.0.5 for both units
to delete the wording ``except where specific written relief has been
granted by the Commission pursuant to 10 CFR 50, Section
50.55a(g)(6)(i).'' This change, which is consistent with guidance in
the November 1993 draft NUREG-1482, ``Guidelines for Inservice Testing
at Nuclear Power Plants,'' would allow the licensee to implement
certain 10 CFR 50.55a relief requests while the relief requests were
being reviewed by the NRC.
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:
We [the licensee] have evaluated the proposed T/S change and
have determined that the change involves no significant hazards
consideration. Operation of Cook Nuclear Plant in accordance with
the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The inspections required under Section XI are intended to
show the operational readiness of the applicable components, and
exceptions to the Code are allowed. When taking relief from Code
requirements, alternate requirements are developed which provide a
high level of confidence that components will perform their intended
function.
The proposed change does not alter the Code requirements or
lessen our obligations under existing regulations. Its only effect
is to allow implementation of Code relief prior to obtaining NRC
written approval. The proposed T/S change is consistent with NUREG-
1431, and, as such, has been found to be acceptable by the NRC.
Therefore, we believe that implementation of this change will not
involve a significant increase in the probability or consequences of
a previously analyzed incident.
(2) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed amendment does not create the possibility of a new
or different kind of accident from any previously evaluated. Typical
relief requests involve using alternative testing methods or
increasing the time interval between tests. Each proposed
alternative must assure that the component will perform its intended
function. The proposed change involves no physical changes to the
plant; therefore, we believe that implementation of this change will
not introduce a new of different kind of accident than previously
analyzed.
(3) Involve a significant reduction in a margin of safety.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The inspections required under Section XI are intended to
show the operational readiness of the applicable components, and
exceptions to the Code are allowed. When taking relief from Code
requirements, alternate requirements are developed which provide a
high level of confidence that components will perform their intended
function.
The proposed change does not altar the Code requirements or
lessen our obligations under existing regulations. Its only effect
is to allow implementation of Code relief prior to obtaining NRC
written approval. The proposed T/S change is consistent with NUREG-
1431, and, as such, has been found to be acceptable by the NRC.
Therefore, we believe that implementation of this change will not
result in a significant reduction of the margin of safety.
NRC staff has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon

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

Date of amendment request: November 14, 1994
Description of amendment request: The proposed license amendment
would revise Technical Specification 4.5.1.e.2.e) to reduce the leak
rate test pressure for the Automatic Depressurization System (ADS)
nitrogen receiving tanks from 385 psig to 365 psig. This pressure
reduction would be made to reduce potential degradation of the rupture
disk installed on each ADS nitrogen receiving tank during periodic leak
testing of the receiving tanks. Plant operating experience has shown
that leak rate testing at 385 psig occasionally results in inadvertent
failure of the rupture disks. Testing at the reduced pressure would be
consistent with the manufacturer's recommendations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The ADS is required to effect or support the safe shutdown of
the reactor. This is accomplished by the blowdown of steam from a
group of seven designated ADS SRVs [safety/relief valves] to the
suppression pool. The proposed change to the test pressure does not
affect any accident precursors. Therefore, the proposed change
cannot increase the probability of an accident previously evaluated.
In the event the nitrogen gas supply from the nitrogen gas
storage tanks is lost, a minimum nitrogen pressure of 334 psig in
the ADS nitrogen receiver tanks assures a five-day supply of
nitrogen to the ADS accumulators. The proposed change to
Surveillance Requirement 4.5.1.e.2.e) would decrease the leak rate
test pressure of the ADS nitrogen receiver tanks from 385 psig to
365 psig. Since the proposed test pressure remains well above the
design minimum pressure of 334 psig, the surveillance test continues
to ensure that the actual leakage of the safety related ADS
accumulator pneumatic supply system is bounded by the leakage
assumptions contained in the system design. In addition, the
surveillance test continues to ensure that the ADS nitrogen receiver
tanks are capable of providing a 5-day supply of nitrogen to the ADS
accumulators. Therefore, the proposed change does not significantly
increase the consequences of a previously evaluated accident.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed leak rate test pressure of 365 psig for the ADS
nitrogen receiver tanks is above the minimum design pressure to
assure a 5-day supply of nitrogen is available to the ADS
accumulators if makeup from the high pressure nitrogen gas storage
tanks is lost. With the proposed change, the ADS will continue to
perform its safety function of effecting and supporting the safe
shutdown of the reactor. The nitrogen receiving tank test pressure
is not a precursor for any new or different accident and the change
does not affect the operation of the system in any way.
Accordingly, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The operation of the ADS SRVs, in conjunction with the LPCI [low
pressure coolant injection] mode of RHR [residual heat removal
system] and/or the LPCS [low pressure core spray] system, functions
as an alternative to the HPCS [high pressure core spray system] for
protection against fuel cladding damage upon a small break loss-of-
coolant accident. The blowdown of steam by these SRVs depressurizes
the reactor, allowing injection by the low-pressure coolant
injection sources. With the proposed change, the ADS will continue
to perform its intended safety function of effecting and supporting
the safe shutdown of the reactor as an alternate to the HPCS. The
proposed test pressure of 365 psig remains well above the minimum
acceptable pressure for the nitrogen receiver tanks of 334 psig.
Therefore, the change will not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Michael J. Case

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

Date of amendment request: November 30, 1994
Description of amendment request: The proposed amendment would
modify the Technical Specifications by adding a footnote to Limiting
Conditions for Operation (LCOs) 3.8.1.1.b and 3.8.1.2.b which will
denote that 24,000 gallons of fuel oil is capable of supporting the
operation of one emergency diesel generator (EDG) for at least 4 days
and the other EDG for 1 hour with the EDGs loaded to the continuous
rated load of 2750 kW.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed changes do not involve a significant hazards
consideration because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases
Section 3/4.8 will revise the Millstone Unit No. 2 design
requirements regarding the volume of EDG fuel oil which is required
to be stored onsite. The new rationale indicates that 24,000 gallons
of safety-related fuel oil would support the operation of one EDG
for at least four days with the other EDG running for at least one
hour. These run-times assume the EDGs are loaded to the continuous
rated loading of 2750 kW.
The proposed changes have no effect on EDG operation and
reliability. They provide additional operational flexibility,
because the EDG loading can be varied without the EDG minimum run-
time being altered. Also, an EDG run-time of at least four days
provides significant time to replenish fuel oil from onsite and
offsite sources even in the event of a hurricane or seismic event.
Based on the above, there is no effect on the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The EDGs are required to operate in response to a loss of
offsite power. The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b
and Bases Section 3/4.8 do not change the manner in which the EDGs
respond to a design basis accident. Also, the proposed changes do
not introduce any new failure mechanisms. Therefore, the proposed
changes do 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.
The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases
Section 3/4.8 have no effect on EDG operation and reliability. They
provide additional operational flexibility, because the EDG loading
can be varied without the EDG minimum run-time being altered.
An EDG run-time of at least four days provides significant time
to replenish EDG fuel oil via onsite or offsite sources even in the
event of a hurricane of seismic event. EPIP 4400 requires that the
need to order EDG fuel oil be evaluated within four hours of a loss
of offsite power event. Also, the high reliability of the electrical
grid and the high probability that offsite power would be restored
within 24 hours reduces the need to rely on extended EDG operation.
Millstone Unit No. 2 has more margin than is indicated by the
new design requirements. The EDG run-time will be significantly
greater than four days, because the electrical loading on the EDGs
will be less than the continuous rated loading, and electrical loads
will be shed through normal recovery actions following a design
basis accident.
Based on the above, 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee

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

Date of amendment requests: July 11, 1994
Description of amendment requests: The proposed amendments would
change license condition 2.C.4 of each license to conform to the
standard fire protection license condition as stated in Generic Letter
(GL) 86-10, ``Implementation of Fire Protection Requirements.'' In
addition, the amendments would delete the fire protection program
elements from the Technical Specifications and incorporate, by
reference, the NRC-approved Fire Protection Program and major
commitments, including the fire hazards analysis, into the Updated
Safety Analysis Report. Guidance for these proposed changes is also
provided in GL 88-12, ``Removal of Fire Protection Requirements from
Technical Specifications.''
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[s] will not involve a significant
increase in the probability or consequences of [an] accident
previously evaluated.
The requested changes are administrative in nature in that they
move fire protection requirements from the Technical Specifications
to the Fire Protection Program and associated implementing
procedures following the guidance provided in GL 86-10 and GL 88-12.
The requested changes will not revise the requirements for fire
protection equipment operability, testing or inspections. The
amendment would give added responsibility to the Operations
Committee for review of the Fire Protection Program in accordance
with the guidance given in GL 86-10 and 88-12 including special
reporting requirements associated with limiting conditions for
operation for fire protection systems.
The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do they affect
any assumptions or conditions in any of the accident analyses. Since
the accident analyses remain bounding, their radiological
consequences are not adversely affected.
Therefore, the probability or consequences of an accident
previously evaluated are not affected.
(2) The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The requested changes are administrative in nature in that they
move fire protection requirements from the Technical Specifications
to the Fire Protection Program and associated implementing
procedures following the guidance provided in GL 86-10 and GL 88-12.
The requested changes will not revise the requirements for fire
protection equipment operability, testing or inspections. The
amendment would give added responsibility to the Operations
Committee for review of the Fire Protection Program in accordance
with the guidance given in GL 86-10 and 88-12 including special
reporting requirements associated with limiting conditions for
operation for fire protection systems.
The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be
created.
(3) The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The requested changes are administrative in nature in that they
move fire protection requirements from the Technical Specifications
to the Fire Protection Program and associated implementing
procedures following the guidance provided in GL 86-10 and GL 88-12.
The requested changes will not revise the requirements for fire
protection equipment operability, testing or inspections. The
amendment would give added responsibility to the Operations
Committee for review of the Fire Protection Program in accordance
with the guidance given in GL 86-10 and 88-12 including special
reporting requirements associated with limiting conditions for
operation for fire protection systems.
Therefore, a significant reduction in the margin of safety would
not be involved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
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: John N. Hannon

Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: November 11, 1994
Description of amendment request: The proposed amendment to the
Technical Specifications (TSs) would make administrative changes to TS
5.2 and 5.5. These changes reflect organizational changes in OPPD
senior management, delete specific titles of personnel on the Safety
Audit and Review Committee (SARC) and Plant Review Committee (PRC), and
make changes to SARC reviews and audits.
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) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are administrative changes to reflect
organizational changes in Omaha Public Power District (OPPD) Senior
Management, remove specific titles from the membership of the Plant
Review Committee (PRC) and the Safety Audit and Review Committee
(SARC), add minor clarifications to SARC reviews and audits and
delete statements concerning the frequency of SARC audits from the
Technical Specifications (TS).
The proposed change to revise the overall corporate
responsibility for plant nuclear safety from the Senior Vice
President to Vice President is administrative in nature as it only
reflects an organizational change. Section 12 of the Updated Safety
Analysis Report describes the management structure and reporting
responsibilities of OPPD. Section 12 provides an organizational
chart to differentiate the Vice President in charge of nuclear
activities from other Vice Presidents within OPPD. Therefore,
changing the corporate reporting responsibility does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes to the membership of the PRC and SARC are
administrative in nature since only the specific titles of the
members are being removed from the TS. The management level and
expertise of personnel who are PRC or SARC members is not being
changed. The review of plant operations is still required to be in
compliance with ANSI N18.7-1976 and Regulatory Guide 1.33, Revision
2, as committed to in the Fort Calhoun Station Quality Assurance
(QA) Program. Any changes in the QA Program which reduce the
effectiveness of the program must be approved by the NRC in
accordance with 10 CFR 50.54(a)(3). Therefore, the proposed changes
to the membership of the PRC and SARC do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Clarifications of SARC reviews and audits and the deletion of
SARC audit frequencies from the TS are administrative changes. The
audit frequencies are required by the NRC approved QA Program and
any changes that could reduce the effectiveness of the QA Program
must be approved by the NRC in accordance with 10 CFR 50.54(a)(3).
Therefore, the clarifications and deletion of the specific audit
frequencies do not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes are administrative in nature to reflect
organizational changes in OPPD Senior Management, remove specific
titles from the membership of the PRC and SARC, provide minor
clarifications of SARC reviews and audits and delete statements
concerning the frequency of SARC audits from the TS. The proposed
changes do not revise any equipment setpoints, change the manner in
which any plant equipment is operated, or propose any new operating
modes. Therefore, the proposed changes do 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.
The proposed changes revise organizational and administrative
requirements contained within the Administrative Controls section of
the TS. The proposed changes do not revise any equipment setpoints,
change the manner in which any plant equipment is operated, or
propose any new operating modes. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut Avenue, NW., Washington, DC 20009-5728
NRC Project Director: Theodore R. Quay

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: October 28, 1994
Description of amendment request: The proposed changes to the
Technical Specifications (TS) for the two units would add reference
20 (Unit 1) and reference 18 (Unit 2) to Section
6.9.3.2 as ``PL-NF-90-001, Supplement 1, 'Application of Reactor
Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating
Changes and Use of RETRAN MOD 5.1', September 1994''. These changes
would add changes to the methodology that the licensee is using to
perform its nuclear fuel reload analysis for the two units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Incorporation of these proposed minor changes into PP&L's NRC
approved methodology for performing reload licensing analysis is
considered to be an enhancement to the currently approved
methodology. Upgrading of the RETRAN code allows for taking
advantage of state-of-the-art technology, while utilization of the
generic correlation for the LOFWH event supports consistency in
licensing analysis performance. Results of incorporating these
changes will not significantly increase the probability or the
consequences of an accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
As stated above, the incorporation of these minor changes are
considered enhancements, allowing PP&L to more efficiently and cost
effectively continue to perform future reload licensing analysis.
Therefore, the incorporation of these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
III. Involve a significant reduction in a margin of safety.
In addition to the extensive testing perform by EPRI, PP&L has
performed its own comparison tests utilizing RETRAN MOD005.1 in
place of MOD004 for four licensing transients that use the RETRAN
code. Results of this comparison were essentially the same for both
codes and support this proposed change. Also, the Loss of Feedwater
Heating event is not a limiting event for establishing MCPR
Operating Limits for Susquehanna. Therefore, the incorporation of
these changes will have no impact on current safety margins, nor
will they involve a significant reduction in the margin to 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: October 28, 1994
Description of amendment request: The proposed changes to the
Technical Specifications (TS) for the two units would make a number of
administrative changes. These would include changing the title of the
positions of Superintendent of Plant to the Vice President-Nuclear
Operations, and changing the title of Vice-President-Nuclear Operations
to Senior Vice President-Nuclear for the listing of the assignment of
certain duties in various sub-sections of Section 6.0 of the TS. Other
proposed changes would be the deletion of a number of footnotes
indicating times, dates, and events that are no longer applicable, the
addition of a footnote to Section 6.5.1.2 indicating that the Station
Duty Manager shall act as a PORC [Plant Operations Review Committee]
chairman in the absence of the Vice President-Nuclear Operations, and
the change of the Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report in Table 4.11.2.1.2-1
footnote g.
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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposal to change the title of Superintendent of Plant to
Vice President-Nuclear Operations and Vice President-Nuclear
Operations to Senior Vice President-Nuclear (for certain duties) is
administrative in nature and does not compromise the minimum
qualifications or training required for these positions. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change for the removal of footnotes that reference
periods of time, dates, and events that have since past is justified
based on the fact that they are no longer applicable. Because
operators must perform unnecessary applicability reviews on these no
longer applicable footnotes, removing the footnotes decreases the
potential for confusion and incorrect actions. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to add a footnote indicating the Station
Duty Manager shall act as PORC chairman in the absence of the Vice
President-Nuclear Operations will ensure continuous leadership of
PORC and will enhance performance by ensuring a responsible
individual is available during all shifts. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Changing the Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report was previously approved
in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating
this change into footnote g of Table 4.11.2.1.2-1 will maintain
accuracy and consistency. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposal to change the title of Superintendent of Plant to
Vice President-Nuclear Operations and Vice President-Nuclear
Operations to Senior Vice President-Nuclear (for certain duties) is
administrative in nature and does not compromise the minimum
qualifications or training required for these positions. Also, the
change does not diminish the responsibilities or functions of these
positions. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change for the removal of footnotes that reference
periods of time, dates, and events that have since past is justified
based on the fact that they are no longer applicable. Because
operators must perform unnecessary applicability reviews on them,
removing the no longer applicable footnotes decreases the potential
for confusion and incorrect actions, thereby enhancing the safe
operation of Susquehanna SES. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to add a footnote indicating the Station
Duty Manager shall act as PORC chairman in the absence of the Vice
President-Nuclear Operations will ensure continuous leadership of
PORC and will enhance performance by ensuring a responsible
individual is available during all shifts. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Changing Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report was previously approved
in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating
this change into footnote g of Table 4.11.2.1.2-1 will maintain
accuracy and consistency. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
For the reasons discussed in items I and II above, as well as
the enclosed Safety Assessment, 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: October 28, 1994
Description of amendment request: The amendment would delete the
requirements for chlorine detection and the associated Bases from the
Technical Specifications for each unit as a result of the removal of
bulk quantities of gaseous chlorine from the Susquehanna Steam Electric
Station. Specifically, Sections 3.3.7.8 and the associated Surveillance
Requirements in Section 4.3.7.8 would be deleted. In addition, Bases 3/
4.3.7.8 would also be deleted.
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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Review of the various design basis accidents identified in
Chapter 15 of the Susquehanna SES Final Safety Analysis Report
(FSAR) concluded that none of these accidents are affected by
deletion of the chlorine detection requirements from Technical
Specifications. With the elimination of bulk quantities of gaseous
chlorine from use at Susquehanna SES the probability of control room
inhabitability due to a gaseous chlorine release has actually
decreased. Therefore, this proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change involves only the deletion of the chlorine
detection system Technical Specifications based upon a plant
modification to remove gaseous chlorine as a biocide from
Susquehanna SES and replace it with a nonoxidizing biocide. The
release of chlorine from an off-site source is bounded by Reg. Guide
1.95 in that manual isolation capability for the control room
ventilation system is acceptable. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed change would not alter the margins of safety
provided in the existing FSAR analysis (Sections 2.2.3.1.3 and 6.4)
for chlorine release events since the basis for the existing margin
of safety, which are the Reg. Guide 1.95 requirements, are not
altered by the change. As stated above, since gaseous chlorine is no
longer used for open cooling water treatment at Susquehanna SES and
since the nonoxidizing biocide is relatively nontoxic to humans,
safety margin has actually increased. 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: November 11, 1994
Description of amendment request: The amendment would extend the
Main Turbine Valve surveillance test interval from a weekly basis to no
greater than 92 days for all Main Turbine Stop, Control, and Combined
Intermediate Valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed Technical Specification change to a quarterly
turbine inlet valve surveillance test interval is based on
maintaining the turbine missile generation probability within the
NRC criteria as stated in Table 3.1 of NUREG-1048. This, combined
with the NRC acceptable strike-and-damage probability as specified
in NUREG-1048, will keep the probability of unacceptable damage to
safety-related structures, systems, and components from turbine
missiles acceptably low (i.e., -7) . Thus, the NRC
acceptable risk rate of -7/yr. is not changed and there is
no increase in the probability of an accident previously evaluated.
The proposed Technical Specification change to the turbine inlet
valve surveillance interval does not effect the sequence of events
or the consequences of an accident previously evaluated. The
surveillance interval does not affect the strike and damage scenario
of an accident previously evaluated. Thus, the radiological
consequences of an accident previously evaluated will not be
increased.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed Technical Specification change to the turbine inlet
valve surveillance interval does not affect the surveillance test
characteristics. There are no new surveillance testing requirements.
Surveillance testing of these valves does not create the possibility
for a new or different kind of accident from any accident previously
evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed Technical Specification change to the turbine inlet
valve surveillance interval is based on maintaining the same margin
of safety as previously determined by the NRC and does not reduce
the margin of safety. In fact, the reduction in the testing rate
will reduce the potential for testing related transients, which have
been credited with causing 18 reactor scrams in the period 1985
through 1992.
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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 14, 1994

Description of amendment request: The proposed changes relocate
audit topics and frequencies, Nuclear Review Board review requirements
and requirements associated with the independent Safety Engineering
Group function from the Peach Bottom Atomic Power Station, Units 2 and
3 Technical Specifications to licensee controlled documents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:1)
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the changes relocate requirements from the TS to licensee
controlled documents consistent with the NRC Final Policy Statement
on TS Improvements. Any changes to the licensee controlled documents
will be evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59
as appropriate. Therefore, these changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because the changes will not alter the plant or the manner in which
the plant is operated. The changes will not involve a design change
or introduce any new failure modes. The changes will not alter
assumptions made in the safety analysis and licensing basis.
Adequate control of information will be maintained. Therefore, these
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
2) The proposed changes do not involve a significant reduction
in a margin of safety because they have no impact on any safety
analysis assumptions. The requirements to be transposed from the TS
to licensee controlled documents are the same as the existing TS.
Any future changes to licensee controlled documents will be
evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59 as
appropriate. Because the proposed changes are consistent with NUREG-
1433, as modified by approved generic change BWOG-09, and the change
controls for proposed relocated details and requirements provide an
equivalent level of regulatory authority, revising the TS to reflect
the approved level of detail and requirements ensures no 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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz

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

Date of amendment request: November 15, 1994 (TS 350)
Description of amendment request: The proposed change would remove
the frequency for each of the audits specified in the administrative
controls section of the technical specifications (TS). The requirements
to perform the audits would be retained, but the frequency for their
performance would be controlled by a requirement to be added to the
Nuclear Quality Assurance Plan. This would require that the audits
listed in the TS be performed on a biennial frequency. In addition, the
proposed change would remove the requirement to perform site
Radiological Emergency Plan and Physical Security/Safeguard Contingency
Plan reviews and audits from the TS, since these requirements presently
exist in the respective Plans.
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:
TVA has concluded that operation of BFN units 1, 2, and 3 in
accordance with the proposed change to the technical specifications
does not involve a significant hazards consideration. TVA's
conclusion is based on its evaluation in accordance with 10 CFR
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).
TVA's conclusion is based on the following:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by this Technical Specification change which only
affects review and audit frequencies. This Technical Specification
change will not impact the function or method of operation of plant
equipment. Thus, there is not a significant increase in the
probability of a previously analyzed accident due to this change. No
systems, equipment, or components are affected by the proposed
change. Thus, the consequences of a malfunction of equipment
important to safety previously evaluated in the UFSAR are not
increased by this change.
The proposed change only affects review and audit frequencies.
As such, the proposed change has no impact on accident initiators or
plant equipment, and thus, does not affect the probabilities or
consequences of an accident.
Therefore, we conclude that this change does not significantly
increase the probabilities or consequences of an accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since program audits do not contribute to
accident initiation, a change related to audit functions cannot
produce a new accident scenario or produce a new type of equipment
malfunction. Also, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation, and, thus, does not create the possibility of a new or
different kind of accident. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change concerning conduct of reviews and audits
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are no affected by this proposed
change.
Therefore, use of the proposed Technical Specification would not
involve any 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: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

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

Date of amendment request: November 15, 1994 (TS 94-12)
Description of amendment request: The proposed change would remove
the frequency for each of the audits specified in the administrative
controls section of the technical specifications (TS). The requirements
to perform the audits would be retained, but the frequency for their
performance would be controlled by a requirement to be added to the
Nuclear Quality Assurance Plan. This would require that the audits
listed in the TS be performed on a biennial frequency. In addition, the
proposed change would remove the requirement to perform site
Radiological Emergency Plan and Physical Security/Safeguard Contingency
Plan reviews and audits from the TS, since these requirements presently
exist in the respective Plans.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The standards used to arrive at a determination that a Technical
Specification change request involves no significant hazards
consideration are included in the Commission's regulations, 10 CFR
50.92, which states that no significant hazards considerations are
involved if the 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. Each standard is addressed as
follows:
1. Operation of the facility in accordance with the proposed
technical specifications would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by the Technical Specification change which only
affects review and audit frequencies. This Technical Specification
change will not impact the function or method of operation of plant
equipment. Thus, there is not a significant increase in the
probability of a previously analyzed accident due to this change. No
systems, equipment, or components are affected by the proposed
changes. Thus, the consequences of a malfunction of equipment
important to safety previously evaluated in the FSAR are not
increased by this change.
The proposed change only affects review and audit frequencies.
As such, the proposed change has no impact on accident initiators or
plant equipment, and thus, does not affect the probabilities or
consequences of an accident.
Therefore, we conclude that this change does not significantly
increase the probabilities or consequences of an accident.
2. Operation of the facility in accordance with the proposed
technical specifications would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since program audits do not contribute to
accident initiation, a change related to audit functions cannot
produce a new accident scenario or produce a new type of equipment
malfunction. Also, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation, and, thus, does not create the possibility of a new or
different kind of accident. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3. Operation of the facility in accordance with the proposed
technical specifications would not involve a significant reduction
in a margin of safety.
The proposed change concerning conduct of reviews and audits
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are no affected by this proposed
change.
Therefore, use of the proposed Technical Specification would not
involve any reduction in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, 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: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

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

Date of amendment request: September 8, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 4.2.2.2, 4.2.2.4, and 6.9.19 to
incorporate a penalty in the Core Operating Limit Report (COLR) to
account for FQ increases greater than 2 percent between
measurements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes to the Technical Specifications do not
involve a significant hazards consideration because operation of
Callaway Plant in accordance with these changes would not:
1) Involve a significant increase in the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report. There is no increase in the probability of
occurrence or the consequences of an accident. The removal of
FQ(Z) penalty values from the Callaway Plant Technical
Specifications and the creation of cycle-specific FQ(Z) values
in the COLR has no influence or impact on the probability or
consequences of any accident previously evaluated. The cycle-
specific FQ(Z) values, although not in Technical
Specifications, will be followed in the operation of the Callaway
Plant. The proposed amendment still requires exactly the same
actions to be taken when or if FQ(Z) limits are exceeded as is
required by current Technical Specifications.
2) Create a possibility of a new or different kind of accident
from any previously evaluated in the safety analysis report. There
is no new type of accident or malfunction created and the method and
manner of plant operation will not change. As stated earlier, the
removal of the cycle-specific FQ(Z) value has no influence or
impact, nor does it contribute in any way to the probability or
consequences of an accident. No safety-related equipment, safety
function, or plant operation will be altered as a result of this
proposed change. The cycle-specific FQ(Z) values are calculated
using NRC approved methods. The Technical Specifications will
continue to require operation within the required FQ(Z) limits
and appropriate actions will be taken when or if limits are
exceeded.
3) Involve a significant reduction in a margin of safety. This
is based on the fact that no plant design changes are involved and
the method and manner of plant operation remains the same. The
margin of safety is not affected by change and removal of FQ(Z)
penalty values from the Technical Specifications. The margin of
safety presently provided by current Technical Specifications
remains unchanged. The current FQ(Z) limits remain unchanged
and the current safety analysis limits remain valid and unaffected
by this change. The proposed amendment continues to require
operation within the core limits as obtained from the NRC approved
design methodology and appropriate actions to be taken when or if
FQ(Z) limits are violated remain unchanged.
Given the abo

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A94-31196. Public record. Not legal advice.