[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