[Federal Register Volume 63, Number 91 (Tuesday, May 12, 1998)]
[Notices]
[Pages 26213-26216]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-12526]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-244]


Rochester Gas and Electric Corporation; Notice of Consideration 
of Issuance of Amendment to Facility Operating License, Proposed No 
Significant Hazards Consideration Determination, and Opportunity for a 
Hearing

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an amendment to Facility Operating License No. 
DRP-18 issued to Rochester Gas and Electric Corporation (the licensee) 
for operation of the R.E. Ginna Nuclear Power Plant located in Wayne 
County, New York.

[[Page 26214]]

    The proposed amendment would revise the Ginna Station Improved 
Technical Specifications (ITS) to reflect a planned modification to the 
spent fuel pool (SFP) storage racks. Specifications associated with SFP 
boron concentration, fuel assembly storage, and maximum limit on the 
number of fuel assemblies which can be stored in the SFP would be 
revised.
    Before issuance of the proposed license amendment, the Commission 
will have made findings required by the Atomic Energy Act of 1954, as 
amended (the Act) and the Commission's regulations.
    The Commission has made a proposed determination that the amendment 
request involves no significant hazards consideration. Under the 
Commission's regulations in 10 CFR 50.92, this means that operation of 
the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The design basis events considered for the spent fuel pool 
include both external events and postulated accidents in the pool. 
The external events considered are tornado missiles and seismic 
events. The evaluation of the postulated impact of a tornado missile 
is detailed in Sections 3, 4, and 6 of Reference 1 [see application 
dated March 31, 1997]. The structural evaluation indicates that 
there are no gross distortions of the racks or any adverse effects 
upon plant structures or equipment. The radiological consequences of 
this event indicate that offsite doses are ``well within'' the 10 
CFR 100 limits.
    The structural evaluation is detailed in Section 3 of Reference 
1 [see application dated March 31, 1997]. Current state of the art 
methods are used in the structural analysis. The evaluation of the 
storage racks is based on a conservative interpretation of the ASME 
[American Society of Mechanical Engineers] Boiler and Pressure 
Vessel Code. The evaluation of the spent fuel pool is based on a 
conservative interpretation of requirements set forth in the 
American Concrete Institute, Code Requirements for Nuclear Safety 
Related Concrete Structures, and American Institute of Steel 
Construction, Specification for Structural Steel Buildings. The 
spent fuel storage system was designed to meet all applicable 
structural criteria for normal (Level A), upset (Level B), and 
faulted (Level D) conditions as defined in NUREG-0900, SRP [Standard 
Review Plan] 3.8.4, Appendix D. The following loadings were 
considered: dead weight, seismic, thermal, stuck fuel assembly, drop 
of a fuel assembly, and tornado missile impact. Load combinations 
were performed in accordance with SRP 3.8.4, Appendix D. Given the 
evaluated seismic events, the changes in the final position of the 
racks are small as compared to the initial position prior to the 
seismic event. The maximum closure of gaps is such that no 
significant changes in gaps results during any single seismic event. 
Furthermore, the combined gap closures resulting from a combination 
of 5 OBEs [Operating Basis Earthquakes] and 1 SSE [Safe Shutdown 
Earthquake] show that there are no rack-to-rack or rack-to-wall 
impacts. These evaluations conclude that under these postulated 
events, the stored fuel assemblies are maintained in a stable, 
coolable geometry, and a subcritical configuration.
    As described in the bases for LCO [Limiting Condition for 
Operation] 3.7.12 and 3.7.13, the postulated accidents in the spent 
fuel pool are divided into two categories. The first are those 
involving a loss of cooling in the spent fuel pool. The thermal-
hydraulic analysis for the maximum expected decay heat loads is 
described in Section 5 of Reference 1 [see application dated March 
31, 1997]. The proposed modification does not change the 
configuration of the available spent fuel cooling systems, the 
limiting design conditions for maximum decay heat load which occurs 
during a full core offload, or the existing requirement to maintain 
pool temperature below 150  deg.F. Utilizing the three available 
spent fuel cooling systems, Ginna Station maintains full redundancy 
during high heat load conditions. The decay heat load to the spent 
fuel pool is maintained within the capacity of the operating cooling 
system by appropriately delaying fuel offload from the reactor. 
Should a failure occur on the operating cooling system, the 
resulting heat rates allow sufficient time to place a standby 
cooling system in service before the pool design limit temperature 
is exceeded. Increases in spent fuel pool temperature, with the 
corresponding decrease in water density and void formation from 
boiling, will result in a decrease in reactivity due to the decrease 
in moderation effects. In addition, the analysis demonstrates that 
the storage rack geometry and required fuel storage configurations 
result in a k<INF>eff</INF> [less than or equal to] .95 assuming no 
soluble boron allowing for the potential of makeup to the pool with 
unborated water if credit is taken in Region 2 for minimal 
availability of boraflex panels installed on the storage rack. (Note 
that concerns with boraflex degradation are discussed later in this 
evaluation).
    The second category is related to the movement of fuel 
assemblies and other loads above the spent fuel pool. The limiting 
accident with respect to reactivity is the fuel handling accident 
which is analyzed in Section 4 of Reference 1 [see application dated 
March 31, 1997]. For both the incorrectly transferred fuel assembly 
(placed in an unauthorized location) or a dropped fuel assembly, the 
positive reactivity effects resulting are offset by the negative 
reactivity from the required minimum soluble boron concentration. 
The resulting k<INF>eff</INF> is shown to be less than 0.95 if 
credit is taken in Region 2 for minimal availability of boraflex 
panels installed on the storage racks. The radiological consequences 
of a fuel assembly drop remain as described in Section 15.7.3 of the 
UFSAR [updated final safety analysis report] and as discussed in 
Section 6 of Reference 1 [see application dated March 31, 1997]. 
Loads in excess of a fuel assembly and its handling tool are 
administratively prohibited from being carried over spent fuel. 
There are no changes anticipated for either the fuel handling 
equipment of the auxiliary building overhead crane due to the 
proposed modification to the fuel storage racks. The modification is 
scheduled for the Year 1998 to be performed while Ginna Station is 
operating. Movement of heavy loads around the spent fuel pool are 
controlled by the requirements of NUREG-0612 and the regulatory 
guidelines set forth in NRC Bulletin 96-02 (see Section 3 of 
Reference 1 [see application dated March 31, 1997]). Spent fuel 
casks and storage racks (during removal and installation) will be 
moved using the auxiliary building crane and lifting attachments 
satisfying the single failure proof criteria of NUREG-0554, 
obviating the need to determine the consequences for this accident.
    Due to boraflex degradation within the spent fuel pool, credit 
must be temporarily taken for soluble boron to maintain 
k<INF>eff</INF> [less than or equal to] 0.95. There is no increase 
in the probability of a loss of spent fuel pool cooling or fuel 
handling accident as a result of crediting soluble boron. The spent 
fuel pool is normally maintained at a boron concentration level 
greater than that proposed, including during fuel movement. 
Therefore, there is no effect on plant systems or spent fuel pool 
activities than which are currently in effect. The proposed boron 
concentration level is also equivalent to that required by LCO 3.9.1 
during MODE 6 such that no boron dilution event is expected to occur 
within the pool during refueling operations when the reactor coolant 
system and spent fuel pool are hydraulically coupled.
    Crediting soluble boron does not increase the consequences of an 
accident. As described in the bases for LCO 3.7.12, increases in 
spent fuel pool temperature, with the corresponding decrease in 
water density and void formation from boiling, will generally result 
in a decrease in reactivity due to the decrease in moderation 
effects. The only exception are temperature bands where positive 
reactivity is added as a result of the high boron concentration. 
This effect is bounded by the reactivity added as a result of a 
misloaded fuel assembly. With respect to the more limiting dropped 
fuel assembly accidents, boraflex neutron absorber panels were 
originally assumed in the criticality analysis. Requiring a high 
concentration of soluble boron in place of boraflex panels ensures 
that the spent fuel pool remains subcritical with k<INF>eff</INF> 
[less than or equal to] 0.95 for these accidents. Fuel assembly 
movement will continue to be controlled in accordance with plant 
procedures and LCO

[[Page 26215]]

3.7.13 which specifies limits on fuel assembly storage locations. 
Periodic surveillances of boron concentration will be required every 
7 days with level verified every 7 days during fuel movement per LCO 
3.7.11. Due to the large inventory within the spent fuel pool, 
dilution of the soluble boron within the pool is very unlikely 
without being detected by operations personnel during auxiliary 
operator rounds or available level detection systems. There is also 
a large margin between the required boron concentration to maintain 
the pool subcritical k<INF>eff</INF> [less than or equal to] 0.95 
and the proposed value (approximately 900 ppm).
    Based on the above, it is concluded that the proposed changes do 
not significantly increase the probability or consequences of any 
accident previously analyzed.
    2. Operation in accordance with the proposed changes does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed modification does not alter the function of any 
system associated with spent fuel handling, cooling or storage. The 
proposed changes do not involve a different type of equipment or 
changes in methods governing normal plant operation. The additional 
restrictions placed on the acceptable storage locations for spent 
fuel are consistent with the type of restriction that previously 
existed. The potential violation of these restrictions (incorrectly 
transferred fuel assembly) are analyzed as discussed above. The 
rerack design, analysis, fabrication, and installation meet all the 
appropriate NRC regulatory requirements, and appropriate industry 
codes and standards.
    Crediting soluble boron within the spent fuel pool in place of 
boraflex neutron absorber panels does not create the possibility of 
a new or different kind of accident since the spent fuel pool is 
normally maintained with high boron concentrations. Assuming a boron 
dilution event to the level required to reach k<INF>eff</INF> [less 
than or equal to] 0.95 conditions within the spent fuel pool would 
require either overfill of the pool or a controlled feed and bleed 
process with unborated water. In both cases, greater than 105,000 
gallons of unborated water would be required to reach 
k<INF>eff</INF> > 0.95. There is no source of unborated water of 
this size available to reach the spent fuel pool under procedural 
control or via a pipe break other than a fire water system pipe 
break or SW leak through the spent fuel pool heat exchangers. 
However, there are numerous alarms available within the control room 
to indicate this condition including high spent fuel pool water 
level and sump pump actuations within the residual heat removal pump 
pit (lowest location in the Auxiliary Building). Auxiliary operators 
also perform regularly scheduled tours within the Auxiliary 
Building. This provides sufficient time to terminate the event such 
that there is no credible spent fuel pool dilution accident.
    Based on the above, the change does not create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in the margin of 
safety.
    The Licensing Report enclosed as Reference 1 [see application 
dated March 31, 1997] addresses the following considerations: 
nuclear criticality, thermal-hydraulic, and mechanical, material, 
and structural. Results of these evaluations demonstrate that the 
changes associated with the spent fuel reracking does not involve a 
significant reduction in the margin of safety as summarized below:

Nuclear Criticality

    The established regulatory acceptance criterion is that 
k<INF>eff</INF> be less than or equal to 0.95, including all 
uncertainties at the 95/95 probability/confidence level, under 
normal and abnormal conditions. The methodology used in the 
evaluation meets NRC requirements, and applicable industry codes, 
standards, and specifications with credit taken in Region 2 for the 
previously installed boraflex panels. In addition, the methodology 
has been reviewed and approved by the NRC in recent nuclear 
criticality evaluations. Specific conditions which were evaluated 
include misloading of a fuel assembly, drop of a fuel assembly 
(shallow, deep drops, and side drops), pool water temperature 
effects, and movement of racks due to seismic events. Results 
described in Section 4 of Reference 1 [see application dated March 
31, 1997] document that the criticality acceptance criterion is met 
for all normal and abnormal conditions.

Thermal-Hydraulic

    Conservative methods and assumptions have been used to calculate 
the maximum temperature of the fuel and the increase of the bulk 
pool water temperature in the spent fuel pool under normal and 
abnormal conditions. The methodology for performing the thermal-
hydraulic evaluation meets NRC regulatory requirements. Results from 
the thermal-hydraulic evaluation show that the maximum temperature 
at the hottest fuel assembly, intact or consolidated canister, is 
less than the temperature for nucleate boiling condition. The 
effects of cell blockage on the maximum temperature of intact fuel 
and consolidated canisters were evaluated. Results described in 
Section 5 of Reference 1 [see application dated March 31, 1997] show 
that adequate cooling of the intact or consolidated fuel is assured. 
In all cases, the existing spent fuel pool cooling system will 
maintain the bulk pool temperature at or below 150  deg.F by 
delaying core offload from the reactor.

Mechanical, Material, and Structural

    The primary safety function of the spent fuel pool and the racks 
is to maintain the spent fuel assemblies in a safe configuration 
through all normal and abnormal loads. Abnormal loadings which have 
been considered in the evaluation are: seismic events, the drop of a 
fuel assembly, the impact of a tornado missile, a stuck assembly, 
and the drop of a heavy load. The mechanical, material, and 
structural design of the new spent fuel racks is in accordance with 
NRC regulatory requirements (including the NRC OT Position dated 
April 14, 1978, [NRC letter to all power reactor licensees dated 
April 14, 1978] and addendum dated January 18, 1979), and applicable 
industry standards. The rack materials are compatible with the spent 
fuel pool environment and fuel assemblies. The material used as a 
neutron absorber (borated stainless steel) has been approved by the 
American Society for Testing and Materials (ASTM), and licensed 
previously by the NRC for use as a neutron absorber at Indian Point 
3, Indian Point 2, and Millstone 2. The structural evaluation 
presented in Section 3 of Reference 1 [see application dated March 
31, 1997] documents that the tipping or sliding of the free-standing 
racks will not result in rack-to-rack or rack-to-wall impacts during 
seismic events. The spent fuel assemblies will remain intact and the 
criticality criterion of k<INF>eff</INF> [less than or equal] to 
0.95 is met if credit is taken in Region 2 for previously installed 
boraflex panels.
    Soluble boron within the spent fuel pool provides a significant 
negative reactivity such that k<INF>eff</INF> is maintained [less 
than or equal to] 0.95. The proposed surveillance frequency will 
ensure that the necessary boron concentration is maintained. A boron 
dilution event which would remove the soluble boron from the pool 
has been shown to not be credible.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received. 
Should the Commission take this action, it will publish in the Federal 
Register a notice of issuance and provide for opportunity for a hearing 
after issuance. The Commission expects that the need to take this 
action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of

[[Page 26216]]

Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D59, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 
4:15 p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for hearing and 
petitions for leave to intervene is discussed below.
    By June 11, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and at the local public 
document room located at the Rochester Public Library, 115 South 
Avenue, Rochester, New York 14610. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to Nicholas S. Reynolds, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005, attorney for the 
licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for hearing will not 
be entertained absent a determination by the Commission, the presiding 
officer or the presiding Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment dated March 31, 1997, supplemented June 18, 
1997, October 10, 1997, October 20, 1997, November 11, 1997, December 
22, 1997, January 15, 1998, January 27, 1998, March 30, 1998, April 23, 
1998, and April 27, 1998, which are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room 
located at the Rochester Public Library, 115 South Avenue, Rochester, 
New York 14610. This notice supersedes the March 31, 1997, application 
published on April 30, 1997 (62 FR 23502) in its entirety.

    Dated at Rockville, Maryland, this day of May 1998.

    For the Nuclear Regulatory Commission.
Guy S. Vissing,
Senior Project Manager, Project Directorate I-1, Division of Reactor 
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 98-12526 Filed 5-11-98; 8:45 am]
BILLING CODE 7590-01-P