[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

[[Page 3428]]

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