[Federal Register Volume 60, Number 60 (Wednesday, March 29, 1995)]
[Notices]
[Pages 16181-16196]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-20329]



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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
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 March 3, 1995, through March 17, 1995. The 
last biweekly notice was published on March 15, 1995.

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 
[[Page 16182]] 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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By April 28, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room 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 petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
[[Page 16183]] telephone number, date petition was mailed, plant name, 
and publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to 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, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: February 24, 1995
    Description of amendment request: The proposed change would remove 
Section 4.3 from the Technical Specifications (TS) because the primary 
system testing following opening is already performed in accordance 
with the American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code, as implemented in the licensee's inservice inspection 
program as required by TS 4.0.1.
    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:
    This change does not involve a significant hazards consideration 
for the following reasons.
    1. The requested change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. This requested change will provide consistency between 
our Technical Specifications (TS) and 10 CFR 50.55a which requires 
testing in accordance with Section XI of the ASME Boiler and 
Pressure Vessel Code. The requirements contained in TS Section 4.3 
were placed into TS prior to incorporation of Section XI into the 
ASME Boiler and Pressure Vessel Code. The NRC and industry have 
since recognized the ASME Boiler and Pressure Vessel Code, Section 
XI as the appropriate testing program. Adequate assurance of primary 
system integrity will be provided since primary system testing will 
continue to be controlled and performed in accordance with the rules 
for inservice inspections provided by ASME Boiler and Pressure 
Vessel Code, Section XI as implemented by our approved In-Service 
Inspection (ISI) Program, as required by TS Section 4.0.1. 
Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.
    2. The requested change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The requested change deletes the current TS requirements 
for primary system testing by recognizing that we will continue to 
perform required testing consistent with 10 CFR 50.55a and ASME 
Boiler and Pressure Vessel Code, Section XI, as implemented by our 
approved ISI Program, as required by TS Section 4.0.1. This 
requested change does not involve the addition or modification of 
plant equipment, nor does it alter the design or operation of plant 
systems. Therefore, the requested change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The requested change does not involve a significant reduction 
in a margin of safety. The requested change deletes the current TS 
Section 4.3 requirements for primary system testing and maintains 
the margin of safety by continuing to perform required testing in 
accordance with 10 CFR 50.55a and ASME Boiler and Pressure Vessel 
Code, Section XI, as implemented by our approved ISI Program, as 
required by TS Section 4.0.1. Therefore, the requested 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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: March 3, 1995
    Description of amendment request: The proposed amendment would 
eliminate the requirement to perform periodic measurement testing of 
the response times for selected pressure and differential pressure 
sensors. The requirement that reactor trip and engineered safety 
feature response time functions be within their specified limit at 
least once per 18 months will be verified instead of demonstrated. The 
associated bases section for response time requirements will be changed 
to allow the sensor response time portion of the channel response time 
to use historical records, testing results, or vendor supplied 
engineering specifications. No other changes to response time methods 
are included in this change.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment does not result in a condition where the 
design, material, or construction standards that were
    applicable prior to the change are altered nor does it modify 
any system interface. The same Reactor Trip System and Engineered 
Safety Features Actuation System instrumentation is being used; the 
time response allocations/modeling assumptions in the Final Safety 
Analysis Report (FSAR) Chapter 15 analyses are still the same; only 
the method of verifying time response is changed. The proposed 
activity will not change, degrade, or prevent actions or alter any 
assumptions previously made in evaluating the radiological 
consequences of an accident described in the FSAR. Therefore, there 
would be no 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 amendment does not alter the performance of the 
pressure and the differential pressure transmitters used in the 
plant protection systems. The sensors will still have response time 
verified by test before placing the sensor in operational service 
and after any maintenance that could affect response time. Changing 
the method of periodically verifying instrument response for certain 
sensors (assuring equipemt operable) from time response testing to 
calibration and channel checks will not create any new accident 
initiators or scenarios. Periodic surveillance of these instruments 
will detect significant degradation in the sensor response 
characteristic. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. [[Page 16184]] 
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed amendment to [sic] does not affect the total system 
response time assumed in the safety analysis. The periodic system 
response time verification method for selected pressure and 
differential pressure sensors is modified to allow use of actual 
test data or engineering data. The method of verification still 
provides assurance that the total system response is within that 
defined in the safety analysis, since calibration tests will detect 
any degradation which might significantly affect sensor response 
time. Therefore, 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 
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: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of amendment request: May 20, 1994, as supplemented February 
2, 1995
    Description of amendment request: The proposed amendment would 
permit the licensee to use an alternate repair criteria (ARC), 
designated as the F* criteria. Use of the F* criteria would 
allow tubes with otherwise pluggable indications, to remain in service 
as long as the indications are below the designated minimum distance of 
the F* criteria. The F* criteria for Byron and Braidwood 
defines a length of 1.7 inches of undegraded expanded tube within the 
tubesheet as the minimum distance acceptable for implementing the ARC. 
Below the F* length, a circumferential tube defect can exist and 
the tube can remain in service. The proposed amendment will change the 
plugging limit definition and would exclude plugging steam generator 
tubes with indications that satisfy the F* criteria. The F* 
criteria maintains the structural integrity of the degraded tube as the 
primary pressure boundary and allows the tube to remain in service for 
heat transfer and core cooling.
    This alternate repair criteria qualification is documented in 
Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P 
Revision 1, ``W-D4 F* Qualification Report,'' which is included as 
part of the licensee's submittal. The staff's proposed no significant 
hazards consideration determination for the requested change was 
published on July 6, 1994 (59 FR 34659). In response to the staff's 
request for additional information by letter dated February 2, 1995, 
the licensee revised their previous submittal.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The supporting qualification report for subject criteria 
demonstrates that the presence of the tubesheet will enhance the 
tube integrity in the region of the tube-to-tubesheet roll 
expansions by precluding tube deformation beyond its initial 
expanded outside diameter. The resistance to a tube rupture is 
strengthened by the presence of the tubesheet in that region. The 
results of hardrolling of the tube into tubesheet provides a 
mechanical leak limiting seal between the tube and the tubesheet. A 
tube rupture cannot occur because the contact between the tube and 
the tubesheet does not permit sufficient movement of tube material.
    The type of degradation for which the F* criteria has been 
developed (cracking with a circumferential orientation) can 
theoretically lead to a postulated tube rupture event provided that 
the postulated through-wall circumferential crack exists near the 
top of the tubesheet. An evaluation including analysis and testing 
has been done to determine the resistive strength of the expanded 
tubes within the tubesheet. This evaluation provides the basis for 
the acceptance criteria for tube degradation subject to the F* 
criteria. The F* length of roll expansion is sufficient to 
preclude tube pullout from tube degradation located below the 
F* distance, regardless of the extent of the tube degradation. 
The Technical Specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. For consistency 
with current offsite dose limits, the site allowable leakage limit 
during a MSLB has been conservatively calculated to be 12.8 gpm for 
Byron and 9.1 gpm for Braidwood, which includes the accident leakage 
from IPC in addition to the accident leakage from F* on the 
faulted steam generator and the operational leakage limit. The 
operational leakage limit of Specification 3.4.6.2.c in each of the 
three remaining intact steam generators shall include the 
operational leakage from F*. As a requirement for operation 
following application IPC, the projected distribution of crack 
indications over the operating period must be verified to result in 
primary to secondary accident leakage less than the site allowable 
leakage limit. Thus, the consequences of a MSLB remain unchanged.
    The tube rupture and pullout is fully bounded by the existing 
steam generator tube rupture analysis included in the UFSAR. The 
leakage testing of the roll expanded tubes indicates that for tube 
expansion lengths approximately equal to the * distance, any 
postulated primary to secondary leakage from * tubes would be 
insignificant. The proposed alternate repair criteria does not 
adversely impact any other previously evaluated design basis 
accident.
    The leakage from an F* tube would be limited by the tube-
to-tubesheet interface since this leak would occur below the 
secondary face of the tubesheet. Qualification testing and previous 
experience indicate that normal and faulted leakage is well below 
Technical Specification and administrative limits creating no 
increase in the consequences associated with tube rupture type 
leakages. The UFSAR analyzed accident scenarios are still bounding 
since the normal and faulted leak rates are well within the normal 
operating limit of 150 gallons per day. This conclusion is 
consistent with previous F* programs approved and used at other 
operating plants.
    All of the design and operating characteristics of the steam 
generator and connected systems are preserved since the F* 
criteria utilizes the ``as rolled'' tube configuration that exists 
as part of the original steam generator design. The F* joint 
has been analyzed and tested for design, operating, and faulted 
condition loadings in accordance with Regulatory Guide 1.121 safety 
factors. The potential for a tube rupture is not increased from the 
original submittal as demonstrated in the qualification analyses and 
testing completed in the BWNT report.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    B. The proposed changes do not create the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    Implementation of the proposed F* criteria does not 
introduce any changes to the plant design basis. Use of the criteria 
does not provide a mechanism to initiate an accident outside of the 
region of the expanded portion of the tube. In the unlikely event 
the failed tube severed completely at a point below the F* 
region, the remaining F* joint would retain engagement in the 
tubesheet due to its length of expanded contact within the tubesheet 
bore. This engagement length would prevent any interaction of the 
severed tube with neighboring tubes. Any hypothetical accident as a 
result of any tube degradation in the expanded region of the tube 
would be bounded by the existing tube rupture accident analysis. 
Tube bundle structural integrity will be maintained. Tube bundle 
[[Page 16185]] leak tightness will be maintained such that any 
postulated accident leakage from F* tubes will be negligible 
with regard to offsite doses.
    Therefore, there is not a potential for creating the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The use of the F* criteria has been demonstrated to 
maintain the integrity of the tube bundle commensurate with the 
requirements of Regulatory Guide 1.121 and the primary to secondary 
pressure boundary under normal and postulated accident conditions. 
Acceptable tube degradation for the * criteria is any 
degradation indication in the tubesheet region, more than the 
F* distance from the secondary face of the tubesheet or the top 
of the last hardroll contact point whichever is further into the 
tubesheet. The safety factors used in the verification of the 
strength of the degraded tube are consistent with the safety factors 
in the ASME Boiler and Pressure Vessel Code and Regulatory Guide 
1.121 used in steam generator design. The * distance has been 
verified by various testing to be greater than the length of the 
roll expanded tube-to-tubesheet interface required to preclude both 
tube pullout and significant leakage during normal and postulated 
accident conditions. The protective boundaries of the steam 
generator continue to be maintained with the use of the F* 
criteria. A tube with the indication of degradation previously 
requiring removal from service can be kept in service through the 
F* criteria. Since the joint is contained within the tubesheet 
bore, there is no additional risk associated with the previously 
analyzed tube rupture event. The leak testing acceptance criteria 
are based on the primary to secondary leakage limit in the Technical 
Specifications and the leakage assumptions used in the UFSAR 
accident analyses.
    Implementation of the alternate repair criteria will decrease 
the number of tubes which must be taken out of service with tube 
plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS 
flow margin; thus, implementation of the F* criteria will 
maintain the margin of flow that would otherwise be reduced in the 
event of increased plugging or sleeving.
    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: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee 
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

    Date of amendment request: November 22, 1994, as supplemented 
January 30, March 2, and March 13, 1995.
    Description of amendment request: This request was previously 
published in the Federal Register on February 15, 1995 (60 FR 8746). It 
is being renoticed to provide clarification to the scope of the 
original request. The amendments would revise Technical Specification 
(TS) 3.8 to establish restricted loading patterns and associated burnup 
criteria for placing fuel in the Oconee spent fuel pools. In addition, 
the Design Features sections associated with the reactor and fuel 
storage would be revised. These changes are necessary to address two 
new fuel designs which have increased initial fuel enrichment and 
therefore cannot be stored in the spent fuel pools under existing TS or 
loaded into the reactor. An administrative change would be made to TS 
6.9.1 to include spent fuel pool boron concentration in the Core 
Operating Limits Report. Other administrative changes would be made in 
the Design Features section to make the specification consistent with 
wording in the standard TS. Finally, the two additional supplements to 
the original request are referenced herein.
    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:
    Standard 1. The proposed amendments will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Each accident analysis addressed in the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to changes in 
Cycle 15 parameters to determine the effect of the Cycle 16 reload 
and to ensure that the acceptance criteria of the FSAR safety 
analyses remain satisfied. The transient evaluation of Cycle 16 is 
considered to be bounded by previously accepted analyses. Section 7 
of the Reload Report addresses ``Accident and Transient Analysis'' 
for this core reload.
    There is no increase in the probability or consequences of an 
accident due to the spent fuel storage restrictions proposed in this 
amendment request. It has been shown that the calculated, worst case 
keff for this area is [less than or equal to] 0.95 under all 
conditions. There is no increase in the probability of a fuel drop 
accident in the SFP [spent fuel pool] since the mass of the new 
assemblies is not significantly different from the mass of the old 
assemblies. The likelihood of other accidents, previously evaluated 
and described in the FSAR, is also not affected by the proposed 
changes. In fact, it could be postulated that since the increase in 
fuel enrichment will allow for extended fuel cycle lengths, there 
will be a decrease in fuel movement and the probability of an 
accident may actually be reduced. There is also no increase in the 
consequences of a fuel rod drop accident in the SFP since the 
fission product inventory of individual fuel assemblies will not 
change significantly as a result of increasing the initial 
enrichment. In addition, no change to safety related systems is 
being made. Therefore, the consequences of a fuel rupture accident 
remain unchanged. In addition, it has been shown that Keff all 
conditions. Therefore, the consequences of a criticality accident in 
the SFP remain unchanged as well. The above analysis ensures that 
the proposed reload amendment request will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The analyses performed in support of this reload are in 
accordance with the NRC approved methods delineated in Specification 
6.9.2. The predicted operating characteristics of Oconee 3 Cycle 16 
are similar to previously licensed designs. The Mark B10T and Mark 
B11 fuel assembly designs remain mechanically compatible with all 
fuel handling equipment. Therefore, no new or different kind of fuel 
handling accident is created by the proposed amendment request.
    Section 15.11 of the Oconee FSAR states that the refueling boron 
concentration is maintained such that a criticality accident during 
refueling is not considered credible. The proposed amendment request 
continues to assure that a criticality accident in the SFP or during 
refueling is not credible. The double contingency principle 
discussed in ANSI N-16.1-1975 and the April 1978 NRC letter allows 
credit for soluble boron under other abnormal or accident 
conditions, since only a single accident need be considered at one 
time. Thus, by requiring a minimum boron concentration in the SFP, a 
criticality accident caused by violating the SFP storage 
restrictions is not considered credible. Therefore, the proposed 
amendment request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The Oconee 3 Cycle 16 design was performed using the NRC 
approved methods given in Specification 6.9.2. The safety limits for 
Oconee 3 Cycle 16 are unchanged from previous cycles. The limits and 
margins summarized in the Oconee 3 Cycle 16 Reload Report are well 
within the allowable limits and requirements, and reflect no 
reductions to any margins 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 
[[Page 16186]] satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: December 14, 1993, as supplemented by 
letter dated March 3, 1995.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by removing the reactor 
vessel material specimen withdrawal schedule and by updating the 
reactor coolant system pressure-temperature (P-T) curves. The specimen 
withdrawal schedule will be relocated to the Updated Final Safety 
Analysis Report (UFSAR). The original Notice was published on January 
19, 1994 (59 FR 2867).
    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:
    Although the Reactor Vessel material specimens withdrawal 
schedule will be removed from the Technical Specifications, the 
Technical Specifications bases will continue to provide background 
information on the use of the data obtained from material specimens. 
Also, updates to the schedule will continue to be submitted to the 
NRC for approval prior to implementation.
    Operating the plant in accordance with the new, updated P-T 
Curves will assure preserving the structural integrity of the 
reactor vessel over the life of the plant. The pressure and 
temperature limits were developed in accordance with 10 CFR [Part] 
50 Appendix G requirements.
    Removing the requirements associated with the previous exemption 
to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative 
change.
    Therefore, the proposed changes will not significantly increase 
the probability or consequences of any accident previously 
evaluated.
    Removal of the Reactor Vessel material specimen schedule from 
the Technical Specifications has no impact on accidents at the 
plant. Updates to the schedule will still be required to be 
submitted to the NRC prior to implementation per Section II.B.3 of 
Appendix H to 10 CFR Part 50.
    Also, updates to the P-T Curves will not create a new or 
different type [of] accident. The reactor vessel beltline P-T limits 
were revised applying the general guidance of the ASME Code, 
Appendix G procedures with the necessary margins of safety for 
heatup, cooldown and inservice hydro test conditions.
    The change to TS 4.4.8.1.2 items a & b is purely administrative.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Removal of the schedule for Reactor Vessel material specimen 
withdrawal from the Technical Specifications does not impact the 
margin of safety. The schedule will continue to receive NRC review 
and approval prior to implementation of updates to the schedule.
    Updates to the P-T Curves are provided to preserve the margin to 
[sic] safety to assure that when stressed under operating, 
maintenance and testing the boundary behaves in a non-brittle manner 
and the probability of rapidly propagating fracture is minimized.
    The change to TS 4.4.8.1.2 items a & b is purely administrative.
    Therefore, the proposed changes will not result in 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: February 22, 1995
    Description of amendment request: The proposed changes are 
administrative in nature in that reference to an ``automatic'' 
containment air lock tester will be deleted from TS 4.6.1.3. The 
automatic airlock tester is no longer being used.
    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 amendment is administrative in nature in that the 
revision will eliminate the wording associated with optional use of 
the personnel airlock automatic leakage tester. The requirement for 
testing the personnel airlock at a pressure greater than or equal to 
Pa for at least 15 minutes remains unchanged. The acceptance 
criteria of personnel airlock seal leakage less than 0.01 La is 
also unchanged. The automatic leakage tester is not an accident 
initiator nor a part of the success path(s) which function to 
mitigate accidents evaluated in the plant safety analyses. The 
proposal does not involve any changes to the configuration or method 
of operation of any plant equipment that is used to mitigate the 
consequences of an accident, nor does it alter any assumptions or 
conditions in the plant safety analyses. Therefore, 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.
    (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 accident previously evaluated.
    The proposed amendment to remove the reference to the personnel 
airlock automatic tester from the technical specifications will not 
introduce any new failure modes or system interactions, nor will it 
require the installation of any new or modified equipment. The 
requirement to leak test the personnel air locks will not be 
changed. Thus, 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 accident previously evaluated.
    (3)Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment is administrative in nature in that it 
eliminates the reference to the personnel airlock automatic leakage 
tester but does not alter the surveillance and acceptance criteria 
for such testing. Seal leakage testing is performed in accordance 
with an approved plant procedure which allows use of either an 
automatic tester or a portable testing cart. The automatic leakage 
tester is not used to actuate safety related equipment, provide 
interlocks, or perform plant control functions. The conditions 
evaluated in the plant accident and transient analyses do not 
involve this tester. The proposed change does not alter the basis 
for any technical specification that is related to the establishment 
of, or the maintenance of, a nuclear safety margin. Therefore, 
operation of the facility in accordance with the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based on the above discussion and the supporting Evaluation of 
Technical Specification changes, FPL has determined that the 
proposed license amendment involves no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this [[Page 16187]] 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: February 27, 1995
    Description of amendment request: The proposed amendment will 
change Table 3.3-3 and 3.3-4 to accommodate an improved coincidence 
logic and relay replacement for the 4.16 kV Loss of Voltage Relays. 
Actions required for certain trip units with the number of operable 
channels one less than the total number of channels will also be 
changed. In addition, the format used to state the time delay for the 
4.16 kV Degraded Voltage trip unit will be revised.
    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 change will result in a better overall posture of 
the plant under degraded/loss of voltage conditions. The design 
upgrade for the 4.16 kV Loss of Voltage system is more reliable, has 
inherently higher accuracy, and is easier to maintain and calibrate 
in the field. The coincidence logic will eliminate the spurious 
plant trip potential from the existing design. Restating the maximum 
time delay for the 4.16 kV Degraded Voltage (coincident with SIAS 
[safety injection actuation signal]) protective relays in a ``less 
than'' format will assure that the transfer of power to the on-site 
sources occurs before the level of voltage becomes injurious to the 
equipment under accident conditions, and will ensure that stripping 
of the emergency power busses and loading of the EDG (s) [emergency 
diesel generators] will occur within the time allowed by original 
design criteria. The maximum allowed time delay for this function is 
not being increased, and the time delay assumed in the accident 
analyses for connecting the emergency bus to the diesel generator 
will not be exceeded. Therefore, operation of the facility in 
accordance with the proposed amendment will not involve a 
significant increase in the probability 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 accident previously evaluated.
    The proposed amendment does not change the operation, function 
or modes of plant operation. The ability of the loss of power and 
degraded grid voltage protection scheme to properly transfer from 
the off-site to the on-site power sources is being maintained. The 
relays in the improved design of the 4.16 kV Loss of Voltage 
function are of the type presently being used in identical 
applications at both St. Lucie plant units. No new hazards are 
created or postulated which may cause an accident different from any 
accident previously analyzed. The modifications will result in a 
more sensitive protection scheme allowing continuous operation 
without unnecessary challenges to the safety systems, and will 
continue to provide adequate protection to all the safety equipment. 
Therefore, 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 accident previously evaluated.
    (3)Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The capability of the loss of power and degraded grid voltage 
protection scheme is enhanced by the changes being proposed and is 
confirmed by the existing surveillance requirements. The planned 
modifications to the 4.16 kV Loss of Voltage function will result in 
a more sensitive undervoltage detection system and reduce the 
possibility of spurious actuation. The maximum time assumed in the 
safety analyses for connecting each Emergency Bus to its dedicated 
Emergency Diesel Generator is not being changed, and assurance that 
separation from a degraded off-site power source will occur before 
this time interval is exceeded during accident conditions will be 
maintained by the proposed amendment. Accordingly, the margin of 
safety is not affected. Therefore, operation of the facility in 
accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS [Technical Specifications] Changes, FPL 
has concluded that this proposed license amendment involves 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 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: February 27, 1995
    Description of amendment request: The proposed amendment will 
modify surveillance requirement (SR) 4.9.8.1 and 4.9.8.2 to allow a 
reduction in the required minimum shutdown cooling flow rate under 
certain conditions during operational MODE 6. In addition, the format 
of the SR will be changed to clarify the intent of the stated 
surveillances.
    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.
    Operation of the SDCS [shutdown cooling system] is not an 
accident initiator and, therefore, does not significantly increase 
the probability of an accident previously evaluated.
    The proposed change will allow a plant configuration needed to 
perform maintenance activities on LPSI [low-pressure safety 
injection]/SDCS headers by isolating one injection flow line for an 
operable SDCS train during certain MODE 6 conditions. In the event 
of a failure or unavailability of the alternate SDCS train, this 
configuration could result in the proposed minimum flow rate.
    The proposed change only modifies the minimum required flow 
rate, and does not affect the probability of this event. FPL has 
evaluated the proposed value of reactor coolant flow and has shown 
that the bases for the existing LCO [limiting condition for 
operation] will continue to be satisfied. Therefore, there are no 
significant increases in the consequences of any event from the 
proposed change. No other system interactions are involved related 
to previously evaluated accidents, and the proposed change has no 
adverse effect on any other system performance.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant increase in the 
probability 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 accident previously evaluated.
    The proposed change does not affect the normal operation of the 
plant. No new [[Page 16188]] systems are introduced and there is no 
adverse effect on any other system configuration or performance. The 
change will, however, allow isolation of one SDCS injection flow 
path for maintenance activities in MODE 6 under controlled 
conditions. The failure of the alternate SDCS train does not create 
a new accident and has been further evaluated in the reduced flow 
configuration, and shown to meet all the TS bases requirements. 
Therefore, operation of the facility in accordance with the proposed 
amendment will not create 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 significant reduction in a margin of 
safety.
    The safety considerations related to the proposed change are 
described in the bases to TS [Technical Specification] 3/4.9.8. FPL 
has evaluated the proposed reduction in SDCS flow requirement, under 
stated conditions, and has shown that the proposed flow rate meets 
all the TS bases requirements involving decay heat removal, boron 
dilution, and stratification. Established acceptance criteria 
providing margins of safety are not being changed by the proposed 
amendment. Therefore, operation of the facility in accordance with 
the proposed amendment would not involve a significant reduction in 
a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS Changes, FPL has concluded that this 
proposed license amendment involves 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 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:Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, 
Georgia

    Date of amendment request: March 14, 1995
    Description of amendment request: Georgia Power Company (GPC or the 
licensee) has proposed a temporary change to Hatch Unit 2 Technical 
Specification (TS) Required Action 3.3.6.1.F.1, and associated Bases. 
The proposed change would add a note to the Primary Containment 
Isolation Instrumentation actions to permit the drywell and wetwell 
purge valves which are isolated by the drywell radiation monitor signal 
to be opened with one inoperable drywell radiation monitor. The note 
will expire prior to startup from the Hatch Unit 2 refueling/
maintenance outage scheduled in the fall of 1995, at which time the 
radiation monitor can be repaired or replaced. Should the unit be 
forced into a cold shutdown of sufficient duration (i.e., drywell de-
inerted), the inoperable radiation monitor will be repaired at that 
time. The TS containment sections allow these valves to be opened for 
inerting, de-inerting, and pressure control. However, with radiation 
monitor 2D11-K621B inoperable, the primary containment isolation 
instrumentation TS require the valves be closed until the unit achieves 
a cold shutdown condition. Without the ability to open these valves 
until cold shutdown, pressure control and de-inerting are difficult.
    The purpose of the high drywell radiation primary containment 
isolation signal is to limit fission product release following a 
postulated loss-of-coolant accident (LOCA) with significant fuel 
damage. It is one of several signals which isolate the primary 
containment vent and purge valves. A high drywell pressure signal will 
not only shut down the reactor and generate a LOCA signal, it will also 
isolate these valves.
    High drywell radiation indicates possible gross failure of the fuel 
cladding. The generation of this isolation signal is not credited in 
any accident or transient analysis. Chapter 15 of the Hatch Unit 2 
Final Safety Analysis Report (FSAR) discusses the radiological 
consequences of a postulated large break LOCA with fuel failure to show 
conformance to 10 CFR Part 100 and 10 CFR Part 50, Appendix A. This 
analysis is not affected.
    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 change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Opening the containment purge and vent valves with an inoperable 
drywell radiation monitor will not increase the probability of any 
previously evaluated accident. The fact that the monitor cannot send 
an automatic isolation signal will not significantly affect the 
consequences of an accident. The function of the primary containment 
isolation signal is to detect and limit release of fission products 
following significant fuel damage. The generation of this isolation 
signal is not credited in any accident or transient analysis. 
Chapter 15 of the Unit 2 FSAR evaluates the radiological 
consequences of a postulated design basis LOCA with non-mechanistic 
fuel damage. This licensing evaluation shows conformance to the 
radiological limits presented in 10 CFR 100 and 10 CFR 50, Appendix 
A. The results of this analysis are not affected since the valves 
are otherwise operable and receive isolation signals from other 
instrumentation.
    2. The 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 the installation of any new 
equipment, or the modification of any equipment designed to prevent 
or mitigate the consequences of accidents or transients. Therefore, 
the change has no effect on any accident initiator, and no new or 
different type of accidents are postulated to occur.
    3. The proposed amendment does not result in a significant 
reduction in the margin of safety.
    As discussed in Item 1 above, the assumptions and results of the 
licensing evaluations remain unchanged. Therefore, the margin of 
safety is not significantly affected.
    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:Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: February 28, 1995
    Description of amendment request: Technical Specification (TS) 
Section 6.5.1.12 would be revised to delete the requirement to render 
determinations in writing with regard to whether or not activities 
listed in TS Sections 6.5.1.2 and 6.5.1.5 constitute an unreviewed 
safety question.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the [[Page 16189]] issue of no significant 
hazards consideration, which is presented below:
    . Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed change removes the requirement to 
render determinations in writing with regard to whether or not 
opposed changes to the Technical Specifications and investigations 
of violations of Technical Specifications constitute an unreviewed 
safety question. This change is considered an administrative change 
to remove a requirement which is not relevant to these activities 
and which is also consistent with the BWR Revised Standard Technical 
Specifications (NUREG 1433). Existing requirements to perform 
Technical and Independent Safety Reviews of these activities are not 
affected. Therefore, the proposed amendment does not significantly 
increase the probability of occurrence or the 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 accident previously evaluated. The 
proposed change is considered administrative since it removes a 
requirement which is not relevant to the affected activities, and 
which is also consistent with the BWR Revised Standard Technical 
Specifications Administrative Controls for Review and Audit. 
Existing requirements to perform Technical and Independent Safety 
Reviews for the affected activities are not changed. Therefore, this 
change has no effect on the possibility of creating a new or 
different king of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed change removes a requirement which is not 
relevant to the affected activities. Existing Technical 
Specification requirements to perform Technical and Independent 
Safety Reviews for the affected activities are not changed and 
therefore, will continue to ensure that such activities properly 
address nuclear safety and safe plant operation. Therefore, it is 
concluded that operation of the facility in accordance with the 
proposed amendment does not involve a 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:Ocean County Library, Reference 
Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Phillip F. McKee

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: February 15, 1995
    Description of amendment request: The proposed amendment would 
modify (by relocation to the Technical Requirements Manual) Technical 
Specification (TS) 3/4.3.3.7, Chemical Detection Systems, and TS 3/
4.8.4.1, Electrical Equipment Protective Devices - Containment 
Penetration Conductor Overcurrent Protective Devices, and the 
associated Bases.
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to Technical Specification 3.3.3.7, Chemical 
Detection Systems and 3.8.4.1, Electrical Equipment Protective 
Devices-Containment Penetration Conductor Overcurrent Protective 
Devices, is of an administrative nature in that the listed Technical 
Specifications and Bases will be relocated in entirety to the 
Technical Requirements Manual (TRM). Any future changes to the 
relocated requirements will be in accordance with 10CFR 50.59 and 
approved station procedures. Whether the listed Technical 
Specifications and Bases are located in Technical Specifications or 
the Technical Requirements Manual has no effect on the probability 
or consequences of any accident previously evaluated.
    The proposed change does not alter the assumptions previously 
made in the listed Technical Specifications. The proposed change 
allows the Commission and South Texas more effective use of 
personnel resources to control requirements that meet the four 
Criteria in the Final Policy Statement. The proposed change will not 
change the dose to workers.
    Since the probability of a [sic] accident is unaffected by the 
administrative relocation of the listed Technical Specifications, 
and the doses are not affected and do not exceed acceptance limits, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to Technical Specification 3.3.3.7, Chemical 
Detection Systems and 3.8.4.1, Electrical Equipment Protective 
Devices-Containment Penetration Conductor Overcurrent Protective 
Devices, is of an administrative nature in that the listed Technical 
Specifications and Bases will be relocated in entirety to the 
Technical Requirements Manual (TRM). Any future changes to the 
relocated requirements will be in accordance with 10CFR 50.59 and 
approved station procedures. Whether the listed Technical 
Specifications and Bases are located in Technical Specifications or 
the Technical Requirements Manual has no effect on any previously 
evaluated accident. It does not represent a change in the 
configuration or operation of the plant and, therefore, does not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The proposed change to Technical Specification 3.3.3.7, Chemical 
Detection Systems and 3.8.4.1, Electrical Equipment Protective 
Devices-Containment Penetration Conductor Overcurrent Protective 
Devices, is of an administrative nature in that the listed Technical 
Specifications and Bases will be relocated in entirety to the 
Technical Requirements Manual (TRM). Any future changes to the 
relocated requirements will be in accordance with 10CFR 50.59 and 
approved station procedures. The margin of safety is not reduced 
when the requirements are relocated to a Licensee-controlled 
document because the requirements to change a License Basis Document 
via the 10CFR 50.59 process ensures the same questions concerning 
the margin to safety required for a License Amendment are asked. The 
major difference is the time and expense required for the License 
Amendments. Therefore, this proposed change does not significantly 
reduce 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., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: February 15, 1995 [[Page 16190]] 
    Description of amendment request: The proposed amendment would 
modify Technical Specification 4.6.2.3.a.2 (and associated Bases) to 
reflect the reactor containment fan cooler flow rate assumed in the 
accident analyses and to specify that this flow is provided by the 
component cooling water system.
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to Technical Specification 4.6.2.3.a.2 is to 
reflect the cooling water temperature assumed in the accident 
analyses. The revised Technical Specification surveillance 
requirement will change the cooling water flow rate requirement to 
each Reactor Containment Fan Cooler from greater than or equal to 
550 gallons per minute to greater than or equal to 1800 gallons per 
minute.
    The proposed change, which will result in an increased 
acceptance criteria for the flow to the Reactor Containment Fan 
Coolers, is not indicative of accident initiators. The change will 
ensure that the surveillance requirement reflects the flow rate 
value assumed in the South Texas Project accident analyses and that 
the design and operability requirements of equipment important to 
safety are ensured.
    The accident mitigation features of the plant are not affected 
by the proposed change since the change reflects the original 
assumptions made in the design of the accident mitigation features 
of the South Texas Project. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

    2. Does the proposed change create the possibility of 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 previously evaluated in the Safety 
Analysis Report because all the accidents were analyzed with a flow 
rate of 1800 gallons per minute to the Reactor Containment Fan 
Cooler.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    There will be no adverse affects on margins of safety since a 
more stringent surveillance requirement will be applied to the 
Reactor Containment Fan Cooler. The Technical Specification 
operability and surveillance requirements are not reduced but rather 
made more restrictive by this proposed change. The change ensures 
that the margin of safety originally intended for the Reactor 
Containment Fan Coolers is maintained.
    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., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: February 13, 1995
    Description of amendment request: The proposed amendment would 
delete the audit frequency requirements from the Duane Arnold Energy 
Center Technical Specifications (TS) and add them to the Quality 
Assurance Program Description located in the Updated Final Safety 
Analysis Report.
    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 does not involve a change in the 
probability or consequences of an accident previously evaluated. No 
physical changes will occur as a result of this amendment. The 
change is administrative in nature and does not impact the operation 
of the plant or the plant's response to any accident. Because it 
will allow management the flexibility to adjust the audit 
frequencies based upon the performance of the program or 
organization being audited, the overall performance of the 
organization will be improved.
    (2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No physical changes will occur as a result of this 
amendment. The change is administrative in nature and does not 
affect the operation or design of the plant; therefore, it does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The audits will continue to be 
performed to provide assurance of conformance to the applicable 
requirements.
    (3) The proposed amendment will not reduce the margin of safety. 
No physical changes will occur as a result of this amendment. The 
change is administrative in nature and does not affect the operation 
or design of the plant. Safety limits and limiting safety system 
settings are not affected by this proposed amendment. The amendment 
removes requirements for frequency of audits from the TS, thus 
permitting more effective scheduling of audits based on performance 
and the status of the activities audited. This should result in a 
more effective audit program that will contribute to an improvement 
in the overall performance of the organization.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis 
& Bockius, 1800 M Street, N. W., Washington, D. C. 20036-5869NRC Acting 
Project Director: John N. Hannon

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of amendment request: February 10, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.3.2.1, ``Control Rod Block 
Instrumentation,'' to revise two surveillance requirements and their 
associated notes for the Rod Withdrawal Limiter (RWL) mode of the Rod 
Pattern Control System. These changes will conform these requirements 
to their original bases and eliminate the potential for unnecessary 
power reductions.
    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 changes are consistent with the Rod Withdrawal 
Error (RWE) analysis presented in Clinton Power Station (CPS) 
Updated Safety Analysis Report (USAR) Section 15.4.2. The proposed 
changes do not result in any change to plant equipment or operation; 
only the plant conditions for which the Rod Withdrawal Limiter (RWL) 
function(s) are required to be tested are being revised. The 
proposed changes continue to ensure that the RWL is OPERABLE and 
tested to ensure that continuous control rod withdrawals remain 
within the assumptions of the RWE analyses. The proposed changes 
have no impact on the probability of occurrence of a RWE event. 
Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    (2) The proposed changes do not result in any changes to plant 
equipment or operation; only the plant conditions for which the RWL 
[[Page 16191]] function(s) are required to be OPERABLE and tested 
are being revised. The proposed changes continue to ensure that the 
RWL is OPERABLE and tested to ensure that continuous control rod 
withdrawals remain within the assumptions of the RWE analyses. As a 
result, no new failure modes are introduced. The proposed changes 
are clearly within the limits of plant operation as described in the 
USAR and the RWE analyses. Therefore, the proposed changes cannot 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) The proposed changes revise the testing requirements to be 
consistent with the testing required prior to Amendment No. 95. The 
proposed changes ensure that the RWL is OPERABLE and tested to 
ensure that continuous control rod withdrawals remain within the 
assumptions of the RWE analyses. The proposed changes are clearly 
within the limits of plant operation as described in the USAR and 
the RWE analyses. Therefore, the proposed changes do not involve a 
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:Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur, 
Illinois 62525.
    NRC Acting Project Director: John N. Hannon

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine 
YankeeAtomic Power Station, Lincoln County, Maine

    Date of amendment request: February 14, 1995
    Description of amendment request: The proposed amendment would 
change responsibility for audits of the emergency and security plans 
and their implementing procedures. Audit responsibility would change 
from the licensee's Nuclear Safety Audit and Review (NSAR) Committee 
and the Plant Operation Review Committee (PORC), to the respective 
emergency and security plans. The proposed amendment is consistent with 
the guidance of NRC Generic Letter 93-07, Modification of the Technical 
Specification Administrative Control Requirements for Emergency and 
Security Plans, dated December 28, 1993.
    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 NRC staff has reviewed the licensee's analysis 
against the Standards of 10 CFR 50.92(c). A summary of the licensee's 
analysis is presented below:
    1. The proposed amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    The proposed changes do not have a direct effect on the physical 
plant or the maintenance of the physical plant, but would improve the 
safe operation of the plant by reducing the administrative burden of 
PORC and NSAR. This change would allow a better focus of management 
resources to the operational safety oversight of plant activities. The 
requirement to review, audit, document, control, and submit for 
regulatory review, the Emergency Plan and the Security Plan and their 
implementing procedures, is defined by regulation and remains 
unchanged. The proposed changes will not, of themselves, result in any 
reduction in the effectiveness of either the Emergency Plan or the 
Security Plan to protect the health and safety of the public. The 
proposed changes, therefore, will not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any previously evaluated.
    This change is administrative in nature and does not change or 
modify the physical plant or maintenance of the physical plant. 
Applicable regulations continue to enforce the requirement for review 
and audit by individuals not responsible for implementation of the 
existing programs. Consequently, independent oversight of the programs 
and procedures is not compromised by these proposed changes. The 
possibility of a new or different accident from any previously 
evaluated as a result of future changes in the implementation of the 
Security or Emergency Plans is not created.
    3. The proposed amendment would not involve a significant 
reduction in a margin of safety.
    The proposed changes will revise the administrative 
responsibilities of the PORC and NSAR committees allowing a better 
focus of resources on operational safety reviews. The requirements 
of the applicable Federal and State regulations ensure the continued 
effective oversight of the implementation of Security and Emergency 
Plans. Consequently, the adoption of the proposed changes would not 
involve a significant reduction in a margin of safety.
    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: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, Maine 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, Maine 04011
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: October 18, 1994, as supplemented 
February 21, 1995.
    Description of amendment request: The following changes requested 
in the October 18, 1994, submittal were published in Federal Register 
on November 9, 1994 (59 FR 35876). The proposed amendment would require 
three Type A overall Integrated Containment Leakage Tests be conducted 
at approximately equal intervals during shutdowns during each 10 year 
service period. For the third Type A test for the second 10-year 
period, it would be conducted during the thirteenth refueling outage 
extending the second 10-year service period to the end of the 
thirteenth refueling outage. The amendment would also change the 
Containment Leakage Bases by reflecting the conditions of a proposed 
exemption to 10CFR50, Appendix J, that would remove the requirement 
that the third Type A test for each 10-year period be conducted when 
the plant is shutdown for the 10-year plant inservice inspection.
    By letter dated February 21, 1995, the licensee withdrew the action 
related to conducting the third Type A test for the second 10-year 
period during the thirteenth refueling outage and the reference to a 
proposed exemption to 10 CFR 50, Appendix J, that would remove the 
requrement that the third Type A test for each 10-year period be 
conducted when the plant is shutdown for the 10-year plant inservice 
inspection. The following basis for the proposed no significant hazards 
consideration determination relates to the February 21, 1995, 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 
[[Page 16192]] consideration, which is presented below:
    The proposed change does not involve a SHC because the change 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    Type A tests are performed to ensure that the total leakage from 
containment does not exceed the maximum allowable primary 
containment leakage rate at the design pressure. This assures 
compliance with the dose limits of 10CFR100.
    The proposed change to Surveillance Requirement 4.6.1.2.a of the 
Millstone Unit No. 2 Technical Specifications will increase the 
flexibility for scheduling the Type A tests. It does not modify the 
maximum allowable leakage rate at the design containment pressure, 
does not impact the design basis of the containment, and does not 
make any physical or operational changes to existing plant 
structures, systems, or components.
    Historically, Type A tests have a relatively low failure rate 
where Type B and C testing (local leakage rate tests) could not 
detect the leakage path. Most Type A test failures are attributed to 
failures of Type B or C components (containment penetrations and 
isolation valves). Type B and C components are tested per 
Surveillance Requirement 4.6.1.2.d of the Millstone Unit No. 2 
Technical Specifications. These tests are required to be conducted 
at intervals no greater than 24 months. These local leakage rate 
tests provide assurance that containment integrity is maintained. 
The Type B and C tests will continue to be performed in accordance 
with the requirements of Surveillance Requirement 4.6.1.2.d.
    The previous Type A, B, and C tests demonstrate that Millstone 
Unit No. 2 has maintained control of containment integrity by 
maintaining a conservative margin between the acceptance criterion 
and the ``As-Found'' and ``As-Left'' leakage results. Based on this, 
the Millstone Unit No. 2 containment is considered to be in sound 
condition.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.1.2.a of the Millstone Unit No. 2 Technical 
Specifications does not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change to Surveillance Requirement 4.6.1.2.a of the 
Millstone Unit No. 2 Technical Specifications will increase the 
flexibility in scheduling the Type A tests. It does not make any 
physical or operational changes to existing plant structures, 
systems, or components. In addition, the proposed change does not 
modify the acceptance criteria for the Type A tests. Maintaining the 
leakage through the containment boundary to the atmosphere within a 
specific value ensures that the plant complies with the requirements 
of 10CFR100. The containment boundary serves as an accident 
mitigator; it is not an accident initiator. Therefore, the proposed 
change to Surveillance Requirement 4.6.1.2.a does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The proposed change to Surveillance Requirement 4.6.1.2.a of the 
Millstone Unit No. 2 Technical Specifications will increase the 
flexibility for scheduling the Type A tests. It does not modify the 
maximum allowable leakage rate at the design containment pressure, 
does not impact the design basis of the containment, and does not 
make any physical or operational changes to existing plant 
structures, systems, or components.
    Based on the above, 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 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 Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: February 1, 1995
    Description of amendment request: This Technical Specification (TS) 
change would modify the applicable operational conditions for the 
secondary containment isolation radiation monitors located on the 
refueling floor and for the radiation monitor located in the railroad 
access shaft. Specifically, for the refueling floor exhaust duct and 
wall exhaust duct radiation monitors, the proposed change would modify 
the applicable operational condition during specific control rod 
testing evolutions which are core alterations and would indicate that 
the operability requirement does not apply during shutdown margin 
demonstrations. For the railroad access shaft exhaust duct radiation 
monitor, the change to the TS would modify the applicable operational 
condition to address plant evolutions involving irradiated fuel 
transfer within the railroad accessshaft and above the access shaft 
with the equipment hatch open.
    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:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    a. The proposed change to the applicable operational condition 
for the refueling floor process radiation monitors does not affect 
the probability of the design basis accidents. The monitors function 
in response to an airborne radioactivity concentration in the 
unfiltered air from the Zone III exhaust system and provide 
isolation signals which limit offsite doses to within regulatory 
limits. As such, there is no correlation between monitor operability 
and accident probability. The monitors act to mitigate the offsite 
effects of airborne contamination producing accidents, they are not 
potential accident initiators.
    The proposed change does not result in a significant increase in 
the consequence of the design basis accidents. The postulated event 
associated with control rod related CORE ALTERATIONS which could 
result in increased Zone III airborne radioactivity concentrations 
is criticality resulting from a single control rod withdrawal, 
resulting in release of fission products. The probability of an 
unintended criticality from a single control rod withdrawal is low, 
and the potential for this criticality to result in fuel failure 
under shutdown conditions is even more remote. Withdrawal of a 
single control rod is an analyzed evolution during which time 
adequate design and operating controls exist to preclude 
criticality. However, in the unlikely event criticality should 
occur, the potential offsite effects would not be significant. 
Localized criticality involving a leaking rod, or criticality 
induced fuel failure, are the postulated mechanisms by which an 
increase in Zone III airborne radioactivity could be attained. 
Neither of these postulated, but very unlikely events, will result 
in radioactive release in excess of 10CFR100 limits. Any release 
would be monitored by instrumentation in the Reactor Building vent 
stack required to be OPERABLE at all times. In addition, Area 
Radiation Monitors are installed on the refueling floor to 
supplement the refueling floor process radiation monitors by 
providing radiological information to plant operators. Operators can 
use the vent stack and/or ARM information to manually initiate 
secondary containment isolation if radiological conditions warrant 
this action. Emergency Operating Procedures direct operator action 
in the event of higher than normal radiation readings.
    b. The proposed change to the applicable operational condition 
for the railroad access shaft process radiation monitor does not 
affect the probability of the design basis accidents. The monitor 
functions in response to an airborne radioactivity concentration in 
the unfiltered air from the Zone III exhaust [[Page 16193]] system 
and provides isolation signals which limit offsite doses to within 
regulatory limits. As such, there is no correlation between monitor 
operability and accident probability. The monitor acts to mitigate 
the offsite effects of airborne contamination producing accidents, 
it is not a potential accident initiator.
    The proposed change does not result in a significant increase in 
the consequence of the design basis accidents. The design intent of 
the railroad access shaft process radiation monitor is to monitor 
radiation in the unfiltered air from the Zone III railroad access 
shaft exhaust system, and provide signals which automatically 
isolate the Zone III portion of the secondary containment, start the 
Standby Gas Treatment System, and start the Recirculation System 
(Zone III) on a high radiation condition within the access shaft. 
This function is intended to limit the consequences of a fuel 
handling accident in the railroad access shaft. The monitor has no 
significant capability to react to a CORE ALTERATION related 
transient, or one resulting from operations with the potential to 
drain the reactor vessel. The design intent of the monitor is 
maintained under the proposed change, as the proposed change focuses 
monitor operability on conditions when irradiated fuel is in the 
railroad access shaft or above it with the railroad access shaft 
cover open.
    For the above stated reasons, the applicable operational 
condition for Refueling Floor Exhaust Duct High Radiation Monitors, 
Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft 
Exhaust Duct Radiation Monitor can be modified without significantly 
increasing the probability or consequences of an accident previously 
evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The Refueling Floor Exhaust Duct High Radiation Monitors, Wall 
Exhaust Duct Radiation Monitors, and the Railroad Access 
ShaftExhaust Duct Radiation Monitor function in response to an 
airborne radioactivity concentration in the unfiltered air from the 
Zone III exhaust system and provide isolation signals which limit 
offsite doses to within regulatory limits. As such, there is no 
correlation between monitor operability and the potential for 
creating new or different accident scenarios. The monitors act to 
mitigate the offsite effects of airborne contamination producing 
accidents, they are not potential accident initiators.
    For the above stated reasons, the applicable operational 
condition for Refueling Floor Exhaust Duct High Radiation Monitors, 
Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft 
Exhaust Duct Radiation Monitor can be modified without creating the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    a. The proposed change to the applicable operational condition 
for the refueling floor process radiation monitors does not involve 
a significant reduction in the margin of safety. The postulated 
event associated with control rod related CORE ALTERATIONS which 
could result in increased Zone III airborne radioactivity 
concentrations is criticality resulting from a single control rod 
withdrawal under shutdown conditions. There are multiple barriers to 
protect against the postulated event of criticality from a single 
rod withdrawal. Technical Specifications, plant operating 
procedures, and plant design control the withdrawal of control rods 
to minimize the potential for an inadvertent criticality event 
during shutdown. In addition, a fuel loading verification is 
performed, per procedure, on the as loaded core configuration to 
ensure that the fuel is loaded correctly. Each reload core is 
designed such that there is at least a 99.9% probability with a 95% 
confidence that the core will not be critical as a result of a 
single control rod withdrawal. The safety margin associated with a 
potential criticality event from a single control rod withdrawal, 
under shutdown conditions, is not impacted by the proposed change.
    In the unlikely event that control rod manipulations resulted in 
reactor criticality, adequate protective measures are provided by 
core monitoring instrumentation required to be operable in OPCON 5. 
Under this scenario, assuming the inadvertent control rod withdrawal 
resulted in a significant reactivity addition, the Reactor 
Protection System (RPS) would respond by inserting all control rods 
via the Scram function. The RPS monitors for recriticality during 
OPCON 5 with SRMs (per Technical Specification Section 3.9.2), and 
IRMs. The safety margin associated with RPS response to a 
criticality event, under shutdown conditions, is not impacted by the 
proposed change.
    Assuming that a criticality did occur as a result of a single 
control rod withdrawal, any increase in Zone III airborne 
radioactivity from a previously failed assembly located in the 
vicinity of the withdrawn control rod or a fuel rod failure 
associated with the control rod withdrawal would not result in an 
offsite dose exceeding regulatory limits. Assuming that criticality 
occurs following core loading and verification (i.e. 20 
days after shutdown), the offsite dose as a result of the release of 
fission products from a single failed fuel rod would be much less 
than 1% of the applicable site boundary limits. In addition, the 
failure of four complete fuel assemblies (i.e. nearly equal to 300 
fuel rods in the bundles surrounding the withdrawn control rod) 
would not result in offsite dose exceeding the applicable regulatory 
limits. Failure of more than four complete fuel assemblies due to 
the withdrawal of a single control rod in OPCON 5 is not considered 
credible. In fact, given the initial conditions of this event (i.e. 
cold, zero power, subcritical) and the reactivity characteristics of 
the fuel (i.e. negative fuel temperature reactivity coefficient) it 
is very unlikely that a criticality of this nature would result in 
failure of any fuel rods. Although the refueling floor process 
radiation monitors would not be OPERABLE, Zone III airborne 
radioactivity concentrations can be independently detected with Area 
Radiation Monitors (ARMs) which are located on the refueling floor. 
These monitors provide control room indication, and would alert 
operators to changing radiological conditions on the refueling 
floor. In addition to providing personnel notification, the ARMs act 
as a supplement to the process radiation monitors in detecting 
abnormal migrations of radioactive material in or from the process 
streams. Operators can manually initiate secondary containment 
isolation based on ARM input. The Emergency Operating Procedures 
require the operators to take appropriate actions on higher than 
normal radiation readings. Moreover, any airborne radioactivity 
leakage from Zone III would be monitored via instrumentation in the 
Reactor Building vent stack required to be OPERABLE at all times; 
local alarms, remote recording, and main control room and Technical 
Support Center alarms are provided. Operators can manually initiate 
secondary containment isolation based on exhaust sample readings. 
Due to the bounding regulatory limits and the redundant monitoring 
and operator response capabilities, the safety margin associated 
with the potential for offsite airborne radioactive release, under 
shutdown conditions, is not significantly impacted by the proposed 
change.
    b. The elimination of operability requirements associated with 
CORE ALTERATIONS, operations with the potential to drain the reactor 
vessel, and other irradiated fuel moves not associated with the 
railroad access shaft, do not affect the ability of the railroad 
access shaft process radiation monitor to implement its design 
function. As such, the current operability requirements for the 
monitor which involve evolutions in areas other than the railroad 
access shaft do not contribute to the margin of plant safety; thus 
eliminating these operability requirements will not reduce the 
margin of plant safety.
    For the above stated reasons, the applicable operational 
condition for Refueling Floor Exhaust Duct High Radiation Monitors, 
Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft 
Exhaust Duct Radiation Monitor can be modified without 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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz [[Page 16194]] 

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: February 2, 1995
    Description of amendment request: This amendment would change the 
Technical Specifications for the units to increase the licensed 
discharge fuel assembly exposure for SPC 9X9-2 fuel from 40 to 45 GWD/
MTU.
    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 changes do not:
    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    PP&Ls technical basis for increasing the licensed discharge 
exposure limit as proposed is documented in PL-NF-94-005-P-A. The 
technical basis includes onsite fuel inspections, fuel design 
analyses and evaluations, and an in-reactor fuel assembly extended 
exposure demonstration. In response to NRC concerns on fuel failures 
at higher exposures, very conservative analyses were performed for 
the CRDA [control rod drive assembly] assuming very low failure 
thresholds, and offsite dose calculation results were shown to be 
well within regulatory limits, even at a failure threshold of 30 
cal/gm. The NRC has previously reviewed and approved all of the 
above information, and inspection results have met all approved 
criteria.
    An evaluation of FSAR [Final Safety Analysis Report] design 
basis events was performed to determine the impact of the proposed 
increase in fuel exposure. The LOCA [loss-of-coolant accident] 
analysis performed in support of PP&Ls Power Uprate efforts 
incorporated the effects of higher exposure and LHGR [linear heat 
generation rate]. From a radiological release perspective, the Power 
Uprate evaluations of LOCA, MSLB [main steam line break], CRDA, and 
refueling accidents each bound the potential impacts of extended 
exposure fuel.
    Those reload analyses deemed necessary to confirm that the above 
conclusions remain valid will be performed on a cycle-specific 
basis.
    Based on the above, the proposed action will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed action will increase the residence time of fuel 
within the Susquehanna reactors. The potential consequences of this 
action remain solely with the fuels ability to perform within 
specified limits during the increased duty, and were reviewed in I 
above. All required evaluations involving fuel impacts have been 
previously evaluated.
    Based on the above, the proposed action cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    III. Involve a significant reduction in a margin of safety.
    The proposed action will allow increasing the licensed discharge 
fuel assembly exposure limit, resulting in increases in the fuel rod 
LHGR and LHGR for APRM [average power range monitor] Setpoints, 
which are controlled via the Technical Specifications and the Core 
Operating Limits Report.
    The discussion in I. above delineates the evaluations performed 
to support this action. It concludes that neither the probability 
nor the consequences of events previously evaluated will be 
affected. Operator performance will not be affected, because the 
operators only monitor the ratio of the fuel LHGR to the fuel design 
limit. No other potentially impacted safety margins have been 
identified.
    Based on the above, the proposed 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:Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: February 10, 1995
    Description of amendment request: The proposed amendment would 
modify the Susquehanna Steam Electric Station, Unit 1 and 2 Technical 
Specifications (TS) to (1) extend the allowable out-of-service times 
(AOTs) for maintenance and repair and the surveillance test intervals 
(STIs) between channel functional tests for the following groups of 
instruments: reactor protection systems instrumentation (TS 3.3.1), 
isolation actuation instrumentation (TS 3.3.2), emergency core cooling 
system actuation instrumentation (TS 3.3.3), ATWS (anticipated 
transient without scram) recirculation pump trip system instrumentation 
(TS 3.3.4.1), end-of-cycle recirculation pump trip system 
instrumentation (TS 3.3.4.2), reactor core isolation cooling system 
(RCIC) actuation instrumentation (TS 3.3.5), control rod block 
instrumentation (TS 3.3.6), radiation monitoring instrumentation (TS 
3.3.7.1), and feedwater/main turbine trip system actuation 
instrumentation (TS 3.3.90); (2) change the required actions and AOTs 
for the instruments listed above to make requirements consistent with 
supporting analysis in General Electric topical reports and change 
additional actions required to prevent extended AOTs from resulting in 
extended loss of instrument function; (3) change the required actions 
and AOTs for the instruments listed above for instrumentation 
associated with the ADS (automatic depressurization system), 
recirculation pump trip, and pump suction lineup for HPCI (high 
pressure core injection) and RCIC; (4) change applicability 
requirements and required actions for the reactor vessel water level-
low, level 3 function that isolates the RHR (residual heat removal) 
system shutdown cooling system so that the function is required to be 
operable in operational conditions 3, 4, and 5 to prevent inadvertent 
loss of reactor coolant via the RHR shutdown cooling system; (5) remove 
notes in Table 3.3.2-1, 3.3.2-2, and 4.3.1-1 related to maintenance on 
leak detection temperature detectors and remove the note to TS 3.3.6 
for Unit 1 related to a previous relief from TS 3.0.4; and reformat, 
renumber, and/or reword existing requirements to incorporate the 
changes listed above.
    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:
    I. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS changes increase the AOTs and STIs for actuation 
instrumentation intended to detect or mitigate accidents; establish 
required actions consistent with NUREG-1433 for some instruments 
that are more specific but equivalent to existing required actions; 
establish new requirements to prevent inadvertent loss of reactor 
coolant via the RHR Shutdown Cooling System during OPERATIONAL 
CONDITIONS 3, 4 and 5; and, eliminate notes that were intended to 
provide one time only exemptions from certain requirements. The 
proposed changes affect only those Technical Specification 
requirements that govern operability, required actions and routine 
testing of plant instruments that detect or mitigate accidents. The 
proposed changes do [[Page 16195]] not affect any equipment or 
requirements that are assumed to be initiators of any analyzed 
events. Therefore, the proposed changes will not involve an increase 
in the probability of occurrence of an accident previously 
evaluated.
    The proposed changes will not increase the consequences of an 
accident previously evaluated because the changes will not involve 
any physical changes to plant systems, structures, or components 
(SSC), or the manner in which these SSC are operated, maintained, 
modified, tested or inspected. The proposed changes will not alter 
the operation of equipment assumed to be available for the 
mitigation of accidents or transients by the plant safety analysis 
or licensing basis. The proposed changes extend the intervals 
between required performances of routine instrument testing. The 
proposed changes also modify time limits allowed for operation with 
inoperable instrument channels in situations when an inoperable 
instrument channel would not prevent actuation of the associated 
equipment. These changes are based on the demonstrated reliability 
of these instruments and are justified by the analysis in References 
1 through 8 [See February 10, 1995 application]. The small increases 
in the probability that the proposed changes will result in an 
equipment actuation failure has been determined in References 1 
through 8 [See February 10, 1995 application] to be offset by safety 
benefits such as a reduction in the number of inadvertent 
actuations, a reduction in wear due to excessive testing, and better 
utilization of plant personnel and resources. These changes will not 
allow continuous plant operation with plant conditions such that a 
single failure will result in a loss of any safety function.
    Proposed changes to required actions and completion times for 
instrumentation associated with the ADS initiation, Recirculation 
Pump Trip, and pump suction lineup for HPCI and RCIC make the 
required actions and completion times consistent with NUREG-1433, 
Standard Technical Specifications for General Electric Plants, BWR/
4, Revision 0 (Reference 12). These changes are also consistent with 
the assumptions used in References 1 through 8 [See February 10, 
1995 application]. Therefore, these changes establish or maintain 
adequate assurance that components are operable when necessary for 
the prevention or mitigation of accidents or transients and that 
plant variables are maintained within limits necessary to satisfy 
the assumptions for initial conditions in the safety analysis. In 
addition, the proposed change provides the benefit of avoiding an 
unnecessary shutdown transient when appropriate measures are 
available to compensate for the inoperable instrumentation. 
Therefore, the proposed changes will not increase the consequences 
of an accident previously evaluated.
    There is no significant increase in the probability or 
consequences of an accident previously evaluated resulting from 
changes that reformat, renumber, and/or reword existing requirements 
to incorporate the changes above or from the removal of notes that 
were intended for one time only use and are no longer applicable.
    II. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change will not involve any physical changes to 
plant systems, structures, or components (SSC), or the manner in 
which these SSC are operated, maintained, modified, tested, or 
inspected. The changes in normal plant operation are consistent with 
the current safety analysis assumptions. Therefore, this change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    III. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed TS changes: increase the AOTs and STIs for 
actuation instrumentation intended to detect or mitigate accidents; 
establish required actions consistent with NUREG-1433 for some 
instruments that are more specific but equivalent to existing 
required actions; establish new requirements to prevent inadvertent 
loss of reactor coolant via the RHR Shutdown Cooling System during 
Operational Conditions 3, 4 and 5; and, eliminate notes that were 
intended to provide one time only exemptions from certain 
requirements.
    There is no significant reduction in the margin of safety 
resulting from changes to the minimum surveillance test intervals 
(STIs) and allowable out-of-service times (AOTs) for the testing 
and/or repair of instrumentation. This conclusion is based on the 
demonstrated reliability of these instruments and is justified by 
the analysis in References 1 through 8 [See February 10 1995 
application]. The small increases in the probability that the 
proposed changes will result in an equipment actuation failure has 
been determined in References 1 through 8 [See February 10, 1995 
application] to be offset by safety benefits such as a reduction in 
the number of inadvertent actuations, a reduction in wear due to 
excessive testing.
    These changes will not allow continuous plant operation with 
plant conditions such that a single failure will result in a loss of 
any safety function.
    There is no significant reduction in the margin of safety 
resulting from changes to required actions and completion times for 
instrumentation associated with the ADS initiation, Recirculation 
Pump Trip, and pump suction lineup for HPCI and RCIC. These changes 
make the required actions and completion times consistent with 
NUREG-1433, Standard Technical Specifications for General Electric 
Plants, BWR/4. These changes are also consistent with the 
assumptions used in References 1 through 8 [See February 10, 1995 
application]. Therefore, these changes establish or maintain 
adequate assurance that components are operable when necessary for 
the prevention or mitigation of accidents or transients and that 
plant variables are maintained within limits necessary to satisfy 
the assumptions for initial conditions in the safety analysis. In 
addition, the proposed change provides the benefit of avoiding an 
unnecessary shutdown transient when appropriate measures are 
available to compensate for the inoperable instrumentation. 
Additionally, the proposed required actions ensure that actions to 
mitigate loss of single failure tolerance are initiated within 24 
hours (12 hours for RPS) in accordance with the results of the 
analyses in References 1 through 8 [See February 10, 1995 
application] and action to mitigate a loss of instrument function is 
initiated within 1 hour. Therefore, these changes will not allow 
continuous plant operation with plant conditions such that a single 
failure will result in a loss of any safety function. The 
Pennsylvania Power & Light Company performed reviews that confirmed 
the analyses in References 1 through 8 [See February 10, 1995 
application] are applicable to SSES and that there would be no 
effect on the identification of excessive instrument setpoint drift 
as a result of increasing the minimum interval between instrument 
functional tests from monthly to quarterly.
    There is no significant reduction in the margin of safety 
resulting from changes that reformat, renumber, and/or reword 
existing requirements to incorporate the changes above or from the 
removal of notes that were intended for one time only use and are no 
longer applicable.
    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas 
Company,Delmarva Power and Light Company, and Atlantic City 
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: January 13, 1995
    Description of amendment request: The proposed changes concern a 
revision to the frequency of calibration for the Local Power Range 
Monitor (LPRM) signals from every 6 weeks to every 2000 Megawatt Days 
per Standard Ton (MWD/ST).
    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: [[Page 16196]] 
    1. The proposed change does not involve a significant increase 
in the probability of consequences of an accident previously 
evaluated.
    This change does not affect the operation of any equipment. The 
change does not affect the fundamental method by which the LPRMs are 
calibrated. The increased time between required LPRM calibrations 
does not affect either the initiator of any accident previously 
evaluated or any equipment required to mitigate the consequences of 
an accident, or the isotopic inventory in the fuel. Thus, the change 
does not increase either the probability or the radiological 
consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not introduce a new mode of plant 
operation and does not involve the installation of any new equipment 
or modifications to the plant. Therefore, it does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The GETAB determination of the Maximum Critical Power Ratio 
(MCPR) Safety Limit allows a maximum total nodal uncertainty of the 
TIP readings (of which the LPRM Update uncertainty is a part) of 
8.7%. The change in LPRM calibration frequency results in an LPRM 
Update uncertainty of 4.2% nodal power. This, combined with the 
other uncertainties which comprise the total TIP readings 
uncertainty, yields a total TIP readings uncertainty of less than 
the allowed 8.7%. Thus the change in LPRM calibration frequency will 
not affect the MCPR Safety Limit.
    The LPRMs are utilized as input to the APRM and RBM systems. The 
primary safety function of the APRM system is to initiate a scram 
during core-wide neutron flux transients before the actual core-wide 
neutron flux level exceeds the safety analysis design basis. This 
prevents fuel damage from single operator errors or equipment 
malfunctions. The APRMs are calibrated at least twice per week to 
the plant heat balance, utilize a radially and axially diverse group 
of LPRMs as input and are utilized to detect changes in average, not 
local, power changes. Therefore, the effects of decreasing the LPRM 
calibration frequency on the APRM system responses will be minimal 
due to any individual LPRM drift being practically canceled out (due 
to diversity of input) and/or due to the frequent recalibration of 
the APRMs to an independent power calculation (the heat balance). 
Thus, decreasing the LPRM calibration frequency will not 
significantly impact the performance of the APRM system's scram 
function, and there is no impact on transient delta-CPRs.
    The RBM system is utilized in the mitigation of a Rod Withdrawal 
Error (RWE). The RBM system is designed to prevent the operator from 
increasing the local power significantly when withdrawing a control 
rod. On each selection of a control rod, the average of the 
assigned, unbypassed LPRMs is adjusted to equal a 100% reference 
signal for each of the two RBM channels. Each RBM channel 
automatically limits the local thermal margin changes by limiting 
the allowable change in local average neutron flux to the RBM 
setpoint. If the local average neutron flux change is greater than 
that allowed by the RBM setpoint, within either RBM channel, the rod 
withdrawal permissive is removed preventing further movement. Since 
the change in local neutron flux is calculated from the change in 
the average of the LPRM readings, and calibrated on every rod 
selection to the reference signal, offsets in individual LPRM 
readings due to calibration differences are effectively eliminated 
for a given RBM setpoint. Therefore, the constraints on the 
withdrawal of any given rod are unchanged and there will not be any 
increase in RWE delta-CPR.
    Since the MCPR Safety Limit is unaffected and the delta-CPR 
values are unchanged, the cycle CPR limits are unchanged. Therefore, 
the change in the frequency of LPRM calibration does not result in a 
reduction in a margin of safet