[Federal Register Volume 60, Number 187 (Wednesday, September 27, 1995)]
[Notices]
[Pages 49963-49968]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-23929]



-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457]


Commonwealth Edison Company; Notice of Consideration of Issuance 
of Amendments to Facility Operating Licenses, Proposed no Significant 
Hazards Consideration Determination, and Opportunity for A Hearing

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of amendments to Facility Operating License Nos. 
NPF-37, NPF-66, NPF-72, and NPF-77, issued to Commonwealth Edison 
Company for operation of the Byron Station, Units 1 and 2, located in 
Ogle County, Illinois and Braidwood Station, Units 1 and 2, located in 
Will County, Illinois.
    The proposed amendments would revise the present voltage-based 
repair criteria in the Byron 1 and Braidwood 1 Technical Specifications 
(TSs). These proposed revisions would raise the lower voltage limit 
from its present value of 1.0 volt to 3.0 volts; there would no longer 
be an upper voltage limit.
    The Braidwood 1 TSs were revised by License Amendment No. 54, 
issued on August 18, 1994, to add voltage-based repair criteria to the 
existing steam generator (SG) tube repair criteria. The Byron 1 TSs 
were revised in a similar manner by License Amendment No. 66, issued on 
October 24, 1994.
    The voltage-based repair criteria in the subject TSs are applicable 
only to a specific type of SG tube degradation which is predominantly 
axially-oriented outer diameter stress corrosion cracking (ODSCC). This 
particular form of SG tube degradation occurs entirely within the 
intersections of the SG tubes with the tube support plates (TSPs).
    The present voltage values for the ODSCC repair criteria are based 
on the assumption of a ``free span'' exposure of the SG tube flaw; 
i.e., no credit is given for any constraint against burst or leakage, 
which may be provided by the presence of the TSPs. This approach is, in 
turn, based on the assumption that under postulated accident 
conditions, the TSPs may be displaced sufficiently by blowdown 
hydrodynamic loads such that a SG tube flaw which was fully confined 
within the thickness of the TSP prior to the accident would then be 
fully exposed. This approach was first advanced by the NRC staff in a 
draft generic letter issued on August 12, 1994, which was subsequently 
modified slightly and issued as Generic letter (GL) 95-05, ``Voltage-
Based Repair Criteria For Westinghouse Steam Generator Tubes Affected 
by Outside Diameter Stress Corrosion Cracking,'' dated August 3, 1995. 
The previous license amendments related to the issue of ODSCC were 
based to a large extent on the draft generic letter cited above.
    The fundamental difference between the pending proposal to raise 
the lower voltage repair limit to 3.0 volts and the methodology 
contained in GL 95-05, is that the licensee proposes to install certain 
modifications to the SG internal structures, thereby limiting to a 
small value, the maximum displacement of the TSPs under accident 
conditions. The proposed structural modifications consist of expanding 
a limited number of SG tubes only on the hot leg side of the TSP, at 
each of the intersections of the tubes with the TSPs. The purpose of 
this approach would be to greatly reduce the probability of SG tube 
burst under postulated accident conditions by several orders of 
magnitude. There would be a negligible impact on the primary-to-
secondary SG tube leakage under accident conditions.
    While the voltage-based repair criteria for ODSCC flaws are 
applicable only to Byron 1 and Braidwood 1, the pending request for 
license amendments involves all four units in that both stations have a 
common set of TSs.
    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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The previously evaluated accidents of interest are steam 
generator tube burst and main steam line break [MSLB]. Their 
potential impact on public health and safety due to the change in SG 
tube plugging criteria proposed in this amendment request is very 
low as discussed below. Tube burst related to the types of cracks 
under 

[[Page 49964]]
consideration is precluded during normal operating plant conditions 
since the tube support plates are adjacent to the degraded regions 
of the tube in the tube to tube support plate crevices.
    During accident conditions, i.e., MSLB, the tubes and TSP may 
move relative to each other, which can expose a crack length portion 
to freespan conditions. Testing has shown that the burst pressure 
correlates to the crack length that is exposed to the freespan, 
regardless of the length that is still contained within the TSP 
bounds.
    Therefore, a more appropriate methodology has been established 
for addressing leakage and burst considerations that is based on 
limiting potential TSP displacements during postulated MSLB events, 
thus reducing the freespan exposed crack length to minimal levels. 
The tube expansion process to be employed in conjunction with this 
TS change is designed to provide postulated TSP displacements that 
result in negligible tube burst probabilities due to the minimal 
freespan exposed crack lengths.
    Thermal hydraulic modeling was used to determine TSP loading 
during MSLB conditions. A safety factor was conservatively applied 
to these loads to envelope the collective uncertainties in the 
analyses. Various operating conditions were evaluated and the most 
limiting operating condition was used in the analyses. Additional 
models were used to verify the thermal hydraulic results.
    Assessment of the tube burst probability was based on a 
conservative assumption that all hot-leg TSP intersections (32,046) 
contained throughwall cracks equal to the postulated displacement 
and that the crack lengths were located within the boundaries of the 
TSP. Alternatively, it was assumed that all hot-leg TSP 
intersections contained throughwall cracks with length equal to the 
thickness of the TSP. The postulated TSP motion was conservatively 
assumed to be uniform and equal to the maximum displacement 
calculated.
    The total burst probability for all 32,046 throughwall 
indications given a uniform MSLB TSP displacement of 0.31'' is 
calculated to be 1 x 10-5. This is a factor of 1000 less than 
the Generic Letter 95-05 burst probability limit of 1 x 10-2. 
Therefore, the functional design criteria for tube expansion is to 
limit the TSP motion to 0.31'' or less. However, the design goal for 
tube expansion limits the TSP MSLB motion to less than 0.1'', which 
results in a total tube burst probability of 1 x 10-10 for all 
32,046 postulated throughwall indications. Additional tubes will be 
expanded to provide redundancy to the required expansions.
    The structural limit for the hot-leg SG tube repair criteria 
with tube expansion is based on axial tensile loading requirements 
to preclude axial tensile severing of the tube. Axially oriented 
ODSCC does not significantly impact the axial tensile loading of the 
tube, therefore, the more limiting degradation mode with respect to 
affecting the tube structural limit at TSPs is cellular corrosion. 
Tensile tests that measure the force required to sever a tube with 
cellular corrosion and uncorroded cross sectional areas are used to 
establish the lower bound structural limit. Based upon these tests, 
a lower bound 95% confidence level structural voltage limit of 37 
volts was established for cellular corrosion. This limit meets the 
Regulatory Guide (RG) 1.121, ``Basis for Plugging Steam Generator 
Tubes,'' structural requirements based upon the normal operating 
pressure differential with a safety factor of 3.0 applied. Due to 
the limited database supporting this value, the structural limit was 
conservatively reduced to 20 volts. Accounting for voltage growth 
and Non-Destructive Examination (NDE) uncertainty, the full [interim 
plugging criteria] IPC upper limit exceeds 10 volts. However, for 
added conservatism a single voltage repair limit for hot-leg 
indications is specified in this request. All hot-leg indications 
with bobbin coil probe voltages greater than the hot-leg voltage 
repair limit will be plugged or repaired.
    The freespan tube burst probability must be calculated for the 
cold-leg TSP indications to be within the requirements of Generic 
Letter 95-05. The freespan structural voltage limit is calculated 
using correlations from the database described in Generic Letter 95-
05, with the inclusion of the recent Byron and Braidwood tube pull 
results. This structural limit is 4.75 volts. The lower voltage 
repair limit for cold-leg indications continues to be 1.0 volt. The 
upper voltage repair limit for cold-leg indications will be 
calculated in accordance with Generic Letter 95-05. Since flow 
distribution baffle indications are to be repaired to the 40% depth 
criteria, no leakage or burst analyses are required for these 
indications.
    Per Generic Letter 95-05, MSLB leak rate and tube burst 
probability analyses are required prior to returning to power and 
are to be included in a report to the Nuclear Regulatory Commission 
(NRC) within 90 days of restart. If allowable limits on leak rates 
and burst probability are exceeded, the results are to be reported 
to the NRC and a safety assessment of the significance of the 
results is to be performed prior to returning the steam generators 
to service.
    A postulated MSLB outside of containment but upstream of the 
Main Steam Isolation Valve (MSIV) represents the most limiting 
radiological condition relative to the IPC. The ODSCC voltage 
distribution at the TSP intersections are projected to the end of 
the cycle and MSLB leakage is calculated.
    A site specific calculation has determined the allowable MSLB 
leakage limit for Byron Unit 1 and Braidwood Unit 1. These limits 
use the recommended dose equivalent Iodine-131 transient spiking 
values consistent with NUREG-0800, ``Standard Review Plan'' and 
ensure site boundary doses are within a small fraction of the 10 CFR 
100 requirements. The projected MSLB leakage rate calculation 
methodology described in WCAP-14046, ``Braidwood Unit 1 Technical 
Support for Cycle 5 Steam Generator Interim Plugging Criteria,'' and 
WCAP 14277, ``SLB Leak Rate and Tube Burst Probability Analysis 
Methods for ODSCC at TSP Intersections,'' will be used to calculate 
end-of-cycle (EOC) leakage. This method includes a Probability Of 
Detection (POD) value of 0.6 for all voltage amplitude ranges and 
uses the accepted leak rate versus bobbin voltage correlation 
methodology (full Monte Carlo) for calculating leak rate, as 
described in Generic Letter 95-05. The database used for the leak 
and burst correlations is consistent with that described in Generic 
Letter 95-05 with the inclusion of the Byron Unit 1 and Braidwood 
Unit 1 tube pull results. The EOC voltage distribution is developed 
from the POD adjusted beginning-of-cycle (BOC) voltage distributions 
and uses Monte Carlo techniques to account for variances in growth 
and uncertainty.
    The Electric Power Research Institute (EPRI) leak rate 
correlation has been used. It is based on free span indications that 
have burst pressures above the MSLB pressure differential. There is 
a low but finite probability that indications may burst at a 
pressure less than MSLB pressure. With limited TSP motion due to 
tube expansion, the tube is constrained by the TSP and tube burst is 
precluded. However, the flanks of the crack open up to contact the 
Inside Diameter (ID) of the TSP hole and result in a primary-to-
secondary leak rate potentially exceeding that obtained from the 
EPRI correlation. This phenomenon is known as an Indication 
Restricted from Burst (IRB) condition.
    ComEd has performed laboratory testing to determine the bounding 
leak rate obtainable in an IRB condition. The bounding leak rate 
value was then applied in a leak rate calculation methodology that 
accounts for the MSLB leak rate contribution from IRB indications to 
the total MSLB leak rate calculated as described above. Results 
indicate that the IRB contribution to the total leak rate value is 
negligible, however, ComEd will conservatively add a leakage 
contribution due to IRBs in addition to the leakage calculated in 
accordance with Generic Letter 95-05. When this is done, the dose at 
the site boundary resulting from the predicted leakage is shown to 
be a small fraction (less than 10%) of 10 CFR 100 limits.
    Modification of the Byron and Braidwood Specifications for 
conformance with Generic Letter 95-05 requirements is primarily 
administrative and does not significantly increase the probability 
of any accidents previously evaluated. For Braidwood, the changes 
decrease the allowed burst probability from 2.5 x 10-2 to 
1.0 x 10-2. This change is in the conservative direction. Byron 
Station has previously incorporated this requirement.
    In addition, defense in depth is provided by lowering the Unit 1 
[reactor coolant system] RCS dose equivalent I-131 limit from 1.0 
Ci/gm to 0.35 Ci/gm. Based on current predictions 
of MSLB leakage at the time of SG replacement, the lower RCS dose 
equivalent I-131 limit also ensures that the resulting 2-hour dose 
rates at the Braidwood and Byron site boundaries will not exceed an 
appropriately small fraction of 10 CFR 100 dose guideline values.
    For these reasons, an increase in the IPC voltage repair limit 
to a maximum of 3.0 volts for the hot-leg support plate 
intersections does not adversely affect steam generator tube 
integrity and results in acceptable dose consequences. By 
effectively eliminating tube burst at hot-leg TSP intersections, the 
likelihood of a tube rupture is substantially reduced and the 
probability of occurrence of an accident previously evaluated is 
reduced. 

[[Page 49965]]

    This conclusion is not affected by recent foreign and domestic 
plant SG experiences. As the following evaluation shows, these 
experiences are not relevant to Byron and Braidwood. A foreign unit 
detected eddy current signal distortions in one area of the top tube 
support plate during a 1995 inspection. The steam generators had 
been chemically cleaned in 1992. Visual inspection showed that a 
small section of the top support plate had broken free and was 
resting next to the steam generator tube bundle wrapper. The support 
plate showed indications of metal loss. The chemical cleaning 
process used by the foreign unit was developed by the utility and 
differs significantly from the modified EPRI/SGOG process performed 
at Byron Unit 1 in 1994.
    The foreign process, coupled with specific application of the 
process, resulted in tube support plate corrosion of up to 250 mils 
compared to a maximum of 2.16 mils (11 mils maximum allowed) 
measured at Byron. During the Byron eddy current inspection 
performed after the chemical cleaning, no distortion of the tube 
support plate signals was reported. Therefore, these differences in 
cleaning processes imply that this foreign experience is irrelevant 
to the effects of the chemical cleaning process on the TSPs at 
Byron.
    A number of units have experienced TSP cracking associated with 
severe tube denting due to TSP corrosion at the tube to TSP crevice. 
WCAP 14273, Section 12.4, shows that a diametral reduction of 65 
mils is required to develop stress levels above yield in the TSP 
ligaments at dented intersections. The bobbin voltage associated 
with a 1 mil radial dent is 20 to 25 volts.
    Although, Byron Unit 1 and Braidwood Unit 1 have not seen 
corrosion-induced denting, an appropriately sized bobbin probe will 
be used as a go/no-go gauge to assess hot-leg dents, if they occur 
in the future. If a tube has a dent at a hot-leg intersection that 
fails to pass the go/no-go test probe, cold-leg repair criteria will 
be applied to the affected tube and the adjacent tubes. In this way, 
any indications at these locations will be treated as free-span 
indications for the purposes of burst and leakage evaluation, which 
is bounded by the existing 1.0 volt IPC analysis. IPC repair limits 
will not be applied to tubes with dents> 5.0 volts since they could 
mask a 1.0 volt signal. Tubes with corrosion-induced dents> 5.0 
volts and those tubes adjacent to such a tube will not be selected 
for tube expansion to preclude adverse effects of the failure of 
such a tube on limiting TSP displacement. Therefore, the denting 
experience at other plants is not relevant to Byron and Braidwood.
    A foreign utility's steam generators have experienced cracking 
at the top tube support plate. The cause of the cracking appears to 
be the configuration of the single anti-rotation device, connected 
between the steam generator shell and wrapper, and the wrapper 
internals. The single anti-rotation device carries the full load 
associated with wrapper to shell motion. This rotational load is 
believed to be transferred to the TSP via the wrapper internals. The 
Byron/Braidwood Unit 1 steam generator design (D-4) uses three anti-
rotation devices to spread the rotational load. The D-4 wrapper 
internals are configured such that this load is not directly 
transmitted to the TSP.
    No top support plate cracking has been detected at Byron Unit 1 
or Braidwood Unit 1 and very few (<1%) of the indications seen at 
Byron and Braidwood to date have been at the top TSP elevation.
    Nevertheless, an analysis was performed to assess the impact of 
cracking of the top support plate. The results show an increase in 
top support plate deflection for a very limited number of tubes to 
greater than the 0.10'' limit used in the 3.0 volt IPC analysis. The 
deflections of the lower support plates also increase, but remain 
within the 0.10'' limit. Thus, hot-leg indications in a cracked top 
TSP continue to be bounded by the existing analysis. ComEd will 
develop an inspection plan for the SG internals to identify if 
indications detrimental to the load path exist. If the inspection 
determines that indications detrimental to the integrity of the load 
path necessary to support the 3 volt IPC are found, the results are 
to be reported to the NRC and a safety assessment of the 
significance of the results is to be performed prior to returning 
the steam generators to service.
    A domestic utility reported several distorted TSP signals over 
the past three refueling outage tube inspections. It was determined 
that these signals were associated with the TSP geometry in an area 
where an access cover is welded into the TSP. These signal 
distortions are not attributed to TSP cracking or degradation. Since 
the distorted signals were due to TSP geometry which did not 
indicate or result in a defect of the TSP, there is no increase in 
the probability or consequences of an accident previously evaluated 
due to Byron Unit 1 and Braidwood Unit 1 steam generator TSP 
geometries which may result in distorted eddy current signals.
    One foreign unit observed a dislocation of the tube bundle 
wrapper when they were unable to pass sludge lancing equipment 
through a handhole in the wrapper. The dislocation appears to be a 
result of improper attachment of the wrapper to the support 
structure. Steam generator sludge lance operations have been 
successfully performed on Byron Unit 1 and Braidwood Unit 1 which 
indicates that no problem with wrapper attachment exists. The 
foreign unit's wrapper support design is significantly different 
than that used on Byron Unit 1 and Braidwood Unit 1. Therefore, a 
similar wrapper dislocation will not occur and the foreign 
experience is not applicable to Byron and Braidwood.
    Therefore, the proposed amendment does not result in any 
significant increase in the probability or consequences of an 
accident previously evaluated within the Byron Unit 1 and Braidwood 
Unit 1 Updated Final Safety Analysis Report (UFSAR).
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed steam generator tube plugging 
criteria with tube expansion 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 tube support plate elevations as ODSCC does not extend 
beyond the thickness of the tube support plates. Neither a single 
nor multiple tube rupture event would be expected in a steam 
generator in which the plugging criteria has been applied.
    The tube burst assessment involves a Monte Carlo simulation of 
the site specific voltage distribution to generate a total burst 
probability that includes the summation of the probabilities of 1 
tube bursting, 2 tubes bursting, etc. For the hot-leg TSP 
intersections, the maximum total probability of burst, by design, is 
estimated to be 1 x 10-10 with all tube expansions functional.
    Accounting for the unlikely event of expansion failures, a 
sufficient number of redundant expansions exist to ensure that the 
burst probability remains below 1 x 10-5. This includes the 
conservative assumption that all 32,046 hot-leg TSP intersections 
contain throughwall indications. This level of burst probability is 
considered to be negligible when compared to the Generic Letter 95-
05 limit of 1 x 10-2.
    In addressing the combined effects of Loss Of Coolant Accident 
(LOCA) + Safe Shutdown Earthquake (SSE) on the SG as required by 
General Design Criteria (GDC) 2, it has been determined that tube 
collapse may occur in the steam generators at some plants. The tube 
support plates may become deformed as a result of lateral loads at 
the wedge supports located at the periphery of the plate due to the 
combined effects of the LOCA rarefaction wave and SSE loadings. The 
resulting pressure differential on the deformed tubes may cause some 
of the tubes to collapse. There are two issues associated with SG 
tube collapse. First, the collapse of SG 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 throughwall cracks in tubes could 
progress to throughwall cracks during tube deformation or collapse. 
The tubes subject to collapse have been identified via a plant 
specific analysis and excluded from application of the voltage-based 
criteria. This analysis is included in revision 3 to WCAP-14046 
which was submitted to the NRC June 19, 1995.
    ComEd will continue to apply a maximum primary-to-secondary 
leakage limit of 150 gallons per day (gpd) through any one SG at 
Byron and Braidwood to help preclude the potential for excessive 
leakage during all plant conditions. The RG 1.121 criterion for 
establishing operational leakage limits that require plant shutdown 
are based on detecting a free span crack prior to resulting in 
primary-to-secondary operational leakage which could potentially 
develop into a tube rupture during faulted plant conditions. The 150 
gpd limit provides for leakage detection and plant shutdown in the 
event of an unexpected single crack leak associated with the longest 
permissible free span crack length.
    Tube burst is precluded during normal operation due to the 
proximity of the TSP to 

[[Page 49966]]
the tube and during a postulated MSLB event with tube expansion. The 
150 gpd limit provides a conservative limit for plant shutdown prior 
to reaching critical crack lengths should significant crack 
extension unexpectedly occur outside the thickness of the TSP.
    Lowering the Unit 1 RCS dose equivalent I-131 limit from 1.0 
Ci/gm to 0.35 Ci/gm is conservative and provides a 
defense in depth approach to implementation of this IPC.
    Based on current predictions of MSLB leakage at the time of SG 
replacement, the lower RCS dose equivalent I-131 limit also ensures 
that the resulting 2-hour dose rates at the Braidwood and Byron site 
boundaries will not exceed an appropriately small fraction of 10 CFR 
100 dose guideline values.
    Modification of the Byron and Braidwood Specifications for 
conformance with Generic Letter 95-05 requirements is primarily 
administrative and will not alter the plant design basis. For 
Braidwood, the decrease in the allowed burst probability from 
2.5 x 10-2 to 1.0 x 10-2 is conservative. Byron Station 
has previously incorporated this requirement.
    With implementation of an increased IPC voltage repair limit (up 
to a maximum of 3.0 volts) using tube expansion for the hot-leg 
support plate intersections, steam generator tube integrity 
continues to be maintained through inservice inspection, tube repair 
and primary-to-secondary leakage monitoring. By effectively 
eliminating tube burst at hot-leg TSP intersections, the potential 
for multiple tube ruptures is essentially eliminated. Therefore, the 
possibility of a new or different kind of accident from any 
previously evaluated is not created.
    ComEd has evaluated industry experiences with TSP degradation, 
eddy current signal distortions, and component misalignment. Eddy 
current signal distortions due to TSP geometry are not indicative of 
TSP degradation and do not result in any kind of accident.
    The component misalignment experienced by one unit is not 
applicable to Byron Unit 1 or Braidwood Unit 1 and, thus, will not 
result in any kind of accident. Specific limitations, as discussed 
above, will be applied to indications at hot-leg intersections which 
contain dents. These limitations ensure that integrity of the SG 
tubes is maintained consistent with current analyses should tube 
denting or TSP cracking occur. Application of the 3.0 volt hot-leg 
IPC to Byron Unit 1 and Braidwood Unit 1, with the limitations 
specified, will not result in the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of the voltage-based, bobbin coil, tube support plate 
elevation plugging criteria with tube expansion at Byron Unit 1 and 
Braidwood Unit 1 is demonstrated to maintain steam generator tube 
integrity commensurate with the criteria of RG 1.121. RG 1.121 
describes a method acceptable to the NRC staff for meeting GDC 14, 
15, 31, and 32 by reducing the probability or the consequences of 
steam generator tube rupture.
    This is accomplished by determining an eddy current inspection 
voltage value which represents a limit for leaving a SG tube in 
service. Tubes with ODSCC voltage indications beyond this limiting 
value must be removed from service by plugging or repaired by 
sleeving. Upon implementation of an increased IPC voltage repair 
limit (up to a maximum of 3.0 volts) for the hot-leg, even under the 
worst case conditions, the occurrence of ODSCC at the tube support 
plate elevations has been evaluated and shown not to present a 
credible potential for a steam generator tube rupture event during 
normal or faulted plant conditions. The End Of Cycle (EOC) 
distribution of crack indications at the tube support plate 
elevations will be confirmed to result in acceptable primary-to-
secondary leakage during all plant conditions such that radiological 
consequences are not adversely impacted.
    Addressing RG 1.83 considerations, implementation of the 
increased hot-leg tube support plate intersection bobbin coil 
voltage-based repair criteria is supplemented by enhanced eddy 
current inspection guidelines to provide consistency in voltage 
normalization and a 100% eddy current inspection sample size at the 
affected tube support plate elevations.
    For the leak and burst assessments, the population of 
indications in the voltage distribution is dependant on the POD 
function. The purpose of the POD function is to account for 
indications that may not be identified by the data analyst.
    In implementing this proposed IPC, ComEd will use the 
conservative Generic Letter 95-05 POD value of 0.6 for all voltage 
amplitude ranges.
    Lowering the Unit 1 RCS dose equivalent I-131 limit from 1.0 
Ci/gm to 0.35 Ci/gm is conservative and provides a 
defense in depth approach to implementation of this IPC. Based on 
current predictions of MSLB leakage at the time of SG replacement, 
the lower RCS dose equivalent I-131 limit also ensures that the 
resulting 2-hour dose rates at the Braidwood and Byron site 
boundaries will not exceed an appropriately small fraction of 10 CFR 
100 dose guideline values.
    Modification of the Byron and Braidwood Specifications for 
conformance with the Generic Letter 95-05 requirements is primarily 
administrative and will not reduce any safety margins. For 
Braidwood, the decrease in the allowed burst probability from 
2.5x10-2 to 1.0x10-2 is conservative. Byron Station has 
previously incorporated this requirement.
    Implementation of the tube support plate elevation repair limits 
will decrease the number of tubes which must be repaired. The 
installation of steam generator tube plugs or sleeves reduces the 
RCS flow margin. Thus, implementation of the interim plugging 
criteria will maintain the margin of flow that would otherwise be 
reduced in the event of increased tube plugging.
    As discussed previously, ComEd has evaluated industry 
experiences with TSP degradation, eddy current signal distortions, 
and component misalignment. Eddy current signal distortions at tube 
support plates will be evaluated to attempt determination of the 
cause of the distortion. A signal distortion alone will not result 
in reduction in the margin of safety. The foreign unit that 
experienced the component misalignment was of a significantly 
different design than the Byron Unit 1 and Braidwood Unit 1 steam 
generators. Analysis of the design differences shows that component 
misalignment of that type is not applicable to Byron Unit 1 or 
Braidwood Unit 1 and, thus, will not result in a reduction in the 
margin of safety.
    Specific limitations, as discussed previously, will be applied 
to indications at hot-leg intersections which contain dents. These 
limitations conservatively treat indications as freespan to ensure 
that integrity of the SG tubes is maintained consistent with current 
analyses should tube denting or TSP cracking occur. Also, tubes with 
large dents (> 5.0 volts) and tubes adjacent to these dented tubes 
will not be used for tube expansion to ensure success of tube 
support plate motion limitation under accident conditions. 
Application of the 3.0 volt hot-leg IPC to Byron Unit 1 and 
Braidwood Unit 1, with the limitations specified, will not result in 
a reduction in a margin of safety.
    Thus, the implementation of this amendment does not result in a 
significant reduction in a margin of safety.
    Therefore, based on the above evaluation, ComEd has concluded 
that these changes involve no significant hazards considerations.

    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 amendments 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 

[[Page 49967]]
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, NW., Washington, DC.
    The filing of requests for hearing and petitions for leave to 
intervene is discussed below.
    By October 27, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendments to the subject facility 
operating licenses 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 rooms which for Byron is located at the Byron Public Library 
District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; and for 
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481. 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 amendments 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 amendments.
    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 amendments.
    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 Mr. Robert A. Capra: 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 Michael I. 
Miller, Esquire; Sidley and Austin, One First National Plaza, Chicago, 
Illinois 60603, 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 amendments dated September 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 rooms which for Byron is located at the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
and for Braidwood, the Wilmington Public 

[[Page 49968]]
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

    Dated at Rockville, Maryland, this 19th day of September 1995.

    For the Nuclear Regulatory Commission.
M. David Lynch,
Senior Project Manager, Project Directorate III-2, Division of Reactor 
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 95-23929 Filed 9-26-95; 8:45 am]
BILLING CODE 7590-01-P