Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing; Commonwealth Edison Co.

NUCLEAR REGULATORY COMMISSION

[Docket Nos. STN 50-454 and STN 50-455]

Consideration of Issuance of Amendment to Facility Operating

License, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing; Commonwealth Edison Co.

The U.S. Nuclear Regulatory Commission (the Commission) is

considering issuance of an amendment to Facility Operating License Nos.

NPF-37 and NPF-66, issued to Commonwealth Edison Company (the

licensee), for operation of Byron Station, Units 1 and 2, located in

Ogle County, Illinois.

The proposed amendment would revise the technical specifications

(TS) to incorporate a 1.0 volt steam generator tube interim plugging

criteria (IPC) for Unit 1 beginning with Cycle 7, which has begun. This

supplements the information that was published in the Federal Register

on August 31, 1994 (59 FR 45019).

Before issuance of the proposed license amendment, 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 amendment 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.

Consistent with Regulatory Guide (RG) 1.121, ``Basis for

Plugging Degraded PWR Steam Generator Tubes,'' Revision 0, August

1976, the traditional depth-based criteria for SG tube repair

implicitly ensures that tubes accepted for continued service will

retain adequate structural and leakage integrity during normal

operating, transient, and postulated accident conditions. It is

recognized that defects in tubes permitted to remain in service,

especially cracks, occasionally grow entirely through-wall and

develop small leaks. Limits on allowable primary-to-secondary

leakage established in Technical Specifications ensure timely plant

shutdown before the structural and leakage integrity of the affected

tube is challenged.

The proposed license amendment request to implement voltage

amplitude [steam generator] SG tube support plate Interim Plugging

Criteria for Byron Unit 1 Cycle 7 meets the requirements of RG

1.121. The IPC methodology demonstrates that tube leakage is

acceptably low and tube burst is a highly improbable event during

either normal operation or the most limiting accident condition, a

postulated main steam line break (MSLB) event. Requesting a single

cycle applicability is more conservative than the guidance contained

in the draft Generic Letter on voltage-based repair criteria issued

for comment on August 12, 1994.

Adequate SG tube leakage integrity during normal operating

conditions is assured by limiting allowable primary-to-secondary

leakage to 150 gpd per SG or 600 gpd total. Currently, this limit is

administratively controlled.

However, a license amendment request was submitted on June 3,

1994, to incorporate this limit into the Byron Technical

Specifications. During normal operating conditions, the tube support

plate constrains the [outer diameter stress corrosion cracking]

ODSCC affected area of the tube to provide additional strength that

precludes burst. Any leakage of a tube exhibiting ODSCC at the [tube

support plate] TSP is fully bounded by the existing SG tube rupture

analysis included in the Byron [Updated Final Safety Analysis

Report] UFSAR. Therefore, probability of failure of a tube left in

service or consequences of tube failure during normal operating

conditions is not significantly increased by the application of IPC.

During transients, the TSP is conservatively assumed to displace

due to the thermal-hydraulic loads associated with the transient.

This may partially expose a crack which is within the boundary of

the TSP during normal operations to free span conditions. Burst is

therefore conservatively evaluated assuming the crack is fully

exposed to free span conditions. The structural eddy current bobbin

coil voltage limit for free-span burst is 4.54 volts. This limit

takes into consideration a 1.43 safety factor applied to the steam

line break differential pressure that is consistent with RG 1.121

requirements. With additional considerations for growth rate

assumptions and an upper 95% confidence estimate on voltage

variability, the maximum voltage indication that could remain in

service is reduced to 2.7 volts. For added conservatism, the

allowable indication voltage is further reduced in the proposed

amendment to a 1.0 volt confirmed ODSCC indication limit. All

indications greater than 1.0 volt will be subject to an RPC

examination. Tubes with RPC confirmed ODSCC indications will be

plugged or sleeved. Any ODSCC indications between 1.0 volt and 2.7

volts which are not confirmed as ODSCC will be allowed to remain in

service since these indications are not as likely to affect tube

structural integrity or leakage integrity over the next operating

cycle as the indications that are detectable by both bobbin and RPC

inspections.

The eddy current inspection process has been enhanced to address

RG 1.83, ``Inservice Inspection of PWR Steam Generator Tubes,''

Revision 1, July 1975. consideration as well as the EPRI SG

Inspection Guidelines. Enhancements in accordance with NUREG-1477

and Appendix A of the Catawba IPC report (WCAP-13698) are in place

to increase detection of ODSCC indications and to ensure reliable,

consistent acquisition and analysis of data. Based on the

conservative selection of the voltage criteria and the increased

ability to identify ODSCC, the probability of tube failure during an

accident is also not significantly increased due to application of

requested IPC.

For consistency with current offsite dose limits, the site

allowable leakage limit during a MSLB has been conservatively

calculated to be 12.8 gpm. This leakage limit includes maximum

allowable operational leakage from the unaffected SGs and the

accident leakage from the affected SG. As a requirement for

operation following application of IPC, the projected distribution

of crack indications over the operating period must be verified to

result in primary to secondary accident leakage less than the site

allowable leakage limit. Thus, the consequences of a MSLB remain

unchanged.

For an unscheduled mid-cycle inspection as a result of leakage

due to mechanisms other than ODSCC at support plates or some other

cause, the ODSCC indication limit is represented by the following

equation:

TN23SE94.003

where:

V=measured voltage

VBOC=voltage at BOC

t=time period of operation to unscheduled outage

CL=cycle length (full operating cycle length where operating cycle

is the time between two scheduled steam generator inspections)

VSL=4.5 volts

Assuming linear flaw growth from BOC to EOC and a maximum

structural limit of 4.5 volts, the voltage expected for an

identified flaw at any time in the cycle can be predicted. The

allowed voltage limit for an unscheduled inspection, as identified

by the equation given above, reduces the predicted straightline

growth voltage to ensure conservatism in the limit. A flaw which has

not exceeded the predicted voltage growth at any point in the cycle

would not be expected to exceed the structural limit at end of cycle

or negatively impact the burst probability calculated based on

results from the last scheduled inspection. Therefore, it is

acceptable to leave the tube in service.

Therefore, as implementation of the 1.0 volt IPC for Byron Unit

1 Cycle 7 does not adversely affect steam generator tube integrity

and results in acceptable dose consequences, the proposed license

amendment request does not result in any significant increase in the

probability or consequences of an accident previously evaluated

within the Byron Updated Final Safety Analysis Report.

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 SG tube IPC for Byron Unit 1

Cycle 7 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 the tube support plate

elevations since industry experience indicates that ODSCC

originating within the tube support plate does not extend

significantly beyond the thickness of the support plate. This

criteria only applies to ODSCC contained within the region of the

tube bounded by the tube support plate.

In addressing the combined effects of Loss of Coolant Accident

(LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG

(as required by General Design Criterion 2), it has been determined

that tube collapse of select tubes may occur in the SGs at some

plants, including Byron Unit 1. 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 through-wall cracks in

tubes could progress to through-wall cracks during tube deformation

or collapse.

A number of tubes have been identified, in the ``wedge''

locations of the SG TSPs, that demonstrate the potential for tube

collapse during a LOCA + SSE event. Because of this potential, these

tubes have been excluded from application of the voltage-based SG

TSP IPC.

Therefore, neither a single nor multiple tube rupture event

would be expected in a steam generator in which IPC has been

applied.

ComEd has implemented a maximum primary to secondary leakage

limit of 150 gpd through any one SG at Byron to help preclude the

potential for excessive leakage during all 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. The 150 gpd limit provides

adequate leakage detection and plant shutdown criteria in the event

an unexpected single crack results in leakage that is associated

with the longest permissible free span crack length. Since tube

burst is precluded during normal operation due to the proximity of

the TSP to the tube and the potential exists for the crevice to

become uncovered during MSLB conditions, the leakage from the

maximum permissible crack must preclude tube burst at MSLB

conditions. Thus, the 150 gpd limit provides a conservative limit to

prompt plant shutdown prior to reaching critical crack lengths under

MSLB conditions.

Upon implementation of the 1.0 volt IPC for Byron Unit 1 Cycle

7, steam generator tube integrity continues to be maintained through

inservice inspection and primary-to-secondary leakage monitoring.

Therefore, the possibility of a new or different kind of accident

from any previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The use of the voltage based bobbin coil probe SG TSP IPC for

Byron Unit 1 Cycle 7 will maintain steam generator tube integrity

commensurate with the criteria of RG 1.121 as discussed above. Upon

implementation of the criteria, even under the worst case

conditions, the occurrence of ODSCC at the TSP elevations is not

expected to lead to a steam generator tube rupture event during

normal or faulted plant conditions. The distribution of crack

indications at the TSP elevations result in acceptable primary-to-

secondary leakage during all plant conditions and radiological

consequences are not adversely impacted by the application of IPC.

The installation of SG tube plugs and sleeves reduces the RCS

flow margin. As noted previously, implementation of the SG TSP IPC

will decrease the number of tubes which must be repaired by plugging

or sleeving. Thus, implementation of IPC will retain additional flow

margin that would otherwise be reduced due to increased tube

plugging. Therefore, no significant reduction in the margin of

safety will occur during Cycle 7 as a result of the implementation

of this proposed license amendment request.

Although not relied upon to prove adequacy of the proposed

amendment request, the following analyses demonstrate that

significant conservatism exists in the methods and justifications

described above:

Limited Tube Support Plate Displacement

An analysis was performed to verify the extent of limited TSP

displacement during accident conditions (MSLB). Application of

minimum TSP displacement assumptions reduce the likelihood of a tube

burst to negligible levels.

Consideration of limited TSP displacement would also reduce

potential MSLB leakage when compared to the leakage calculated

assuming free span indications.

Probability of Detection

The Electric Power Research Institute (EPRI) Performance

Demonstration Program analyzed the performance of approximately 20

eddy current data analysts evaluating data from a unit with \3/4\''

inside diameter and 0.049'' wall thickness tubes. The results of

this analysis clearly show that the detectability of larger voltage

indications is increased which lends creditability for application

of a POD of >0.6 for ODSCC indications larger than 1.0 volt.

Risk Evaluation of Core Damage

As part of ComEd's evaluation of the operability of Byron Unit 1

Cycle 7, a risk evaluation was completed. The objective of this

evaluation was to compare core damage frequency under containment

bypass conditions, with and without the interim plugging criteria

applied at Byron Unit 1.

The total Byron core damage frequency is estimated to be 3.09E-5

per reactor year with a total contribution from containment bypass

sequences of 3.72E-8 per reactor year according to the results of

the current individual plant evaluation (IPE). Operation with the

requested IPC resulted in an insignificant increase in core damage

frequency resulting from MSLB with containment bypass conditions.

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 amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received.

Should the Commission take this action, it will publish in the Federal

Register a notice of issuance and provide for opportunity for a hearing

after issuance. The Commission expects that the need to take this

action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11455 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

20555.

The filing of requests for hearing and petitions for leave to

intervene is discussed below.

By October 24, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

public document room located at the Byron Public Library, 109 N.

Franklin, Byron, Illinois 61010. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

Atomic Safety and Licensing Board, designated by the Commission or by

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of hearing or an

appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 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; Sidney and Austin, One First National

Plaza, Chicago, Illinois 60690, 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

applications for amendment dated September 7, 1994, and September 17,

1994, (two letters) which are available for public inspection at the

Commission's Public Document Room, the Gelman Building, 2120 L Street,

NW., Washington, DC 20555 and at the local public document room located

at the Byron Public Library, 109 N. Franklin, P.O. Box 434, Byron,

Illinois 61010.

Dated at Rockville, Maryland, this 21st day of September 1994.

For the Nuclear Regulatory Commission.

Robert A. Capra,

Director Project Directorate III-2, Division of Reactor Projects--III/

IV, Office of Nuclear Reactor Regulation.

[FR Doc. 94-23730 Filed 9-22-94; 8:45 am]

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