[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]
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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 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
