[Federal Register Volume 60, Number 147 (Tuesday, August 1, 1995)]
[Notices]
[Pages 39189-39192]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-18805]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-327 and 328]
Sequoyah Nuclear Plant Units 1 and 2; 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 Nos.
DPR-77 and DPR-79 issued to the Tennessee Valley Authority (the
licensee) for operation of the Sequoyah Nuclear Plant, Units 1 and 2,
located in Soddy Daisy, Tennessee.
The proposed amendments would incorporate new requirements
associated with steam generator tube inspections and repair in the
Sequoyah Nuclear Plant, Units 1 and 2 Technical Specifications. The new
requirements would establish alternate steam generator tube plugging
criteria at the tube support plate intersections.
Before issuance of the proposed license amendments, 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 amendments 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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Testing of model boiler specimens for free-span tubing (no tube
support place restraint) at room temperature conditions shows burst
pressures in excess of 5,000 pounds per square inch (psi) for
indications of outer diameter stress corrosion cracking with voltage
measurements as high as 19 volts. Burst testing performed on
intersections pulled from SQN with up to a 1.9-volt indication shows
measured burst pressure in excess of 6,600 psi at room temperature.
Burst testing performed on pulled tubes from other plants with up to
7.5-volt indications shows burst pressures in excess of 5,200 psi at
room temperatures. Correcting for the effects of temperature on
material properties and minimum strength levels (as the burst
testing was done at room temperature), tube burst capability
significantly exceeds the safety-factor requirements of NRC
Regulatory Guide (RG) 1.121.
Tube burst criteria are inherently satisfied during normal
operating conditions because of the proximity of the tube support
plate (TSP). Since tube-to-tube support plate proximity precludes
tube burst during normal operating conditions, use of the criteria
must retain tube integrity characteristics that maintain a margin of
safety of 1.43 times the bounding faulted condition steam line break
(SLB) pressure differential. During a postulated SLB, the TSP has
the potential to deflect during blowdown following a main SLB,
thereby uncovering the TSP intersections.
Based on the existing database, the RG 1.121 criterion requiring
maintenance of a safety factor of 1.43 times the SLB pressure
differential on tube burst is satisfied by \7/8\-inch-diameter
tubing with bobbin coil indications with signal amplitudes less than
8.82 volts (WCAP-13990), regardless of the indicated depth
measurement. A 2.0-volt plugging criterion (resulting in a projected
end-of-cycle [EOC] voltage) compares favorably with the 8.82-volt
structural limit considering the extremely slow apparent voltage
growth rates and few numbers of indications at SQN. Using the
established methodology of RG 1.121, the structural limit is reduced
by allowances for uncertainty and growth to develop a beginning of
cycle (BOC) repair limit that would preclude indications at EOC
conditions that exceed the structural limit. The nondestructive
examination (NDE) uncertainty component is 20.5 percent, and is
based on the Electric Power Research Institute (EPRI) alternate
repair criteria (ARC).
Test data indicates that tube burst cannot occur within the TSP,
even for tubes that have 100 percent throughwall electro-discharge
machining notches, 0.75 inch long, provided that the TSP is adjacent
to the notched area. Because of the few number of indications at
SQN, the EPRI methodology of applying a growth component of 35
percent per effective full power year (EEPY) will be used. Near-term
operating cycles at SQN are expected to be bounded by 1.23 years,
therefore, a 43 percent growth component is appropriate. When these
allowances are added to the BOC alternate plugging criteria (APC) of
2.0 volts in a deterministic bounding EOC voltage of approximately
3.26 volts for a Cycle 7, operation can be established. A 5.56-volt
deterministic safety margin exists (8.82 structural limit--3.26-volt
EOC equal 5.56-volt margin).
For the voltage/burst correlation, the EOC structural limit is
supported by a voltage of 8.82 volts. Using this structural limit of
8.82 volts, a BOC maximum allowable repair limit can be established
using the guidance of RG 1.121. The BOC maximum allowable repair
limit should not permit the existence of EOC indications that exceed
the 8.82-volt structural limit. By adding NDE uncertainty allowances
and an allowance for crack growth to the repair limit, the
structural limit can be validated. Therefore, the maximum allowable
BOC repair limit (RL) based on the structural limit of 8.82 volts
can be represented by the expressions:
RL+(0.205 x RL)+(0.43 x RL)=8.82 volts, or, the maximum
allowable BOC repair limit can be expressed as,
RL=8.82-volt structural limit/1.64=5.4 volts.
This RL (5.4 volts) is the appropriate limit for APC
implementation to repair bobbin indications greater than 2.0 volts
independent of rotating pancake coil (RPC) confirmation of the
indication. This 5.4-volt upper limit for non-confirmed RPC calls is
consistent with other recently approved APC programs (Farley Nuclear
Plan, Unit 2).
The conservatism of the growth allowance used to develop the
repair limit is shown by the most recent SQN eddy current data. Two
tubes plugged in Unit 1 during the last outage had less than one
volt of growth over the past five operating cycles. Only seven tubes
in Unit 2 required repair because of outside diameter stress
corrosion cracking (ODSCC) at the TSP intersections.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main
SLB outside of containment, but upstream of the main steam isolation
valve (MSIV), represents the most limiting radiological condition
relative to the APC. Implementation of the APC will determine
whether the distribution of cracking indications at the TSP
intersections is projected to be such that primary-to-
[[Page 39190]]
secondary leakage would result in site boundary doses within a small
fraction of the 10 CFR part 100 guidelines. A separate analysis has
determined this allowable SLB leakage limit to be 4.3 gallons per
minute (gpm) in the faulted loop. This limit uses the TS reactor
coolant system (RCS) Iodine-131 activity level of 1.0 microcuries
per gram dose equivalent Iodine-131 and the recommended Iodine-131
transient spiking values consistent with NUREG-0800. The analysis
method is WCAP-14277, which is consistent with the guidance of the
NRC draft generic letter (GL) and will be used to calculate EOC
leakage. Because of the relatively low number of indications at SQN,
it is expected that the actual leakage values will be far less than
this limit. Additionally, the current Iodine-131 levels at SQN range
from about 25 to 100 times less than the TS limit.
Application of the criteria requires the projection of
postulated SLB leakage, based on the projected EOC voltage
distribution for Cycle 8 operation. Projected EOC voltage
distribution is developing using the most recent EOC eddy current
results and a voltage measurement uncertainty. Data indicates that a
threshold voltage of 2.8 volts would result in throughwall cracks
long enough to leak at SLB condition. The draft GL requires that all
indications to which the APC are applied must be included in the
leakage projection. Tube pull results from another plant with \7/8\-
inch tubing with a substantial voltage growth database have shown
that tube wall degradation of greater than 40 percent throughwall
was readily detectable either by the bobbin or RPC probe.
The tube with maximum throughwall penetration of 56 percent (42
average) had a voltage of 2.02 volts. The SQN Unit 1 pulled tube had
a 1.93-volt indication with a maximum depth of 91 percent and did
not leak at SLB condition. Based on the SQN pulled tube and industry
pulled tube data supporting a lower threshold for SLB leakage of 2.8
volts, inclusion of all APC intersections in the leakage model is
quite conservative. The ODSCC occurring at SQN is in its earliest
stages of development. The conservative bounding growth estimations
to be applied to the expected small number of indications for the
upcoming inspection should result in very small levels of predicted
SLB leakage. Historically, SQN has not identified ODSCC as a
contributor to operational leakage.
I order to assess the sensitivity of an indication's BOC voltage
to EOC leakage potential, a Monte Carlo simulation was performed for
a 2.0-volt BOC indication. The maximum EOC voltage (at 99.8 percent
cumulative probability) was found to be 4.8 volts. The leakage
component from an indication of this magnitude, using either the
NUREG-1477 or EPRI leakage models, is 0.12 or 0.028 gpm,
respectively.
Therefore, as implementation of the 2.0-volt APC does not
adversely affect steam generator (S/G) tube integrity and
implementation will be shown to result in acceptable dose
consequences, the proposed amendment does not result in 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 previously analyzed.
Implementation of the proposed S/G tube APC does not introduce
any significant changes to the plant design basis. Use of the
criteria does not provide a mechanism that could result in an
accident outside of the region of the TSP elevations; no ODSCC is
occurring outside the thickness of the TSP. Neither a single or
multiple tube rupture event would be expected in a S/G in which the
plugging criteria is applied (during all plant conditions).
TVA will implement a maximum leakage rate limit of 150 gallon
per day per S/G to help preclude the potential for excessive leakage
during all plant conditions. The SQN TS limits on primary-to-
secondary leakage at operating conditions include a maximum of 0.42
gpm (600 gallons per day [gpd]) for all S/Gs, or, a maximum of 150
gpd for any one S/G. The RG 1.121 criterion for establishing
operational leakage rate limits that require plant shutdown is based
upon leak-before-break considerations to detect a free-span crack
before potential tube rupture during faulted plant conditions. The
150-gpd limit should provide for leakage detection and plant
shutdown in the event of the occurrence of an unexpected single
crack resulting in leakage that is associated with the longest
permissible crack length. RG 1.121 acceptance criteria for
establishing operating leakage limits are based on leak-before-break
considerations such that plant shutdown is initiated if the leakage
associated with the longest permissible crack is exceeded. The
longest permissible crack is the length that provides a factor of
safety of 1.43 against bursting at faulted conditions maximum
pressure differential. A voltage amplitude of 8.82 volts for typical
ODSCC corresponds to meeting this tube burst requirement at a lower
95 percent prediction limit on the burst correlation coupled with
95/95 lower tolerance limit material properties. Alternate crack
morphologies can correspond to 8.82 volts so that a unique crack
length is not defined by the burst pressure versus voltage
correlation. Consequently, typical burst pressure versus through-
wall crack length correlations are used below to define the
``longest permissible crack'' for evaluating operating leakage
limits.
The single through-wall crack lengths that result in tube burst
at 1.43 times the SLB pressure differential and the SLB pressure
differential alone are approximately 0.57 inch and 0.84 inch,
respectively. A leak rate of 150 gpd will provide for detection of
0.4-inch-long cracks at nominal leak rates and 0.6-inch-long cracks
at the lower 95 percent confidence level leak rates. Since tube
burst is precluded during normal operation because of the proximity
of the TSP to the tube and the potential exists for the crevice to
become uncovered during SLB conditions, the leakage from the maximum
permissible crack must preclude tube burst at SLB conditions. Thus,
the 150-gpd limit provides for plant shutdown before reaching
critical crack lengths for SLB conditions. Additionally, this leak-
before-break evaluation assumes that the entire crevice area is
uncovered during blowdown. Partial uncover will provide benefit to
the burst capacity of the intersection.
As S/G tube integrity upon implementation of the 2.0-volt APC
continues to be maintained through in-service inspection and
primary-to-secondary leakage monitoring, the possibility of a new or
different kind of accident from any accident previously evaluated is
not created.
3. Involve a significant reduction in a margin of safety.
The use of the voltage based APC at SQN is demonstrated to
maintain S/G tube integrity commensurate with the criteria of RG
1.121. RG 1.121 describes a method acceptable to the NRC Staff for
meeting General Design Criteria (GDC) 14, 15, 31, and 32 by reducing
the probability or the consequences of S/G tube rupture. This is
accomplished by determining the limiting conditions of degradation
of S/G tubing, as established by in-service inspection, for which
tubes with unacceptable cracking should be removed from service.
Upon implementation of the criteria, even under the worst-case
conditions, the occurrence of ODSCC at the TSP elevations is not
expected to lead to a S/G 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
radiological consequences are not adversely impacted.
In addressing the combined effects of loss-of-coolant accident
(LOCA), plus safe shutdown earthquake (SSE) on the S/G component (as
required by GDC 2), it has been determined that tube collapse may
occur in the S/Gs at some plants. This is the case as the TSP may
become deformed as a result of lateral loads at the wedge supports
at the periphery of the plate because of the combined effects of the
LOCA rarefaction wave and SSE loadings. Then, the resulting pressure
differential on the deformed tubes may cause some of the tubes to
collapse.
There are two issues associated with S/G tube collapse. First,
the collapse of S/G 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 partial through-wall cracks in tubes could progress to through-
wall cracks during tube deformation or collapse.
Consequently, since the leak-before-break methodology is
applicable to the SQN 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 line break or the
pressurize surge line break. LOCA loads for the primary pipe breaks
were used to bound the conditions at SQN for smaller breaks. The
results of the analysis using the larger break inputs show that the
LOCA loads were found to be of insufficient magnitude to result in
S/G tube collapse or significant deformation. The LOCA, plus SSE
tube collapse evaluation performed for another plant with Series 51
S/Gs using
[[Page 39191]]
bounding input conditions (large-break loadings), is applicable to SQN.
Therefore, at SQN, no tubes will be excluded from using the voltage
repair criteria due to deformation of collapse of S/G tubes
following a LOCA plus an SSE.
Addressing RG 1.83 considerations, implementation of the bobbin
probe voltage based interim tube plugging criteria of 2.0 volt is
supplemented by: (1) Enhanced eddy current inspection quidelines to
provide consistency in voltage normalization, (2) a 100 percent eddy
current inspection sample size at the TSP elevations, and (3) RPC
inspection requirements for the larger indications left in service
to characterize the principal degradation as ODSCC.
As noted previously, implementation of the TSP elevation
plugging criteria will decrease the number of tubes that must be
repaired. The installation of S/G tube plugs reduces the RCS flow
margin. Thus, implementation of the alternate plugging criteria will
maintain the margin of flow that would otherwise be reduced in the
event of increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin of safety.
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 amendments 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 amendments involve 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 Pubic
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By August 31, 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 Chattanooga-Hamilton County Library, 1101
Broad Street, Chattanooga, Tennessee 37402. 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
[[Page 39192]]
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 Frederick J. Hebdon: 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 General
Council, Tennessee Valley Authority, ET 11H, 400 West Summit Hill
Drive, Knoxville, Tennessee 37902, 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 July 19, 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 Chattanooga-Hamilton County Library, 1101
Broad Street, Chattanooga, Tennessee 37402.
Dated at Rockville, MD, this 26th day of July 1995.
For the Nuclear Regulatory Commission,
David E. LaBarge, Sr.
Project Manager, Project Directorate II-3, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 95-18805 Filed 7-31-95; 8:45 am]
BILLING CODE 7590-01-M