[Federal Register Volume 59, Number 119 (Wednesday, June 22, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-15266]


[[Page Unknown]]

[Federal Register: June 22, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-461]

 

Illinois Power Co.; 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. 
NPF-62, issued to the Illinois Power Company (the licensee), for 
operation of the Clinton Power Station, Unit 1, located in DeWitt 
County, Illinois.
    The proposed amendment would modify Technical Specification 3/
4.4.3.1, ``Reactor Coolant System Leakage--Leakage Detection Systems,'' 
to permit continued plant operation with inoperable drywell floor drain 
sump flow rate monitoring instrumentation. Continued plant operation 
would be permitted until the first time the plant is required to be 
brought to COLD SHUTDOWN after July 10, 1994.
    Technical Specification 3/4.4.3.1 requires that systems capable of 
monitoring unidentified reactor coolant system leakage rates remain 
operable. Rector coolant system leakage that falls on the drywell 
floors is channeled through the floor drains and enters the drywell 
floor drain sump. Prior to entering the floor drain sump, water passes 
through the drywell floor drain sump flow monitoring instrumentation 
where the instantaneous flow rates and total integrated flow are 
measured. The flow monitoring instrumentation consists of a V-notch 
weir box containing a capacitance probe. Water flows through a V-notch 
water level which is directly proportional to the flow through the weir 
box. Thus, flow through the V-notch is equal to the sump inlet flow 
rate. The capacitance probe is calibrated to correspond to the incoming 
flow rate and provides a continuous control room indication of the 
unidentified reactor coolant system leakage rate. An alarm is generated 
when the technical specification limit of 5 gpm of unidentified leakage 
occurs. The V-notch weir box instrumentation meets the accuracy and 
sensitivity requirements of Regulatory Guide 1.45 for drywell floor 
drain sump flow monitoring.
    The licensee began to observe questionable readings from the 
indicated drywell floor drain sump inlet flow and subsequently declared 
the drywell floor drain sump monitoring instrumentation inoperable on 
June 10, 1994. Technical Specification 3.4.3.1 permits 30 days of 
continuous plant operation provided the drywell floor drain sump flow 
rate is monitored and determined by alternative means at least once 
every 8 hours.
    All efforts by the licensee to restore the drywell sump inlet flow 
monitoring instrumentation to operable status have been unsuccessful. 
The instrument loop has been recalibrated and equipment external to the 
drywell has been verified to be operating properly. The only option 
remaining for the licensee is to enter the drywell in order to examine 
the V-notch weir box and associated capacitance probe. However, the V-
notch weir box is located in a keyway beneath the reactor vessel and 
inside the biological shield wall. Due to the high radiation and 
temperatures in this location, a plant shutdown would be required 
before personnel would be able to reach the instrumentation.
    In a letter dated June 20, 1994, the licensee requested that this 
amendment application be treated as an emergency because unless 
approved, technical specifications would require a plant shutdown. The 
licensee stated that such action would be necessary to preclude an 
unnecessary plant transient and related plant risk associated with a 
plant shutdown. Due to time constraints, sufficient time is not 
available to permit the customary public notices in advance of this 
action.
    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) The proposed change does not affect any initiators of any 
previously evaluated accidents. Additionally, the proposed change 
involves equipment that only provides indication and therefore, it 
cannot increase the probability of any accident previously 
evaluated.
    As stated in Updated Safety Analysis Report (USAR) Section 
7.7.1.24.1, no credit is taken in the safety analysis for operation 
of or operator reliance upon the leakage detection monitoring 
instrumentation associated with the drywell sumps. Notwithstanding, 
the drywell floor drain sump flow monitoring system provides the 
capability to detect and measure leakage from unknown sources of 
leakage in the drywell. The drywell floor drain sump inlet flow 
monitoring V-notch weir box instrumentation is designed to meet the 
accuracy requirements of Regulatory Guide 1.45. This instrumentation 
does not provide any automatic action or control functions. In 
addition to the V-notch system, drywell floor drain sump flow rates 
can be determined by using the sump pump pump-out timers, cycle 
counters and level switches. In addition, unidentified leakage into 
the drywell is monitored by a flow rate meter in the condensate 
discharge line from the drywell air coolers and by a particulate and 
a gaseous radiation monitoring channel of the drywell fission 
product monitor. While the drywell fission product monitor does not 
provide a quantitative leakage rate, it is sensitive enough to 
provide plant operators with early indication of an unanticipated 
increase in unidentified leakage. Furthermore, a number of other 
parameters are monitored with appropriate instrumentation to provide 
the plant operators with indirect indication of increases in 
unidentified leakage. These parameters include drywell pressure and 
drywell temperature. These alternative methods of detecting 
increases in unidentified leakage rates provide operators with 
sufficient information to take appropriate action to respond to an 
increase in leakage. Based on the above, Illinois Power concludes 
that the proposed change will not increase the consequences of any 
accident previously evaluated.
    (2) The proposed change does not involve any modification to 
plant structures or components and only involves equipment that 
provides indication of leakage to the plant operators. The affected 
equipment does not provide any automatic action or control 
functions. As a result, the proposed change does not involve a 
change in the operation of the plant, nor does it introduce any new 
failure modes. Therefore, this proposed change cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) The margin of safety associated with the instrumentation 
affected by the proposed change may be related to the limits on 
unidentified leakage. As stated in the Bases for Technical 
Specification 3/4.4.3.2. ``The allowable leakage rates from the 
reactor coolant system have been based on the predicted and 
experimentally observed behavior of cracks in pipes . . . The 
evidence obtained from experiments suggests that for leakage 
somewhat greater than that specified for unidentified leakage the 
probability is small that the imperfection or crack associated with 
such leakage would grow rapidly. With respect to Intergranular 
Stress Corrosion Cracking (IGSCC) related cracks in service 
sensitive austenitic stainless steel piping however, an additional 
limit on the allowed increase in unidentified leakage (within a 24-
hour period or less) is imposed in accordance with NRC Generic 
Letter 88-01, `NRC Position on IGSCC in BWR Austenitic Stainless 
Steel Piping,' since an abrupt increase in the unidentified leakage 
could be indicative of leakage from such a source.'' The proposed 
change does not alter any of these limits on the unidentified 
leakage.
    As previously described, flow rates into the drywell floor drain 
sump can be determined based on the indicated run time for the sump 
pumps and the known pump flow rates or by monitoring the sump fill-
up times and considering the volume corresponding to the current 
level control band. These alternate methods are sufficient to 
determine whether unidentified leakage in the drywell exceeds the 5 
pgm limit and whether changes in this leakage exceed the limit of a 
2 gpm increase in any 24-hour period or less.
    Additionally, with respect to the ability to detect changes in 
unidentified leakage rates, in addition to the V-notch system, 
drywell floor drain sump flow rates can be determined by using the 
sump pump pump-out timers, cycle counters and level switches. In 
addition, unidentified leakage into the drywell is monitored by a 
flow rate meter in the condensate discharge line from the drywell 
air coolers and by a particulate and a gaseous radiation monitoring 
channel of the drywell fission product monitor. While the drywell 
fission product monitor does not provide a quantitative leakage 
rate, it is sensitive enough to provide plant operators with early 
indication of an unanticipated increase in the unidentified leakage 
rate involving reactor coolant. Furthermore, a number of other 
parameters are monitored with appropriate instrumentation to provide 
the plant operators with indirect indication of increases in 
unidentified leakage. These parameters include drywell pressure and 
drywell temperature.
    As stated above, the drywell floor drain sump flow monitoring 
instrumentation does not provide any automatic action or control 
functions. Further, as stated in USAR Section 7.7.1.24.1, no credit 
is taken in the safety analysis for operation of or operator 
reliance upon the leakage detection monitoring instrumentation 
associated with the drywell sumps.
    In light of all the above, Illinois Power concludes that the 
proposed change does not involve a reduction in the margin of 
safety.

    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, 11555 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 July 22, 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 Vespasian Warner Public Library, 
120 West Johnson Street, Clinton, Illinois, 61727. 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 particularly 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 providing 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 significant 
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 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 John Hannon, Director, Project 
Directorate III-3; 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 Sheldon Zabel, Esq., Schiff, 
Hardin and Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 
60606, 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 June 20, 1994, which is 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 Vespasian Warner Public Library, 
120 West Johnson Street, Clinton, Illinois 61727.

    For the Nuclear Regulatory Commission.
Douglas V. Pickett,
Acting Director, Project Directorate III-3, Division of Reactor 
Projects-III-IV, Office of Nuclear Reactor Regulation.
[FR Doc. 94-15266 Filed 6-21-94; 8:45 am]
BILLING CODE 7590-01-M