ML20086T699
| ML20086T699 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/27/1995 |
| From: | Hagan J Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20086T701 | List: |
| References | |
| LCR-95-07, LCR-95-7, LR-N95102, NUDOCS 9508030127 | |
| Download: ML20086T699 (9) | |
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Pubhc Service Dectnc and Gas Cornpany Joseph J. Hagan Public Service Electnc and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339 1200 JUL 2 71995
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LR-N95102 LCR 95-07 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 Gentlemen:
REQUEST FOR LICENSE AMENDMENT HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 In accordance with the requirements of 10CFR50.90, Public Service Electric and Gas Company (PSE&G) hereby submits a request for amendment of Facility Operating License NPF-57 for Hope Creek Generating Station.
The License Amendment proposes to revise the Hope Creek Generating Station Technical Specifications to incorporate updated pressure vs. temperature operating limit curves contained in Technical Specification Figure 3.4.6.1-1.
The changes are a result of data obtained from the first set of specimen capsules removed during Refueling Outage 5.
The report supporting these changes was previously submitted to the NRC in Letter LR-N95071 dated June 9, 1995.
In addition, the License Amendment proposes to revise Technical Specification Surveillance Requirement 4.4.6.1.3 based on implementation of Regulatory Guide 1.99, Rev. 2 in accordance with Generic Letter 88-11.
A description of the requested amendment, supporting information and analysis of the change, and the basis for a no significant hazards consideration are provided in Attachment 1.
The Technical Specification pages affected by the proposed change are f
marked-up in Attachment 2.
Pursuant to the requirements of 10CFR50.91(b) (1), PSE&G has provided a copy of this amendment request to the State of New Jersey.
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vs 9500030127 950727 fDR ADOCK 05000354 PDR U
Document Control Desk 2
JUL 2 71995 IR-N95102 Upon NRC approval of the proposed change, PSE&G requests that the amendment be made effective upon issuance, but will be implemented within 60 days to provide sufficient time for associated administrative activities.
Please contact us should you have any questions regarding this submittal.
Sincerely, Attachments (2)
Mr. T. T. Martin, Administrator - Region I U.
S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr.
D. H. Moran Licensing Project Manager - Hope Creek U.
S.
Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. R. Summers (SO9)
Senior Resident Inspector Mr. K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 l
95 4933 t
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9 REF:
LR-N95102 STATE OF NEW JERSEY
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SS.
COUNTY OF SALEM
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J. Hagan, being duly sworn according to law deposes and says:
I am Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Hope Creek Generating Station, are true to the best of my knowledge, information and belief.
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Subscribed and Sworn to before me this ki day of [uEMn, 1995 dD A r
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Ndtary Public of New Jersey My Commission expires on 17 N 95
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i LR-N95102 LCR 95-07 i
t ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION AND BASES CHANGE REQUEST FOR LICENSE AMENDMENT l
HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE-NO. NPF-57 DOCKET NO. 50-354 1
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LR-N95102 LCR 95-07 1
I.
Descriotion of the Chance I
This amendment request revises the Reactor Coolant System pressure vs. temperature operating limit curves contained in Technical Specification Figure 3.4.6.1-1.
The amendment proposes
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to add threa new pressure vs temperature curves for (1) 3 hydrostati: or leak testing, (2) heatup by non-nuclear means, cooldown following a nuclear shutdown and low power physics tests, and (3) operations with a critical core other than low power physics tests to replace Figure 3.4.6.1-1.
i This amendment request also revises Surveillance Requirement
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4.4.6.1.3 criteria for recalculating pressure and temperature limits based on the implementation of Regulatory Guide 1.99, Rev.
2,
" Radiation Embrittlement of Reactor Pressure Vessel materials".
The change proposes to delete specific reference to i
the plate material heat number and weld material heat number for determining the actual shift in reference temperature, and delete reference to the plate material used for predicting the shift in reference temperature calculated in accordance with Regulatory Guide 1.99, Rev.
1.
The associated Technical Specification Bases is also revised to include material data for the most limiting plate and beltline weld material as determined by Regulatory Guide 1.99, Rev.
2.
II.
Reason for the Proposed Chance The updated pressure-temperature curves contained in Technical Specification Figure 3.4.6.1-1 reflect Reactor Vessel material changes identified from analysis of the first set of surveillance capsule specimens removed from Hope Creek Generating Station during Refueling Outage 5.
The pressure-temperature curves contain the predicted shift in reference temperature determined in accordance with Regulatory Guide 1.99, Rev. 2 for the most limiting beltline plate material 5K3025-1.
The specimen capsule surveillance report was submitted to the NRC by Letter LR-N95071 dated June 9, 1995.
Since the most limiting materials were not included in the surveillance capsules based on Regulatory Guide 1.99, Rev.
1, the NRC required that PSE&G recalculate the pressure-temperature limits based on the greater of the following (See Hope Creek Safety Evaluation Report (NUREG-1048):
- 1) The actual shift in reference temperature for plate material from heat SK3238-1 and weld metal 510-01205 as determined by Charpy Impact test, or
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LR-N95102 LCR 95-07
- 2) The predicted shift in reference temperature for plate material from heat SK3025-1 as determined by Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Material" (Rev 1).
This requirement for utilizing the greater of the actual or predicted shift in referenced temperature to update the pressure-temperature limits was included in Surveillance Requirement 4.4.6.1.3.
In response to NRC Generic Letter 88-11, PSE&G committed to use the methods contained in Regulatory Guide 1.99, Rev.
2.
The revised methodology contained in Regulatory Guide 1.99, Rev. 2 resulted in changes to what material would be used (i.e., nickel in lieu of phosphorus), and how the Adjusted (RT Regulatory Guide 2 provides gui,)is calculated.
Reference Temperature dance for determining the actual or 1.99, Rev.
i predicted shift in reference temperature to be utilized when updating the pressure-temperature curves.
The actual or predicted reference temperature shift for the most limiting beltline plate or weld material that results in the maximum End of Life (EOL) RT,7 are included in the pressure / temperature curves.
Therefore, PSE&G proposes to delete reference to the plate and weld material contained in TS 4.4.6.1.3 based on the commitment to Reg. Guide 1.99, Rev. 2 made in response to GL 88-11.
The Bases has also been changed to reflect the most limiting plate and weld material fracture toughness data as currently calculated in accordance with Regulatory Guide 1.99, Rev.
2, and to reflect the results of the analysis of the specimen capsule removed from Hope Creek during Refueling Outage # 5.
III. Justification for the Proposed Chance 10CFR50, Appendix H requires a surveillance program to monitor changes in fracture toughness properties of Reactor Vessel materials due to irradiation.
Surveillance capsules are periodically removed and analyzed to determine the irradiation effects.
Based on the removal of the first set of specimen capsules, a change is required to the pressure-temperature curves contained in Technical Specification Figure 3.4.6.1-1.
These curves define the pressure and temperature limits to assure the prevention of vessel failures.
The revised curves reflect the analysis results of the specimens removed from the Hope Creek vessel which were forwarded to the NRC by Letter LR-N95071 dated June 9, 1995, and reflect the most limiting shift in reference temperature determined in accordance with Regulatory Guide 1.99, Rev.
2.
1
NLR-N95102 LCR 95-07 Regulatory Guide 1.99, Rev. 2 resulted in changes to what limiting materials would be used (i.e., nickel in lieu of phosphorus) and how the reference temperature shift and Adjusted Reference Temperature are calculated for Reactor Vessel beltline and weld materials.
This changed the most limiting beltline weld material used in determining the pressure-temperature limits to heat D53040/1125-02205.
This weld material is contained in the surveillance capsules.
The most limiting beltline plate material remained 5K3025-1 which is not in the specimen capsule.
The actual or predicted reference temperature shift for the most limiting vessel material is 53.5'F.
The reference temperature shift is predicted based on the most limiting beltline plate 5K3025-1 that has a maximum End of Life (EOL) adjusted reference temperature (RT,)
of 72.5'F.
The revised pressure-temperature curves include the predicted shift in reference temperature that was calculated for beltline plate 5K3025-1.
Therefore, PSE&G proposes to delete reference to the plate and weld material contained in TS 4.4.6.1.3 based on the commitment to Reg. Guide 1.99, Rev.
2.
The Bases has also been changed to reficct the fracture toughness data for the most limiting plate 5K3025-1 and beltline weld D53040/1125-02205 material as contained in the specimen capsule surveillance report that was submitted to the NRC by Letter LR-N95071 dated June 9, 1995.
IV.
Sionificant Hazards Consideration PSE&G has, pursuant to 10CFR50.92, reviewed the proposed changes to determine whether our request involves a significant hazards consideration.
We have determined that operation of Hope Creek Generating Station in accordance with the proposed change:
- 1. Will not involve a significant increase in the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated.
The proposed changes assure that the existing safety limits are not exceeded due to changing Reactor Vessel conditions.
These changes reflect the latest material testing results in accordance with 10CFR50, Appendix G.
The proposed changes to the pressure and temperature limits do not increase the probability of nonductile failures.
The proposed changes to the surveillance requirement and the associated changes to the Bases to include a commitment to the methodology of Regulatory Guide 1.99, Rev. 2 ensures that the most limiting Reactor Vessel material is used in the determination of the pressure-temperature operating limits.
LR-N95102 LCR 95-07 Therefore, it may be concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated.
- 2. Will not create the possibility of a new or different kind of accident from any previously evaluated.
No physical plant modifications or new operating configurations result from these changes.
These changes do not adversely affect the design or operation of any system or component important to safety, rather they establish limits to assure that operations remain within acceptable safety boundaries.
Therefore, the possibility of a new or different kind of accident from any previously evaluated will not be created.
- 3. Will not involve a significant reduction in a margin of safety.
Analysis of the capsule specimens has concluded that the Reactor Vessel has sufficient fracture toughness for continued safe operation, provided that operation remains within acceptable i
pressure-temperature limits.
The revised pressure-temperature curves define these acceptable pressure-temperature limits during plant operation.
The proposed changes maintain the existing margins of safety by modifying the operating limits based on the i
most limiting of the actual reference temperature shifts.
This new limit considered analytical results of the capsule specimens, or a predicted shift considering the most limiting pressure vessel material.
Changes to the Surveillance Requirement criteria and the associated Bases to include a commitment to the l
methodology contained in Regulatory Guide 1.99, Rev. 2 will ensure that the most limiting plate or beltline weld material will be utilized in the determination of the pressure-temperature limits.
Therefore, it is concluded that the proposed changes will not involve a significant reduction in a margin of safety.
Conclusion Based upon the above, we have determined that the proposed amendment does not involve a Significant Hazards Consideration.
e LR-N95102 LCR 95-07 ATTACHMENT 2 INSERTS AND MARKED-UP PAGES REQUEST FOR LICENSE AMENDMENT HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 i
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