[Federal Register Volume 60, Number 156 (Monday, August 14, 1995)]
[Notices]
[Pages 41904-41906]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-20112]




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DEPARTMENT OF JUSTICE
[Docket No. 50-315]


Indiana Michigan Power Co.; Notice of Consideration of Issuance 
of Amendment to Facility Operating License, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an amendment to Facility Operating License No. 
DPR-58, issued to Indiana Michigan Power Company (the licensee), for 
operation of the Donald C. Cook Nuclear Plant, Unit 1, located in 
Berrien County, Michigan.

    The proposed amendment would modify technical specifications 
4.4.5.4 and 4.4.5.5, on steam generators, to allow for repair of hybrid 
expansion joint sleeves under redefined repair boundary limits.

    The licensee requested this change on an exigent basis because: (1) 
The change is associated with steam generator tube repairs during the 
Unit 1 refueling outage currently in progress, and (2) the empirical 
data compiled from the Kewaunee Nuclear Plant steam generator tube 
pulls in March 1995 is the primary support for this amendment and the 
final implications and conclusions from assessment of that data are 
just now being formulated. The Unit 1 tube repairs are currently 
scheduled to begin on August 29, 1995.

    The NRC staff has reviewed and concurred with the licensee's 
reasons for requesting this amendment on an exigent basis.

    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.
    Pursuant to 10 CFR 50.91(a)(6), for amendments to be granted under 
exigent circumstances the NRC staff must determine 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) Operation of the CNP [Donald C. Cook Nuclear Plant] unit 1 
in accordance with the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Mechanical testing has shown that the inherent structural 
strength of the HEJ [hybrid expansion joint] provides sufficient 
integrity such that the tube rupture capability recommendations of 
RG [Regulatory Guide] 1.121 are met, even for instances of 100% 
throughwall, 360 deg. circumferentially oriented degradation in the 
HEJ hardroll lower transition region. Structural integrity 
recommendations consistent with RG 1.121 are supplied for all tube 
degradation 1.1 inch or greater below the bottom of the HEJ hardroll 
upper transition. Based on test data, a bounding SLB [steam line 
break] leak rate of 0.033 gpm for indications between 1.1 and 1.3 
inch below the bottom of the hardroll upper transition is applied. 
As the leakage data base is expanded and statistical basis 
established, this SLB leakage allowance may be reduced. For 
indications existing greater than 1.3 inch below the bottom of the 
hardroll upper transition, SLB event leakage can be neglected.
    Additional prevention from tube rupture is inherently provided 
by the HEJ geometry. For RCS [reactor coolant system] release rates 
to exceed the normal makeup capacity of the plant, approximately 120 
gpm, the tube must be postulated to experience a complete 
circumferential separation at the lower transition, and become 
axially displaced by 3 to 3.25 inches, resulting in complete 
geometric disassociation between the tube and sleeve resulting in 
sufficient flow area to support leakage of 120 gpm. During the 1989 
plug top release event at North Anna unit 1, primary to secondary 
release rates were calculated to be less than 80 gpm, for a flow 
area approximately 4 times larger than the flow area created by a 
tube which has axially displaced by about 1.25 to 1.5 inch. Analysis 
of the steam generator indicates that at a 95% cumulative 
probability, the tube would experience an axial displacement of less 
than the 1.1 inch boundary. At this level of axial displacement, a 
ring of metal to metal contact would remain between the tube and 
sleeve, and leakage would be far less than 120 gpm. Projected 
leakage at this point is expected to be less than 2.5 gpm. 
Therefore, implementation of the proposed repair boundary will not 
result in tube rupture, even for a tube postulated to not behave as 
predicted by the available test and pulled tube data.
    The proposed technical specification change to support the 
implementation of the HEJ sleeve tube repair boundary for parent 
tube degradation in the HEJ hardroll lower transition region does 
not adversely impact any other previously evaluated design basis 
accident or the results of accident analyses for the current 
technical specification minimum reactor coolant system flow rate. 
Plugging limit criteria are established using the guidance of RG 
1.121. Furthermore, per RG 1.83 recommendations, the sleeved tube 
assembly can be monitored through periodic inspections with present 
eddy current techniques.
    (2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the repair boundary will not introduce 
significant or adverse changes to the plant design basis. Mechanical 
testing of degraded sleeve joints supports the conclusions of the 
calculations that the sleeve retains structural (tube burst) 
capability consistent with RG 1.121. As with [the] initial 
installation of sleeves, implementation of the alternate criteria 
cannot interact with other portions of the RCS. Any hypothetical 
accident as a result of potential tube degradation in the HEJ 
hardroll lower transition region of the tube is bounded by the 
existing tube rupture accident analysis. Neither the sleeve design 
nor implementation of the tube repair boundary defined in Attachment 
4 [Westinghouse Electric Corporation Proprietary Report, WCAP-14446] 
affects any other component or location of the tube outside of the 
immediate area repaired. In addition, as the installation of sleeves 
and the impact on current plugging level analyses is accounted for, 
any postulation that the alternate repair criteria for parent tube 
degradation in the HEJ hardroll lower transition creates a new or 
different type of accident is not supported.
    (3) The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The safety factors used in the establishment of the HEJ sleeved 
tube alternate repair boundary for the disposition of indications in 
the hardroll lower transition of potentially degraded parent tubes 
are consistent with the safety factors in the ASME Boiler and 
Pressure Vessel Code used in steam generator design. Based on the 
sleeved tube geometry, it is unrealistic to consider that 
application of the repair boundary could result in single tube leak 

[[Page 41905]]
rates exceeding the normal makeup capacity during normal operating 
conditions. The repair boundary established in Attachment 4 has been 
developed using the methodology of RG 1.121. The performance 
characteristics of postulated degraded parent tubes of HEJ tube/
sleeve joints have been verified by testing to retain structural 
integrity and preclude significant leakage during normal and 
postulated accident conditions. Testing indicates that postulated 
circumferentially separated tubes which the repair boundary 
addresses would not experience axial displacement during either 
normal operation or SLB conditions. The existing offsite dose 
evaluation performed for CNP unit 1 in support of the voltage based 
plugging criteria for axial ODSCC [outside diameter stress corrosion 
cracking] at TSP [tube support plate] intersections established a 
faulted loop primary to secondary leak rate of 12.6 gpm using 
technical specification dose equivalent Iodine-131 activity levels. 
Following implementation of the criteria, postulated leakage from 
all sources must not exceed 12.6 gpm in the faulted loop. 
Maintenance of this limit will ensure that offsite doses would not 
exceed the currently accepted limit of 10% of the 10 CFR [Part] 100 
guidelines. The repair boundary uses a conservatively established 
``per indication'' leak rate for estimation of SLB leakage. This 
leak rate is applied to all indications left in service as a result 
of the tube repair boundary, including non-throughwall indications 
and a limited number of indications of circumferential throughwall 
extent.
    For a postulated indication whose performance is not 
characteristic of the test and pulled tube data, and which would 
experience axial displacement at the 95% cumulative probability 
value following a postulated SLB event with no operator 
intervention, leakage would not be expected to result in an 
uncontrolled release of reactor coolant in excess of normal makeup 
capacity.
    For the three pulled tubes and nearly 1,000 crack indications 
detected in the field, there were no instances of degradation of 
elevations, (multiple expansion transitions) on either side of the 
hardroll expansion in the same tube. This includes no instances of 
non-detected degradation in the upper hydraulic and hardroll upper 
expansion transitions for the pulled tubes. One tube was identified 
in the most recent Kewaunee inspection with two separate 
circumferential crack elevations within the hardroll lower 
transition. Rapidly occurring degradation would not be expected at 
the upper transitions, based partly on the field inspection results. 
The available inspection results include two inspection programs 
(1994 and 1995) at Kewaunee and one at Point Beach unit 2 (1994). 
Through these three inspection programs, approximately 11,000 HEJ 
sleeved tubes have been inspected using advanced probes.
    The portions of the installed sleeve assembly which represent 
the reactor coolant pressure boundary can be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirements of Regulatory Guide 1.83.

    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 15 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 15-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 15-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. The Commission expects that the need to 
take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC 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 August 29, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street NW., Washington, DC, and at the local public 
document room located at the Maud Preston Palenske Memorial Library, 
500 Market Street, St. Joseph, Michigan 49085. 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 a 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 

[[Page 41906]]
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 the amendment is issued before the expiration of the 30-day 
hearing period, the Commission will make a final determination on the 
issue of no significant hazards consideration. If a hearing is 
requested, 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, 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 John N. Hannon: 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 Gerald Charnoff, 
Esq., Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., 
Washington, DC 20037, attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for hearing will not 
be entertained absent a determination by the Commission, the presiding 
officer or the presiding Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment dated August 4, 1995, which is available for 
public inspection at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street NW., Washington, DC, and at the local public 
document room, located at the Maud Preston Palenske Memorial Library, 
500 Market Street, St. Joseph, Michigan 49085.

    Dated at Rockville, Maryland, this 9th day of August 1995.

    For The Nuclear Regulatory Commission.
Tae Kim,
Project Manager, Project Directorate III-1, Division of Reactor 
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 95-20112 Filed 8-11-95; 8:45 am]
BILLING CODE 7590-01-P