ML17255A705
| ML17255A705 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/15/1984 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17255A704 | List: |
| References | |
| NUDOCS 8403210143 | |
| Download: ML17255A705 (19) | |
Text
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 NOTICE OF CONSIDERATION'OF ISSUANCE OF AMENDMENT TO PROVISIONAL OPERATING LICENSE AND PROPOSED NO SIGNIFICANT.HAZARDS CONSIDERATION DETERMINATION AND OPPORTUNITY FOR HEARING The United States Nuclear Regulatory Commission (the Commission) is considering issuance of'n amendment to Provisional Operating License No.
'I DPR-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.
The proposed amendment to the Technical Specifications (TS) would incorporate certain provisions to permit a planned fuel design change from Exxon Nuclear Company (ENC) fuel to the Westinghouse 14X14 9-grid Optimized n
'Fuel Assembly (OFA) fuel with core loadings ranging from approximately a
15%
OFA and 85%
ENC fueled core to a core fueled entirely with OFA.
The proposed changes to the TS would allow (1) a positive moderator temperature coefficient (MTC) (+5 pcm/'F) up to 70% power, (2) a reduction in shutdown margin at EOC from 1900 pcm to 1800 pcm, (3) a change in the F> limits at less than 100~ power, (4) a change in the core protection limits (OTAT and OPDT setpoint equations),
and (5) a deletion of'he limits on Target Axial Offset.
The changes to the core physics parameters and.thermal character-istics are required to account for increased enrichment and improved neutronic char'acteristics of the fuel and control 'assemblies.
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.
eoosafoa4S sOOSie PDR-ADOCK 05000244 I p PDR
7590-01 The 14X14 OFA design meets the same basic requirements and criteria which were approved by the NRC for the 17X17 OFA design in t1ay 1981.
In addition there are four 14X14 7-grid demonstration assemblies that have
.completed two cycles of operation (established burnup 20,000 N<D/HTU).
Post-test examination at the completion of the first cycle of irradiation indicated no abnormalities.
- However, one demonstration assembly at the end of the second cycle of irradiation was damaged and removed.
It was concluded that the cause of the damage was an isolated event and not a generic OFA design problem (see Letter Report IT-83-222, "Failure Investigation of Point Beach Unit 2 OFA Rods," July 1983).
The demonstration assemblies will have experienced approximately 35,000 NWD/MTU of burnup in 1984.
.A'safety analysis was prepared which used parameters which will bound those to be experienced during the transition which will apply to subsequent cycles with the OFA.
The safety analysis was composed of a summary of the mechanical, thermal-hydraulic and accident analysis and detailed results of the non-LOCA and LOCA analysis.
Those analyses incorporated the proposed changes to the Technical Specifications and showed that the applicable design criteria for the Exxon and OFA were satisfied.
r Based on the results of evaluation/analyses and tests it was concluded that:
(1) The Westinghouse OFA's are mechanically and hydraulically compatible with the ENC fuel assemblies, control rods, and reactor internals interfaces.
All NRC approved design criteria for the Westinghouse OFA's are satisfied.
(2) Generally changes in the nuclear characteristics because of the transition from ENC to OFA fuel will lie within the cycle-to-cycle variations observed for past fuel reload design.
The moderator temperature'oefficient is the most significant exception to this.
Since the Hydrogen/Uranium ratio
r I~
7590-01 is larger for OFA, the moderator temperature coefficient is more positive than observed in past Westinghouse fueled R;E. Ginna cores.
This has been
,accounted for in the accident evaluations.
(3) Demonstration experience with Westinghouse OFAs containing Zircaloy grids provides reason to expect satisfactory operation from OFA Zircaloy grids.
(4) The proposed technical specifications changes are applicable to cores containing any combination of OFA and ENC fuel and plant operating limitations will be satisfied with these proposed changes.
(5)
A reference was established upon which to base V
future cycle safety evaluations for Westinghouse OFA reload fuel.
Four mixed oxide assemblies (MOX) will remain in the core for'Cycle 14.
.These assemblies are mechanically identical,to the Westinghouse HIPAR design used as reload fuel to Ginna prior to Cycle 8.
Exxon previously has performed a safety analysis and concluded on a best estimate basis that in a mixed core configuration the flow to each assembly was within one percent of the core average.
Applying a DNBR penalty equivalent to a decrease in one percent of flow to.. the minimum DNBR for Exxon fuel calculated by Westinghouse indicates that sufficient margin to the design DNBR limit exists.
The Commission has provided guidance concerning the application of standards for making a no significant hazards consideration determination by providing certain examples (April 6,
- 1983, 48 FR 14870).
The requested action does not fall within the scope of the examples of such actions.
Thus the bases for considering that the no significant hazards consideration deter-.
- mination, as specified in 10 CFR 50.92, have been met are delineated below.
Moderator Tem erature Coefficient (MTC
'The present Technical Specifications (TSs) require that the MTC be negative or positive at all times when the reactor is critical.
The
0 7590-01 proposed TS allows a +5 pcm/'F NTC below 70K of rated
- power, changing to a 0 pcm/'F MTC at 70% power. and above.
The MTC'is only a portion of the overall temperature 'coefficient.
Other factors which. contribute to the overall temperature coefficient compensate for the positive MTC at power levels below 70~ thus maintaining the overall temperature coefficient negative at all times.
The NTC is incorporated as an assumption into the analysis of Design I
Basis Events.
These events assume certain transients occur which challenge the safety systems of the plant.
A change in the MTC has no effect on the probability of such event.
Westinghouse has demonstrated that the results of these transients satisfy the acceptance criteria, therefore, the consequences of the accident are unchanged.
The possibility of a new accident is not created by the change in a V
core physics parameter such at NTC.
r Westinghouse analysis has shown that the limiting transients for which the assumption of positive MTC is conservative are those initiated from 1004 power.
However, the proposed TSs state at this 'condition the NTC must still be negative which is unchanged from previous analyses.
Westinghouse has incorporated the positive NTC into their transient analysis and has demonstrated that appropriate margin exists between the results and the acceptance criteria.
.Therefore, there is no significant reduction in the margin of safety.
Reduction in EOC Shutdown Mar in r
The EOC shutdown margin was incorporated into the Westinghouse transient analysis.
The limiting Design Basis Accident affected by the EOC shutdown margin is the Large Steam Break at OX power.
A change in this
7590-01 parameter has no effect on the probability of this event.
Westinghouse has demonstrated that the results of this transient satisfy the 'acceptance criteria.
Therefore, the consequences of the event are unchanged.
The change in the EOC shutdown margin is a small change in a known quantity and does not involve the possibility of a new or different kind of accident from any accident previously evaluated.
The Westinghouse transient analysis has incorporated the change in shutdown margin into the Large Steam Break.
THe results satisfy the acceptance criteria for Condition IV events.
Therefore, there is no reduction in the margin of safety.
Change in the F>H Limits at Less than 100Ã Power N
The limit on F~ has been increased and the equation format changed to be continuous over all power ranges.
The modification of tlie F> limit for'artial
- power, a 0.3 multiplier in the equation, has similarly been previously reviewed and approved by the NRC for other facilities.
The proposed change does not significantly increase the probability or consequences of an existing
- accident, create a
new accident, or significantly'educe the margin of. safety.
Chan e in Core Protection Limits (OTAT and OPBT)
Westinghouse has updated the correlations for the core protection limits using current methodology.
The new limits were incorporated into their analysis of Design Basis Events.
The results of these analyses show that the acceptance criteria were satisfied.
Therefore, the consequences of accidents previously evaluated are not increased.
The possibility of' new accident being created is not created by a change in core protection limits.
7590-01 Westinghouse has incorporated these new protection limits into their transient analysis and the results satisfy the acceptance criteria.
The purpose of these limits are unchanged, i.e., to protect against fuel
-centerline melting and Departure from Nucleate Boiling.
This change does not involve a significant reduction in the margin of safety.
A Deletion in the Limits of Target Axial Offset (TAO Westinghouse assumes that the worst case power distributions for their transient analysis that they predict would bound any that would occur during operation.
For every reload design, Westinghouse assures that the potential worst case power distribution for that reload does not exceed those assumed in the transient analysis.
Therefore, a limitation on TAO is unnecessary.
I Based on the above aforementioned, this change has no effect on either the probability or consequences of an accident.
This change does not create the possibility of a different kind of accident.
Since Westinghouse evaluates the worst case power distribution for every reload to assure it is bounded by those.
assumed in the transient
- analysis, and the transient analysis results show that the acceptance criteria is met, there is no significant reduction in the margin of safety.
For the foregoing reasons the Commission's staff proposes to determine that operating 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.
Therefore, the staff proposes to determine that the proposed license amendment would involve no significant hazards considerations.
7590-01 l
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.
The Commission will not normally make a final determination unless it receives a request for a hearing.
Comments should be addressed to the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, llashington, D.C; 20555, Attn:
Docketing and Service Branch.
By April 20, 1984, the licensee may file a request for a hearing:with respect to issuance of the amendment to the subject provisional operating
'icense 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 petition for leave to intervene.
Request f'r a hearing and petitions 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. 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 52.714, a petition for leave to intervene shall set forth with'articularity 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 >vith particular reference to the following factors:
( I) the nature
7590-01 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 qffect ot any order which
'may be entered in the proceeding on the petitioner's interest.
The petition should also identify the specific aspects 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 fifteen (15) days prior to the first prehearing conference scheduled in the proceed-ing, but such an amended petition must satisfy the specificity requirements described above.
'Not later than fifteen (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'o be litigated in the matter, and the bases for each contention set forth with reasonable specificity.
Contentions shall be limited to matters within the scope of the amendment under consideration.
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.
0 7590-01 If a hearing is requested, the Commission will make a final determination on the issue of.no significant hazards consideration.
The final determination W
will serve to decide when the hearing is held.
4 If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it 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 involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.
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 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.
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, D.C. 20555, Attention:
Docketing and Service
- Branch, or may be delivered to the Commission's Public Document
- Room, 1717 H Street, N.W. Washington, D.C., by the above date.
Where petitions are filed during k
7590-01
~
~
~
~ the last ten (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 (800) 325-6000 [in Missouri (800) 342-6700j.
The Western
.Union operator should be given Datagram Identification Number 3737 and the following message addressed to Dennis M. Crutchfield:
petitioner's name and telephone number; date petition was mailed; plant name; and publication date
'I 4
and page number of this FEDERAL REGISTER notice.
A copy of the petition should also be sent to the Executive Legal Director, U.S Nuclear Regulatory Commission, Washington, D.C.
- 20555, and to Harry H. Yoigt, Esquire,
- LeBoeuf, Lamb, Leiby and MacRae, 1333 New Hampshire
- Avenue, N,W., Suite 1100, Washington, D.C.
20036, 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 Atomic Safety and Licensing Hoard designated to rule on the petition and/or request, that the petitioner has made a substantial showing of good cause for the granting of a late petition and/or request.
That determination will be base'd upon a balancing of the factors specified in 10 CFR 2.714(a)(l)(i)-(v) and 2;714(d).
7590-01 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, 1717 H Street, N.l<., llashington, D.C., and.at the Rochester Public Library, 115 South Avenue, Rochester, New York 14604.
Dated at Bethesda, Maryland, this 15 day of March 1984.
FOR THE NUCLEAR REGULATORY COMMISSION enny
- 1. Crutch 1eld, ief Operating Reactors Br ch 85 Division of Licensing
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- ,50 UNITEO STATES IMUC1 EAR R EGULATORY COMMISSION AT0MIcsAFETY AND LIcENSING 80ARD PANEL WASHINGTON. 0 C. 20555 a/4 REQUEST FOR REPORTING SERVIC NORK ORDER NO. LB- )25
CANCELLED -
)IAY 2;",
)224 NAN AND TYPE OF PROCEEDING:(SECOND PREHEARINB CONFERENCE)
DOCyET NO(S). 50-2OII-OIA ROCHES1ER GAS 2 ELECTRIC CORPORATION (R,E, GINNA NUCLEAR STATION, UNIT ¹ ).)
LOCATION OF HEARING:
U.S.
CouRTHousE, CouRTRooM Iot 2 100 STATE ST, ROCHESTER, N,Y, 14614 PERSON TO CONTACT AT HEARING LOCATION: F ~ LE FTS (?16)
PHONE NO.:
263-6263
~DATE S 13 JUNE 1984 (WED.)
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LOCATION OF HEARING:
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CouRTHousE, CouRTRooM 0 2 100 STATE ST,>
ROCHESTER>
N,Y, 14614 PERSON TO CONTACT AT HEARING LOCATION: "R~ LE
~DATE S 13 JuNE 1984 (khD.)
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,'iD~ COMBO
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WHERE Hearing Site Headquarters Other (This section,'f'r ASLBP use only)
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SECURITY PERSON:
NR, WAYNE BERNSIDE, 286-SSr FTS/301-027-0330,.
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+0 UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMICSAFETY AND LICENSING BOARD PANEL WASHINGTON, D.C. 20555 REQUEST FOR REPORTING SERVICE WORK ORDER NO. LB-I25 NAME AND TYPE OF PROCEEDING:(SECOND PREHEARING CONFERENCE)
DOCKET NO(S). 50-2(g(~LA ROCHESTER ClAS B ELECTRIC CORPORATION (R E GINNA NUCLEAR STATION> UNIT I) 1)
LOCATION OF HEARING:
U,S, COURTHOUSES COURTROOM ¹ 2 100 STATE ST, ROCHESTER N, Y, 14614 PERSON TO CONTACT AT HEARING LOCATION FRED LESTER FTS (716)
PHONE NOaI 2Q-62@
~DATE 5 33 JUNE 1984 (RD.)
TIME S From-To OF HEARING 09:30 A M 04'%
M " ("MAY BE DISMISSED EARLIER )
ASLBP DOCKET CLERK:
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e >ver orsg1na to et es a, ary an:
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Hearing Site Headquarters Other BOARD STAFF(ELD OTHERS III WHERE RIG C
Sa kjILDERp BETHESDA> ND BY UE a
(This section for ASLBP use only)
So cial Instructions for Distribution P
SIGNATURE OF PREPARER DATE OF ORAL REQUEST TO REPORTING CO.
DATE OF WRITTEN CONFIRMATION BY:
DATE AND PERSON CONTACTED 18 APRIL 1984 (OVER)
WORK ORDER NO.
125 DATE:
18 APRIL 19%
LODGING ARRANGEMENTS FOR BOARD AND TELEPHOl'lE NUMBER (Include;ng Area Code):
OTHER SPECIAL INSTRUCTIONS:
SECURITY INFORMATION:
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I SECURITY PERSON:
WAYNE BERNS IDE> 286-SSr FTS/501-427-LI330,