[Federal Register Volume 72, Number 16 (Thursday, January 25, 2007)]
[Notices]
[Pages 3427-3429]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-1087]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-272]
PSEG Nuclear Llc, Exelon Generation Company, LLC; Notice of
Consideration of Issuance of Amendment to Facility Operating License,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (NRC or the Commission) is
considering issuance of an amendment to Facility Operating License No.
DPR-70 issued to PSEG Nuclear LLC (the licensee) for operation of the
Salem Nuclear Generating Station (Salem), Unit No. 1, located in Salem
County, New Jersey.
The amendment request proposes a one-time change to the Technical
Specifications (TSs) regarding the steam generator (SG) tube inspection
and repair required for the portion of the SG tubes passing through the
tubesheet region. Specifically, for Salem Unit No. 1 refueling outage
18 (planned for spring 2007) and the subsequent operating cycle, the
proposed TS changes would limit the required inspection (and repair if
degradation is found) to the portions of the SG tubes passing through
the upper 17 inches of the approximate 21-inch tubesheet region.
Before issuance of the proposed license amendment, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act), and the Commission's regulations.
The Commission has made a proposed determination that the amendment
request involves no significant hazards consideration. Under the
Commission's regulations in Title 10 of the Code of Federal Regulations
(10 CFR), Section 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; (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, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Of the accidents previously evaluated, the proposed changes only
affect the steam generator tube rupture (SGTR) event evaluation and
the postulated steam line break (SLB) accident evaluation. Loss-of-
coolant accident (LOCA) conditions cause a compressive axial load to
act on the tube. Therefore, since the LOCA tends to force the tube
into the tubesheet rather than pull it out, it is not a factor in
this amendment request. Another faulted load consideration is a safe
shutdown earthquake (SSE); however, the seismic analysis of Model F
steam generators has shown that axial loading of the tubes is
negligible during an SSE.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) below 17 inches from the top of the
tubesheet is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region.
For the SGTR event, the required structural margins of the steam
generator tubes will be maintained by the presence of the tubesheet.
Tube rupture is precluded for cracks in the
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hydraulic expansion region due to the constraint provided by the
tubesheet. Therefore, the performance criteria of NEI [Nuclear
Energy Institute] 97-06, Rev. 2, ``Steam Generator Program
Guidelines'' and the Regulatory Guide (RG) 1.121, ``Bases for
Plugging Degraded PWR [pressurized-water reactor] Steam Generator
Tubes,'' margins against burst are maintained during normal and
postulated accident conditions. The limited inspection length of 17
inches supplies the necessary resistive force to preclude pullout
loads under both normal operating and accident conditions. The
contact pressure results from the hydraulic expansion process,
thermal expansion mismatch between the tube and tubesheet and from
the differential pressure between the primary and secondary side.
Therefore, the proposed change does not result in a significant
increase in the probability or consequence of a[n] SGTR.
The probability of a[n] SLB is unaffected by the potential
failure of a SG tube as the failure of a tube is not an initiator
for a[n] SLB event. SLB leakage is limited by leakage flow
restrictions resulting from the crack and tube-to-tubesheet contact
pressures that provide a restricted leakage path above the
indications and also limit the degree of crack face opening compared
to free span indications. The leak rate during postulated accident
conditions would be expected to be less than twice that during
normal operation for indications near the bottom of the tubesheet
(including indications in the tube end welds) based on the
observation that while the driving pressure increases by about a
factor of two, the flow resistance increases with an increase in the
tube-to-tubesheet contact pressure. While such a decrease is
rationally expected, the postulated accident leak rate is bounded by
twice the normal operating leak rate if the increase in contact
pressure is ignored. Since normal operating leakage is limited to
0.10 gpm [gallons per minute] (150 gpd [gallons per day]), the
attendant accident condition leak rate, assuming all leakage to be
from indications below 17 inches from the top of the tubesheet would
be bounded by 0.187 gpm. This value is bounded by the 0.35 gpm leak
rate assumed in Section 15.4.2, ``Major Secondary System Pipe
Rupture'' of the Salem Unit 1 Updated FSAR [Final Safety Analysis
Report (UFSAR)].
Based on the above, the performance criteria of NEI-97-06, Rev.
2 and draft RG 1.121 continue to be met and the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not introduce any changes or mechanisms
that create the possibility of a new or different kind of accident.
Tube bundle integrity is expected to be maintained for all plant
conditions upon implementation of the limited tubesheet inspection
depth methodology. The proposed changes do not introduce any new
equipment or any change to existing equipment. No new effects on
existing equipment are created nor are any new malfunctions
introduced.
Therefore, based on the above evaluation, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change maintains the required structural margins of
the steam generator tubes for both normal and accident conditions.
NEI 97-06, Rev. 2 and RG 1.121 are used as the basis in the
development of the limited tubesheet inspection depth methodology
for determining that steam generator tube integrity considerations
are maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC staff for meeting General Design Criteria 14,
15, 31, and 32 by reducing the probability and consequences of an
SGTR. RG 1.121 concludes that by determining the limiting safe
conditions of tube wall degradation beyond which tubes with
unacceptable cracking, as established by inservice inspection,
should be removed from service or repaired, the probability and
consequences of a[n] SGTR are reduced. This RG uses safety factors
on loads for tube burst that are consistent with the requirements of
Section III of the ASME [American Society of Mechanical Engineers
Boiler and Pressure Vessel] Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Reference 1 [Westinghouse
Report WCAP-16640-P, ``Steam Generator Alternate Repair Criteria for
Tube Portion Within the Tubesheet at Salem Unit 1,'' August 2006]
defines a length of non-degraded expanded tube in the tubesheet that
provides the necessary resistance to tube pullout due to the
pressure induced forces (with applicable safety factors applied).
Application of the limited tubesheet inspection depth criteria will
not result in unacceptable primary-to-secondary leakage during all
plant conditions.
Plugging of the steam generator tubes reduces the reactor
coolant flow margin for core cooling. Implementation of the 17[-
]inch inspection length at Salem Unit 1 will result in maintaining
the margin of flow that may have otherwise been reduced by tube
plugging.
Based on the above, it is concluded that the proposed changes do
not result in any reduction of margin with respect to plant safety
as defined in the [UFSAR] or bases of the plant Technical
Specifications.
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.
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 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example, in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D59, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Documents may be examined, and/or copied for a fee, at the NRC's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309,
[[Page 3429]]
which is available at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly-available records will be accessible from
the Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or a presiding officer designated by the Commission or by
the Chief Administrative Judge of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner intends to rely to establish those facts or expert opinion.
The petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, 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.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(c)(1)(i) through (viii).
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to Jeffrie J. Keenan,
Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ
08038, attorney for the licensee.
For further details with respect to this action, see the
application for amendment dated January 18, 2007, which is available
for public inspection at the Commission's PDR, located at One White
Flint North, File Public Area O1 F21, 11555 Rockville Pike (first
floor), Rockville, Maryland. Publicly-available records will be
accessible from the ADAMS Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.
Persons who do not have access to ADAMS or who encounter problems in
accessing the documents located in ADAMS, should contact the NRC PDR
Reference staff by telephone at 1-800-397-4209, 301-415-4737, or by e-
mail to [email protected].
Dated at Rockville, Maryland, this 19th day of January, 2007.
For the Nuclear Regulatory Commission.
Richard B. Ennis,
Senior Project Manager, Plant Licensing Branch I-2, Division of
Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
[FR Doc. E7-1087 Filed 1-24-07; 8:45 am]
BILLING CODE 7590-01-P