[NukeNet] PSEG WANTS TO CUT BACK ON STEAM GENERATOR TUBE INSPECTIONSr

[Norm Cohen]

I'd hate to let this go by without a fight. Marv, Dave, Paul G,
any technical thoughts?  Same question to Jim H and Paul P. on K-N

Norm


>From the Federal Register:

 

NRC:  PSEG Nuclear Llc, Exelon Generation Company, LLC; Notice of
 
                         FR Doc E7-1087
[Federal Register: January 25, 2007 (Volume 72, Number 16)]
[Notices] [Page 3427-3429] From the Federal Register Online via
GPO Access [wais.access.gpo.gov] [DOCID:fr25ja07-76]
 
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,
HEARINGDOCKET at NRC.GOV; 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

 
OGCMailCenter at nrc.gov. 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 pdr at nrc.gov. 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