ML18012A551
| ML18012A551 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/07/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18012A550 | List: |
| References | |
| NUDOCS 9703130382 | |
| Download: ML18012A551 (9) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMM(SSION WASHINGTON, D.C. 20555-0001 S
FETY EVA UATION BY THE OFFIC OF NUCLEAR REACTOR REGULATION CAROLINA POWER 8( LIGHT CON A SHEA ON HARRIS NUCLEAR OWER PLA T UNI 1
DOC E
- 0. 50-400
- 1. 0 INTRODUCTION By letter dated December 30, 1996, the Carolina Power
& Light Company (the licensee) submitted a request for changes to the Shearon Harris Nuclear Power Plant, Unit 1
(SHNPP), Technical Specifications (TS).
The requested changes would revise (1) chemistry data shown on TS Figures 3.4-2 and 3.4-3 for TS 3/4.4.9, "Pressure/Temperature Limits" and (2) its associated Bases.
In the December 30, 1996, submittal, the licensee submitted slight changes to some of the best estimate chemistry values for the beltline materials.
These changes resulted from cooperative data sharing activities in response to Generic Letter (GL) 92-01, Revision 1, Supplement 1.
Specifically, the licensee reviewed available reactor pressure vessel (RPV) beltline material
- data, and identified plants that have the same weld heat identifications as those contained in the Harris vessel (sister plants).
In addition to sharing data with sister plants, the licensee also shared data with the Westinghouse Owners Group (WOG) in order to determine best estimate chemistry for the beltline materials.
As a result of these assessments, the nickel content of the limiting material increased from 0.45X to 0.46X.
This slight change in nickel content does not affect the pressure-temperature (P-T) limits in the SHNPP TS.
- However, some documentation changes in TS Figures 3.4-2 and 3.4-3 are necessary.
The amendment request also revises the associated Bases to
'eflect chemistry revisions and minor material property changes for the beltline weld fabricated from weld wire heat 5P6771.
In addition, the licensee submitted revisions to the Bases in order to comply with 10 CFR 50, Appendix G rule changes.
- 2. 0 EVALUATION The staff evaluates the P-T limits based on the following NRC regulations and guidance:
Appendix G to 10 CFR Part 50; GL 88-11 and 92-01; Regulatory Guide (RG) 1.99, Rev. 2; and Standard Review Plan (SRP) Section 5.3.2.
Appendix G
to 10 CFR Part 50 requires that P-T limits for the reactor vessel must be at least as conservative as those obtained by Appendix G to Section XI of the American Society of Mechanical Engineers (ASHE) Boiler and Pressure Vessel 9703i30382 970307 PDR ADOCK 05000400 PDR
Code.
GL 88-11 requires that licensees use the methods in RG 1.99, Rev.
2, to predict the effect of neutron irradiation on the adjusted reference temperature (ART) of reactor vessel materials.
The ART is defined as the sum of initial nil-ductility transition reference temperature (RT>>,) of the
- material, the increase in RT>> caused by neutron irradiation, and a margin to account for uncertainties in the prediction method.
The increase in RT>>~ is calculated from the product af a chemistry factor (CF) and a fluence factor.
The CF may. be calculated using credible surveillance data,.obtained by the licensee's surveillance
- program, as directed by Position 2 of RG 1.99, Rev.
2.
If credible surveillance data are not available, the CF is calculated dependent upon the amount of copper and nickel in the vessel material as specified in Table 1 of RG 1.99, Rev.
2.
GL 92-01 requires licensees to submit reactor vessel materials data, which the staff uses in the review of the P-T limits submittals.
SRP 5.3.2 provides guidance on calculation of the P-T limits using linear elastic fracture mechanics methodology specified "in Appendix G to Section III of the ASIDE Code.
The linear elastic fracture mechanics methodology postulates sharp surface defects that are normal to the direction of maximum stress and have a depth of one-fourth of the reactor vessel beltline thickness (1/4T) and a length of 1-1/2 times the beltline thickness.
The critical locations in the vessel for this methodology are the 1/4T and 3/4T locations, which correspond to the maximum depth'f the postulated inside surface and outside surface defects, respectively.
The limiting material in the Shearon Harris reactor vessel is the intermediate shell plate that was fabricated using heat number A9153-1.
, The licensee determined that the slight increase in the nickel content of the limiting material from 0.45X to 0.46X did not affect the P-T limits.
The staff performed independent calculations of the CF values for the beltline materials using the revised best estimate chemistry and the methodology in RG 1.99, Revision 2.
Based on-these calculations, the staff verified that the CF of the limiting material remained unchanged.
Since the revised best estimate chemistry does not affect the limiting material, it will not affect the P-T limits.
Therefore, Amendment 38 dated August 20, 1993, which approved the P-T limits curves to ll effective full power years (EFPY), remains valid.
Some beltline materials experienced a slight reduction in copper and nickel content which resulted in a CF that remained the same or could have been reduced.
For those CF values that could have been
- reduced, the licensee elected to maintain the current conservative CF value.
Assessment of the newly acquired data resulted in the determination of a revised dropweight temperature (T,) for the weld fabricated from weld wire heat SP6771.
The original T>>, viue was -20'F, and the revised value is
-80'F as determined from the unirradiated surveillance weldment test data.
However, the revised T>>, value does not affect the initial reference temperature since it is based on the surveillance weldment temperature for the Charpy 50 ft-lb value minus 60 F.
The current initial reference temperature
value of -20'F remains valid, and the weld is not the limiting reactor vessel beltline material.
Therefore, the P-T limits are not affected.
Based on the above, the staff finds the proposed revisions to the applicable Bases sections for In-Service Leak
& Hydrotests (ISLH), and the ASME Code Section to be used for the development of P-T limits acceptable since the revisions are consistent with the amended rule for 10 CFR 50, Appendix G, that was effective January 18, 1996.
The staff also notes that the revisions to the best estimate chemistry do not affect any significant results relative to the withdrawal schedule or testing of any surveillance capsules.
E.
~TATE A
EE ATT In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRON ENTAL CONSIDERATION The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 4342).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
- 5. 0 CONCLUSION The Commission has concluded, based on the considerations discussed
- above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
E.E
~AE TATIIEEE Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temperature Limits
~8
V 3.
Code of Federal Regulat':ons, Title 10, Part 50, Appendix G, Fracture Toughness Requirements 5.
6.
Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations; July 12, 1988 ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components December 30, 1996, Letter from W.
R. Robinson to USNRC Document Control Desk,
Subject:
Shearon Harris Nuclear Power Plant Request for License Amendment RCS Pressure/Temperature Limits 7.
. August 20, 1993, Letter from N ~
Le to W.
ST Orser,
Subject:
Issuance of Amendment No. 38 To Facility Operating License No. NPF-63 Regarding Technical Specifications Change to Pressure-Temperature Limits-Shearon Harris Nuclear Power Plant, Unit 1
(TAC NO. M85876) 8.
November 16, 1995, Letter from J.
W. Donahue for W.
R. Robinson to USNRC Document Control Desk,
Subject:
Shearon Harris Nuclear Power Plant
Response
to Generic Letter 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity Principal Contributor:
Andrea Lee Date March 7, 1997
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