[Federal Register Volume 61, Number 74 (Tuesday, April 16, 1996)]
[Notices]
[Pages 16651-16654]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-9325]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-498]
Houston Lighting and Power Company, et al.; 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 (the Commission) is
considering issuance of an amendment to Facility Operating License No.
NPF-76 issued to Houston Lighting and Power Company, et. al., (the
licensee) for operation of the South Texas Project (STP), Unit 1,
located in Matagorda County, Texas. The original application dated
January 22, 1996, was previously published in the Federal Register on
February 28, 1996, (61 FR 7552). That application was supplemented by
letter dated April 4, 1996.
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 steam generator voltage-based repair
criteria for the tube support plate (TSP)/tube intersections for Unit
1.
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 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, 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 current structural limit of 4.7 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 next projected end-of-cycle (EOC) voltage compares favorably
with the current structural limit considering the EPRI voltage
growth rate for indications at STP. Using the methodology of 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 Generic Letter 95-05 is approximately
20 percent. The growth allowance will be 30 percent/EPFY [effective
full power year] or a STP Unit 1 plant specific growth value, to be
calculated in accordance with Generic Letter 95-05, which ever is
greater. The use of 30%/EPFY growth is conservative when compared to
the actual STP growth experience. Each succeeding cycle upper
voltage repair limit will also be conservatively established based
on Generic Letter 95-05 methodology. By adding NDE uncertainty
allowances and a growth allowance to the repair limit, the
structural limit can be validated.
The upper voltage repair limit could be applied to bobbin coil
voltages between the lower and upper repair limits to leave such
indications in service independent of RPC [rotating pancake coil-
probe] confirmation. However, RPC confirmed indications will be
conservatively removed from service consistent with Generic Letter
95.05.
Leakage Considerations
As part of the implementation of voltage-based repair criteria,
the distribution of EOC degradation 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 the applicable dose limits of 10 CFR 100 and GDC [General
Design Criterion] 19. This limit was calculated using the Technical
Specification RCS [reactor coolant system] Iodine-131 transient
spiking values consistent with NUREG-0800. Application of the
voltage-based repair criteria requires the projection of postulated
MSLB [main steamline break] leakage based on the projected EOC
voltage distribution
[[Page 16652]]
from the beginning of cycle voltage distribution. Projected EOC
voltage distribution is developed using the most recent EOC eddy
current results and a voltage measurement uncertainty. Draft NUREG-
1477 and Generic Letter 95-05 require that all indications, to which
voltage-based repair criteria is applied, must be included in the
leakage projection.
The projected MSLB leakage rate calculation methodology
prescribed in Westinghouse WCAP-14277 or Generic Letter 95-05 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 bobbin
coil eddy current test uncertainties, voltage growth, and an assumed
probability of detection (POD) of 0.6. The fitted log-logistic
probability of leakage correlation will be used to establish the STP
MSLB leak rate for each cycle. This leak rate will be used for
comparison with a bounding allowable leak rate in the faulted loop
which would result in radiological consequences which are within the
dose limits of 10 CFR 100 for offsite doses and GDC 19 for control
room doses. Due to the relatively low voltage levels of indications
at STP to date and low voltage growth rates, it is expected that the
actual calculated leakage values will be far less than this limit
for each successive cycle.
Therefore, implementation of voltage-based repair criteria does
not adversely affect steam generator tube integrity and the
radiological consequences will remain below the limits of 10 CFR 100
and GDC 19. 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 voltage-
based repair 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 nor
multiple tube rupture event would likely occur in a steam generator
where voltage-based repair criteria has been applied.
Specifically, STP will implement, for Unit 1, a maximum leakage
rate of 150 gpd 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 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 based on LBB considerations such that plant shutdown is
initiated if permissible degradation is exceeded.
The predicted EOC leakage for STP is based on 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 degradation 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 voltage-based repair criteria 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 voltage-based
repair criteria is demonstrated to maintain steam generator tube
interity in accordance with the requirements of RG 1.121. RG 1.121
describes a method acceptable to the NRC staff for meeting GDCs 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 degradation 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
indications at the TSP elevations for each successive cycle will be
confirmed to result in acceptable primary-to-secondary leakage
during all plant conditions.
In addressing the combined effects of loss of coolant accident
(LOCA) and safe shutdown earthquake (SSE) on the steam generators,
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 tubes 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 degradation 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.
The analysis results in Framatome Technologies, Inc. Topical
Report, BAW 10204P, 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 voltage-
based repair criteria. Thus, existing tube integrity requirements
apply to these tubes and the margin of safety is not reduced. 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. 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 known outer
diameter stress corrosion cracking (ODSCC) indications at each
cycle. 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 bobbin
voltages greater than 1.0 volt. This further establishes the
principal degradation morphology as ODSCC.
Implementation of voltage-based repair criteria at TSP
intersections will decrease the number of tubes which must be
repaired at each subsequent inspection. Since the installation of
tube plugs, to remove ODSCC degraded tubes from service, reduces the
RCS flow margin, voltage-based repair criteria implementation will
help preserve the margin of flow.
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 the dose limits of 10 CFR 100
for offsite doses and GDC 19 for control room doses. 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 three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
[[Page 16653]]
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 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.
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, N.W., Washington,
D.C.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By May 15, 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, N.W., Washington, D.C., and at the local
public document room located at the Wharton County Junior College, J.
M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488. 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 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 party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to William D. Beckner, Director, Project Directorate
IV-1: 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
[[Page 16654]]
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to Jack R. Newman, Esq., Newman & Holtzinger, P.C., 1615 L Street, NW.,
Washington, DC 20036, attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the Commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
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 dated January 22, 1996, as supplemented by
letter dated April 4, 1996, which are 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
located at the Wharton County Junior College, J.M. Hodges Learning
Center, 911 Boling Highway, Wharton, TX 77488.
Dated at Rockville, Maryland, this 9th day of April 1996.
For The Nuclear Regulatory Commission.
Thomas W. Alexion,
Project Manager, Project Directorate IV-1, Division of Reactor Projects
III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 96-9325 Filed 4-15-96; 8:45 am]
BILLING CODE 7590-01-P
