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

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 (-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