ML20116P169
| ML20116P169 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/21/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20116P167 | List: |
| References | |
| NUDOCS 9608230170 | |
| Download: ML20116P169 (6) | |
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i pm00 9 y-t UNITED STATES s
j NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20066-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.127 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS. INC.. ET AL.
GRAND GULF NUCLEAR STATION. UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
i By letters dated May 2 and 31, 1996, Entergy Operations, Inc. (the licensee) submitted a surveillance specimen program evaluation with a revised capsule withdrawal schedule for NRC review and approval.
In the letter of M;y 2, 1996, the licensee also submitted a technical report from the General Electric Company and stated that the report provided the details for the conditions that exist at Grand Gulf Nuclear Station, Unit 1, (GGNS) and the basis for the proposed revised withdrawal schedule for GGNS (Ref.1).
The capsules in the reactor vessel contain reactor vessel material specimens to be irradiated by the neutron fluence during plant power operations. A capsule is periodically removed from the reactor vessel in accordance with Section III.B.3 of Appendix H to 10 CFR Part 50 (Ref. 2), and the material specimens are tested to provide an indication of the effect of the neutron radiation on the material properties of the reactor vessel. The schedule for withdrawing the capsules must be approved by the Nuclear Regulatory Commission.
The first capsule at GGNS was withdrawn at 8 effective full power years (EFPY) during refueling outage (RFO) 7 on May 7, 1995. The revised withdrawal schedule proposed by the licensee would allow the first capsule to be returned to the vessel during RF0 8, scheduled to begin in October 1996, and retained as a standby capsule. The first capsule specimens have not been tested. As proposed by the licensee, the first capsule for testing would be either capsule number 2 or 3, and would be removed at 24 EFPY.
2.0 EVALUATION Appendix H endorses American Society for Testing and Materials (ASTM) standard E185, " Surveillance Tests for Nuclear Reactor Vessels" and states that "The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased.
Later Editions of ASTM E185 may be used, but including only those editions through 1982."
9608230170 960821 PDR ADOCK 05000416 P
. The number of GGNS specimen holders was determined per ASTM E185-73 (Ref. 3).
The three specimen holders were designed, built and analyzed to Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), 1971 Edition, with Addenda through Winter 1972.
Each holder has 12 Charpy V-notch (CVN) specimens of the weld, base metal and heat-affected zone (HAZ) for a total of 36 specimens. A set of unirradiated specimens are kept, as well as archive material, for additional testing in order to provide baseline information.
1 GGNS is defined as an ASTM E185-73 Case "A" plant since the vessel has a shift j
in the reference null-ductility temperature (ART will be exposed to a neutron fluence of less tha,n 5x10 aor) of less jhan 100*F and 1
n/cm over the design lifetime of the plant. The current testing schedule requires that the first specimen holder be removed at 8 EFPY, the second at 24 EFPY, and the testing and reporting is to be performed in accordance with the more recent ASTM E185-82 (Ref. 4).
If the ASTM E185-82 requirements were applied to determine the schedu wall fluence is 5x10(e the,first capsule should be withdrawn when the vessel n/cm, or when the ART,or reaches 50*F whichever is first. The GGNS vessel wall is unlikely to reach the conditions described above during the design lifetime of the plant, therefore, early capsule withdrawal is not critical for continued operation of the plant, j
In response to Generic Letter (GL) 92-01, Supplement 1 (Ref. 5), a study was performed by General Electric Company for the Boiling Water Reactor (BWR)
Vessel and Internals Project (VIP) on the copper levels present in BWR beltline materials (Ref. 6). The purpose was to verify plants with significant variation in the repcrted copper levels. GGNS was determined to be consistent with reported values with no significant variation in the reactor vessel material (e.g., 0.02-0.06% copper).
The licensee used Regulatory Guide (RG) 1.99, Revision (Rev.) 2 to calculate the shift RT and adjusted reference temperature (ART) values for all GGNg belt {ine mateNals. The fluence used to evaluate the 32 EFPY ART was 2.5 x10 n/cm. The resulting predicted values of RT shift indicate that the vessel will not experience a large shift over vess,o'l life. A comparison e
was made between calculated shift and fluence values and actual surveillance data from other BWR's in order to confirm the conservative predicted shift plus margin values that were used to modify the surveillance program schedule.
The results for BWR/6's, including GGNS, show a small shift for capsules removed at EFPY similar to Grand Gulf's current schedule and at higher fluence level s.
Based on the data, the measured shift for GGNS would be conserva-tively bound by the RG 1.99, Rev. 2 calculations.
The shift in RT that results from surveillance testing is used to determine thecrackarresE1 fracture toughness (K ).
pressure-temperature (P-T) limits curv,,s.
Th,, is used to calculate the K
e e current P-T limits curves for GGNS are calculated with the 10 EFPY shift in RT 'ted using the conservative The limiting condition isthepressuretest,andtheP-TcurveiscalcuYa.
lower bound static crack initiation fracture toughness (K ) (where K is ic ic
i approximately 2.4 times K Vessel fracture toughness is not a significant concern for GGNS over the,$)1fe of the vessel.
The BWR Owner's Group (BWROG) began a supplemental surveillance program (SSP) in the late 1980's that was designed to significantly increase the amount of BWR surveillance data in a systematic manner. The BWROG's reasons for beginning the program were the following:
There are a smaller number of capsules per plant and fewer BWRs than pressurized water reactors (PWRs)
There are not much BWR surveillance data at higher fluences, nor would there be for many years RG 1.99, Rev. 2 imposed some hardships on pressure testing for BWRs, some of which might be relieved if a better understanding of BWR embrittlement phenomenon were obtained.
Supplemental capsules were installed in Cooper and Oyster Creek, and specimen withdrawal is planned for 1996, 2000, and 2002. The results will be the equivalent of S4 additional surveillance capsules compared to approximately 25 which have been tested. The materials used were selected to bound the range of chemistries in BWR beltline materials, and in most cases are BWR beltline materials.
The GGNS surveillance plate and weld material, including the limiting material, are among the materials in the capsules. At least one of these materials is in each of the seven capsules in the SSP. Results will be developed which will provide information on all the GGNS plate and weld surveillance materials, and will be directly applicable to the GGNS collected between 5x10'ppecifically, the capsules, when tested, will have surveillance program.
n/c/ and 2x10'8 n/cm fluence, which bounds the end 2
of life (E0L) fluence for the GGNS vessel.
Since the expected shift is low, the first surveillance program testing should be at a time when the majority of the shift in the vessel RT,7 has been achieved.
Early testing may yield shift in RT 7 values that are not i
j distinguishable from the data scatter. Anomalous shift is not a major concern because, if it were to occur, the BWROG SSP will identify any greater than expected shift.
l The staff used'pG 1.99 calculational methods to verify that the 8 EFPY fluence value (2.2 x10 n/cm ) in combination with the low copper values (0.02-0.06%)
result in predicted values that are not distinguishable from thgdata, scatter.
The staff also verified that the 24 EFPY fluence value (6.9 x10 n/cm ) would result in predicted values that are more likely to be distinguishable from the data scatter.
The licensee determined the revised surveillance schedule by examining the fracture toughness decrease as a function of shift, and used that shift to determine the appropriate EFFY for removal and testing of the first capsule.
. Fracture toughness at the beginning of plant life was 200 ksiVin and at 32 EFPY is 80.1 ksiVin. Therefore, the change in fracture toughness over the design life of the plant is 119.9 ksiVin. A value of 75% of the predicted fracture toughness change (i.e., (0.75)(119.1) = 89.9 ksiVin over the design life) was selected as an appropriate criterion.
If a significant shift is to occur, this value is large enough to ensure its detectability. Therefore, the first surveillance capsule should be removed when the 75% criterion has been met, which is at 200-89.9 ksiVin - 110.1 ksiVin. This change in fracture toughness is expected to be achieved when the shift has reached a value of 28.5'F (from a plot of K versus Predicted Shift). The licensee then used the shift value of 28.5'E to determine that the capsule will experience a similar shift at about 24 EFPY (from a plot of Predicted Shift vs. EFPY).
Testing the first capsule at 24 EFPY would more likely yield shift in RT,
values that are distinguishable from the data scatter.
At this time, the licensee has not made a recommendation for removal of the second capsule. Additional data from the SSP capsules (using the GGNS limiting weld and plate materials) will soon be available. The combination of data from the first capsule and the SSP will be used to develop the appropriate schedule for the second capsule.
The staff compared the licensee's proposed withdrawal schedule to the recommended withdrawal schedule in ASTM E185-82, which is referenced in Appendix H to 10 CFR Part 50.
The staff concludes that the proposed withdrawal schedule satisfies ASTM E185-82 and, therefore, the schedule complies with Appendix H to 10 CFR Part 50.
Based on the above review of the licensee's surveillance specimen program evaluation, the staff concludes that the proposed extension of the withdrawal schedule is acceptable because of the following:
The CGNS reactor vessel has low copper and fluence values, Actual BWR/6 surveillance data has shown small shifts for capsules removed at EFPY similar to the current GGNS schedule, Vessel fracture toughness is not a significant concern for GGNS over the e
life of the vessel, and j
GGNS participates in the BWROG SSP which can provide detection of any e
unusual RT., shifts.
As a result of the low copper (0.02-0.06%) and 8 EFPY fluence (2.2x10'7 n/cm )
values, testing of the capsule that was removed at 8 EFPY may yield shift in RT, values that are not distinguishable from the data scatter.
The proposed withdrawal schedule satisfies ASTM E185-82 and, therefore, complies with Appendix H to 10 CFR Part 50 and is acceptable.
. l
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Mississippi State l
official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIR m l
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 and changes surveillance requirements. 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 (61 FR 31179). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 1
51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 QNCLUSION 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.
6.0 REFERENCES
1.
GE Report GE-NE-81301807-OlR1, " Surveillance Specimen Program Evaluation for Grand Gulf Nuclear, Station," April 1996 2.
Code of Federal Regulations, Title 10, Part 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements," December 1995 3.
American Society for Testing and Materials, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,"
ASTM E 185-73, 1973.
American Society for Testing and Materials, " Standard Recommended 4.
l Practice for Surveillance Tests for Nuclear Reactor Vessels,"
ASTM E 185-82, 1982.
5.
Generic Letter 92-01, Revision 1, Supplement 1 " Reactor Vessel Structural l
Integrity," May 19, 1995 i
i
. 6.
BWR Vessel and Internals Project, EPRI Report TR-105908, " Bounding Assessment of BWR/2-6. Reactor Pressure Vessel Integrity Issues (BWRVIP-08)," November 1995 Principal Contributor: Andrea Lee Date:
August 21, 1996
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