[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