[Federal Register Volume 60, Number 48 (Monday, March 13, 1995)]
[Notices]
[Pages 13478-13481]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-6067]



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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-498]


Houston Lighting and Power Co., City Public Service Board of San 
Antonio, Central Power and Light Co., City of Austin, TX; 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-6, issued to Houston Lighting & Power Company, et al., (the 
licensee) for operation of the South Texas Project (STP), Unit 1, 
located in Matagorda County, Texas.
    The proposed amendment would modify the steam generator tube 
plugging criteria in Technical Specification 3/4.4.5, Steam Generators, 
and the allowable leakage for Unit 1 in Technical Specification 3/
4.4.6.2, Operational Leakage, and the associated Bases.
    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 proposed 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 shows 
typical burst pressures in excess of 5000 psi for indications of 
outer diameter stress corrosion cracking with voltage measurements 
at or below the structural limit of 4.0 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 [[Page 13479]] from STP Unit 1 with a 0.51 volt 
indication yielded a measured burst pressure of 8900 psi at room 
temperature. It is noted that the industry burst pressure tests do 
not reflect the effects of temperature and material properties in 
terms of the realized reduction in strength. However, even 
correcting for the effects of temperature on material properties 
(which represents about 80% of the strength at room temperature from 
ASME [American Society of Mechanical Engineers] Code Section III 
Appendix 1 values) yields effective burst pressures of about 4000 
psi which is above the RG [Regulatory Guide] 1.121 limit of 3790 psi 
(1.43 times the MSLB [main steam line break] pressure differential) 
at 4.0 volts. The STP Unit 1 data point (0.51 volt) would yield an 
effective burst pressure of about 7100 psi, which is well above the 
95% lower tolerance limit (LTL) prediction per the burst correlation 
data used. Additional benefit is realized during normal operation 
since the proximity of the TSP [tube support plate] will reinforce 
the tube, further reducing the likelihood of tube burst.
    The projected end-of-cycle (EOC) voltage compares favorably with 
the 4 volt structural limit considering the EPRI [Electric Power 
Research Institute] voltage growth rate for indications at STP. 
Using the methodology of the NRC Draft Generic Letter 94-XX, 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 EPRI is approximately 21 percent. The EPRI recommended 
growth allowance of 35 percent/EFPY [effective full power years] is 
also applied. This growth value is conservative for STP Unit 1 based 
on previous inspection history. By adding NDE uncertainty allowances 
and a crack growth allowance to the repair limit, the structural 
limit can be validated. Therefore, the maximum allowable BOC repair 
limit (RL) based on the structural limit of 4 volts can be 
represented as:

RL + (0.20  x  RL) + (0.53*  x  RL) = 4 volts, which yields RL of 
2.3 volts.

    *The 35% growth rate for 1 EFPY was scaled up to the cycle 
length used at South Texas.

    This repair limit (2.3 volts) reasonably could be applied for 
APC [alternate plugging criteria] implementation to repair bobbin 
indications greater than the 1.0 volt criterion specified by NRC 
Generic Letter 94-XX and is independent of RPC [rotating pancake 
coil] confirmation of the indications. Houston Lighting & Power has 
chosen to use a steam generator tube upper repair limit of 2.3 volts 
to assess tube integrity for those bobbin indications which are 
above 1.0 volt but do not have confirming RPC calls. This 2.3 volt 
upper limit for non-confirmed RPC calls is consistent with the NRC 
Generic Letter 94-XX which establishes 2.7 volts as the upper limit 
for 3/4 tubing. Since the upper bound for repair of non-confirmed 
RPC is limited to a value far less than the structural limit 
associated with full alternate criteria, the establishment of the 
repair limits are determined to be reasonable and conservative with 
respect to the industry pulled tube data base used.

Leakage Considerations

    As part of the implementation of APC, the distribution of EOC 
cracking 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 applicable dose limits (10 
CFR 100, NUREG-0800 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. Applications of the APC criteria requires the projection 
of postulated MSLB leakage based on the projected EOC voltage 
distribution for the beginning of the cycle. Projected EOC voltage 
distribution is developed using the most recent EOC eddy current 
results and a voltage measurement uncertainty. Draft NUREG-1477 
requires that all indications to which APC is applied must be 
included in the leakage projection.
    The projected MSLB leakage rate calculation methodology 
prescribed in EPRI TR-100407 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 ECT [eddy current testing] 
uncertainties, voltage growth, and an assumed probability of 
detection (POD) of 0.6 for a 1.0 volt repair limit. The fitted 
logarithmic function probability of leakage correlation will be used 
to establish the STP MSLB leak rate used for comparison with a 
bounding allowable leak rate in the faulted loop which would result 
in radiological consequences which are within applicable dose 
limits. Due to the relatively low voltage levels of indications at 
STP and low voltage growth rates, it is expected that the actual 
calculated leakage values will be far less than this limit. 
Currently, the leakage projected for EOC-05 at STP Unit 1 is 0.02 
gpm [gallons per minute] (<21 gpd [gallons per day]) which is 
negligible in comparison to the allowable limit.
    Therefore, implementation of APC does not adversely affect steam 
generator tube integrity and implementation will be shown to result 
in acceptable does consequences. The proposed amendment does not 
result in any increase in the probability of 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 alternate 
plugging 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 occur in a steam generator where 
APC has been applied.
    Specifically, Houston Lighting & Power 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 1.121 criterion for 
establishing operational leakage rate limits governing plant 
shutdown be based [sic] 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 the permissible crack is exceeded.
    The predicted EOC leakage for STP is based on a 35% growth rate 
and does not take credit for the TSP proximity during normal 
operation. The total current projected leakage for EOC 05 is 20.5 
gpd for the limiting SG (C) at STP Unit 1 which is considerably less 
than the 150 gpd limit. Thus, the 150 gpd limit provides for plant 
shutdown prior to reaching critical crack 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 APC repair 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 APC is demonstrated 
to maintain steam generator tube integrity 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 cracking 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 leak to a steam generator tube rupture event 
during normal or faulted plant conditions. The EOC distribution of 
crack indications at the TSP elevations will be confirmed to result 
in acceptable primary-to-secondary leakage during all plant 
conditions and that radiological consequences are not adversely 
impacted.
    In addressing the combined effects of loss of coolant accident 
(LOCA) and safe [[Page 13480]] shutdown earthquake (SSE) on the 
steam generator component (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 [loss of 
coolant accident] rarefaction wave and SSE loadings. The resulting 
secondary-to-primary pressure differential on the deformed tubes may 
cause some of the tube 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 on 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 cracks 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.
    Consequently, 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. The limiting LOCA 
event becomes either the accumulator, RHR [residual heat removal], 
or the pressurizer surge line break. The analysis identifies tube 
located adjacent to wedge regions that are subject to potential 
collapse during combined LOCA and SSE. These tubes will be excluded 
from application of APC. Thus, existing tube integrity requirements 
apply to these tubes and the margin of safety is not reduced.
    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 outer diameter 
stress corrosion cracking (ODSCC) indications. 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 larger 
indications left inservice. This further establishes the principal 
degradation morphology as ODSCC.
    Implementation of APC at TSP intersections will decrease the 
number of tubes which must be repaired. Since the installation of 
tube plugs (to remove ODSCC degraded tubes from service) reduces the 
RCS flow margin, APC implementation will help preserve the margin of 
flow that would otherwise be reduced.
    The projected EOC primary-to-secondary leakage rate allowed is 
bounded by a leak rate which limits the radiological consequences of 
a EOC MSLB to within applicable dose limits. 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 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, Washington, DC.
    The filing of requests for hearing for petitions for leave to 
intervene is discussed below.
    By April 12, 1995, 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, 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, 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 be set forth with particularly 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 [[Page 13481]] 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 award 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 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 March 1, 1995, which is 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, Texas 77488.

    Dated at Rockville, Maryland, this 7th day of March 1995.

    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. 95-6067 Filed 3-10-95; 8:45 am]
BILLING CODE 7590-01-M