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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A96-4342

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** February 28, 1996
- **Citation:** 61 FR 7542

## Text

NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 5, 1996, through February 15, 1996.
The last biweekly notice was published on February 14, 1996 (61 FR
5809).

[[Page 7543]]

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 March 29, 1996, 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

[[Page 7544]]
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, Units Nos.
1, 2, and 3, Maricopa County, Arizona

Date of amendments request: December 20, 1995.
Description of amendments request: The proposed amendment would
change the instrumentation setpoint for the reactor trip and main steam
isolation signal (MSIS) actuation on low steam generator pressure from
greater than or equal to 919 psia with an allowable value of greater
than or equal to 911 psia to greater than or equal to 895 psia with an
allowable value of greater than or equal to 890 psia.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment does not involve any change to the method
of operation of any plant equipment that is used to mitigate the
consequences of an accident. The proposed change only affects the
instrument setpoint for steam generator low pressure reactor trip
and MSIS actuation. The proposed setpoint meets the requirement of
ensuring a reactor trip and MSIS actuation prior to steam generator
pressure reaching the analytical limits even under worst-case
accident conditions. Thus, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed amendment does not alter any of the assumptions or
bounding conditions currently in the UFSAR [updated final safety
analysis report] and meets the requirement of ensuring a reactor
trip and MSIS actuation prior to steam generator pressure reaching
the analytical setpoint under worst-case accident conditions. As a
result, the proposed amendment does not involve a significant
increase in the consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve any change to the method of
operation of any plant equipment that is used to mitigate the
consequences of an accident. Accordingly, no new failure modes have
been defined for any plant system or component important to safety
nor has any new limiting failure been identified as a result of the
proposed change. The intent of the proposed change is to increase
the margin between normal operating parameters and trip setpoints.
This minimizes the possibility of unnecessary challenges to safety
systems improving the safety of operation. The method of protecting
the facility for an excess steam demand event remains unchanged and
therefore, the amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is the implementation of a setpoint value
which was derived using methodologies endorsed by Revision 2 of NRC
Regulatory Guide 1.105, ``Instrument Setpoints.'' The new setpoint
ensures that sufficient margin exists below the full load operating
value for steam pressure so as not to interfere with normal plant
operation, but still high enough to provide the required protection
(reactor trip and main steam line isolation) in the event of an
excessive steam demand event. The new setpoint ensures that safety
margins are maintained within the results of existing calculations.
The margin of safety between the analyzed trip value and the point
at which safety analysis results become unacceptable remain
unchanged since the analytical setpoints are not affected by the
amendment. The new setpoint resulted from the reduced instrument
uncertainty and will ensure that the reactor trip and MSIS actuation
on low steam generator pressure will occur before the analyzed value
and hence, this change does not involve a significant reduction in
the 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
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.

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

Date of amendments request: January 5, 1996.
Description of amendments request: The proposed amendment would
revise paragraph 2.C.(1) of the operating licenses and Section 1.26 of
the TS for each of the three PVNGS Units to increase the authorized 100
percent reactor core power (rated thermal power) from 3800 megawatts
thermal (Mwt) to 3876 Mwt, an increase of 2 percent. The proposed
amendment would also revise TS 4.1.1.4, TS 3.1.3.4, and TS 3.2.6
(Figure 3.2-1) to lower the allowable reactor coolant system cold leg
temperature limits for each of the three PVNGS Units, and revise TS
3.4.2.1 and TS 3.4.2.2 to lower the pressurizer safety valve setpoints
for Units 1 and 3 to support the increased power operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment does not change the method of operation
or modify the plant configuration other than minor changes in
equipment setpoints. Thus no increase in the probability of an
accident is created by this amendment. System and programmatic
reviews have been performed on the nuclear

[[Page 7545]]
steam supply system controls, reactor coolant system mechanical, steam
generator mechanical, balance of plant systems, and fire protection,
equipment qualification, and probabilistic risk assessment programs.
The conclusion of these reviews was that operation in accordance
with the changes proposed in this amendment was acceptable and posed
no significant risk to the health and safety of the public. The
analyses supporting this amendment demonstrate that the consequences
of events using the changes specified in the amendment are within
the criteria which are the current licensing basis for the PVNGS
Units. Therefore the amendment, as proposed, does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not modify the configuration of the
units except for minor equipment setpoints. No equipment changes and
no new methods of plant operation are being proposed, therefore, no
new failure modes are introduced by the proposed amendment. The
setpoint changes proposed have been evaluated and shown to be
acceptable in providing their design function. The increased rated
thermal power and associated changes have been incorporated into the
safety analysis performed in support of this amendment request and
the results have been shown to be similar to those previously
obtained. No possibility of a new or different kind of accident from
any accident previously evaluated will be created as a result of the
proposed amendment.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The changes proposed were evaluated in the safety analysis
performed to justify the amendment request. Although the
consequences of some events increased slightly, the results continue
to meet the criteria which form the PVNGS licensing basis. The
programmatic and system reviews provide further assurance of the
capability of the units to continue to operate safely with the
changes proposed in this amendment. Therefore the amendment 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
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: January 29, 1996.
Description of amendment request: The proposed change would revise
the technical specifications (TS) table 4.1-3, item 4 to change the
frequency of main steam safety valve (MSSV) testing to that specified
in NUREG-1431, the improved ``Standard Technical Specifications,
Westinghouse Plants'' (one third of the MSSVs each refueling outage).
In addition, the licensee proposed adding the MSSV test acceptance
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 presented below:

1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Neither the valves' nor the system's configuration or functions
are being altered. The valves' setpoints and their ``as-left''
range, +/-1%, will not be changed. The changes are to the testing
frequency and the ``as found'' tolerance of the MSSV setpoint.
The proposed changes in testing frequency and the higher
tolerance are in the less conservative direction, but are not
significant for several reasons. First, the new standards are based
on the American Society of Mechanical Engineers (ASME) Boiler and
Pressure Vessel Code. The new standards have been accepted by the
nuclear industry and the NRC, and are referenced in the improved
Standard Technical Specifications. Based on a discussion with the H.
B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2 MSSV
manufacturer (i.e., Crosby), HBRSEP, Unit No. 2 has not experienced
more problems with the Crosby MSSVs than the nuclear industry in
general, thus, the new level of safety will be equivalent to that of
the nuclear industry. Second, if a MSSV does fail the surveillance
test, the proposed TS will require additional MSSVs to be tested.
This requirement provides assurance that testing will reveal
possible generic problems. The impact of the tolerance on the
Chapter 15 accidents was analyzed and found to be within acceptable
limits.
Since no Updated Final Safety Analysis Report (UFSAR) Chapter 15
accident analysis is significantly impacted by the proposed changes,
there would be no increase in the consequences of an accident
previously evaluated. The testing in accordance with the ASME Boiler
and Pressure Vessel Code will provide an adequate level of assurance
that the MSSVs will be able to perform their intended function;
therefore the probability of a previously evaluated accident is not
increased.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No new systems or equipment are involved with the proposed
changes; and the plant's configuration and operational procedures
are unaffected. Since the proposed changes do not impact the plant's
operation, it can not create a new or different kind of accident.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The change in testing frequency is in a less conservative
direction, but it is based on the ASME Code and the improved
Standard Technical Specifications. Since HBRSEP, Unit No. 2 has not
experienced a greater number of failures associated with these MSSVs
than the nuclear industry in general, the decrease in the MSSV
testing frequency will not significantly impact the margin of
safety. Also, analyses have been performed that demonstrate that the
impact of the setpoint tolerance change on the UFSAR Chapter 15
accident analysis results is not significant. 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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: David B. Matthews.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: January 31, 1996.
Description of amendment request: The proposed change would revise
the Technical Specifications section 4.4 to allow the use of 10 CFR
Part 50, Appendix J, Option B, Performance-Based Containment Leakage
Rate Testing. A new TS section 6.12 is proposed to describe the
containment leakage rate testing program, committing to meet 10 CFR
50.54(o) and 10 CFR Part 50, Appendix J, Option B for type A tests; and
to meet 10 CFR part 50, Appendix J, Option A, for types B and C tests.
The bases would be changed to reflect the proposed changes.

[[Page 7546]]

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 change does not involve a significant hazards
consideration for the following reasons.
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2
Type A testing history provides substantial justification for the
proposed test schedule change to one test in a 10 year period. Three
Structural Integrity Tests (SITs) and seven Integrated Leak Rate
Tests (ILRTs) have been performed with acceptable results. Previous
testing has affirmed the acceptable reliability of the containment
structure to minimize leakage as designed, and provides assurance
that its performance to continuously function as designed is not
challenged due to this test schedule extension to once in 10 years.
Therefore, this proposed change to the TS that revises the Type
A testing frequency does not involve an increase in the probability
of an accident previously evaluated.
This proposed change to revise the test schedule frequency does
not impact nor alter the design of any system, structure or
component. The limit on allowable leakage is not increased. Type A
testing provides periodic verification of the leak tight integrity
of the containment and the systems and components that penetrate the
containment structure.
NUREG-1493, ``Performance-Based Containment Leak-Test Program,''
provides the technical basis for the NRC's rulemaking to revise
containment leakage testing requirements for nuclear power reactors
in 10 CFR 50, Appendix J. Section 10.1.2 of NUREG-1493, ``Summary of
Technical Findings, Leakage-Testing Intervals,'' states the
following.
1. Reducing the frequency of Type A tests (ILRTs) from the
current three per 10 years to one per 20 years was found to lead to
an imperceptible increase in risk. The estimated increase in risk is
very small because ILRTs identify only a few potential containment
leakage paths that cannot be identified by Type B and C testing, and
the leaks found by Type A tests have been only marginally above
existing requirements.
2. Given the insensitivity of risk to containment leakage rate
and the small fraction of leakage paths detected solely by Type A
testing, increasing the interval between ILRTs is possible with
minimal impact on public risk.
Therefore, based on the previous Type A test results, the
proposed change does not involve a significant increase in the
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change only incorporates the performance based
testing approach authorized in 10 CFR 50, Appendix J, Option B, and
is justified based on previous plant-specific Type A test results.
Plant structures, systems, and components will not be operated in a
different manner as a result of this proposed change and no physical
modifications to equipment are involved. The interval extensions
allowed by Option B of 10 CFR 50, Appendix J, do not have the
potential for creating the possibility of new or different type of
accidents from those previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change does not change the allowable leak rate from
the containment, it only allows an extension of the interval between
the performance of Type A leak rate testing. NUREG-1493, which
provides the technical basis for the NRC's rulemaking to revise
containment leakage testing requirements for nuclear power reactors
in 10 CFR 50, Appendix J. Section 10.1.2 of NUREG-1493, ``Summary of
Technical Findings, Leakage-Testing Intervals,'' states the
following.
``1. Reducing the frequency of Type A tests (ILRTs) from the
current three per 10 years to one per 20 years was found to lead to
an imperceptible increase in risk. The estimated increase in risk is
very small because ILRTs identify only a few potential containment
leakage paths that cannot be identified by Type B and C testing, and
the leaks found by Type A tests have been only marginally above
existing requirements.
2. Given the insensitivity of risk to containment leakage rate
and the small fraction of leakage paths detected solely by Type A
testing, increasing the interval between ILRTs is possible with
minimal impact on public risk.''

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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: David B. Matthews.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: January 29, 1996.
Description of amendment request: The proposed change would revise
the technical specifications (TS) to: (1) add TS 4.6.1.5 to provide
criteria for 24-hour full-load testing of the emergency diesel
generators (EDGs) to be performed during each refueling outage; (2)
revise TS 4.6.1.2 to allow testing of the EDG protective bypasses
listed in TS 3.7.1.d to be done independent of the safety injection or
loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the
EDG protective bypass inspection and a requirement to inspect the EDGs
at least once every refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The proposed changes do not involve a significant hazards
consideration for the following reasons.
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability of an accident previously evaluated. The proposed
changes require additional testing of the EDGs and will change the
requirement for when the protective bypasses are tested. The
function of the EDGs remains unchanged. Since the additional testing
involves the EDGs, which are required to mitigate an accident and
are not involved in the initiation of an accident, the proposed
changes will not increase the probability of an accident.
The proposed changes do not involve a significant increase in
the consequences of an accident previously evaluated. The proposed
changes require additional testing to verify the reliability of the
EDGs and to show the EDGs can withstand maximum accident loading
conditions. The proposed changes will also require the testing of
the EDG protective bypasses to be accomplished during EDG outages
and not during the SI/LOOP testing during a refueling outage. The
ability of the EDGs to perform their accident mitigation function
remains unchanged. Therefore, the proposed changes will not increase
the consequences of an accident.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not create the possibility of a new kind
of accident from any previously evaluated. The proposed changes are
an enhancement to the EDG testing requirements. The most significant
change will require additional testing of the EDGs to demonstrate
adequate reliability and to determine if the EDGs can withstand
maximum accident loading conditions. The remaining changes will
augment the TS to allow on-line EDG inspections and testing. Since
the function of the EDGs remains unchanged and they are not the
initiator of an accident, the proposed changes will not

[[Page 7547]]
create the possibility of a new kind of accident from any previously
evaluated.
The proposed changes do not create the possibility of a
different kind of accident from any accident previously evaluated.
The proposed changes require additional testing of the EDGs (i.e.,
the 24 hour full-load test) and revise the requirement for testing
the EDG protective bypasses during the SI/LOOP testing. The
additional testing of the EDGs will demonstrate sufficient
reliability and determine if the EDGs can withstand maximum accident
loading conditions. The EDG protective bypasses will be statically
tested during an EDG outage thus preventing possible damage to
equipment from a transient if the protective bypass fails. The
function of the EDGs remains unchanged by these proposed changes.
Since the EDGs are required to mitigate an accident and are not the
initiators of an accident, the proposed changes will not create a
different kind of accident from any kind of accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes do not reduce the margin of safety as
defined in the TS. The proposed changes are being submitted as an
enhancement to the testing requirements outlined in the TS. The
changes include additional testing, revising the requirement to test
the engine protective bypasses during the SI/LOOP testing and
clarification of the periodicity of inspecting the EDGs. The
additional testing demonstrates increased reliability and determines
that the EDGs can cope with maximum accident loading. The remaining
proposed changes provide clarification as to when the EDG
inspections and testing are required. The ability of the EDGs to
perform their function will not be reduced. Therefore, the margin of
safety will not be reduced by the proposed changes.

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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: David B. Matthews.

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

Date of amendment request: December 6, 1995.
Description of amendment request: The proposed amendment would
change the technical specifications of these plants to incorporate 10
CFR Part 50, Appendix J, ``Primary Reactor Containment Leakage Testing
for Water-Cooled Power Reactors'', Option B.
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:

ComEd proposes to revise Byron Nuclear Power Station, Units 1
and 2 (Byron), and Braidwood Nuclear Power Station, Units 1 and 2
(Braidwood) Technical Specification (TS) Section 3/4.6.1, ``Primary
Containment,'' and the associated Bases to reflect recent changes to
Appendix J to 10 CFR 50, ``Primary Reactor Containment Leakage
Testing for Water-Cooled Power Reactors.'' The proposed revisions
include:
1. Adding TS Definitions 1.15.a for the maximum allowable
primary containment leakage rate (La) and 1.20.a for the
maximum calculated primary containment pressure (Pa). The
redundant definitions throughout TS Section 3/4.6.1 are deleted,
2. Adding numerous statements throughout TS Section 3/4.6.1 that
leak rate testing is performed in accordance with Regulatory Guide
(RG) 1.163, Revision 0, ``Performance-Based Containment Leak-Test
Program,'' and its referenced documents,
3. Deleting TS requirements that are taken verbatim from 10 CFR
50, Appendix J. The specific requirements will be placed in the
containment leakage rate test program in accordance with RG 1.163,
and its referenced documents, and
4. Clarifying Technical Specification Surveillance Requirement
(TSSR) 4.6.1.1.a for consistency with NUREG-1431, Revision 1,
``Standard Technical Specifications for Westinghouse Plants.''
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
10 CFR 50, Appendix J, has been amended to include provisions
regarding performance-based leakage testing requirements (Option B).
Option B allows plants with satisfactory Integrated Leak Rate
Testing (ILRT) performance history to reduce the Type A testing
frequency from three tests in ten years to one test in ten years.
For Type B and Type C tests, Option B allows plants to reduce
testing frequency based on the leak rate test history of each
component. In addition, Option B establishes controls to ensure
continued satisfactory performance of the affected penetrations
during the extended testing interval. To be consistent with the
requirements of Option B to 10 CFR 50, Appendix J, ComEd proposes to
include appropriate changes to the TSs that incorporate the
necessary revisions.
Some of the proposed changes represent minor curtailments to
current TS requirements, but are based on the requirements specified
by Option B to 10 CFR 50, Appendix J. Any such changes are
consistent with the current plant safety analyses and have been
determined to represent sufficient requirements for the assurance of
the reliability of equipment assumed to operate in the safety
analyses, or provide continued assurance that specified parameters
associated with containment integrity remain within their acceptance
limits. The other proposed changes maintain consistency with those
requirements specified by Option B to 10 CFR 50, Appendix J and are
consistent with the current plant safety analyses. Implementation of
these changes will provide continued assurance that specified
parameters associated with containment integrity will remain within
their acceptance limits, and as such, will not significantly
increase the probability or consequences of a previously evaluated
accident.
The associated systems affecting the leak rate integrity are not
assumed in any safety analyses to initiate any accident sequence;
therefore, the probability of occurrence of any accident previously
evaluated is not increased. In addition, the proposed changes to the
limiting conditions for operation and surveillance requirements for
such systems are consistent with the current 10 CFR 50, Appendix J,
requirements. The proposed changes maintain an equivalent level of
reliability and availability for all affected systems.
Maintaining allowable leakage within the analyzed limit assumed
for the accident analyses does not adversely affect either the
onsite or offsite dose consequences. Furthermore, containment
leakage is not an accident initiator. As such, there is no adverse
impact on the probability of accident initiators. Thus, there is no
significant increase in the probability or occurrence of any
previously analyzed accident, or increase the consequences of any
previously analyzed accident.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Option B of 10 CFR 50, Appendix J, specifies, in part, that a
Type A test may be conducted at a periodic interval based on the
performance of the overall containment system. Type A tests measure
both the containment system overall integrated leakage rate at the
containment pressure boundary and system alignments assumed during a
large break loss-of-coolant accident (LOCA), and demonstrate the
capability of the primary containment to withstand an internal
pressure load. The acceptable leakage rates are specified in the
TSs. For Type B and C tests, intervals are proposed for
establishment based on the performance history of each component.
Acceptance criteria for each component are based upon demonstration
that the leakage rates at design basis pressure conditions for
applicable penetrations are within the limits specified in the TSs.
The proposed changes reflect the requirements specified in the
amended 10 CFR 50, Appendix J, and are consistent with the current
plant safety analyses. Some minor curtailments of current TS
requirements are

[[Page 7548]]
based on generic guidance or similarly approved provisions for other
plants. These changes do not involve revisions to the design of the
plant. Some of the changes may involve revision in the testing of
components at the plant; however, these are in accordance with the
current plant safety analyses and provide for appropriate testing or
surveillance that is consistent with Option B to 10 CFR 50, Appendix
J. The proposed changes will not introduce new failure mechanisms
beyond those already considered in the current plant safety
analyses.
No new modes of operation are introduced by the proposed
changes. Surveillance requirements are changed to reflect
corresponding changes associated with Option B to 10 CFR 50,
Appendix J. The proposed changes maintain at least the present level
of operability of any such system that affects plant containment
integrity. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated. The associated systems that affect plant leak
rate integrity related to the proposed amendment are not assumed to
initiate any accident sequence. In addition, the proposed
surveillance requirements for any such affected systems are
consistent with the current requirements specified within the TSs
and are consistent with the requirements of Option B to 10 CFR 50,
Appendix J. The proposed surveillance requirements maintain an
equivalent level of reliability and availability of all affected
systems and, therefore, do not affect the consequences of any
previously evaluated accident. As such, the probability of systems
associated with leak rate test integrity failing to perform their
intended function is unaffected by the proposed limiting conditions
for operation and surveillance requirements.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The provisions specified in Option B to 10 CFR 50 Appendix J,
allows changes to Type A, B, and C test intervals based upon the
performance of past leak rate tests. The effect of extending
containment leak rate test intervals is a corresponding increase in
the likelihood of containment leakage. The degree to which intervals
can be extended has a direct impact on the potential effect on
existing plant safety margins and the public health and safety that
can occur due to an increased likelihood of containment leakage.
Changing Type A, B, and C test intervals from those currently
provided in the TS to those provided for in 10 CFR 50, Appendix J,
Option B, slightly increases the risk associated with Type A, B, and
C specific accident sequences. Historical data suggest that
increasing the Type C test interval can slightly increase the
associated risk; however, this is compensated by the corresponding
risk reduction benefits associated with reduction in component
cycling, stress, and wear associated with increased test intervals.
In addition, when considering the total integrated risk, which
includes all analyzed accident sequences, the additional risk
associated with increasing test intervals is negligible.
The proposed changes are consistent with those provisions
specified in Option B of 10 CFR 50, Appendix J, and are consistent
with current plant safety analyses. In addition, these proposed
changes do not involve revisions to the design of the plant. As
such, the proposed individual changes will maintain the same level
of reliability of the equipment associated with containment
integrity, assumed to operate in the plant safety analysis, or
provide continued assurance that specified parameters affecting
plant leak rate integrity, will remain within their acceptance
limits. Therefore, the proposed changes provide continued assurance
of the leakage integrity of the containment without adversely
affecting the public health and safety and, as such, will not
significantly reduce existing plant safety margins.
The proposed changes are based on United States Nuclear
Regulatory Commission (USNRC) accepted provisions and maintain
necessary levels of system or component reliability affecting plant
containment integrity. The performance-based approach to leakage
rate testing concludes that the impact on public health and safety
due to revised testing intervals is negligible. The proposed changes
will not reduce the availability of systems associated with
containment integrity when they are required to mitigate accident
conditions; therefore, the proposed changes do not involve a
significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut

Date of amendment requests: December 4, 19, 19, 20, 20, and 20,
1995.
Description of amendment request: Each proposed amendment would
change the surveillance requirement frequency from the current once per
18-month interval to once per 24-month which is the proposed length of
a Haddam Neck refueling cycle. The changes pertain to the following
equipment:
December 4, 1995, Reactivity control systems flow paths, rod
position indication system, and Rod drop time.
December 19, 1995, Containment Air Recirculation System.
December 19, 1995, Main steam line (MSL) Code Safety Valves self
actuation, auxiliary feedwater system, service water system, snubber
testing, feedwater isolation valve actuation, and primary auxiliary
building cleanup system.
December 20, 1995, reactor coolant system (RCS) interlock,
containment sump, High Pressure Safety Injection Pump and Low Pressure
Safety Injection autostart and alignment, containment spray, and PH
control.
December 20, 1995, Trip actuating devices and channel trips,
reactor trip system, reactor trip system instrumentation, and accident
monitoring instrumentation.
December 20, 1995, RCS flow indicators, Loop stop valve interlock,
Pressurizer code safety valves, Emergency power supply for the
pressurizer heaters, Containment main sump and volume control tank
(VCT) level monitoring system, RCS pressure boundary valves, Low
temperature overpressure protection (LTOP) system, and RCS vent path.
Basis for proposed no significant hazards consideration
determination: The Commission has made a proposed determination that
the amendment request involves 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. As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to surveillance requirements of the Haddam
Neck Plant Technical Specifications extend the frequency for checking
the operability of the affected components/equipment. The proposal
would extend the frequency from at least once per 18 months to at least
once each refueling interval (i.e., nominal 24-months).

[[Page 7549]]

Changing the frequency of surveillance requirements from at least
once per 18 months to at least once each refueling interval does not
change the basis for the frequency. The frequency was chosen because of
the need to perform this verification under the conditions that apply
during a plant outage, and to avoid the potential of an unplanned
transient if the surveillance were conducted with the plant at power.
The proposed changes do not alter the intent or method by which the
surveillance are conducted, do not involve any physical changes to the
plant, do not alter the way any structure, system, or component
functions, and do not modify the manner in which the plant is operated.
As such, the proposed changes in the frequency of surveillance
requirements will not degrade the ability of the equipment/components
to perform its safety function.
Additional assurance of the operability of the components/equipment
is provided by additional surveillance requirements (e.g., monthly or
quarterly surveillance).
Equipment performance over the last four operating cycles was
evaluated to determine the impact of extending the frequency of
surveillance requirements. This evaluation included a review of
surveillance results, preventive maintenance records, and the frequency
and type of corrective maintenance. It concluded that there is no
indication that the proposed extension could cause deterioration in the
condition or performance of any of the subject components.
In addition to the substantive changes, there are format changes
which are merely editorial and because format changes produce no
physical change they do not influence the probability or consequences
of accidents.
Since the proposed changes only affect the surveillance frequency
for safety systems that are used to mitigate accidents, the changes
cannot affect the probability of any previously analyzed accident.
While the proposed changes can lengthen the intervals between
surveillance, the increases in intervals has been evaluated and it is
concluded that there is no significant impact on the reliability or
availability of the safety system and consequently, there is no impact
on the consequences on any analyzed accident.
2. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes to surveillance requirements of the Haddam
Neck Plant Technical Specifications extend the frequency for verifying
the operability of the affected components/equipment. The proposal
would extend the frequency from at least once per 18 months to at least
once each refueling interval (nominal 24 months).
Changing the frequency of surveillance requirements from at least
once per 18 months to at least once each refueling interval does not
change the basis for the frequency. The frequency was chosen because of
the need to perform this verification under the conditions that apply
during a plant outage, and to avoid the potential of an unplanned
transient if the surveillance were conducted with the plant at power.
In addition to the substantive changes, there are format changes
which are merely editorial and because format changes produce no
physical change they do not influence the probability of new or
different types of accidents.
The proposed changes do not alter the intent or method by which the
surveillance are conducted, do not involve any physical changes to the
plant, do not alter the way any structure, system, or component
functions, and do not modify the manner in which the plant is operated.
As such, the proposed changes cannot create the possibility of a new or
different kind of accident from any previously evaluated.
3. The changes do not involve a significant reduction in a margin
of safety.
The proposed changes to surveillance requirements of the Haddam
Neck Plant Technical Specifications extend the frequency for verifying
the operability of the components/equipment. The proposal would extend
the frequency from at least once per 18-months to at least once each
refueling interval (24-months).
In addition to the substantive changes, there are format changes
which are merely editorial and because format changes produce no
physical change they do not influence the margin of safety.
The proposed changes to surveillance frequency are still consistent
with the basis for the frequency, and the intent or method of
performing the surveillance is unchanged. Further, the current
inservice testing requirements and the previous history of reliability
of the system provides assurance that the changes will not affect the
reliability of the auxiliary feedwater system. Thus, it is concluded
that there is no impact on the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County and Northeast Nuclear Energy Company, et al.,
Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear Power
Station, Units 1, 2, and 3, New London County, Connecticut

Date of amendment request: June 6, 1995 (published August 2, 1995,
60 FR 39434), as supplemented November 22, 1995.
Description of amendment request: The proposed amendments will
modify the size of the Plant Operations Review Committee (PORC) which
will collectively have the experience and expertise in various areas of
plant operation, and will clarify the composition of the Site
Operations Review Committee (SORC).
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 (SHC), which is presented below:

. . . These proposed changes do not involve an SHC because the
changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The plant operations review committee (PORC) is an oversight
group and helps to ensure that the units are operated in a safe
manner. To accomplish this the PORCs provide their recommendations
on the safety related activities to the Vice President--Haddam Neck
Plant for Haddam Neck and to the respective Nuclear Unit Directors
for Millstone. Each Millstone Unit has its own PORC. It is proposed
that the members of the Millstone PORCs be selected by the
respective Nuclear Unit Director based on their knowledge and
expertise in specific key plant functions. The Millstone Station has
one site operations review committee (SORC). The SORC is also an
oversight group whose charter is to advise the Senior Vice
President--Millstone Station on all matters related to nuclear
safety at the Millstone site. The Haddam Neck Plant, being a single
unit site, has one PORC, which advises the Vice President--Haddam
Neck Plant. The members of the Haddam Neck Plant PORC will be
selected by the Vice President--Haddam Neck Plant based on their
knowledge and expertise in specific key

[[Page 7550]]
plant functions. The PORC and SORC add to the defense-in-depth concept
provided by the design, operation, maintenance, and quality
oversight by promoting excellence through the conduct of their
affairs and by maintaining a diligent watch over their
responsibilities.
These administrative changes will revise the composition section
of the technical specifications for the PORC members. Millstone Unit
individuals will be appointed by the Nuclear Unit Directors if the
individual meets one or more of the following areas of expertise:
Plant Operations, Engineering, Reactor Engineering, Maintenance,
Instrumentation and Controls, Health Physics, Chemistry, Work
Planning and Control, and Quality Services. The Haddam Neck Plant,
due to its broader scope of review also include an individual
experienced in Security and specific expertise in Electrical
Maintenance and Mechanical Maintenance. The individuals who will
serve on PORC shall continue to meet the criteria of ANSI N18.1-1971
along with the qualification requirements contained in the technical
specifications. This approach is consistent with the standard
technical specifications and NUREG 0800, Section 13.4. For SORC at
the Millstone Station, the method of identifying who shall serve as
Vice Chairperson has been modified for clarity. Finally, the
individual who shall represent Quality and Assessment Services shall
be modified to allow a qualified member of Quality and Assessment
Services to serve on SORC.
The remaining portions of the technical specifications related
to PORC and SORC are not being revised.
These modifications broaden the unit committee participation and
reflect current organizational positions and will not increase the
probability of occurrence or the 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 administrative enhancements to the composition of
the PORC and Millstone Station SORC will not affect the way in which
the units are physically operated. These administrative changes to
PORC and SORC continue to meet the guidelines of ANSI N18.7-1976.
The modifications to PORC and SORC continue to allow these groups to
provide a thorough review of activities at the units.
The proposed modification does not impact any initiating events,
and therefore, cannot create the possibility of any new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
These proposed administrative changes will not impact the margin
of safety provided by PORC and SORC. The PORC and SORC will continue
to be staffed by qualified individuals experienced in the operation
of the plants. These administrative changes will modify how the
composition of the PORC and SORC members are presented in the
technical specifications, but will not adversely impact their
ability to review and comment on operations at the units.

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: Russell Library, 123 Broad
Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and
the Learning Resources Center, Three Rivers Community-Technical
College, 574 New London Turnpike, Norwich, CT 06360, for Millstone 1,
2, and 3.
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.

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

Date of amendment request: November 22, 1995 (NRC-95-0124).
Description of amendment request: The proposed amendment would
modify the allowed out-of-service time for one onsite alternating
current (ac) electrical power division from 72 hours to 7 days. The
proposed amendment would also eliminate accelerated testing and special
reports as a result of diesel generator surveillance failures in
accordance with Generic Letter 94-01, ``Removal of Accelerated Testing
and Special Reporting Requirements for Emergency Diesel Generators,''
dated May 31, 1994.
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. Changing the out-of-
service time, surveillance frequency and reporting requirements for
emergency diesel generators (EDGs) will not affect the initiation of
an accident, since EDGs are not associated with any accident
initiation mechanism. The proposed changes will not impact the plant
design or method of EDG operation. The increased out-of-service time
has been evaluated to have only a small impact on plant risk.
Performing the EDG inspections during plant operations will decrease
plant risk during plant outages. Deleting the accelerated testing
provisions will not affect the consequences of an accident since the
implementation of a maintenance and monitoring program for EDGs
consistent with the provisions of the maintenance rule will assure
EDG performance as discussed in Generic Letter 94-01. Deleting
reporting requirements has no impact on consequences of an accident
since reporting has no accident effect. Based on the amount of
electrical system redundancy, the small increase in plant risk
during operations and the decrease in plant risk during outages,
this change will not result in a significant increase in the
probability or consequences of an accident.
2. The proposed changes do not create the possibility of a new
or different accident from any previously evaluated. The proposed
changes do not modify the plant design or method of diesel
operation. Therefore, no new accident initiator is introduced, nor
is a new type of failure created. For these reasons, no new or
different type of accident is created by these changes.
3. The proposed changes do not involve a significant reduction
in a margin of safety. Since implementation of a maintenance program
for the EDGs consistent with the Maintenance Rule will ensure that
high EDG performance standards are maintained, the accelerated
testing schedule is not needed to maintain the margin of safety.
Deleting reporting requirements has no impact on safety or margin of
safety. Increasing the allowed out-of-service time for one division
of onsite AC power will slightly increase EDG unavailability during
plant operation. However, this change does not impact the redundancy
of offsite power supplies, the allowed out-of-service time if both
divisions are inoperable, or the ability to cope with a station
blackout event. This request also does not change the Action
statement for AC electrical power systems required when the plant is
shutdown. The increase in core damage frequency was assessed to be
small by an evaluation using the plant PSA [probabilistic safety
assessment] for the operating condition. Enabling the diesel
generator inspections to be performed on-line will improve safety
while shutdown by reducing EDG out-of-service time during outages.
For these reasons, 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: John N. Hannon.

[[Page 7551]]

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

Date of amendment request: December 21, 1995 (NRC-95-0133).
Description of amendment request: The proposed amendment would
implement Option B of the recently revised 10 CFR Part 50 Appendix J in
a manner consistent with Regulatory Guide 1.163, ``Performance-Based
Containment Leak Test Program,'' and industry guidance contained in NEI
94-01, Revision 0, ``Industry Guideline for Implementing Performance-
Based Option of 10 CFR 50, Appendix J,'' with the exception of
previously approved exemptions which the licensee wishes to remain in
effect. The previously approved exemptions are for reduced pressure for
testing MSIVs [main steam isolation valves] and testing of LPCI [low
pressure coolant injection] isolation valves in accordance with
Technical Specification (TS) 4.4.3.2.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change implements the new Option B of 10 CFR Part
50 Appendix J on performance-based containment leakage testing. The
proposed change does not involve a change to the plant design or
operation. As a result, the proposed change does not affect any
parameters or conditions that contribute to the initiation of any
accidents previously evaluated. Thus, the proposed change cannot
increase the probability of any accident previously evaluated.
The proposed change potentially affects the leak-tight integrity
of the containment structure designed to mitigate the consequences
of a loss-of-coolant accident (LOCA). The function of the
containment is to maintain functional integrity during and following
the peak transient pressures and temperatures which result from any
loss-of-coolant accident (LOCA). The containment is designed to
limit fission product leakage following the design basis LOCA.
Because the proposed change does not alter the plant design, only
the frequency of measuring Type A, B, and C leakage, the proposed
change does not directly result in an increase in containment
leakage. However, decreasing the test frequency can increase the
probability that an increase in containment leakage could go
undetected for an extended period of time. Test intervals will be
established based on the performance history of components being
tested. The risk resulting from the proposed changes is
characterized as follows, based primarily on the results contained
in NUREG-1493 [''Performance-Based Containment Leakage Test
Program''], the principal Technical Support Document used by the NRC
as the basis for the Appendix J final rule (Reference 9 [of
application]) and the NRC's Final Regulatory Impact Analysis as
contained in SECY-95-181 [Final Regulatory Impact Analysis,
Performance-Based Containment Leakage-Test Program (Attachment 2 to
NRC Rulemaking Issue Affirmation, SECY-95-181 dated July 17, 1995,
Final Amendment to 10 CFR 50, Appendix J, ``Containment Leakage
Testing,'' to Adopt Performance-Oriented and Risk-Based Approaches)]
(Reference 10 [of application]):

Type A Testing

NUREG-1493 found that the effect of containment leakage on
overall accident risk is minimal since risk is dominated by accident
sequences that result in failure or bypass of the containment.
Industry wide, ILRTs [integrated leak rate tests] have only
found a small fraction of the leaks that exceed current acceptance
criteria. Only three percent of all leaks are detectable only by
ILRTs, and therefore, by extending the Type A testing intervals,
only three percent of all leaks have a potential for remaining
undetected for longer periods of time. In addition, when leakage has
been detected by ILRTs, the leakage rate has been only marginally
above existing requirements. The Fermi Type A testing confirms the
industry-wide experience that a majority of the leakage experienced
during Type A testing is through components tested by Type B and C
tests.
NUREG-1493 found that these observations, together with the
insensitivity of reactor accident risk to the containment leakage
rate, show that increasing the Type A leakage test intervals would
have a minimal impact on public risk.

Type B and C Testing

NUREG-1493 found that while Type B and C tests can identify the
vast majority (greater than 95 percent) of all potential leakage
paths, performance-based alternatives to current local leakage-
testing requirements are feasible without significant risk impacts.
The risk model used in NUREG-1493 suggests that the number of
components tested would be reduced by about 60 percent with less
than a three-fold increase in the incremental risk due to
containment leakage. Since, under existing requirements, leakage
contributes less than 0.1 percent of overall accident risk, the
overall impact is very small. In addition, the NRC's Final
Regulatory Impact Analysis concluded that while the extended testing
intervals for Type B and C tests led to minor increases in potential
offsite dose consequences, the beneficial expected decrease in
onsite (LLRT [local leak rate testing] & ILRT worker) dose exceeds
(by at least an order of magnitude) the potential off-site dose
consequences.
The editorial change to the bases has no impact on the
probability or consequence of an accident since it is strictly a
correction to achieve consistency between the bases and the
specifications.
Based on the above, DECO [the licensee] has concluded that the
proposed change will not result in a significant increase in the
probability or consequences of any accident previously evaluated.
2. The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. This change involves the reduction of
Type A, B, and C test frequency. Except for the method of defining
the test frequency, the methods for performing the actual tests are
not changed. No new accident modes are created by extending the
testing intervals. No safety-related equipment or safety functions
are altered as a result of this change. Extending the test frequency
has no influence on, nor does it contribute to, the possibility of a
new or different kind of accident or malfunction from those
previously analyzed.
The editorial change to the bases has no effect on any kind of
accident since it is strictly a correction to achieve consistency
between the bases and the specifications.
Based on the above, DECO has concluded that the proposed change
will not create the possibility [of] a new or different kind of
accident previously evaluated.
3. The request does not involve a significant reduction in a
margin to safety.
The proposed change only affects the frequency of Type A, B, and
C testing. Except for the method of defining the test frequency, the
methods for performing the actual tests are not changed. However,
the proposed change can increase the probability that an increase in
leakage could go undetected for an extended period of time. NUREG-
1493 has determined that, under several different accident
scenarios, the increased risk of radioactivity release from
containment is negligible with the implementation of these proposed
changes.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to containment
leakage rate. The containment isolation system is designed to limit
leakage to La, which is defined by the Fermi 2 Technical
Specifications to be 0.5 percent by weight of the containment air
per 24 hours at 56.5 psig (Pa). The limitation on containment
leakage rate is designed to ensure that total leakage volume will
not exceed the value assumed in the accident analyses at the peak
accident pressure (Pa). The margin to safety for the offsite
dose consequences of postulated accidents directly related to the
containment leakage rate is maintained by meeting the 1.0 La
acceptance criteria. The La value is not being modified by this
proposed Technical Specification change.
Except for the method of defining the test frequency, no change
in the method of testing is being proposed. The Type B and C tests
will continue to be done at full pressure (Pa) or greater with
the exception of the Main Steam Isolation Valves, which have an
approved exemption. Other programs are in

[[Page 7552]]
place to ensure that proper maintenance and repairs are performed
during the service life of the primary containment and systems and
components penetrating the primary containment.
The editorial change to the bases has no effect on the margin of
safety since it is strictly an editorial change to achieve
consistency between the bases and the specifications.
As a result, DECO has concluded that the proposed change will
not result in a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: John N. Hannon.

Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
No. 50-498, South Texas Project, Unit 1, Matagorda County, Texas

Date of amendment request: January 22, 1996.
Description of amendment request: The proposed amendment would
modify the steam generator tube plugging criteria in Technical
Specification 3/4.4.5, Steam Generators, and the allowable leakage in
Technical Specification 3/4.4.6.2, Operational Leakage, and the
associated Bases. The amendment would allow the implementation of
alternate steam generator tube plugging criteria for the tube support
plate (TSP)/tube intersections for Unit 1.
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?

Structural Considerations

Industry testing of model boiler and operating plant tube
specimens for free span tubing at room temperature conditions show
typical burst pressures in excess of 5000 psi for indications of
outer diameter stress corrosion cracking with voltage measurements
at or below the structural limit of 4.0 volts. One model boiler
specimen with a voltage amplitude of 19 volts also exhibited a burst
pressure greater than 5000 psi. Burst testing performed on one
intersection pulled from STP Unit 1 in 1993 with a 0.51 volt
indication yielded a measured burst pressure of 8900 psi at room
temperature. Burst testing performed on another intersection pulled
from STP Unit 1 in 1995 with a 0.48 volt indication yielded a
measured burst pressure of 9950 psi at room temperature.
The projected end-of-cycle (EOC) voltage compares favorably with
the 4.7 volt structural limit considering the EPRI [Electric Power
Research Institute] voltage growth rate for indications at STP.
Using the methodology of the NRC Generic Letter 95-05, the
structural limit is reduced by allowances for uncertainty and growth
to develop a beginning-of-cycle (BOC) repair limit which should
preclude EOC indications from growing in excess of the structural
limit. The non-destructive examination (NDE) uncertainty to be
applied per EPRI is approximately 20 percent. The EPRI recommended
growth allowance of 30 percent/EFPY [effective full power year] is
also to be applied. This growth value is conservative for STP Unit 1
based on previous inspection history. By adding NDE uncertainty
allowances and a crack growth allowance to the repair limit, the
structural limit can be validated. Therefore, the maximum allowable
BOC repair limit (RL) based on the structural limit of 4.7 volts can
be represented as:

RL + (0.20 x RL) + (0.45* x RL) = 4.7 volts, which yields RL of 2.85
volts.

* The 30% growth rate for 1 EFPY was scaled up to the cycle
length used at South Texas.

This repair limit (2.85 volts) reasonably could be applied for
APC [alternate plugging criteria] implementation to repair bobbin
indications greater than the 1.0 volt criterion specified by NRC
Generic Letter 95-05 and is independent of RPC [rotating pancake
coil-probe] confirmation of the indications. STP has chosen to use a
steam generator tube upper repair limit of 2.85 volts to assess tube
integrity for those bobbin indications which are above 1.0 volt but
do not have confirming RPC calls. This 2.85 volt upper limit for
non-confirmed RPC calls is consistent with the NRC Generic Letter
95-05. Since the upper bound for repair of non-confirmed RPC is
limited to a value far less than the structural limit associated
with a full alternate criteria, the establishment of the repair
limits are determined to be reasonable and conservative with respect
to the industry pulled tube data base used.

Leakage Considerations

As part of the implementation of APC, the distribution of EOC
cracking indications at the TSP intersections has been used to
calculate the primary-to-secondary leakage which is bounded by the
maximum leakage required to remain within applicable dose limits.
This limit was calculated using the Technical Specification RCS
[reactor coolant system] Iodine-131 transient spiking values
consistent with NUREG-0800. Application of the APC criteria requires
the projection of postulated MSLB [main steam line break] leakage
based on the projected EOC voltage distribution for the beginning of
cycle. Projected EOC voltage distribution is developed using the
most recent EOC eddy current results and a voltage measurement
uncertainty. Draft NUREG-1477 requires that all indications to which
APC is applied must be included in the leakage projection.
The projected MSLB leakage rate calculation methodology
prescribed in EPRI TR-100407 will be used to calculate the EOC
leakage. A Monte Carlo approach will be used to determine the EOC
leakage, accounting for all of the ECT [eddy current testing]
uncertainties, voltage growth, and an assumed probability of
detection (POD) of 0.6 for a 1.0 volt repair limit. The fitted
logarithmic function probability of leakage correlation will be used
to establish the STP MSLB leak rate used for comparison with a
bounding allowable leak rate in the faulted loop which would result
in radiological consequences which are within applicable dose
limits. Due to the relatively low voltage levels of indications at
STP and low voltage growth rates, it is expected that the actual
calculated leakage values will be far less than this limit.
Therefore, implementation of APC does not adversely affect steam
generator tube integrity and implementation will be shown to result
in acceptable dose consequences. The proposed amendment does not
result in any 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?
Implementation of the proposed steam generator tube alternate
plugging criteria for ODSCC [outer diameter stress corrosion
cracking] at the TSP intersections does not introduce any
significant changes to the plant design basis. Use of the criteria
does not provide a mechanism which could result in an accident
outside of the region of the TSP elevations since no ODSCC has been
identified outside the thickness of the TSPs. It is therefore
expected that for all plant conditions, neither a single or multiple
tube rupture event would occur in a steam generator where APC has
been applied.
Specifically, STP will implement, for Unit 1, a maximum leakage
rate of 150 gpd [gallons per day] per steam generator (SG) to help
preclude the potential for excessive leakage during all plant
conditions. The current technical specification limits on primary-
to-secondary leakage at operating conditions are 1 gpm [gallon per
minute] for all steam generators or 500 gpd for any one SG. The RG
[Regulatory Guide] 1.121 criterion for establishing operational
leakage rate limits governing plant shutdown is based upon leak-
before-break (LBB) considerations to detect a free span crack before
potential tube rupture as a result of faulted plant conditions. The
150 gpd limit is intended to provide for leakage detection and plant
shutdown in the event of an unexpected crack propagation resulting
in excessive leakage. RG 1.121 acceptance criteria for establishing
operating leakage limits are

[[Page 7553]]
based on LBB considerations such that plant shutdown is initiated if
the permissible crack is exceeded.
The predicted EOC leakage for STP is based on the calculated
growth rate and does not take credit for the TSP proximity during
normal operation. Thus, the 150 gpd limit provides for plant
shutdown prior to reaching critical crack lengths. Additionally,
this leak-before-break evaluation assumes that the entire crevice
area is uncovered during the secondary side blowdown of a MSLB.
Typically, it is expected for the vast majority of intersections
that only partial uncovery will occur. Thus, the proximity of the
TSP will enhance the burst capacity of the tube.
Steam generator tube integrity is continually maintained through
inservice inspection and primary-to-secondary leakage monitoring.
Any tubes falling outside the APC repair limits are removed from
service. Therefore, the possibility of a new or different kind of
accident from any accident previously developed is not created.
3. Does the change involve a significant reduction in a margin
of safety?
The use of the voltage based bobbin probe for dispositioning
ODSCC degraded tubes within TSP intersections by APC is demonstrated
to maintain steam generator tube integrity in accordance with the
requirements of RG 1.121. RG 1.121 describes a method acceptable to
the NRC staff for meeting GDCs [General Design Criterion] 14, 15,
31, and 32 by reducing the probability or the consequences of steam
generator tube rupture. This is accomplished by determining the
limiting conditions of degradation of steam generator tubing, as
established by inservice inspection, for which tubes with
unacceptable cracking are removed from service. Upon implementation
of the criteria, even under the worst case conditions, the
occurrence of ODSCC at the TSP elevation is not expected to lead to
a steam generator tube rupture event during normal or faulted plant
conditions. The EOC distribution of crack indications at the TSP
elevations will be confirmed to result in acceptable primary-to-
secondary leakage during all plant conditions and that radiological
consequences are not adversely impacted.
In addressing the combined effects of loss of coolant accident
(LOCA) and safe shutdown earthquake (SSE) on the steam generator
component (as required by GDC 2), it has been determined that tube
collapse may occur in the steam generators at some plants. This is
the case at STP as the TSP may become deformed as a result of
lateral loads at the wedge supports at the periphery of the plate
due to the combined effects of the LOCA rarefaction wave and SSE
loadings. The resulting secondary-to-primary pressure differential
on the deformed tubes may cause some of the tube to collapse.
There are two concerns associated with steam generator tube
collapse. First, the collapse of steam generator tubing reduces the
RCS flow area through the tubes. The reduction in flow area
increases the resistance to flow of steam from the core during a
LOCA which, in turn, may potentially increase peak clad temperature
(PCT). Second, there is a potential that through wall cracks in
tubes could sufficiently enlarge during tube deformation or
collapse, causing sufficient in-leakage of secondary water back to
the core which dilutes the poisoning effect of boron injection from
the emergency cooling system. Again, an increase in core PCT may
result.
Consequently, since the LBB methodology is applicable to the STP
reactor coolant loop piping, the probability of breaks in the
primary loop piping is sufficiently low that they need not be
considered in the structural design of the plant. The analysis
identified tubes located adjacent to wedge regions that are subject
to potential collapse during combined LOCA and SSE. These tubes will
be excluded from application of APC. Thus, existing tube integrity
requirements apply to these tubes and the margin of safety is not
reduced.
Implementation practices using the bobbin probe voltage based
tube plugging criteria bounds RG 1.83 considerations by:
(1) Using enhanced eddy current inspection guidelines consistent
with those used by EPRI in developing the correlations. This
provides consistency in voltage normalization,
(2) Performing a 100 percent bobbin coil inspection for all hot
leg tube support plate intersections and all cold leg intersections
down to the lowest cold leg tube support plate with outer diameter
stress corrosion cracking (ODSCC) indications. The determination of
the tube support plate intersections having ODSCC indications shall
be based on the performance of at least a 20% random sampling of
tubes inspected over their full length, and
(3) Incorporating RPC inspection for all tubes with larger
indications left in service. This further establishes the principal
degradation morphology as ODSCC.
Implementation of APC at TSP intersections will decrease the
number of tubes which must be repaired. Since the installation of
tube plugs (to remove ODSCC degraded tubes from service) reduces the
RCS flow margin, APC implementation will help preserve the margin of
flow that would otherwise be reduced.
For each cycle the projected EOC primary-to-secondary leak rate
allowed is bounded by a leak rate which limits the radiological
consequences of a EOC MSLB to within applicable dose limits.
Therefore, this change does not involve a significant reduction in
the margin to safety.
It is therefore concluded that the proposed license amendment
request does not result in a significant reduction in the margin of
safety as defined in the plant Final Safety Analysis Report or
Technical Specifications.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: William D. Beckner.

Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
No. 50-498, South Texas Project, Unit 1, Matagorda County, Texas

Date of amendment request: January 22, 1996.
Description of amendment request: The proposed amendment would
modify the steam generator tube plugging criteria in Technical
Specification 3/4.4.5, Steam Generators, and the associated Bases, to
allow the implementation of alternate steam generator tube plugging
criteria for the tube-to-tubesheet joints (known in the industry as F*)
for Unit 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Steam Generator section of Technical
Specifications do not affect any accident initiators or precursors
and do not alter the design assumptions for the systems or
components used to mitigate the consequences of an accident. The
requirements approved by the NRC will not be reduced by this
request. Since F* utilizes the ``as rolled'' tube configuration that
exists as part of the original steam generator design, all of the
design and operating characteristics of the steam generator and
connected systems are preserved. The F* joint has been analyzed and
tested for design, operating and faulted condition loadings in
accordance with Regulatory Guide 1.121 safety factors. At worst
case, a tube leak would occur with the result being a primary to
secondary leak.
Should a tube leak occur, the impact is bounded by the ruptured
tube evaluation submitted by STP for the Unit 1 operating license.
No new or unreviewed accident conditions are created by the use of
F* criteria. The potential for a tube rupture is not increased from
the original submittal, thus there is no impact on accidents
evaluated as the design basis. Therefore use of the F* criteria will
not increase the probability of occurrence of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.

[[Page 7554]]

The use of the proposed F* alternate plugging criteria will not
introduce significant or adverse changes to the plant design basis.
The failure of a tube which remained unplugged in accordance with
the F* criteria would result in a tube leak, which is a previously
analyzed condition. Since this leak would occur below the secondary
face of the tubesheet, its leak rate would be limited by the tube-
to-tubesheet interface. Qualification testing and previous
experience indicates that normal and faulted leakage would be well
below the technical specification limits creating no threat
associated with tube rupture type leakages. This conclusion is
consistent with previous F* programs approved and used at other
operating plants.
However, in the unlikely event the failed tube severed
completely at a point below the F* region, the remaining F* joint
would retain engagement in the tubesheet due to its length of
expanded contact within the tubesheet bore, preventing any
interaction with neighboring tubes. If the tube severs at a point
above the F* region, then it is covered by the tube rupture event as
a part of the UFSAR [Updated Final Safety Analysis Report]. Thus,
the possibility of a new or different type of accident from any
accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Based on previous responses (above), the protective boundaries
of the steam generator are preserved. A tube with degradation can be
kept in service through F* criteria which provided an un-degraded
expanded interface with the tubesheet and which satisfies all of the
necessary structural and leakage requirements in accordance with
Regulatory Guide 1.121 and the Technical Specifications. Since the
joint is constrained within the tubesheet bore there is no
additional risk associated with tube rupture. Since the UFSAR
analyzed accident scenarios remain bounding, the use of an F*
criteria does not reduce the margin of safety.
Thus, these changes do not involve a significant reduction in
the margin of safety. Therefore, based on the above evaluation, STP
has concluded that these changes do not involve any significant
hazards considerations.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner.

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: January 12, 1996 (AEP:NRC:1233).
Description of amendment requests: The proposed amendments would
modify technical specification section 4.4.11 to eliminate the
surveillance requirement (SR) demonstrating operability of the
emergency power supply for the pressurizer power-operated relief valves
(PORVs) and block 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:

Per 10 CFR 50.92, a proposed change does not involve significant
hazards consideration if the change does 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.

Criterion 1

The proposed change is consistent with NUREG-1431 [Standard
Technical Specifications Westinghouse Plants]. Due to the high
reliability and continued testing of the Class 1E power supply, we
conclude that the elimination of the SR will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.

Criterion 2

The proposed change does not involve the addition of any new
plant operation or procedures, and the elimination of the SR is
consistent with NUREG-1431. For these reasons, we believe that the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.

Criterion 3

The proposed change is consistent with NUREG-1431, and it does
not affect the acceptance criteria of any of the other PORV and
block valve tests currently performed. For these reasons, we believe
that the proposed amendment will not involve a significant reduction
in a margin of safety.

The NRC staff has reviewed the licensee's analysis and the
applicable Bases of the Standard Technical Specifications Westinghouse
Plants. The Bases for the applicable surveillance, 3.4.11.4, states
``This Surveillance is not required for plants with permanent 1E power
supplies to the valves.'' 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.

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

Date of amendment request: January 22, 1996.
Description of amendment request: The proposed change relocates the
containment isolation valve (CIV) list, Table 3.6-2, from the Technical
Specifications to the Technical Requirements Manual (TRM). This change
affects Technical Specifications Sections 1.8.1a, 4.6.1.1a, 3.6.3.1,
4.6.3.1.1 and 4.6.3.1.2, and the Basis Section 3/4.6.3. A note at the
bottom of Table 3.6-2 regarding the CIVs that are subject to
administrative control is retained in the Technical Specifications by
relocating it to Sections 1.8.1a and 4.6.1.1a. This change is being
performed in accordance with Generic Letter 91-08, which provides
guidance for removal of component lists from the Technical
Specifications.
Additionally, a change to provide relief in the surveillance
requirement in Section 4.6.1.1a is included. The change allows valves,
blind flanges, and deactivated automatic valves located inside the
containment and are locked, sealed, or otherwise secured in the closed
position to be verified closed during each cold shutdown but not more
often than once per 92 days. The current requirements check the valve
position once per 31 days.
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 (SHC), which is presented below:

Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO)
has reviewed the proposed changes. NNECO concludes that these
changes do not involve a significant hazards consideration (SHC)
since the proposed changes satisfy the criteria in 10CFR50.92(c).
That is, the proposed changes do not:

[[Page 7555]]

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to remove the Containment Isolation Valve
(CIV) list from the Technical Specifications will not result in any
hardware or operating changes. The proposed change is based upon NRC
Generic Letter (GL) 91-08 and merely removes the CIV table and all
references to the table from the technical specifications without
affecting the operability requirements of any of the listed valves.
The technical specifications will continue to require the CIVs to be
operable. Limiting Condition for Operation and surveillance
requirements for the valves will also remain in the technical
specifications. The CIV table will be relocated to the Millstone
Unit No. 2 Technical Requirements Manual (TRM) which is controlled
in accordance with 10CFR50.59.
This change is administrative in nature and does not involve an
increase in the probability or consequence of an accident previously
evaluated. Furthermore, the proposed change does not alter the
design, function, or operation of the valves involved, and therefore
does not affect the probability or consequences of any previously
evaluated accident.
The change to Section 4.6.1.1a that reduces the surveillance
requirement for valves, blind flanges, and deactivated automatic
valves located inside the containment provides consistency with
NUREG-1432, ``Standard Technical Specifications for Combustion
Engineering Plants'' as well as the Technical Specifications of
Millstone Unit No. 3, Haddam Neck Plant, and Seabrook. The
probability or consequences of any previously evaluated accidents
are not affected.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The change to relocate the CIV list from the technical
specifications to the TRM will not impose any different operational
or surveillance requirements, nor will the change remove any such
requirements. Adequate control of information will be maintained.
Furthermore, as stated above, the proposed change does not alter the
design, function, or operation of the valves involved, and therefore
no new accident scenarios are created.
The change to Section 4.6.1.1a that reduces the surveillance
requirement for valves, blind flanges, and deactivated automatic
valves located inside the containment does not alter the design,
function, or operation of the valves involved, and therefore no new
accident scenarios are created.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
it has no impact on any safety analysis assumption. The proposed
changes do not decrease the scope of equipment currently required to
be operable or subject to surveillance testing, nor does the
proposed change affect any instrument setpoints or equipment safety
functions.
The relocation of the valve list is consistent with the guidance
provided in GL 91-08. The change to the surveillance interval is
consistent with NUREG-1432, ``Standard Technical Specifications for
Combustion Engineering Plants'' as well as the Technical
Specifications of Millstone Unit No. 3, Haddam Neck Plant, and
Seabrook. The intent of the technical specification will be met
since the change will not alter function or operability requirements
for any CIV.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee.

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

Date of amendment request: January 17, 1996.
Description of amendment request: The amendment request would
delete a license requirement to submit responses to and to implement
requirements of Generic Letter 83-28, because the requirement has been
completed. Generic Letter 83-28 pertains to the Salem anticipated
transient without scram (ATWS) event.
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 change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
NNECO's proposal to delete License Condition 2.C(4) is an
administrative change. The NRC Staff has accepted Millstone Unit No.
3's responses regarding the actions required by GL 83-28, thus, the
license condition has been met and is no longer necessary. The
proposed change does not affect the configuration, operation, or
performance of any system, structure, or component. Additionally,
the limiting conditions for operation, limiting safety system
settings, and safety limits specified in the Millstone Unit No. 3
Technical Specifications are unchanged. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The NRC Staff has accepted Millstone Unit No. 3's responses
regarding the actions required by GL 83-28, thus, the license
condition has been met and is no longer necessary. The proposed
change to delete License Condition 2.C(4) does not affect the
configuration, operation, or performance of any system, structure,
or component. Additionally, the limiting conditions for operation,
limiting safety system settings, and safety limits specified in the
Millstone Unit No. 3 Technical Specifications are unchanged.
Therefore, this proposed change cannot create the possibility of a
new or different kind of accident from any previously analyzed.
3. Involve a significant reduction in the margin of safety.
The NRC Staff has accepted Millstone Unit No. 3's responses
regarding the actions required by GL 83-28, thus, the license
condition has been met and is no longer necessary. The proposed
change to delete License Condition 2.C(4) does not affect the
configuration, operation, or performance of any system, structure,
or component. Additionally, the limiting conditions for operation,
limiting safety system settings, and safety limits specified in the
Millstone Unit No. 3 Technical Specifications are unchanged.
Therefore, this 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.

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

Date of amendment request: December 22, 1995.
Description of amendment request: The proposed changes will revise
Limerick Generating Station, Units 1 and 2, Technical Specification
3.6.1.8 ``Drywell and Suppression Chamber Purge System,'' increasing
the Drywell and Suppression Chamber Purge System operating time limit
from 90 hours each 365 days to 180 hours each 365 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the

[[Page 7556]]
issue of no significant hazards consideration, which is presented
below:

1. The proposed Technical Specification [TS] changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
These TS changes do not increase the probability of occurrence
of an accident previously evaluated in the SAR [Safety Analysis
Report]. This activity involves changing the allowable operating
limit for the Drywell and Suppression Chamber Purge System from 90
hours each 365 days to 180 hours each 365 days. This change
increases the probability that this system will be in service should
a LOCA [loss of coolant accident] occur, but does not increase the
probability that a LOCA will occur.
Increasing the operating limit for the Drywell and Suppression
Chamber Purge System from 90 hours to 180 hours each 365 days does
not increase the consequences of a LOCA as previously evaluated in
the SAR. These proposed TS changes increase the probability of a
LOCA occurring during the time the Drywell and Suppression Chamber
Purge System is in operation, and therefore, increase the
probability of the failure of the operating SGTS [Standby Gas
Treatment System] filter bank. However, the risk to containment
integrity was previously evaluated and found to be acceptable (UFSAR
[Updated Final Safety Analysis Report] Section 9.4.5.1.2.2 and
WASH--1400 ``Reactor Safety Study'').
Increasing the duration that the vent/purge line isolation
valves may be open does not increase the probability that these
valves will not perform as designed (i.e., close upon receipt of an
isolation signal) in response to a LOCA. However, the changes will
increase the likelihood that the vent and purge valves will be
called on to close. As discussed in UFSAR Section 6.2.4.2, the
containment purge valves have undergone extensive testing and
analyses to demonstrate the operability of these valves following a
LOCA.
In addition to the existing Safety Analysis Report (SAR)
evaluations, a Level 2 PSA [Probabilistic Safety Assessment]
Analysis (containment failure) was performed to determine the
additional risk associated with changing the operating limit from 90
to 180 hours each 365 days. The PSA evaluation conservatively
assumed a 200 hour vent/purge duration per a 365 day period. The
figure of merit evaluated is the large early release frequency
(LERF) which represents the likelihood of containment failure
following core damage that could significantly affect the public
(e.g., release of a large amount of radioactive material early
enough in the accident that evacuation of the public has not
occurred). The 200 hour vent/purge duration increased the LERF
approximately 3% from the base value of 2.57E-8 for all PSA
initiators. This analysis concluded that the increase in risk of
containment failure is well within the bounds of the EPRI
[Electrical Power Research Institute] PSA Applications Guideline for
permanent changes. The same relative increase applies to the large
Design Basis Accident LOCA LERF.
These changes do not directly or indirectly degrade the
performance of any other safety systems (assumed to function in the
accident analysis) below their design basis. The potential for other
equipment failures in the reactor enclosure due to duct-work impact,
impingement, and the resulting environmental conditions was
evaluated. It was concluded that the environmental qualifications
for the LGS equipment are sufficient to ensure operability under the
predicted environmental conditions, and there is no impact or
impingement-related damage to essential equipment. Although the
probability of occurrence of a malfunction of equipment important to
safety is increased, the existing SAR analysis and Level 2 PSA
Analysis demonstrate the increased risk and radiological
consequences are not significant.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This activity does not change the function of the Drywell and
Suppression Chamber Purge System, the containment isolation system,
or SGTS as previously evaluated in the SAR. Changing the duration of
operation of the vent and purge system does not create an accident
initiator not considered in the SAR. Therefore, the possibility of
an accident of a different type is not created.
This activity does not create a failure mode not considered in
the SAR. All possible equipment failures that could occur as a
result of a LOCA during high volume purging have previously been
identified and evaluated in the SAR. Therefore, this activity does
not create the possibility of a different type of malfunction of
equipment important to safety.
Therefore, the proposed TS changes will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The Bases of Technical Specification 3.6.1.8 states that the
intent of the 90 hour per 365 day operating limit for the Drywell
and Suppression Chamber Purge System is to protect the integrity of
the SGTS filters. As discussed above, the requirements specified in
ODCM paragraph 3.3.6 assure the availability of the backup SGTS
filter train during operation of the vent and purge system.
Furthermore, as discussed above, revising the operating limit from
90 hours to 180 hours each 365 days does not involve a significant
increase in risk. The margin of safety as defined in the Bases of
Technical Specification 3.6.1.8 is maintained.
Therefore, the implementation of the proposed TS changes will
not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
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.

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

Date of amendment request: February 6, 1996.
Description of amendment request: The amendments would change the
Technical Specifications to lower the 125 Volt Battery Charger
surveillance amperage from at least 200 amps to at least 170 amps.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment will permit replacement of aging battery
chargers while ensuring these replacement battery chargers will
restore the battery from the design minimum charge to its fully
charged state while supplying normal steady-state loads. This meets
the design basis for the 125V DC system and is consistent with Salem
Unit 1 and 2 commitment to IEEE 308-1971 in UFSAR Section 3A.
The 125V DC battery chargers are not addressed as a contributor
to any accident analyzed in the UFSAR, therefore, changes to the
battery charger output current will not increase the probability of
an accident occurring.
The limiting analyzed accident considered in this proposed TS
amendment is the Loss of Offsite Power coincident with a Loss of
Coolant Accident. This is currently the limiting design duty cycle
for the batteries. The 125V batteries are sized to maintain all
emergency loads for a period of 2 hours without battery chargers.
This is demonstrated by performing the surveillance specified in TS
4.8.2.3.2.f, which is not being changed. Since the chargers are not
required to be available during this 2 hour period, and since the
proposed charging rate will supply the necessary loads following
restoration of AC power, the proposed amendment will have no effect
on the consequences of this accident.
The current limiter is calculated to extend the recharging time
from 20 hours to 30 hours, but this is not considered significant
since two, sequential battery discharge events are not considered
plausible.
PSE&G calculation substantiates the capability of the chargers
to restore the battery from the design minimum charge to its fully
charged state while supplying

[[Page 7557]]
normal steady-state loads following a Station Blackout (SBO) Event
which exceeds the current design duty cycle.
In addition, a review of 125V DC Battery System load profiles
indicated that the battery chargers are capable of supplying
expected loads when restoring the battery from a design minimum
charge state to a fully charged state irrespective of the status of
the plant.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed amendment does not result in any design or physical
configuration changes to the 125V DC system. This change supports
the installation of the replacement chargers and ensures the
chargers are surveilled within the bounds of limiting input
amperage. No changes are being made to the function, design basis,
or operation of the 125V DC system by this proposed change.
Therefore, the proposed amendment will not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The proposed amendment to TS 4.8.2.3.2.e ensures that the
replacement battery chargers have sufficient capacity to restore
each 125V battery from the design minimum charge to its fully
charged state while supplying normal steady-state loads. A margin of
safety is maintained on both the AC input and DC output of the
chargers since the specified current is above that required to
support the 125V DC system and will result in AC current below the
ampacity rating of the battery charger input cables.
Testing to a charger output current of at least 170 amps will
maintain a margin of safety to the current required during actual
worst case normal loading on the 125V DC buses.
Therefore, the proposed amendment will not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.

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

Date of amendment request: February 9, 1996.
Description of amendment request: The proposed amendment would
allow an installed overhead door assembly, to be used in lieu of the
equipment hatch closure, to isolate the hatch opening to the
containment building during fuel movement and core alterations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. Containment
closure is used with respect to the mitigation of fuel handling
accidents, and as such, any change to these requirements will not
affect the probability of an accident. The proposed changes will
also not result in a significant increase in the consequences of an
accident previously analyzed since the technical specification
requirements remain bounded by the fuel handling accident assumption
of no containment closure.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or changes in
the methods governing normal plant operation. The proposed changes
will not impose any new or different requirements. Thus, this change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. Containment closure is not assumed in the accident analyses
for Ginna Station. Also, the proposed change remains acceptable with
respect to SRP [NUREG-800, ``Standard Review Plan for the Review of
Safety Analysis Reports for Nuclear Power Plants, July 1981'']
15.7.4 and GDC [General Design Criterion] 19 requirements.
Therefore, no question of safety is involved, and the change does
not invo

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A96-4342. Public record. Not legal advice.