ML20206J189
| ML20206J189 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/17/1986 |
| From: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20206J177 | List: |
| References | |
| 3454F, NUDOCS 8606270030 | |
| Download: ML20206J189 (6) | |
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GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 Technical Specification Change Request No.149 Docket No. 50-219 Applicant submits, by this Technical Specification Change Request No.149 to the Oyster Creek Nuclear Generating Station Technical Specifications, a change to Specifications 2.3.A.1, 2.3.B, and 3.10.C as well as Figures 3.10-4 and 3.10-5.
l By:
I, J. ? d< -. [A.
e Peter B. F1
'er i.e President a frector Oyster Creek Sworn and subscribed to before me this / 7M day of d ; a,
1986.
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NOTARY PUBLIC 0F NEWJERSEY 8606270030 860617 DR ADOCK 0500 9
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF
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DOCKET N0. 50-219 GPU NUCLEAR CORPORATION )
CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No.149 for the Oyster Creek Nuclear Generating Station Technical Specifications, filed with the United States Nuclear Regulatory Commission on June 17
, 1986, has this day of June 17
, 1986, been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail, addressed as follows:
The Honorable Christopher Connors Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 By:
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Peter B. Fi er Vfc ! President an irector l
Oyster Cre DATED:
June 17, 1986
GPU Nuclear Corporation fh g
gf Post Office Box 388 Route 9 South Forked River.New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
June 17, 1986 The Honorable Christopher Connors Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731
Dear Mayor Connors:
Enclosed herewith is one copy of the Technical Specification Change Request No.149 for the Oyster Creek Nuclear Generating Station Operating License.
This document was filed with the United States Nuclear Regulatory Commission on June 17
,1986.
Very truly yours, k-U V8 4
terB.Fiedle[]
ce President Md Director yster Creek PBF: dam Enclosure i
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0YSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET N0. 50-219 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 149 Applicant hereby requests the Commission to change Appendix A of the license.
Also, pursuant to 10 CFR 50.91, an analysis concerning significant hazards considerations is presented:
1.
Sections to be Changed 2.3. A.1, 2.3.B and 3.10.C as well as Figures 3.10-4 and 3.10-5.
2.
Extent of Change Modify the above sections to accommodate the Cycle 11 Reload.
Specifically, the APRM Scram and Rod Block lines as well as the MCPR and MAPLHGR limits will be changed.
3.
Changes Requested As indicated in the attached revised Technical Specification page 2.3-1, 2.3-2, 2.3-6, 3.10-3, 3.10-7, 3.10-12 and 3.10-13.
4.
Discussion The Cycle 10 core for Oyster Creek consisted of 560 fuel assemblies, the composition of which includad Exxon Type VB assemblies, GE Type P80RB239 assemblies and GE Type P80RB265H assemblies. The Cycle 11 core will consist of Exxon Type VB assemblies, GE Type P80RB239 assemblies, GE Type P8DRB265H assemblies, GE Type P80RB299ZA assemblies and GE Type P8DRB299Z assemblies.
Amendment No. 75 to the Provisional Operating License authorized changes to the Technical Specifications to allow operation of Oyster Creek with the Cycle 10 reload. The NRC's safety evaluation for Amendment No. 75 was based on the staff's comparison of NED0-24195 " General Electric Reload Fuel Application for Oyster Creek" with NE00-24011 " General Electric Boiling Water Reactor Generic Relcad Fuel Application" which has been reviewed and approved by the NRC for reference in the safety analysis of other boiling water reactors. The staff concluded that the procedures employed in the reload design and analysis are essentially the same as those described in the previously approved NED0-240ll and are acceptable.
In addition, the procedures used to establish operating limits are similar to those previously approved and are acceptable.
The safety analyses perfonned in support of the Cycle 11 core design use the methods described by NED0-24195. The attached Appendix D to NED0-24195 is a sununary of results of the Cycle 11 reload core design and safety analysis.
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With respect to the MCPR limit calculated by General Electric (GE), the Rod Withdrawal Error transient (RWE) was the most limiting for Cycle 11 with a maximum Delta-CPR of 0.36.
This represents an increase from the Cycle 10 value which was 0.33.
The main reason for the increase is the fact that the RWE was performed with a Rod Block setpoint of 108% for Cycle 11 while the original analysis for Cycle 10 used the current Rod Block setpoint of 106%. A MCPR value of 1.45 was conservatively chosen.
4 This will not create any operating difficulties since it is expected that Cycle 11 will operate with an MCPR margin of 20% or greater.
j With respect to the APRM Scram and Rod Block lines, the Scram and Rod j
Block lines were modified to allow greater flexibility during startup and power escalation to rated conditions.
4 With respect to the GE calculated MAPLHGR limits, the LOCA/ECCS system responses and heat transfer coefficients remained unchanged from the previous reload licensing evaluations. The only differences between this and the previous analysis were in the Cycle 11 bundle nuclear characteristics. Operating the Cycle 11 bundles within the MAPLHGR limits insure that the peak cladding temperatures will not exceed 2200*F during a LOCA situation.
The proposed Technical Specifications have been based on the results of these analyses and thereby provide reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner.
5.
Determination GPU Nuclear has determined that operation of the Oyster Creek Nuclear Generating Station in accordance with the proposed Technical Specifications does not involve a significant hazard. The change does not:
I 1)
Involve a significant increase in the probability or the consequence of an accident previously evaluated. Methods and procedures which are acceptable to both GPU Nuclear and NRC have been used to perform the design and analysis of the Cycle 11 core. The review has considered tGe fuel mechanical design, nuclear design, thermal-hydraulic design 4
and the transient and accident analysis.
In addition, the Technical Specification limits chosen will ensure that the plant is operated within the boundary prescribed by the analysis.
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- 2) Create the possibility of a new or different kind of accident from any previously evaluated. The alteration of the core is an operational necessity that has been considered in the design of the plant and which is evaluated via methods and procedures that have been agreed upon by GPU Nuclear and the NRC. The resultant Technical Specifications ensure plant operation within the prescribed bounds.
3)
Involve a significant reduction in a margin of safety. The proposed i
Technical Specifications are based on analysis results which were performed in accordance with methods and procedures that have been previously deemed appropriate by GPU Nuclear and NRC. Thus, ensuring that there is not a significant reduction in the margin of safety.
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