[Federal Register Volume 64, Number 134 (Wednesday, July 14, 1999)]
[Notices]
[Pages 38022-38043]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-17750]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from June 19, 1999, through July 2, 1999. The 
last biweekly notice was published on June 30, 1999 (64 FR 35199).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve 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. The basis 
for this proposed determination for each amendment request is shown 
below.
    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 
before action is taken. 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 Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, 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 a hearing and petitions for leave to intervene 
is discussed below.
    By August 13, 1999, 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 for the particular facility involved. 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 a 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

[[Page 38023]]

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 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
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 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 for 
the particular facility involved.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: June 15, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to incorporate the performance-
based 10 CFR 50 Appendix J, Option B for Type A tests (containment 
integrated leakage rate tests). Option B will be implemented for Type A 
testing in accordance with NRC Regulatory Guide 1.163, ``Performance-
Based Containment Leak-Test Program,'' dated September 1995, and 
Nuclear Energy Institute (NEI) Guideline 94-01, Revision 0, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR Part 50, 
Appendix J,'' dated July 26, 1995. Type B and C testing (containment 
penetration leakage tests) will continue to be performed in accordance 
with 10 CFR 50 Appendix J, Option A.
    Basis for proposed no significant hazards consideration 
determination: 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 license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The Harris Nuclear Plant (HNP) Type A testing history provides 
justification for the proposed test schedule change to one test in a 
10 year period. With the successful Type A tests of September 1992 
and May 1997, and a greater than 24 month elapsed time between the 
two tests, CP&L considers the requirement of two consecutive Type A 
tests to have been met. This testing has affirmed the acceptable 
reliability of the containment structure to minimize leakage as 
designed, and provides assurance that its performance to 
continuously function as designed is not challenged due to this test 
schedule extension to once in 10 years.
    This proposed change to revise the test schedule frequency does 
not impact or alter the design of any system, structure or 
component. The limit on allowable leakage is not increased. Type A 
testing provides periodic verification of the leak tight integrity 
of the containment and the components that penetrate the containment 
structure. NUREG-1493, Section 10.1.2, ``Leakage-Testing 
Intervals,'' states that reducing the frequency of Type A tests from 
the current three per 10 years to one per 20 years was found to lead 
to an imperceptible increase in risk.
    Therefore, based on these considerations, and the previous 
plant-specific Type A test results, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes only incorporate the performance-based 
testing approach authorized in 10 CFR 50 Appendix J, Option B, and 
are justified based on previous plant-specific Type A test results. 
Plant structures, systems, and components will not be operated in a 
different manner as a result of these proposed changes and no 
physical modifications to equipment are involved. The interval 
extensions allowed by Option B of 10 CFR 50 Appendix J do not have 
the potential for creating the possibility of a new or different 
type of accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes do not change the allowable leak rate from 
the containment; they only allow an extension of the interval 
between the performance of Type A leak rate testing. NUREG-1493 
provides the technical basis for the NRC's rulemaking to revise 
containment leakage testing requirements for nuclear power reactors 
in 10 CFR 50 Appendix J. NUREG-1493, Section 10.1.2, ``Leakage-
Testing Intervals,'' states that increasing the interval between 
integrated leakage-rate tests is possible with minimal impact on 
public risk.
    Based on these considerations and the previous plant-specific 
Type A test results, the proposed changes do 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

[[Page 38024]]

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.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.
    Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
    Date of amendment request: June 15, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 4.7.D.6 by replacing the leakage limit 
of 11.5 standard cubic feet per hour (scfh) for each main steam 
isolation valve (MSIV) with a limit of 46 scfh on the total combined 
leakage from all four main steam lines.
    Basis for proposed no significant hazards consideration 
determination: 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to the Technical Specifications, Appendix 
A, modifies the allowed MSIV leakage limit to an aggregate value 
with no change to the total allowed leakage rate. This change does 
not affect either the automatic or manual features that would close 
the MSIVs. Performance of the leakage tests do not adversely affect 
any accident previously evaluated. Consequently, this proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The safety function of the MSIVs is to provide a timely steam 
line isolation to mitigate the release of radioactive steam and 
limit reactor inventory loss under certain accident and transient 
conditions. The MSIVs are designed to automatically close whenever 
plant conditions warrant main steam line isolation. Changing the 
leakage limits to include an aggregate value does not affect the 
isolation function. No new equipment will be installed or utilized, 
and no new operating conditions will be initiated as a result of 
this change. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The total allowed leakage rate for all MSIVs remains unchanged 
at 46 scfh. Therefore, there will be no change in the types or 
significant increase in the amounts of any effluents released 
offsite, and, thus, the radiological analyses remain unchanged and 
within the guidelines of 10 CFR 100 and General Design Criteria 19. 
Therefore, these changes do not involve a significant reduction in 
the margin of safety.

    Therefore, based upon the above evaluation, ComEd has concluded 
that these changes involve no significant hazards consideration. 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: May 19, 1999.
    Description of amendment request: The proposed amendments would 
relocate Technical Specification Section \3/4\.4.4, ``Chemistry'' from 
the TS to the Updated Final Safety Analysis Report and Administrative 
Technical Requirements.
    Basis for proposed no significant hazards consideration 
determination: 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:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes simplify the TS, meet regulatory 
requirements for relocated TS, and implement the recommendations of 
the NRC's Final Policy Statement on TS improvements. The Chemistry 
requirements will be relocated to the Updated Final Safety Analysis 
Report (UFSAR) and Administrative Technical Requirement that has 
been incorporated into the UFSAR by reference. Future changes to 
these requirements will be controlled by 10 CFR 50.59. The proposed 
changes are administrative in nature and do not involve any 
modification to any plant equipment or affect plant operation. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any previously 
evaluated accident.
    Consequently, this proposed amendment does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety related system performs 
its function. Therefore, this proposed TS amendment would not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed amendment represents the relocation of current 
requirements that are based on generic guidance or previously 
approved provisions for other stations. The proposed changes are 
administrative in nature and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. The proposed changes have been 
evaluated and found to be acceptable for use at Duane Arnold Energy 
Center and Quad Cities Nuclear Power Station. Since the proposed 
changes are administrative in nature, and are based on NRC accepted 
provisions which have been adopted at other nuclear facilities, and 
maintain the necessary levels of system reliability, the proposed 
changes do not involve a significant 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: June 3, 1999.
    Description of amendment request: The proposed amendment would 
delete sections of the Technical Specifications that no longer apply to 
the Haddam

[[Page 38025]]

Neck Plant's permanently shutdown and defueled condition; increase the 
weight of loads allowed over the spent fuel pool; relocate certain 
definitions and requirements from the Technical Specifications to the 
Technical Requirements Manual (TRM), Connecticut Yankee Quality 
Assurance Program (CYQAP), or the Radiological Effluent Monitoring and 
Offsite Dose Calculation Manual (REMODCM); and correct typographical 
errors, renumber sections, and repaginate the Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The licensee's summary of its analysis is presented 
below:
    The proposed changes do not involve an SHC [significant hazards 
consideration] because the changes would not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    In the present plant configuration, the reactor-related 
accidents previously evaluated (i.e., LOCA [loss-of-coolant 
accident], MSLB [main steamline break], etc.) are no longer 
possible. The accidents previously evaluated that are still 
applicable to the plant are fuel handling accidents and gaseous and 
liquid radioactive releases. The following events are presently 
considered as bounding of all other events:

--Fuel handling and cask drop accidents in the spent fuel building,
--Criticality in the spent fuel pool,
--Loss of spent fuel cooling,
--Resin fire (gaseous release), and
--Rupture of a tank containing radioactive liquid.

    There is no significant increase in the probability of a fuel 
handling accident since refueling operations have ceased, with a 
corresponding decrease in the frequency of fuel movement. The 
radiological consequences of a fuel handling accident, should one 
occur, decrease the longer the spent fuel is allowed to decay. As 
discussed previously, the spent fuel inventory of radioactive iodine 
and noble gases [, with the exception of Kr-85,] have decayed more 
than 20 half-lives since shutdown and are no longer a release 
concern. With this reduced source the results of the fuel handling 
accident show that the filters of Specification 3.9.12 are no longer 
necessary. The allowed weight over the spent fuel pool is still less 
than that previously [evaluated]. Therefore, there has been no 
increase in the probability or consequences of a fuel handling or 
cask drop accident.
    Criticality controls are imposed by specifications \3/4\.9.13 
and \3/4\.9.14 * * * [The requirements of these specifications have 
not been changed.] Therefore, there has been no increase in the 
probability or consequences of a criticality event.
    Spent fuel cooling is maintained by keeping the pool temperature 
below 150 deg.F. Should normal cooling be lost, the availability of 
an abundant supply of water ensures that sufficient time is 
available prior to boiling to restore cooling. This is controlled by 
specifications \3/4\.9.11 and \3/4\.9.16 * * * [The requirements of 
these specifications have not been changed. Technical specification 
\3/4\.9.15 does not apply to the permanently defueled condition of 
the plant. Therefore, there has been no increase in the probability 
or consequences of a loss of cooling event.]
    The probability of a gaseous or liquid radioactive release is 
not changed by the proposed revisions. As the plant undergoes 
decommissioning, the previous limiting events [such as a loss-of-
coolant-accident] are no longer applicable, and previous non-
limiting events [such as a resin fire] now become limiting. These * 
* * events have not changed from how they might have occurred in the 
past. The radiological consequences of a gaseous or liquid 
radioactive release are bounded by the fuel handling accident during 
defueled operation and a spent resin fire during processing of resin 
from the reactor coolant system decontamination. The rupture of a 
tank containing radioactive liquid was assessed and found to be 
bounded by these events. With the plant defueled and permanently 
shutdown, the demands on the radwaste systems are lessened since no 
new radioisotopes are being generated by irradiation or fission. 
Therefore, there is no increase in the probability or consequences 
of a gaseous or liquid radioactive release.
* * * * *
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes are generally of an administrative nature 
and do not have an effect on the physical plant. The events 
considered bound other potential events and are considered the 
limiting cases for potential gaseous or liquid releases to the 
environment.
    With the plant undergoing decommissioning, the types of 
accidents one might be concerned with involve criticality of the 
spent fuel, or draining of the spent fuel pool. None of the proposed 
changes affect the possibility of such an event. Also, none of the 
proposed changes could lead to a radiological release of a greater 
magnitude than for the events considered, such as might occur with 
the accumulation of a greater quantity of radioactive material in 
one location, or with damage to a greater number of fuel assemblies 
than considered in the fuel handling accident.
    The proposed changes restrict the operations that can be 
conducted at the plant, and do not permit any new type of activity 
from what had previously been authorized. The effect on systems, 
structures and components affected by the proposed changes have no 
adverse impact on the storage of fuel nor on the processing of 
radioactive wastes presently at the site. The present set of 
limiting events are a subset of events previously considered. 
Therefore these changes do not create the possibility of a new or 
different kind of accident from any accident previously considered.
* * * * *
    3. Involve a significant reduction in a margin of safety.
    The proposed changes have no impact on the analyses of 
postulated design basis events remaining applicable to the Haddam 
Neck Plant. Analysis of the limiting events show that their 
consequences to the public are within the limits of 10 CFR [Part] 20 
and the EPA PAGs [Environmental Protection Agency Protective Action 
Guides]. The consequences to members of the operating staff are 
within the limits of 10 CFR [Part] 50, Appendix A, General Design 
Criterion 19, ``Control Room Habitability''. Therefore there is no 
reduction in a margin of safety.

* * * * *
    Based on the above, the proposed changes to the operating license 
and technical specifications do not involve a reduction in the margin 
of safety due to the reduced decay heat load, the decay of 
radionuclides since shutdown, and by maintaining the heavy load 
restriction.
    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.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut, 06457.
    Attorney for licensee: Mr. J. A. Ritsher, Ropes & Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Section Chief: Michael T. Masnik.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: June 10, 1999.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to adopt a Ventilation Filter 
Testing Program in TS Section 6.0, ``Administrative Controls,'' and 
remove the specific ventilation filter surveillance requirements from 
TS 3/4.6.4.4, ``Hydrogen Purge System,'' TS 3/4.6.5.1, ``Shield 
Building Emergency Ventilation System,'' and TS 3/4.7.6.1, ``Control 
Room Emergency Ventilation System.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards

[[Page 38026]]

consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power Station 
(DBNPS), Unit Number 1, in accordance with this change would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. The replacement of the specific Technical 
Specification (TS) ventilation filter testing Surveillance 
Requirements for the Containment Hydrogen Purge System 3/4.6.4.4), 
Shield Building Emergency Ventilation System (3/4.6.5.1), and the 
Control Room Emergency Ventilation System (3/4.7.6.1), with a 
reference to the newly created Ventilation Filter Testing Program 
contained in TS Administrative Controls Section 6.8.4.f, Ventilation 
Filter Testing Program, is a removal and relocation of certain TS 
details. The proposed TS 6.8.4.f will, however, add controls to 
maintain similar operation, maintenance, testing and system 
operability for these three ventilation systems. The TS Bases 
changes reflect the use of the Ventilation Filter Testing Program. 
Therefore, it can be concluded that the proposed changes do not 
involve a significant increase in the probability of an accident 
previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. No physical alterations of 
the DBNPS are involved, nor are plant operating methods being 
changed. The proposed changes do not alter the source term, 
containment isolation or allowable radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not change the way the plant is operated. No new or 
different types of failures or accident initiators are being 
introduced by the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because no inputs into the calculation of any Technical 
Specification Safety Limit, Limiting Safety System Settings, 
Technical Specification Limiting Condition for Operation, or other 
previously defined margins for any structure, system, or component 
important to safety are being affected by the proposed changes.

    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.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: May 10, 1999.
    Description of amendment request: The proposed amendment would 
correct the regulation referenced in Section 5.8, ``High Radiation 
Area,'' of the Crystal River Unit 3 (CR-3) Improved Technical 
Specifications (ITS). The ITS currently references 10 CFR 20, Paragraph 
20.1601(2) and (3), whereas the correct reference is 10 CFR 20.1601(c).
    Basis for proposed no significant hazards consideration 
determination: 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed change to the Crystal River Unit 3 (CR-3) Improved 
Technical Specifications (ITS) is editorial in nature. The change 
involves revising the incorrect reference in ITS Section 5.8.1 to 
the correct Code of Federal Regulations reference that pertains to 
controlling access to high radiation areas. The proposed ITS change 
does not involve any change to plant design, operation, maintenance, 
or procedures. As a result, no changes to the plant are being made 
which would impact either the contributors to an accident or to the 
consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    The proposed editorial change to the ITS does not involve any 
changes to any plant structure, system, or component (SSC) or to its 
operation or maintenance. There is no impact on any equipment that 
would be considered as contributors to either new or different 
accidents. Thus, the change to the ITS does not create the 
possibility of a new or different kind of accident.
    3. Involve a significant reduction in a margin of safety?
    The proposed change to the ITS involves a reference change and 
does not involve the design or operation of any plant SSC. No 
changes to the methods for controlling access to high radiation 
areas are proposed. No changes to the methods for controlling 
personnel and/or activities in high radiation areas are proposed. 
Therefore, the proposed change does not involve a significant 
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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A 5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Section Chief: Sheri R. Peterson.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: May 17, 1999.
    Description of amendment request: The proposed amendment would 
revise a note in Surveillance Requirement (SR) 3.3.8.1 in the Crystal 
River Unit 3 Improved Technical Specifications (ITS). The note 
currently states that, when Emergency Diesel Generator (EDG) Loss Of 
Power Start instrumentation is placed in an inoperable status solely 
for performance of this surveillance, entry into associated Conditions 
and Required Actions may be delayed for up to four hours provided the 
two channels monitoring the Function for the bus are OPERABLE or 
tripped. The proposed revision to the note states that entry into the 
Conditions and Required Actions of ITS Section 3.3.8 is not required 
provided the applicable Conditions and Required Actions of ITS Section 
3.8.1, ``Electrical Power Systems, AC Sources--Operating,'' are entered 
for the EDG being made inoperable. The proposed amendment would also 
delete a superceded 60-day surveillance frequency and the associated 
note which indicated that the frequency was not effective after 
November 23, 1997.
    Basis for proposed no significant hazards consideration 
determination: 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?

Proposed Improved Technical Specifications (ITS) Change A--Revision of 
Surveillance Note

    The note in Surveillance Requirement (SR) 3.3.8.1 involves the 
timing for placing Crystal River Unit 3 (CR-3) into the applicable 
Conditions and Required Actions when

[[Page 38027]]

performing the surveillance. The proposed revision will make the 
note consistent with the actual method of performing the 
surveillance at CR-3. The design and testing configuration does not 
allow CR-3 to use the relief provided by the note. As a result, the 
Conditions and Required Actions of ITS Section 3.3.8 are entered at 
the start of the surveillance performance. The design and testing 
configuration requires entry into ITS Section 3.8.1 Conditions prior 
to performing SR 3.3.8.1. The revised note would require entering 
the applicable Conditions and Required Actions of ITS Section 3.8.1 
for one Emergency Diesel Generator (EDG) being inoperable. This 
approach and the proposed note are conservative relative to the 
current note. The proposed note results in no changes to the method 
or to the timing of performing SR 3.3.8.1. Direct entry into ITS 
Section 3.8.1 Conditions will achieve the same final ITS condition 
as if Section 3.3.8 Conditions and Required Actions were entered. 
Therefore, the probability of occurrence and the consequences of any 
accident previously evaluated are unaffected by this change.

Proposed ITS Change B--Deletion of Frequency Note

    The note under Frequency in SR 3.3.8.1 involves the period of 
time that the 60 day surveillance frequency would be in effect. The 
60 day frequency was a temporary extension that was needed to 
implement modifications to the EDG during the 1997 CR-3 design 
outage. This was a one-time extension of the frequency. The note 
indicates this temporary nature of the 60 day frequency. Deleting 
the note is an editorial change since the surveillance has reverted 
back to its 31 day frequency and the note is no longer effective. 
Because the proposed deletion of the note is an editorial change, 
and no change is proposed to the current 31 day frequency, the 
probability of occurrence and the consequences of any accident 
previously evaluated are unaffected by this change.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?

Proposed ITS Change A--Revision of Surveillance Note

    The proposed revision of the note in SR 3.3.8.1 involves only 
the timing of entry into associated ITS Conditions and Required 
Actions. No changes are proposed to the existing ITS Conditions and 
Required Actions. The proposed change is conservative since it will 
require entering the appropriate Conditions and Required Actions 
immediately upon starting SR 3.3.8.1. Changing the timing for entry 
into ITS Conditions and Required Actions does not create the 
possibility of a new or different kind of accident from those 
evaluated previously.

Proposed ITS Change B--Deletion of Frequency Note

    Deletion of the note under SR 3.3.8.1 Frequency is an editorial 
change since the note is no longer effective. The current frequency 
for performing SR 3.3.8.1 is 31 days. This is the same frequency 
that was in effect prior to the one-time, temporary change of the 
frequency to 60 days. The editorial change of deleting the note that 
is no longer effective does not create the possibility of a new or 
different kind of accident from those evaluated previously.
    3. Involve a significant reduction in a margin of safety?

Proposed ITS Change A--Revision of Surveillance Note

    One manner in which a margin of safety related to a Surveillance 
Requirement might be affected would be if entry into a Limiting 
Condition for Operation (LCO) were delayed. The result of a delay in 
entering an LCO would be an increase in the time before a Required 
Action was taken, such as commencing a plant shutdown. Generally, 
such allowed times reflected in ITS Required Actions are based on 
some margin. Increasing the time allowed before starting a certain 
Required Action might result in a reduction of a margin of safety. 
However, the proposed ITS change allows entry into Section 3.8.1 
Conditions immediately rather than after a delay. The proposed ITS 
change does not change the final plant condition required by the 
ITS. Therefore, the proposed ITS change does not result in a 
reduction in a margin of safety.

Proposed ITS Change B--Deletion of Frequency Note

    Another manner in which a margin of safety related to a 
surveillance requirement might be affected would be if the frequency 
of performance were changed. Generally, margin might be reduced if 
the frequency were reduced (i.e., the interval between performing 
surveillances were increased). This proposed editorial change to 
delete the note in SR 3.3.8.1 Frequency does not result in a change 
to the surveillance frequency. Thus, the proposed deletion of the 
note does not affect the existing margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Section Chief: Sheri R. Peterson.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: June 11, 1999.
    Description of amendment request: The proposed amendment makes 
various plant organization title changes.
    Basis for proposed no significant hazards consideration 
determination: 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 facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed changes are administrative in nature and do not affect 
assumptions contained in plant safety analyses, the physical design 
and/or operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. None of 
the proposed changes involve a physical modification to the plant, a 
new mode of operation or a change to the UFSAR [Updated Final Safety 
Analysis Report] transient analyses. No Technical Specification 
Limiting Condition for Operation, Action statement or Surveillance 
Requirement is affected by any of the proposed changes. These 
proposed changes do not reduce the level of qualification, authority 
or accountability associated with the affected Technical 
Specification responsibilities. Further, the proposed changes do not 
alter the design, function, or operation of any plant component. 
Therefore, the proposed amendment does not affect the probability of 
occurrence or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed changes 
are administrative in nature and do not affect assumptions contained 
in plant safety analyses, the physical design and/or modes of plant 
operation defined in the plant operating license, or Technical 
Specifications that preserve safety analysis assumptions. The 
proposed changes do not introduce a new mode of plant operation or 
surveillance requirement, nor involve a physical modification to the 
plant. The proposed changes do not alter the design, function, or 
operation of any plant components. Therefore, the proposed amendment 
does not affect the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a reduction in a margin of safety. The 
proposed changes are administrative in nature. There is no reduction 
in the organization position qualifications, authority and 
accountability associated with the affected Technical Specification 
responsibilities. None of the proposed changes involve a physical 
modification to the plant, a new mode of operation or a change to 
the UFSAR transient analyses. No Technical Specification Limiting 
Condition for Operation, Action statement, or Surveillance 
Requirement is affected. Therefore, the proposed amendment does not 
reduce the margin of safety.


[[Page 38028]]


    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.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: April 12, 1999.
    Description of amendment request: The proposed amendment would 
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
Surveillance Requirement (SR) 3.6.1.3.7 to allow a representative 
sample of reactor instrumentation line excess flow control valves 
(EFCV) to be tested every 24 months, instead of testing each EFCV every 
24 months.
    Basis for proposed no significant hazards consideration 
determination: 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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The current SR frequency requires each reactor instrumentation 
line EFCV to be tested every 24 months. The EFCVs at DAEC are 
designed so that they will not close accidentally during normal 
operation, will close if a rupture of the instrument line is 
indicated downstream of the valve, can be reopened when appropriate, 
and have their status indicated in the control room (reference DAEC 
UFSAR [updated final safety analysis report] 1.8.11). This proposed 
change allows a reduced number of EFCVs to be tested every 24 
months. There are no physical plant modifications associated with 
this change. Industry operating experience demonstrates a high 
reliability of these valves. Neither EFCVs nor their failures are 
capable of initiating previously evaluated accidents; therefore 
there can be no increase in the probability of occurrence of an 
accident regarding this proposed change.
    Instrument lines connecting to the Reactor Coolant Pressure 
Boundary (RCPB) with EFCVs installed also have a flow-restricting 
orifice upstream of the EFCV. The consequences of an unisolable 
rupture of such an instrument line [have] been previously evaluated 
in response to Regulatory Guide (RG) 1.11 (DAEC UFSAR 1.8.1.1). That 
evaluation assumed a continuous discharge of reactor water for the 
duration of the detection and cooldown sequence (3.5 hours). 
Therefore, although not expected to occur as a result of this 
change, the postulated failure of an EFCV to isolate as a result of 
reduced testing is bounded by this previous evaluation. Therefore, 
there is no increase in the previously evaluated consequences of the 
rupture of an instrument line and there is no potential increase in 
the consequences of an accident previously evaluated as a result of 
this change.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This proposed change allows a reduced number of EFCVs to be 
tested each operating cycle. No other changes in requirements are 
being proposed. Industry operating experience demonstrates the high 
reliability of these valves. The potential failure of an EFCV to 
isolate by the proposed reduction in test frequency is bounded by 
the previous evaluation of an instrument line rupture. This change 
will not physically alter the plant (no new or different type of 
equipment will be installed). This change will not alter the 
operation of process variables, structures, systems, or components 
as described in the safety analysis. Thus, a new or different kind 
of accident will not be created.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The consequences of an unisolable rupture of an instrument line 
[have] been previously evaluated in response to RG 1.11 (reference 
DAEC UFSAR 1.8.1.1). That evaluation assumed a continuous discharge 
of reactor water for the duration of the detection and cooldown 
sequence (3.5 hours). The only margin of safety applicable to this 
proposed change is considered to be that implied by this evaluation. 
Since a continuous discharge was assumed in this evaluation, any 
potential failure of an EFCV to isolate postulated by this reduced 
testing frequency is bounded and does not involve a significant 
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.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869
    NRC Section Chief: Claudia M. Craig

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: April 30, 1999.
    Description of amendment request: The proposed amendment would 
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
Surveillance Requirement (SR) 3.4.3.1 to revise the safety function 
lift setpoint tolerance limits for the main safety valves (SVs) and the 
safety/relief valves (SRVs). The current tolerance bands for the SVs 
and SRVs would be revised from -3% to +1% to a new band of plus or 
minus 3%.
    Basis for proposed no significant hazards consideration 
determination: 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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change allows an increase in the as-found SV and 
SRVs safety mode setpoint tolerance, determined by test after the 
valves have been removed from service, from +1%/-3% to plus or minus 
3%.
    The proposed change does not alter the TS requirements on the 
nominal SV or SRV safety mode lift setpoints, the SRV relief mode 
setpoints, the required frequency for the SV or SRV lift setpoint 
tests, or the number of SVs and SRVs required to be operable.
    Consistent with current requirements, this change continues to 
require that these valves be adjusted to within plus or minus 1% of 
their nominal lift setpoints following testing. This change does not 
change the behavior and operation of any SV or SRV and therefore has 
no significant impact on reactor operation. It also has no 
significant impact on response to any perturbation of reactor 
operation including transients and accidents previously analyzed in 
the Updated Final Safety Analysis Report (UFSAR).
    This change does not involve physical changes to the valves, nor 
does it change the operating characteristics or safety function of 
the valves. The proposed TS revision involves no significant changes 
to the operation of any systems or components in normal or accident 
operating conditions and no changes to existing structures, systems, 
or components. Therefore these changes will not increase the 
probability of an accident previously evaluated.
    Generic considerations related to the change in setpoint 
tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure 
Relief Technical Specification Revision Licensing Topical Report,'' 
and were reviewed and approved by the NRC in a Safety Evaluation 
(SE) dated March 8, 1993. The plant specific evaluations, required 
by the NRC's SE and performed to support this proposed change, show 
that there is adequate

[[Page 38029]]

margin to the design core thermal limits and to the reactor vessel 
pressure limits using a plus or minus 3% setpoint tolerance. They 
also show that operation of the high pressure coolant injection 
(HPCI) and reactor core isolation cooling (RCIC) systems will not be 
adversely affected and the containment response from a loss of 
coolant accident will be acceptable. The plant systems associated 
with these proposed changes will still be capable of meeting all 
applicable design basis requirements and retain the capability to 
mitigate the consequences of accidents described in the UFSAR. 
Therefore, these changes will not involve a significant increase in 
the consequences of any accident previously evaluated.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    These proposed changes were developed in accordance with the 
provisions contained in the NRC SE, dated March 8, 1993, for the 
``BWR Owners Group Inservice Pressure Relief Technical Specification 
Revision Licensing Topical Report,'' NEDC-31753P. The revised SV and 
SRV setpoint tolerance limit will not adversely impact the operation 
of any safety related component or equipment. Since the proposed 
changes involve no significant hardware changes, no significant 
changes to the operation of any systems or components, and no 
changes to existing structures, systems, or components, there can be 
no impact on the occurrence of any accident.
    The proposed change would not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change to allow an increase in the SV and SRV safety 
mode setpoint tolerance from +1%/-3% to plus or minus 3% does not 
alter the nominal SV or SRV lift setpoints or the number of SVs or 
SRVs required to be operable. This change does not involve physical 
changes to the valves, nor does it change the operating 
characteristics or the safety function of the valves. The proposed 
change does not involve a physical alteration of the plant. No new 
or different equipment is being installed. There is no alteration to 
the parameters within which the plant is normally operated. As a 
result no new failure modes are being introduced. There are no 
changes in the methods governing normal plant operation, nor are the 
methods utilized to respond to plant transients altered.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety. Establishment of the plus or minus 3% SV and SRV 
setpoint tolerance limit will not adversely impact the operation of 
any safety related component or equipment. Engineering evaluations 
concluded that there are no significant impacts on fuel thermal 
limits, safety related systems, structures or components, and no 
significant impact on the accident analyses associated with the 
proposed changes.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation. The proposed change does not significantly 
impact any safety analysis assumptions or results.

    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.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: May 10, 1999.
    Description of amendment request: The proposed amendment would 
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
Section 2.1.1.2, to revise the Safety Limit Minimum Critical Power 
Ratio (SLMCPR) to support operation with GE-12 fuel with a 10x10 pin 
array.
    Basis for proposed no significant hazards consideration 
determination: 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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    There is no change to any plant equipment other than the fuel. 
The SLMCPR protects the fuel in accordance with the design bases. 
The SLMCPR calculations limit the bundle power to ensure the 
critical power ratio remains unchanged. Therefore, there is not an 
increase in the probability of transition boiling. The basis of the 
SLMCPR calculation remains the same, ensuring that greater than 
99.9% of all fuel rods in the core avoid transition boiling if the 
limit is not violated. Therefore, there is no increase in the 
probability of occurrence of a previously evaluated accident.
    The fundamental sequences of accidents have not been altered. 
The Minimum Critical Power Ratio (MCPR) Operating Limits are 
selected such that potentially limiting accidents do not cause the 
MCPR to decrease below the SLMCPR anytime during the accident. 
Therefore, there is no impact on any of the limiting accidents. 
Therefore there is no increase in the consequences of any accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR values are designed to ensure that fuel damage from 
transition boiling does not occur in at least 99.9% of the fuel rods 
as a result of the limiting postulated accident. The values are 
calculated in accordance with NRC-approved General Electric methods. 
The approved General Electric methods are comprehensive for ensuring 
that fuel designs will perform within acceptable bounds. The SLMCPR 
ensures that the fuel is protected in accordance with the design 
basis. The function, location, operation, and handling of the fuel 
remain unchanged. Therefore, the possibility of a new or different 
kind of accident is not created.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The SLMCPR values do not alter the design or function of any 
plant system. The new values were calculated using NRC-approved 
methods to maintain the same margin of safety as presently exists 
for the prevention of transition boiling. At least 99.9% of the fuel 
rods will avoid transition boiling if the SLMCPR is not violated. 
Therefore, a significant reduction in a margin of safety is not 
involved.

    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.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: May 10, 1999.
    Description of amendment request: The proposed amendment would 
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
to: (1)

[[Page 38030]]

insert NOTE for Limiting Condition for Operation (LCO) 3.7.4 that would 
allow intermittent opening of the control building boundary under 
administrative control; (2) add a CONDITION, REQUIRED ACTION and 
COMPLETION TIME to LCO 3.7.4 for when both standby filter unit (SFU) 
subsystems are inoperable due to inoperable control building boundary 
in MODES 1, 2, and 3; (3) re-letter items in LCO 3.7.4 for consistency; 
and (4) revise LCO 3.7.4 CONDITION D (new CONDITION E) to add ``for 
reasons other than CONDITION B.''
    Basis for proposed no significant hazards consideration 
determination: 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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Requiring the plant to enter LCO 3.0.3 when the control building 
pressure envelope is not intact is excessively restrictive. This 
change provides less restrictive requirements for operation of the 
facility. These less restrictive requirements do not result in 
operation that will increase the probability of initiating an 
analyzed event. The proposed change is acceptable because of the low 
probability (less than 3.04 x 10-8) of a DBA [design 
basis accident] occurring during the 24 hour Completion Time, and 
the availability the SFU system to provide a filtered environment 
(albeit with potential control room in-leakage).
    Intermittent opening of the control building boundary requires 
controls which consist of stationing a dedicated individual at the 
opening who is in continuous communication with the control room. 
This individual will have a method to rapidly close the opening when 
a need for control room isolation is indicated. For entry and exit 
through doors the administrative control is performed by the person 
entering or exiting the area. As a result, the consequences of any 
accident previously evaluated are not significantly increased.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. This change does not involve new or different equipment 
being installed at the facility. The proposed change is acceptable 
because of the low probability (less than 3.04 x 10-8) of 
a DBA occurring during the 24 hour Completion Time, and the 
availability of the SFU system to provide a filtered environment 
(albeit with potential control room in-leakage).
    Intermittent opening of the control building boundary requires 
controls which consist of stationing a dedicated individual at the 
opening who is in continuous communication with the control room. 
This individual will have a method to rapidly close the opening when 
a need for control building isolation is indicated. For entry and 
exit through doors the administrative control is performed by the 
person entering or exiting the area.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety. Requiring the plant to enter LCO 
3.0.3 when the control room ventilation envelope is not intact is 
excessively restrictive. The proposed change is acceptable because 
of the low probability (less than 3.04 x I0-8) of a DBA 
occurring during the 24 hour Completion Time.
    Intermittent opening of the control room boundary requires 
controls which consist of stationing a dedicated individual at the 
opening who is in continuous communication with the control room. 
This individual will have a method to rapidly close the opening when 
a need for control building isolation is indicated. For entry and 
exit through doors the administrative control is performed by the 
person entering or exiting the area.
    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.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 8, 1999.
    Description of amendment request: The proposed change corrects the 
described method by which the Standby Gas Treatment system heaters are 
to be tested. This change is necessary because the reference provided 
in Technical Specification Section 5.5.7e is in error.
    Basis for proposed no significant hazards consideration 
determination: 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:

    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The correction of an error and clarification of a testing 
method does not alter any of the precursors assumed in the CNS 
[Cooper Nuclear Station] accident analysis. The proposed wording for 
testing SGT [Standby Gas Treatment] heaters is in accordance with 
ASME N510-1989, Section 14.5.1, ``Testing of Nuclear Air Treatment 
Systems.'' Since the proposed change does not affect this portion of 
plant design and operation, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change will not create the possibility of a new or 
different kind of accident than evaluated in the Updated Safety 
Analysis Report (USAR). The proposed change does not result in any 
physical change to CNS structures, systems, or components, nor does 
it change the fit, form, or function of any equipment or components 
taken credit for in the accident analyses described in the USAR. 
Therefore, correction of a test reference and specific description 
of the testing method for the SGT heaters does not create the 
possibility of a new or different kind of accident.
    The proposed change will not create a significant reduction in 
the margin of safety. The proposed change does not alter the design 
or administrative controls necessary to ensure the required 
performance of the physical barriers during acticipated operational 
occurrences and postulated accidents. This conclusion is based on 
the fact that the proposed change corrects an erroneous reference, 
conforms to industry standards, and is consistent with past and 
current operating practice at CNS; therefore, the proposed change 
does not create a significant 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.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 15, 1999.
    Description of amendment request: The proposed change would allow 
the use of the service water (SW) system to directly supply cooling 
water to the reactor equipment cooling (REC) system during a loss-of-
coolant accident (LOCA) event. The present maximum allowable REC water 
leakage rate is based on the requirement that there will be sufficient 
water in the REC surge tank to allow the REC system to fulfill its 
safety function for 30 days post-LOCA condition. A proposed Updated 
Safety Analysis Report (USAR) revision would allow Cooper Nuclear 
Station (CNS) to

[[Page 38031]]

revise the maximum allowable REC system leakage during normal power 
operation such that the REC system surge tank would assure that the REC 
would fulfill its safety function for at least the first 7 days 
following a large break LOCA. The SW system would fulfill the safety 
functions of the REC system, if required, for the remaining duration of 
the accident.
    Basis for proposed no significant hazards consideration 
determination: 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. The licensee states that the 
proposed request:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not involve a significant increase in 
the probability of an accident previously evaluated in the USAR 
since there are no hardware changes associated with this USAR 
change. Procedure changes associated with this USAR change are 
limited to direction on which division of SW/REC backup to initiate 
first, and incorporation of new system leakage limits into 
surveillance procedures.
    The proposed change also does not involve a significant increase 
in the consequences of an accident previously evaluated in the USAR. 
This conclusion is based on the safety evaluation (Attachment 2 [of 
the June 15, 1999, application]) which demonstrates that the SW 
system will fulfill the safety functions of the REC system in a post 
LOCA condition and thus the proposed change will not affect the 
performance and reliability of the REC system. The emergency systems 
cooled by the REC system, the ECCS [emergency core cooling] systems 
and their room coolers, will therefore also fulfill their safety 
function when directly supplied by the SW system.
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the USAR. The proposed license amendment does not introduce any new 
equipment or hardware changes. It does, however, allow the SW system 
to perform a different type of function than it is presently 
licensed to perform in a post LOCA condition. This SW system post 
LOCA function has been previously demonstrated to fulfill the 
functions of the REC in a non LOCA emergency shutdown which are the 
same as the functions required following a LOCA.
    3. Does not create a significant reduction in the margin of 
safety.
    The proposed activity does not involve a significant reduction 
in the margin to safety. The safety evaluation (Attachment 2) 
demonstrates that the SW system will perform the required REC post 
LOCA functions. There is an added required operator action which is 
to align the SW system to directly supply cooling water to the REC 
critical loops. As discussed in the safety evaluation [of the June 
15, 1999, application], this action can be performed from the main 
control room utilizing one control switch and there is sufficient 
control room indication for the operator to be alerted to the need 
for the use of service water backup. There is also sufficient time 
for the operator to perform the task. Trending (prior to a 
postulated LOCA) routinely provides the control room operator with 
REC system leakage information. In a post LOCA situation, this 
leakage information would assist the operator in taking timely 
action to initiate the service water back-up before the need is 
alarmed in the control room.

    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.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 15, 1999.
    Description of amendment request: The purpose of the requested 
license amendment is to revise the Updated Safety Analysis Report 
(USAR) to incorporate the latest analysis to demonstrate adequate net 
positive suction head (NPSH) for the low pressure emergency core 
cooling system (ECCS) pumps following a large break loss-of-coolant 
accident (LOCA). Specifically, the change would allow (1) reliance on a 
slightly larger amount of containment overpressure for residual heat 
removal (RHR) and core spray (CS) pump operation during worst-case 
long-term LOCA conditions (greater than 1000 seconds) while still 
maintaining original license margins of 3 and 6 pounds per square inch 
(psi), respectively, for the difference between minimum available 
containment pressure and the pressure required for minimum pump NPSH, 
(2) reliance on a small amount of containment overpressure for CS pump 
runout during worst-case short-term LOCA conditions (less than 10 
minutes) while still maintaining an adequate pressure margin of at 
least 5 psi, and (3) the use of ANS 5.1 decay heat model in the USAR 
Section 5.2.6 as currently presented based on analysis justifying the 
use of this model as described in the amendment request.
    Basis for proposed no significant hazards consideration 
determination: 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. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not involve an increase in the 
probability of an accident previously evaluated in the USAR. There 
are no changes being proposed to the maintenance, operation, or 
design of plant systems or equipment postulated to initiate 
accidents or transients.
    The proposed change does not involve an increase in the 
consequences of an accident previously evaluated in the USAR. This 
conclusion is based on the conclusions of the safety evaluation 
(Attachment 2 [of the June 15, 1999, application]). This safety 
evaluation demonstrates that the containment overpressure is 
sufficiently conservative, and that the calculated margins between 
the available containment overpressure and the overpressure required 
to assure adequate low pressure ECCS pump NPSH are such that ECCS 
pump operation, as credited in the CNS [Cooper Nuclear Station] 
accident analysis, remains unchanged.
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the USAR. The proposed license amendment does not introduce any new 
equipment or hardware changes. The attached safety evaluation 
demonstrates that the only equipment affected by this License 
Amendment are the low pressure ECCS pumps and that these will retain 
their ability to function following a LOCA.
    3. Does not create a significant reduction in the margin of 
safety.
    The proposed activity does not involve a significant reduction 
in the margin of safety. The safety evaluation (Attachment 2) 
demonstrates that, although there is an increased reliance on 
containment overpressure to assure adequate low pressure ECCS pump 
NPSH, there remains sufficient margin to provide confidence that the 
ECCS pumps will operate as required. Sufficient margin is 
demonstrated with the added conservatism of a 2-sigma (2 standard 
deviation) uncertainty in the decay heat model, increased suction 
strainer debris loading, increased RHR heat exchanger tube plugging 
margin, and increases in SW [Service Water] and Suppression Pool 
temperatures. The minimum margin available between available 
overpressure and required overpressure is at least 5 psi for CS 
(just prior to 10 minutes) and at least 3 psi for RHR (well after 10 
minutes).

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 38032]]

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.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: June 23, 1999.
    Description of amendment request: The proposed change to the 
Technical Specifications would increase the allowed outage time for the 
Control Room Air Conditioning Subsystem from 30 days to 60 days, on a 
one-time basis only, for each train.
    Basis for proposed no significant hazards consideration 
determination: 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 operational requirements for the Control Room Air 
Conditioning Subsystems (CRACS) are contained in Technical 
Specification 3.7.6.2 ``Control Room Subsystems Air Conditioning.'' 
This Limiting Condition for Operation (LCO) requires that two 
independent Control Room Air Conditioning Subsystems (trains) be 
operable during all modes of operation. The LCO action statement for 
operational modes 1, 2, 3 and 4, with one Control Room Air 
Conditioning Subsystem inoperable, states: ``restore the inoperable 
system to operable status within 30 days or be in at least Hot 
Standby [Mode 3] within the next 6 hours and in Cold Shutdown within 
the following 30 hours.'' The LCO action statement for operational 
modes 5 and 6 with one Control Room Air Conditioning Subsystem 
inoperable, states: ``restore the inoperable system to operable 
within 30 days or initiate and maintain operation of the remaining 
OPERABLE Control Room Air Conditioning Subsystem or immediately 
suspend all operations involving CORE ALTERATIONS or positive 
reactivity changes.''
    The proposed change adds the following note: ``* For cycle 7, 
the allowable outage time may be extended to 60 days, on a one-time 
basis, for each train, to implement modifications to the control 
room air conditioning subsystems. The provisions of specifications 
3.0.4 and 4.0.4 are not applicable during the implementation of 
modifications to the air conditioning subsystems.''
    This change is a one-time only change to Technical Specification 
3.7.6.2 in order to facilitate the installation of a design change 
to the CRACS during the present operating cycle. This change will 
not affect the existing 30 [day] AOT period presently in place in 
Technical Specification 3.7.6.2 which requires specific actions in 
the event that the CRACS is determined to be inoperable for any 
other reason. The design basis accidents are not affected as a 
result of the proposed one-time change to the Technical 
Specifications. The CRACS are support subsystems which can only 
contribute to the initiation of an accident if the whole function is 
lost. The plant would be required to shutdown before this occurred. 
The proposed change does not adversely affect accident initiators or 
precursors nor alter the design assumptions, conditions, 
configuration of the facility (other than the CRACS) or the manner 
in which the plant is operated nor does it adversely affect the 
response of the plant to a transient or accident. This one-time 
change is to be utilized only during the present operating cycle 
(cycle 7) in order to facilitate the implementation of a design 
change to modify the existing safety-related refrigerant subsystems 
(one train at a time) and replace them with safety-related chilled 
water subsystems. This design change is being implemented to improve 
the overall reliability of the safety-related subsystems.
    The consequences of an extended loss of the operating CRACS and 
the non-safety related chilled water subsystem, during all modes of 
operation, would result in a slow gradual rise in control room 
temperature. The temperature of the control room is normally 
maintained between 70 to 72 deg.F at the discretion of the Unit 
Shift Supervisor utilizing a non-safety-related train of CRACS. In 
the event that the control room temperature increased to a 
temperature greater than 75 deg.F, plant procedures require starting 
other equipment in the non-safety-related chilled water subsystem or 
a safety-related train of CRACS to restore control room temperature 
to its normal operating band. In the unlikely event that the non-
safety-related chilled water subsystems and the operable safety-
related train of CRACS fail during the proposed 60 day AOT period, 
Technical Specification 3.7.6.2 would require that actions be 
commenced to place the plant in a shutdown condition. Additionally, 
alternative actions to reduce control room temperature could also be 
initiated as identified in a plant procedure. It has been 
conservatively determined that safety-related equipment in the 
control room can be operated continuously up to 90\F in an 
environment without affecting the capability of the equipment.
    The exception to specifications 3.0.4 and 4.0.4 as stated in the 
proposed one-time change to Technical Specification 3.7.6.2 will not 
involve an increase in the probability or consequences of an 
accident. TS 3.0.4 prohibits entry into a mode when the conditions 
for the LCO are not met and the associated action(s) requires a 
shutdown if they are not met within a specified time interval. 
Surveillance Requirement 4.0.4 prohibits entry into a mode unless 
the associated surveillance requirement(s) has been performed within 
the stated interval. During the implementation of the modification, 
when one safety-related train of CRACS is inoperable, it is possible 
that a plant shutdown could occur due to reasons unrelated to the 
planned modifications of the CRACS. As stated above, the CRACS are 
support subsystems which do not contribute to the initiation of any 
accident previously evaluated. Entry of the plant into an 
operational mode from a shutdown mode as a result of the proposed 
modification does not adversely affect accident initiators or 
precursors nor alter the design assumptions, conditions, 
configuration of the facility (other than the CRACS) or the manner 
in which the plant is operated nor does it adversely affect the 
response of the plant to a transient or accident. The functions of 
the CRACS to provide a controlled environment inside of the control 
room complex to ensure the comfort of the plant operators and to 
ensure adequate climate conditions for the operability of equipment 
will not be impaired in any way as a result of a plant mode change. 
The remaining actions identified in TS 3.7.6.2 are unchanged as a 
result of the proposed change. The risk significance involved with 
removing a safety-related train of the CRACS during power operation 
or during refueling conditions is low based on the short period (60 
days per train) and consequences of losing this function. The CRACS 
is excluded from modeling in the Seabrook Station Probabilistic Risk 
Assessment (PRA) due to its extremely low risk significance.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
since it is a support system and the loss of function will require a 
plant shutdown. The proposed change adds the following note which 
pertains to both affected action statements: ``* For cycle 7, the 
Allowable Outage Time may be extended to 60 days, on a one-time 
basis, for each train during the implementation of modifications to 
the control room air conditioning subsystems. The provisions of 
specifications 3.0.4 and 4.0.4 are not applicable during the 
implementation of modifications to the air conditioning 
subsystems.'' As previously identified, this change is a one-time 
only change to Technical Specification 3.7.6.2 in order to 
facilitate the installation of a design change to CRACS during the 
present operating cycle.
    The CRACS are support subsystems which do not contribute to the 
creation of a new or different kind of accident from any previously 
evaluated nor is it used to mitigate the consequences of a transient 
or accident. The functions of the CRACS are to provide a controlled 
environment inside of

[[Page 38033]]

the control room complex to ensure the comfort of the plant 
operators and to ensure adequate climate conditions for the 
operability of equipment. The CRACS consists of two independent 
safety-related air conditioning trains that provide cooling of 
recirculated control room air. Due to previous reliability problems 
with the CRACS, an additional non-safety chilled water subsystem has 
been installed to provide control room cooling on a continuous 
basis. Baseload operation of the non-safety related chilled water 
subsystem to provide control room cooling reduces the operational 
load on the safety-related refrigerant trains.
    Implementation of the modification to the CRACS subsystems 
during the 60 day AOT duration in no way affects the availability of 
the non-safety-related chilled water subsystem or the operable 
safety-related train of the CRACS to meet the control room cooling 
requirements. The proposed modification removes freon from the 
control room complex and the quantity of chilled water in the closed 
loop system is too small to become a flood hazard. The consequences 
of an extended loss of the operating CRACS train and the non-safety 
related chilled water subsystem would result in a slow gradual rise 
in control room temperature. In the event that control room 
temperature increased to a temperature greater than 75 deg.F, plant 
procedures require starting either the non-safety-related chilled 
water subsystem or a safety-related train of CRACS to restore 
control room temperature. Additionally, in the unlikely event of a 
loss of the non-safety related chilled water subsystem and the 
operable safety-related train of the CRACS, Technical Specification 
3.7.6.2 would require that actions be taken to place the plant in a 
shutdown condition.
    It has been conservatively determined that safety-related 
equipment in the control room can be operated continuously in an 
environment up to 90 deg.F without affecting the capability of the 
equipment. This proposed change will not affect the existing 30 day 
AOT period presently in place in Technical Specification 3.7.6.2 
which requires specific actions in the event that the CRACS is 
determined to be inoperable for any other reason.
    The exception to specifications 3.0.4 and 4.0.4 as stated in the 
proposed one-time change to Technical Specification 3.7.6.2 will not 
involve the creation of an accident of any type. During the 
implementation of the proposed modification, when one safety-related 
train of CRACS is inoperable, it is possible that a plant shutdown 
could occur due to reasons unrelated to the planned modifications of 
the CRACS. Entry of the plant into an operational mode from a 
shutdown mode as a result of the proposed modification does not 
adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, configuration of the facility or the 
manner in which the plant is operated nor the manner that it 
responds to a transient or accident. The functions of the CRACS to 
provide a controlled environment inside of the control room complex 
to ensure the comfort of the plant operators and to ensure adequate 
climate conditions for the operability of equipment will not be 
impaired in any way as a result of a plant mode change. The 
remaining actions identified in TS 3.7.6.2 are unchanged as a result 
of the proposed change.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed one-time change to Technical Specification 3.7.6.2 
will not involve a significant reduction in the margin of safety. 
The functions of the CRACS are to provide a controlled environment 
inside of the control room complex to ensure the comfort of the 
plant operators and to ensure adequate climate conditions for the 
operability of equipment. The CRACS consists of two independent 
safety-related air conditioning trains that provide cooling of 
recirculated control room air. Additionally, the Seabrook Station 
design incorporates the use of a non-safety chilled water subsystem 
(which is not within the scope of the Technical Specifications) to 
provide baseload cooling of the control room on a continuous basis.
    Implementation of the modification to the CRACS subsystems 
during the 60 day AOT duration does not result in a significant 
reduction in the plant margin of safety. As previously identified, 
the CRACS is a support subsystem and the existing Technical 
Specifications will require a plant shutdown on a loss of function. 
The risk significance involved with removing a safety-related train 
of the CRACS is extremely low based on the short period (60 days per 
train) and the consequences of losing this function. The potential 
that the non-safety-related chilled water subsystem and the operable 
safety-related train of CRACS simultaneously fail during the 
proposed 60 day AOT period of each safety-related train (120 days 
total) is considered unlikely. In the event that control room 
temperature increased to a temperature greater than 75 deg.F, plant 
procedures require starting either the non-safety-related chilled 
water subsystem or a safety-related train of CRACS to restore 
control room temperature. Additionally, in the unlikely event of a 
loss of the non-safety related subsystem and the operable safety-
related train of the CRACS, Technical Specification 3.7.6.2 would 
require that actions be taken to place the plant in a shutdown 
condition. Alternative actions to reduce control room temperature 
could also be initiated as identified in a plant procedure. It has 
been conservatively determined that safety-related equipment in the 
control room can be operated continuously in an environment up to 
90 deg.F without affecting the capability of the equipment.
    The exception to specifications 3.0.4 and 4.0.4 as stated in the 
proposed one-time change to Technical Specification 3.7.6.2 will not 
reduce the margin of safety. During the implementation of the 
proposed modification, when one safety-related train of CRACS is 
inoperable, it is possible that a plant shutdown could occur due to 
reasons unrelated to the planned modifications of the CRACS. Entry 
of the plant into an operational mode from a shutdown mode as a 
result of the proposed modification does not adversely affect 
accident initiators or precursors nor alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. The functions of the CRACS to provide a 
controlled environment inside of the control room complex to ensure 
the comfort of the plant operators and to ensure adequate climate 
conditions for the operability of equipment will not be impaired in 
any way as a result of a plant mode change. The remaining actions 
identified in TS 3.7.6.2 are unchanged as a result of the proposed 
change.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: May 17, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) section 4.4.6.2.2.e to replace the 
reference to American Society of Mechanical Engineers (ASME) Code 
paragraph IWV-3472(b) which pertains to the frequency of leakage rate 
testing for 6-inch, nominal pipe size valves and larger with the 
requirement that the surveillance interval and frequency of 
surveillance leakage rate testing for these valves be performed 
pursuant to the requirements of TS 4.0.5, ``Operations and Surveillance 
Requirements.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided the NRC its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Eliminating the reference to ASME Code paragraph IWV-3427(b) and

[[Page 38034]]

performing pressure isolation valve (PIV) testing pursuant to TS 4.0.5 
does not change the test conditions for PIV leakage testing and is 
consistent with the currently analyzed configurations. This change 
eliminates an unnecessary test requirement and incorporates 
Westinghouse Owner's Group (WOG) Standard Technical Specifications 
(STS) frequency requirements that are deemed to substantially reduce 
the probability of an intersystem loss-of-coolant-accident. This change 
in testing frequency requirements does not affect the accident 
mitigation capabilities of the reactor coolant system (RCS) PIVs. This 
change is bounded by existing accident analyses. Therefore, it is 
concluded that, with the reduced probability of previously analyzed 
accidents, and no effect on accident mitigation, the proposed revision 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Eliminating the IWV-3427(b) trending for 6-inch and larger valves 
(and the accompanying increased frequency testing requirement) does not 
significantly change actual testing frequencies since the frequencies 
continue to be addressed by the remaining TS requirements. This change 
does not affect the ability of a PIV to perform its required RCS 
pressure isolation safety function of limiting RCS leakage to prevent 
overpressure failure of attached low pressure systems. The frequency of 
testing or the testing itself are not initiating events to postulated 
accidents. Therefore, the proposed revision does not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. Involve a significant reduction in the margin of safety.
    There is no impact on the Margin of Safety as defined in the bases 
of any TS, the Updated Final Safety Analysis Report, or other licensing 
basis commitments resulting from the elimination of the reference to 
ASME Code paragraph IWV-3427(b). Periodic surveillances provide 
continued assurance in the capability of safety related equipment to 
perform its design safety (accident mitigating) function and are not 
used to establish the margin of safety for accident mitigation. 
Therefore, the frequency of surveillance testing of the PIVs has no 
impact on the margins of safety assumed in analyzed accidents.
    In its evaluation, NNECO concluded, based on its evaluation as 
required by 10 CFR 50.92, that the proposed revision does not involve a 
SHC.
    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.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment requests: May 13, 1999.
    Description of amendment requests: The proposed amendments would 
modify Technical Specification (TS) 6.2.A.2, ``Onsite and Offsite 
Organizations,'' to reflect a change in the organizational structure 
implemented on March 1, 1999.
    Basis for proposed no significant hazards consideration 
determination: 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 amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change is administrative in nature and does not 
significantly affect any system that is a contributor to initiating 
events for previously evaluated accidents. Neither does the change 
significantly affect any system that is used to mitigate any 
previously evaluated accidents. Therefore, the proposed change does 
not involve any significant increase in the probability or 
consequence of an accident previously evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The proposed change is administrative in nature and does not 
alter the design, function, or operation of any plant component nor 
does the proposed change install any new or different equipment, 
therefore the possibility of a new or different kind of accident 
from those previously analyzed has not been created.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The proposed change is administrative in nature and does not 
involve a significant reduction in the margin of safety associated 
with the safety limits inherent in either the fuel cladding, RCS 
[reactor coolant system] boundary, reactor containment, or other 
structures, systems, or components (SSCs).

    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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Claudia M. Craig.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating 
Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: June 7, 1999.
    Description of amendment request: The proposed amendments, if 
approved, would revise Technical Specifications (TS) Section 3/4.4.3 
and its associated TS Bases to reflect changes to refine and clarify 
the action statement concerning inoperable reactor coolant leakage 
detection systems.
    Basis for proposed no significant hazards consideration 
determination: 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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS changes directly establish the minimum 
acceptable level of Reactor Coolant System (RCS) leakage detection 
instrumentation required to support plant power operations. The 
level of RCS leakage detection capability inherent with the proposed 
TS change will continue to provide acceptable early warning 
detection of potential RCS pressure boundary degradation as required 
under 10 CFR 50.36 (c)(2)(ii) (A) Criterion 1.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes only affect systems associated with the 
detection of accidents involving degradation of the RCS

[[Page 38035]]

pressure boundary. The proposed TS changes do not involve any 
physical changes to plant structures, systems, or components. The 
RCS Leakage Detection Systems will continue to function as designed 
in all modes of operation. No new accident type is created as a 
result of the proposed changes. No new failure mode for any 
equipment is created. The changes are consistent with the guidance 
provided in [Standard Technical Specifications General Electric 
Plants BWR/4 dated April 1995] NUREG-1433, Revision 1, pertaining to 
RCS Leakage Detection.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in the margin of safety.
    The TS Limiting Conditions for Operation (LCO) specify for 
systems and equipment important to safety, the minimum level of 
operability required to permit continued power operation. The 
proposed TS changes revise this minimum level of operability by 
permitting long term plant operation with the removal of the Drywell 
Unit Coolers Condensate Flow Rate Monitoring System from service. 
Currently, this condition would permit the plant to continue to 
operate for up to 30 days. This change is not a reduction in the 
margin of safety since:
    The proposed Technical Specification LCO change for RCS Leakage 
Detection Systems maintains four (4) diverse methods of detecting 
RCS leakage and permits continuous operation with the Drywell Unit 
Coolers Condensate Flow Rate Monitors out of service provided that 
more frequent surveillance checks are provided for the containment 
atmosphere monitoring system. The proposed TS change institutes the 
additional surveillance requirements.
    The LGS reactor coolant pressure boundary was designed to ASME 
Class 1, Seismic Category I design criteria with no special 
dispensation which would warrant such additional RCS leakage 
detection capability or more stringent LCO criteria than those 
generically approved under the Improved Standard Technical 
Specifications.
    Review of the TS Bases Section and UFSAR identified no 
discussions regarding margin of safety for the RCS Leakage Detection 
Systems, which would be reduced by the proposed Technical 
Specification LCO change. It is further demonstrated that an 
acceptable margin of safety exists based on the generic regulatory 
approval of the Improved Standard Technical Specifications which 
will remain bounded by the proposed LGS TS changes.
    Therefore, the proposed TS changes do not involve a significant 
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.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

Portland General Electric Company, Docket No. 50-344, Trojan Nuclear 
Plant, Columbia County, Oregon

    Date of amendment request: August 27, 1998.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating (Possession-Only) License and the 
Permanently Defueled Technical Specifications. Multiple license 
conditions and technical specification requirements are proposed to be 
deleted to reflect the transfer of the nuclear spent fuel from the 
existing 10 CFR Part 50 licensed area to the 10 CFR Part 72 Independent 
Spent Fuel Storage Installation (ISFSI) area.
    Basis for proposed no significant hazards consideration 
determination: 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.
    This proposed amendment reflects removal of the spent nuclear 
fuel from the 10 CFR 50 licensed area and transfer of the spent 
nuclear fuel to the 10 CFR 72 ISFSI licensed area. The probability 
and consequences of accidents associated with storage of spent 
nuclear fuel within the TNP [Trojan Nuclear Plant] ISFSI were 
evaluated as part of PGE's 10 CFR 72 license application. Following 
completion of the transfer of the spent nuclear fuel to the 10 CFR 
72 licensed ISFSI and in light of the revised Appendix A Technical 
Specification, Section 4.2, that precludes storage of spent nuclear 
fuel within the 10 CFR 50 licensed area, the potential for accidents 
associated with the storage and handling of fuel in the 10 CFR 50 
licensed area will be eliminated. Therefore, deleting those 
technical specifications associated with spent nuclear fuel will not 
result in a significant increase in the probability or consequences 
of accidents previously analyzed.
    The proposed license amendment also relocates administrative 
requirements from Section 5.0 of the Technical Specifications to 
topical report PGE-8010, ``TNP Nuclear Quality Assurance Program.'' 
Relocation of administrative requirements follows the guidance 
provided in NRC Administrative Letter 95-06. Relocating these 
administrative requirements will not result in changes in method of 
operation of any plant equipment, therefore these changes will not 
result in a significant increase in the probability or consequences 
of accidents previously evaluated.
    The proposed license amendment will delete the on duty shift 
manning requirements (Technical Specification 5.2.2a). With removal 
of the spent nuclear fuel from the 10 CFR 50 licensed area, there 
are no remaining important to safety systems required to be 
monitored. With removal of the spent nuclear fuel from the 10 CFR 50 
licensed area, there are no remaining credible accidents which 
require the actions of a Shift Manager or non-certified operator to 
prevent occurrence or mitigate consequences. Therefore, deleting the 
shift manning requirements will not result in an increase in the 
probability or consequences of an accident previously analyzed.
    Deleting the Independent Review and Audit Committee (IRAC) is 
also proposed in this license amendment request. The responsibility 
of IRAC is to review and advise the Plant General Manager on matters 
relating to the safe storage of irradiated fuel. Since approval of 
this license amendment request is contingent upon removal of the 
spent nuclear fuel from the 10 CFR 50 licensed area and a revised 
Technical Specification Section 4.2 prevents future storage, 
deleting IRAC will not result in an increase in the probability or 
consequences of an accident previously evaluated.
    This license amendment request proposes to revise and relocate 
License Condition 2.C.(8), Fire Protection, to the TNP Quality 
Assurance Program (PGE-8010). The revised text removes requirements 
associated with making changes that could adversely impact the safe 
storage of irradiated fuel. Following removal of the spent nuclear 
fuel from the 10 CFR 50 licensed area and implementation of the 
proposed revision to Technical Specification Section 4.2, irradiated 
fuel will not be stored within the 10 CFR 50 licensed area so this 
change will not result in an increase in the probability of 
occurrence or consequences of accidents previously analyzed. 
Relocation of the remaining requirements contained in this license 
condition to the TNP Quality Assurance Program (PGE-8010) will 
provide the necessary administrative control to ensure that changes 
to the fire protection program will not increase the likelihood of 
an offsite release of radioactive material due to a fire. Therefore, 
this change will not result in an increase in the probability of 
occurrence or consequence of accidents previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed license amendment reflects the reduced operational 
risks within the 10 CFR 50 licensed area after the spent nuclear 
fuel has been transferred to the ISFSI. In addition, administrative 
controls contained in Section 5.0 of the Technical Specifications 
will [be] relocated to PGE-8010, TNP Nuclear Quality Assurance 
Program. These changes have no impact on plant equipment and only an 
administrative impact on some of the procedures used for operating 
plant equipment, which may still be needed within the 10 CFR 50 
licensed area following the

[[Page 38036]]

transfer of the spent nuclear fuel to the 10 CFR 72 ISFSI license 
area. This proposed amendment does not result in the addition of new 
equipment or result in the alteration of the operation of existing 
structures, systems, or components. Therefore, the proposed changes 
do not create 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 proposed 10 CFR 50 license amendment eliminates those 
technical specifications and license conditions associated with the 
storage of spent nuclear fuel. Following transfer of the spent 
nuclear fuel to the 10 CFR 72 ISFSI, the potential for fuel related 
accidents will be eliminated from the 10 CFR 50 licensed area. 
Therefore, removal of those technical specifications and license 
conditions associated with the safe storage of spent nuclear fuel 
will not involve a significant reduction in a margin of safety.
    Relocating administrative programs in Technical Specification, 
Section 5.0, ``Administrative Controls,'' follows the guidance of 
NRC Administrative Letter 95-06. With the exception of deleting 
those administrative controls associated with storage of spent 
nuclear fuel, the administrative programs will be relocated to the 
TNP Quality Assurance Program (PGE-8010). This administrative 
relocation of requirements does not involve a significant reduction 
in a margin of safety.
    This proposed amendment also requests deleting several license 
conditions and relocating License Condition 2.C.(8), ``Fire 
Protection.'' The deleted license conditions were related to either 
power operations or activities which have been completed and are no 
longer required. Relocating License Condition 2.C.(8), ``Fire 
Protection,'' to the TNP Quality Assurance Program (PGE-8010) will 
continue to maintain the required level of administrative control 
for the fire protection program since changes to PGE-8010 are 
controlled in accordance with the requirements of 10 CFR 
50.54(a)(3). Deleting these license conditions will, therefore, not 
result in 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensee: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Michael T. Masnik.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: June 7, 1999, as supplemented by letter 
dated June 24, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 2.0, Safety Limits and Limiting 
Safety System Settings, TS 3.2.5, DNB [Departure from Nucleate Boiling] 
Parameters, and the associated Bases, and Administrative Controls 
Section 6.9.1.6, Core Operating Limits Report (COLR), by relocating 
cycle-specific reactor coolant system-related parameter limits from the 
TSs to the COLR. This would allow for flexibility to enhance plant 
operating margin and/or core design margins without the need for cycle-
specific license amendment requests.
    Basis for proposed no significant hazards consideration 
determination: 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:

    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment is a programmatic and administrative 
change that does not physically alter safety-related systems, nor 
does it affect the way in which safety-related systems perform their 
functions. Because the design of the facility and system operating 
parameters are not being changed, the proposed amendment does not 
involve an increase in the probability or consequences of any 
accident previously evaluated.
    The cycle-specific limits in the Core Operating Limits Report 
will continue to be controlled by the STP [South Texas Project] 
programs and procedures. Each accident analysis addressed in the 
UFSAR [Updated Final Safety Analysis Report] will be examined with 
respect to changes in the cycle-dependent parameters, which are 
obtained from the use of NRC-approved reload design methodologies, 
to ensure that the transient evaluation of new reloads are bounded 
by previously accepted analyses. This examination, which will be 
conducted per the requirements of 10CFR50.59, will ensure that 
future reloads will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The safety limits imposed in Technical Specification 2.1.1.1 and 
2.1.1.2 are consistent with the values stated in the STP Updated 
Final Safety Analysis Report. The Reactor Coolant System Flow value 
in the Technical Specifications will be changed from the Minimum 
Measured Flow to the Thermal Design System Flow (approved by the 
Nuclear Regulatory Commission in Amendments 97 and 84 on September 
29, 1998) consistent with WCAP-14483-P-A [`Generic Methodology for 
Expanding Core Operating Limits Reports']. This change does not 
involve an increase in the probability or consequences of any 
accident previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Removal of cycle specific variables has no influence or impact 
on, nor does it contribute in any way to the probability or 
consequences of an accident. No safety-related equipment, safety 
function, or plant operation will be altered as a result of this 
proposed change. The cycle specific variables are calculated using 
the NRC-approved methods, and submitted to the NRC to allow the 
staff to continue to trend the values of these limits. The Technical 
Specifications will continue to require operation within the core 
operating limits, and appropriate actions will be required if these 
limits are exceeded. The safety limits imposed in Technical 
Specification 2.1.1.1 and 2.1.1.2 are consistent with the values 
stated in the STP Updated Final Safety Analysis Report. The Reactor 
Coolant System Flow value in the Technical Specifications will be 
changed from the Minimum Measured Flow to the Thermal Design Flow 
(approved by the Nuclear Regulatory Commission in Amendments 97 and 
84 on September 29, 1998) consistent with WCAP-14483-P-A. This 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not involve a significant reduction in 
a margin of safety.
    The margin of safety is not affected by the removal of cycle 
specific core operating limits from the Technical Specifications. 
The margin of safety presently provided by current Technical 
Specifications remains unchanged. Appropriate measures exist to 
control the values of these cycle specific limits. The proposed 
amendment continues to require operation within the core limits as 
obtained from NRC-approved reload design methodologies, and the 
actions to be taken if a limit is exceeded remain unchanged.
    The development of the limits for future reloads will continue 
to conform to those methods described in NRC-approved documentation. 
In addition, each future reload will involve a 10CFR50.59 safety 
review to assure that operation of the unit within cycle-specific 
limits will not involve a significant reduction in the margin of 
safety.
    The safety limits imposed in Technical Specification 2.1.1.1 and 
2.1.1.2 are consistent with the values stated in the STP Updated 
Final Safety Analysis Report. The Reactor Coolant System Flow value 
in the Technical Specifications will be changed from the Minimum 
Measured Flow to the Thermal Design System Flow (approved by the 
Nuclear Regulatory Commission in Amendments 97 and 84 on September 
29, 1998) consistent with WCAP-14483-P-A. This proposed change does 
not involve a significant reduction in the margin of safety.

[[Page 38037]]

    The proposed amendment is a programmatic and administrative 
change that provides assurance that plant operations continue to be 
conducted in a safe manner. As stated previously, the proposed 
amendment does not physically alter safety-related systems, nor does 
it affect the way in which safety-related systems perform their 
functions. Because the design of the facility and system operating 
parameters are not being changed, the proposed amendment does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 
77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of amendment request: June 3, 1999.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to reduce the Allowable Value (Av) 
used for Reactor Vessel Water Level--Low, Level 3 for several 
instrument functions.
    Basis for proposed no significant hazards consideration 
determination: 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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The Reactor Vessel Water Level--Low, Level 3 functions are in 
response to water level transients and are not involved in the 
initiation of accidents or transients. Therefore, reducing the Level 
3 Av does not increase the probability of an accident previously 
evaluated. Additionally, the results of the safety evaluation 
associated with the lowering of the Level 3 Av concludes that the 
previously evaluated transient and accident consequences are not 
significantly affected by the change. Therefore, the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to lower the BFN Units 2 and 3 Reactor 
Vessel Water Level--Low, Level 3 Av does not involve a hardware 
change and the purpose of the Level 3 function is not affected. The 
Level 3 functions will continue to fulfill their design objective. 
Therefore, reduction of the Av does not result in the possibility of 
a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The results of the safety evaluation associated with the 
reducing the BFN Units 2 and 3 Reactor Vessel Water Level--Low, 
Level 3 Av concluded that transient and accident consequences remain 
within the required acceptance criteria. Therefore, the margin of 
safety is not reduced for any event evaluated.

    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.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

    Date of amendment request: June 23, 1999.
    Brief description of amendments: The proposed license amendments 
would change the way in which the Emergency Diesel Generator (EDG) 
automatic trips are tested in Surveillance Requirement (SR) 3.8.1.13. A 
note would also be added to specify the CPSES, Unit 2, test schedule in 
SR 3.8.1.13.
    Basis for proposed no significant hazards consideration 
determination: 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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The emergency diesel generators are used to support mitigation 
of the consequences of an accident and are not considered to be 
initiator of any previously analyzed accident. Revising the 
surveillance to verify the bypass of non-critical EDG trips on both 
LOOP [loss of offsite power] and SI [safety injection] separately 
enhances the ability of the EDG to perform its safety function by 
ensuring continued operation during DBAs [design-basis accidents].
    Therefore, this change will not result in an increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to the surveillance requirement involves an 
EDG start circuit modification. The circuit modification has been 
previously installed on Unit 2 during 2RF04 [CPSES Unit 2, fourth 
refueling outage] for reasons other than the issue associated with 
the FWLB [feedwater line break]. As a part of the Unit 2 
installation a 50.59 evaluation was performed and it was determined 
that the modification did not represent an unreviewed safety 
question. The modification similar to Unit 2 will be implemented on 
Unit 1 and therefore, as concluded in the safety evaluation for the 
original modification, no new failure mechanisms will be introduced 
by the proposed change. The EDGs are designed to provide electrical 
power to equipment important to safety in the event of a loss of 
offsite power. The proposed change to the SR enhances the confidence 
that the EDGs will start and fulfill their safety related function.
    Therefore, this change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed change will not alter any accident analysis 
assumptions, initial conditions, or results. Revising the 
surveillance requirement to verify the EDG trip bypass for the LOOP 
and SI separately will enhance the confidence that the EDG starts as 
assumed in the safety analyses and does not create any new failure 
scenarios and no margin is reduced.
    Therefore, this change does not involve a significant reduction 
in a 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.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gram.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: April 16, 1999, as superseded on June 9, 
1999.

[[Page 38038]]

    Description of amendment request: The licensee proposed clarifying 
the inservice inspection requirements for Vermont Yankee Nuclear Power 
Station regarding the granting of relief from ASME Code requirements by 
the NRC. The licensee also proposed changes to reflect the previous NRC 
approval of the use of ASME Code Case N-560 at Vermont Yankee Nuclear 
Power Station.
    Basis for proposed no significant hazards consideration 
determination: 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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    This change is only an administrative change that: (1) clarifies 
the NRC's authority to grant relief to a specific requirement, and 
(2) conforms the TS language regarding GL 88-01 to agree with the 
NRC's acceptance of ASME Code Case N-560 for use at VY. This 
conclusion is justified in that:
    (a) The pursuit of relief from the ASME code and the imposition 
of alternative requirements are governed by 10CFR50.55a and require 
NRC approval. There are several sections in the regulations under 
which such relief can be granted. The removal of reference to a 
specific section of CFR that may be used to grant relief has no 
effect on plant equipment or its operation.
    (b) Adding words to clarify the relationship between GL 88-01 
and Code Case N-560 eliminates a contradiction in sample selection 
criteria and does not affect any equipment or its operation.
    These changes can be considered administrative in nature and do 
not change any of the accident analyses for the facility. Thus, 
there are no changes to the probability or consequences of accidents 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The revision of the wording in the TS to generalize the granting 
of relief to the ASME code does not result in any changes to the 
plant equipment or its operation. Similarly, adding words to allow 
use of the NRC-approved alternative to the sample selection guidance 
provided in GL 88-01 does not impact plant equipment or its 
operation. These changes are administrative in nature and do not 
result in the creation of any new or different kinds of accidents.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    This change primarily revises the wording in the TS to clarify 
the NRC's authority to grant relief to ASME Section XI requirements. 
The change maintains the requirement for NRC approval to be obtained 
for such relief. Secondly, this change conforms the TS language 
regarding GL 98-01 to agree with a previous relevant NRC disposition 
[Reference (e)]. [The staff notes that reference (e) is an NRC 
letter dated November 9, 1999, which approved the use of Code Case 
N-560 at Vermont Yankee Nuclear Power Station.] These administrative 
changes do not result in a reduction in any 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.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: June 24, 1999
    Description of amendment request: The amendment clarifies the basis 
for the reactor protection system bypass of the turbine stop valve 
(TSV) closure and turbine control valve (TCV) fast closure scram 
signals at low power. The amendment clarifies that the analytical basis 
for this bypass corresponds to a fraction of reactor rated thermal 
power and not other measures of power, for instance, turbine power.
    Basis for proposed no significant hazards consideration 
determination: 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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change clarifies the basis for the reactor 
protection system bypass of the turbine stop valve closure and 
turbine control valve fast closure scram signals. Consideration of 
the bypass function itself only applies to certain pressurization 
transients and not accident analyses.
    The change properly states the basis for the scram bypass and 
relates it to reactor thermal power and precludes potential 
misinterpretation of the basis for the bypass setpoint. Turbine 
power lags reactor power over the range of concern. Therefore, 
changing terminology related to ``power'' to mean ``reactor power'' 
instead of ``turbine power'' is conservative. Accordingly, this 
change can not be less restrictive.
    The low power (TSV closure and TCV fast closure) scram signal 
bypass does not initiate or mitigate any accident considered in the 
Updated Final Safety Analysis Report. This function is enabled at 
higher power to mitigate the effects of the pressurization transient 
which results from TSV closure or TCV fast closure. This change will 
not alter assumptions relative to the initiation or mitigation of 
any accident event.
    This change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
since there is no physical alteration of the plant configuration or 
relaxation of setpoints or operating parameters.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The reactor protection system bypass of the turbine stop valve 
closure and turbine control valve fast closure scram signals is not 
considered an initiator of any accident. This change to clarify the 
basis for applicability of the bypass does not create any new or 
different kind of accident since it does not involve any change in 
the physical configuration of the plant, nor relaxation of setpoints 
or operating parameters.
    VY has determined that the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated because the change merely adds a more 
restrictive interpretation to current terminology.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The change involves reducing the potential for misinterpreting 
the basis for the reactor protection system bypass of the turbine 
stop valve closure and turbine control valve fast closure scram 
signals and consequent potential for nonconservative operation of 
the plant. As a result, the potential for operation of the plant in 
an unsafe condition is reduced, thereby maintaining the margin of 
safety.
    VY has determined that the proposed change does not involve a 
significant reduction in a margin of safety since operation of the 
plant consistent with analytical bases of operation is further 
assured.

    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.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

[[Page 38039]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

Notice of Issuance of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are 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 for 
the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: March 30, 1999.
    Brief description of amendments: The amendments revised license 
conditions in each of the operating licenses to delete those license 
conditions that no longer apply, make an editorial change in the Unit 1 
license, and provide clarifying information regarding the license 
condition in each license concerning equalizer valve restrictions.
    Date of issuance: June 25, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 188 & 185.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the licenses.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24195).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 25, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: March 30, 1999.
    Brief description of amendment: The amendment adds Section 4.0.2 to 
allow a 24-hour grace period for performing inadvertently missed 
surveillance.
    Date of issuance: June 25, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27317).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 25, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: March 30, 1999.
    Brief description of amendment: The amendment adds Section 4.0.2 to 
allow a 24-hour grace period for performing inadvertently missed 
surveillance.
    Date of issuance: June 25, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27317).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 25, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: August 6, 1998, as supplemented 
by letter dated May 18, 1999.
    Brief description of amendment: The amendment approves a change to 
Technical Specification (TS) 3.1.3.2, ``Position Indicator Channels--
Operating,'' which adopts requirements that are consistent with NUREG-
1432, ``Standard Technical Specifications for Combustion Engineering 
Plants.'' In addition, the amendment approves the relocation of TS 
Table 3.8-1, ``Containment Penetration Conductor Overcurrent Protective 
Devices,'' to licensee control procedures in accordance with the 
guidance provided in Generic Letter 91-08, ``Removal of Component Lists 
From Technical Specifications.''
    Date of issuance: June 29, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 208.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56245).
    The May 18, 1999, letter provided clarifying information that did 
not change the scope of the original application and the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 17, 1996, as supplemented by 
letters dated October 22, 1998, and January 12 and February 5, 1999.

[[Page 38040]]

    Brief description of amendment: The amendment extends the 
surveillance test interval for the reactor trip circuit breakers from 
monthly to quarterly and revises the appropriate Bases page.
    Date of issuance: June 29, 1999.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 153.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48261).
    The October 22, 1998, and January 12 and February 5, 1999, letters 
provided additional information that did not extend the scope of the 
original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: October 27, 1998, supplemented 
March 19, 1999.
    Brief description of amendment: This amendment relocates a TS 
surveillance requirement from TS Section /4.6.5.1, ``Shield Building--
Emergency Ventilation System'' to TS Section 3/4.6.5.2, ``Shield 
Building Integrity.'' Administrative and bases changes have also been 
made.
    Date of issuance: June 22, 1999.
    Effective date: June 22, 1999.
    Amendment No.: 233.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64125). The March 19, 1999, supplement to the application did not 
expand the scope of the original application as noticed, and did not 
change the staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida

    Date of application for amendment: August 31, 1998, as revised on 
March 18, 1999.
    Brief description of amendment: The amendment approves changes to 
the Improved Technical Specifications to allow a repair roll process 
which would be used to repair steam generator tubes with defects within 
the upper tubesheet. Changes to inservice inspection and reporting 
requirements and several format and editorial changes were also 
included.
    Date of issuance: June 28, 1999.
    Effective date: As of date of issuance, to be implemented prior to 
commencing Cycle 12 operation.
    Amendment No.: 179.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56249). The revised submittal dated March 18, 1999, expanded the scope 
of the amendment request as originally noticed, and the application was 
renoticed on April 21, 1999 (64 FR 19557).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 28, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: February 7, 1997, as 
supplemented October 24, 1998
    Brief description of amendment: The amendment incorporates changes 
to more accurately reflect current plant design, adopts changes in 
surveillance requirements consistent with the Standard Technical 
Specifications, identifies changes to plant systems and revisions to 
Technical Specifications system descriptions not involving Limiting 
Conditions for Operations, and makes editorial or typographical 
corrections.
    Date of issuance: June 21, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 212.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1998 (63 FR 
14486) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: June 11, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification 6.12.1 to allow use of an alternative high radiation area 
control consistent with Regulatory Guide 8.38.
    Date of issuance: July 1, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 213.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43204) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336, 
and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New 
London County, Connecticut

    Date of application for amendment: December 22, 1998, as 
supplemented March 19, 1999.
    Brief description of amendment: The amendment replaces specific 
titles in Section 6.0 of the Technical Specifications of all three 
Millstone units with generic titles.
    Date of issuance: June 3, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 105, 235, and 171.

[[Page 38041]]

    Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: 
Amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4158). The March 19, 1999, letter provided clarifying information that 
did not change the scope of the December 22, 1998, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: January 4, 1999, as supplemented 
April 7, 1999.
    Brief description of amendment: The amendment changes Technical 
Specifications 3.5.2, ``Emergency Core Cooling Systems--ECCS 
Subsystems--Tavg greater than or less than 300  deg.F;'' 3.6.2.1, 
``Containment Systems--Depressurization and Cooling Systems--
Containment Spray and Cooling Systems;'' 3.7.1.2, ``Plant Systems--
Auxiliary Feedwater Pumps;'' 3.7.3.1, ``Plant Systems--Reactor Building 
Closed Cooling Water System;'' and 3.7.4.1, ``Plant Systems--Service 
Water System.'' The changes were made to the system pump flow 
requirements to incorporate the results of revised hydraulic and 
accident analyses.
    Date of issuance: June 29, 1999.
    Effective date: As of the date of issuance and shall implemented 
within 60 days from the date of issuance.
    Amendment No.: 236.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1999 (64 FR 
2523).
    The April 7, 1999, supplemental letter did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 9, 1999, as supplemented May 3, 
1999
    Brief description of amendments: The amendments delete the 
requirement to have an independent safety engineering group (ISEG) from 
the Technical Specifications and applies the substantive requirements 
now applicable to the ISEG to other organizations and relocates those 
requirements from the Technical Specifications to Chapter 16 of the 
Operational Quality Assurance Plan (OQAP). In the letter of May 3, 
1999, the licensee submitted the changes to Chapter 16 of the OQAP to 
incorporate the substantive Technical Specification requirements 
currently applicable to the ISEG into the OQAP in the form of an 
independent technical review program, and stated that these changes to 
the OQAP will become effective upon approval of the amendments.
    Date of issuance: June 23, 1999. Effective date: June 23, 1999, to 
be implemented within 30 days. Implementation includes incorporating 
the OQAP pages into the OQAP.
    Amendment Nos.: Unit 1-112 ; Unit 2-99.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17030) The May 3, 1999, supplement provided additional clarifying 
information within the scope of the original notice and did not change 
the staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: May 27, 1999, as supplemented May 28, 
1999.
    Brief description of amendments: The amendments add a footnote to 
Technical Specfiication 4.8.2.1e, ``D.C. Sources--Operating,'' which 
would, on a one-time basis for Unit 1 Battery BT1ED2, allow TU Electric 
to substitute a performance discharge test ``...in lieu of the battery 
service test required by Specification 4.8.2.1d, twice within a 60 
month interval.''
    Date of issuance: June 28, 1999.
    Effective date: As of the date of issuance.
    Amendment Nos.: 65 and 65
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (64 FR 31881 dated June 14, 1999). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by July 14, 1999, but 
indicated that if the Commission makes a final determination, any such 
hearing would take place after issuance of the amendments.
    The May 28, 1999, letter provided clarifying information that did 
not change the scope of the original application and the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final NSHC determination are contained in 
Safety Evaluation dated June 28, 1999.
    Attorney for Licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC, 20036.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: February 16, 1999.
    Brief Description of amendments: These amendments revise TS Section 
4.2 for Units 1 and 2. The changes relax the surveillance requirements 
for reactor coolant pump (RCP) flywheels. The flywheels provide 
extended reactor coolant flow coastdown capability if electric power 
for the RCPs is lost. Previously, the flywheel inspections included an 
ultrasonic examination (UT) of areas of high stress

[[Page 38042]]

concentration at the base and keyway every 3 years, and complete UT 
every 10 years. The changes require only a 10-year UT based upon an 
analysis presented in a Westinghouse topical report which has been 
reviewed and accepted by the NRC staff.
    Date of issuance: July 1, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 221 and 221.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24204). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
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 for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By August 13, 1999, 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 for the 
particular facility involved. 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 a 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

[[Page 38043]]

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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the 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).

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Unit 2, Matagorda County, Texas

    Date of amendment request: July 1, 1999.
    Brief description of amendment: The amendment provides for a one-
time change to Technical Specifications 3.3.2 and 3.7.8 for Unit 2 to 
allow all fuel handling building exhaust air system components to be 
inoperable for a period not to exceed 8 hours to facilitate repair of 
the Train B exhaust booster fan.
    Date of issuance: July 2, 1999.
    Effective date: From the date of amendment issuance until July 14, 
1999.
    Amendment No.: Unit 2-100.
    Facility Operating License No. NPF-80: The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated July 2, 1999.
    Local Public Document Room location: Wharton County Junior College, 
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

    Dated at Rockville, Maryland, this 7th day of July 1999.

    For The Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-17750 Filed 7-13-99; 8:45 am]
BILLING CODE 7590-01-P