ML20138K024

Safety Evaluation Supporting Proposed Rev to RPV Surveillance Capsule Withdrawal Schedule

ML20138K024
Person / Time
Site:	Cooper
Issue date:	05/07/1997
From:	NRC (Affiliation Not Assigned)
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ML20138K021	List: ML20138K017 ML20138K024
References
NUDOCS 9705120013
Download: ML20138K024 (7)

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2 NUCLEAR RECULATORY COMMCalON WASHINGTON, D.C. 20066-0001 l.

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED REVISION TO THE RPV SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By letters dated December 13,- 1996, (Reference 1) and April 17, 1997, (Reference 2), the Nebraska Public Power District (the licensee) submitted for NRC review and approval a proposed change to the Cooper Nuclear Station (CNS) reactor vessel material surveillance program withdrawal schedule; a program required by Title 10 of the Code of Federal Regulations, Part 50 (10 CFR Part 50), Appendix H. In this request, the licensee proposed to defer the withdrawal time for-the third CNS surveillance capsule from 15 effective full-power years (EFPYs) of operation to 22 EFPYs of operation.

Title 10 of the Code of Federal Regulations, Part 50, Appendix H requires licensees to withdraw surveillance capsules periodically, according to the capsule withdrawal schedule in the American Society for Testing and Materials (ASTM) E185, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."Section III.B.I. of Appendix H further explains that, "...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."

Additional guidance on the processing of changes to a existing withdrawal schedule has recently been published in the Nuclear Regulatory Comission's Memorandum and Order, CLI-96-13 (Reference 3), in which the Comission reversed the decision of the Atomic Safety and Licensing Board (ASLB) in the matter of the Cleveland Electric Illuminating Company's proposed change to the Perry Nuclear Power Plant's withdrawal schedule. In CLI-96-13, the Comission states that, " Contrary to the assumption made by the Licensing Board, we do not find that all such approvals (i.e., withdrawal schedule changes) are d.g facto license amendments." The Comission affirmed that, "Only those actions falling 'beyond the ambit of the prescriptive authority granted under the license' necessitate a license amendment," and that, "The key consideration should be: did the agency action ' supplement' the existing operating authority

, prescribed in the license?" The staff understands these coments and the remainder of the Comission's Memorandum and Order to indicate that changes to a facility's withdrawal schedule which do not conflict with the requirements ',

in the facility's licensing basis (established to meet the conditions of 10 CFR Part 50, Appendix H) do not require a license amendment.

9705120013 970507 PDR P ADOCK 05000298 ppg

t In its April 17, 1997, letter, the licensee proposed that this change in the i CNS capsule withdrawal schedule does not constitute a change to the facility's l licensing basis and thus does not necessitate a license amendment. The  ;

licensee therefore concluded that it was only required to request the staff's review and approval of this proposed change.

2.0 EVALUATION )

The staff has evaluated the licensee's submittal to verify two aspects of the proposed change. The staff first considered the licensee's technical basis for deferring the withdrawal of the third capsule until 22 EFPY. Next, the staff determined whether a license amendment war or was not required in this case to support the requested change.

2.1 Technical Evaluation l The staff's initial consideration focussed on the technical justification for this change. The licensee noted in their December 13, 1996, submittal that the first two CNS capsules had been withdrawn from the vessel at fluences corresponding to 21% and 25% of the estimated peak end-of-license (E0L) fluence at the 1/4T location within the vessel wall. The peak 1/4T E0L fluence (i.e., the peak E0L fluence at a depth 144 of ghe way through the vessel wall) was given as approximately 1.1 x 10 n/cm. Withdrawal of the third capsule at the end of the current CNS operating cycle (after about 15 EFPY, as required under the current withdrawal schedule) would mean that the third capsule would receive a fluence equivalent to about 36% of the peak 1/4T EOL fluence. Tho licensee concluded that this would provide little additional information on the embrittlement of the CNS vessel. The licensee contended that if removal of the third capsule were delayed until 22 EFPY (an exposure equivalent to 50% of the peak 1/4T E0L fluence), the test results would be more meaningful for assessing the vessel's embrittlement behavior.

The staff agrees with the licensee's conclusion that it would be more technically appropriate to delay withdrawal of the third capsule until a fluence of 50% of the peak 1/4T E0L fluence. Since this data will be used to extrapolate the vessel's embrittlement trend out to E0L, data obtained from specimens irradiated to a higher fraction of the actual vessel E0L fluence should provide a more accurate basis for this extrapolation. The staff also expects that the inherent scatter in the test parameter (the temperature at which 30 ft-lbs is achieved in Charpy impact testing) used to assess the material's embrittlement will have a less significant influence on the trend projection if the data points are spaced further apart in fluence.

In agreeing with this technical basis for deferring capsule withdrawal, the staff also examined what impact this deferral might have on the licensee's ability to conservatively assess the level of embrittlement experienced by the CNS vessel. NRC Regulatory Guide 1.99, Revision 2 (RG 1.99, Rev. 2) provides methods acceptable to the staff for predicting vessel material embrittlement in the presence or absence of surveillance test data, and these methods have been used by the licensee in their assessments. In the absence of test data, RG 1.99, Rev. 2 provides a generic correlation which uses a chemistry factor

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! (CF) based upon the percent copper and percent nickel in a plate or weld to quantify the rate at which the a material embrittles (with a greater CF

! indicating a greater embrittlement rate). Alternatively, surveillance data may instead be used to experimentally determine the CF from the shifts in l Charpy 30 ft-lb temperature at two fluence levels. Obtaining a material- i specific CF is of particular importance if the surveillance data suggest that i a CF greater than the CF from the generic correlation is necessary to bound the material's observed embrittlement behavior.

The test results from the first two CNS capsules demonstrated an embrittlement rate greater than that predicted by the RG 1.99, Rev. 2 generic correlation for both the surveillance plate and the surveillance weld. The staff accepts that the results from the first two capsules already constitute a basis for making embrittlement projections based on the method given' in Position 2.1 of  ;

RG 1.99, Rev. 2. Therefore, the results from the third capsule are not 4

necessary to establish material-specific CFs. The licensee has also examined what effect these material-specific CFs have on the vessel pressure-i temperature (P-T) limits for normal reactor operations and vessel hydrostatic / leak rate testing, as required by 10 CFR Part 50, Appendix G. The

licensee has confirmed (Reference 4) that the P-T limits currently in their Technical Specifications (TS) do provide a conservative bound for operation to

21 EFPY.

Finally, the staff questioned the licensee's use of the peak 1/4T E0L fluence as a reference point for their surveillance program. Reactor vessel

surveillance programs are required to assess trends in material embrittlement in order to ensure that vessel integrity is maintained during all operational conditions. For boiling water reactors like CNS, this entails the s establishment of appropriate P-T limits, as noted above. Since an analysis consistent with the requirements of 10 CFR Part 50, Appendix G assumes a flaw 4 extending to the 1/4T depth in the vessel wall, it is of particular importance to understand the embrittlement behavior of the material to fluence levels i similar to those at the vessel 1/4T location. Therefore, the staff finds that although many surveillance programs use the vessel inside diameter fluence as a reference, it is acceptable for the licensee to use the projected 1/4T l fluence as a reference for establishing its capsule withdrawal schedule.

2.2 Reaulatory Evaluation l

l The staff then assessed whether or not the proposed change to the licensee's withdrawal schedule required a license amendment. The licensee did not

address this question in their original submittal. The staff raised this ,

issue with the licensee during a teleconference on January 30, 1997, and the  !

licensee subsequently submitted additional information (Reference 2) to clarify their position. The licensee has concluded that no license amendment l

" was required to make the proposed change, since it was consistent with the l requirements of the edition of ASTM Standard E185 which was incorporated in '

the facility's license. In verifying this statement, the staff has determined that several regulatory actions modifying the requirements placed on the

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! Cooper reactor vessel surveillance program have occurred since the facility

was first licensed. However, based on the following research, the staff does concur with the licensee's conclusion.

) The CNS reactor vessel was purchased and constructed to the Winter 1966 Addenda of the 1965 ASME Code. This version of the ASME Code referenced the i 1966 edition of ASTM Standard E185 (E185-66) and 10 CFR Part 50, Appendix H

requires that, "the withdrawal schedule must meet the edition of ASTM E185

, that is current on the issue date of the ASME Code to which the reactor vessel 4

was purchased." However, the licensee then incorporated into the TS to the

CNS license a statement which indicated that the surveillance program, "shall L conform to ASTM E185-73 (1973 edition) to the degree possible" (Reference 5).

i The staff has further recognized that since the observed shifts from the first j two surveillance capsules have indicated that the projected shift for CNS i vessel materials at E0L is expected to be more than 100 *F, it would be l required that the licensee use the withdrawal schedule _in Case B of the 1 E185-73 standard. Each of these editions of ASTM E185 would have required the i licensee to maintain a surveillance program based on the removal of three j surveillance capsules from the CNS vessel.

l On July 6,1987, the licensee submitted the results from the testing of the

first CNS surveillance capsule (Reference 6). These results indicated a greater level of embrittlement in the surveillance material th. had been previously predicted. As a result, when the licensee applied to amend the i CNSP-T limit curves and surveillance program in 1987, the staff wrote in its

safety evaluation that it, "recomends that the schedule for withdrawal of the second capsule should be accelerated to 12 EFPY and that schedule for withdrawal of the third capsule should be based on the analysis of the second capsule" (Reference 7). The licensee comitted to this staff recomendation and to meeting the requirements of the 1982 edition of ASTM E185 (E185-82), to the extent practical (Reference 8). The licensee completed this action by removing, testing, and reconstituting the second surveillance capsule in 1991.

With the inclusion of the reconstituted capsule, the licensee now had a surveillance program based on four surveillance capsules and, to that extent, was consistent with E185-82. Since the date of withdrawal for the third CNS capsule was left indeterminate in both the staff's recomendation (Reference 7) and the licensee's comitment (Reference 8), it is the opinion of the staff that the licensee did not comit to meeting the exact withdrawal schedule published in the E185-82, which would recomend that the third capsule be removed at 15 EFPY. The staff further reinforced this point in its safety evaluation addressing the licensee's request to recapture the facility's construction period and to extend the plant's operating license to January 18, 2014, when it' noted, " reconstitution...of a fourth capsule...makes the licensee's surveillance program consistent with the requirements of ASTM E185-82....The withdrawal schedule for the reconstituted capsule and the original third capsule will be determined based upon the analysis of the

[second capsule]" (Reference'9).

Finally, in 1992, the licensee requested and received permission from the staff to remove the actual withdrawal schedule from the plant TS per the 1

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) guidance of Generic Letter 91-01 (Reference 10) and to maintain the schedule in the facility's Updated Safety Analysis Report (USAR). This line-item

i. improvement was in fact promulgated by the staff to alleviate the need for a '

j license amendment if a change in the surveillance schedule were necessary. It

should also be noted that in this submittal (Reference 10), the licensee

! supports the opinion expressed above that the licensee committed only to the -

! withdrawal of the second capsule at 12 EFPY and the reconstitution of that capsule, not to the use of the E185-82 withdrawal schedule.

! Therefore, the staff's position on the licensing basis for the CNS vessel j surveillance program is as follows. To meet the requirements of 10 CFR a Part 50, Appendix H, the licensee's withdrawal schedule must comply with

.E185-66 and the program's test procedures and reporting requirements must comply with E185-82; the withdrawal schedule proposed by the licensee meets i these requirements. However, since the licensee has incorporated E185-73 into

~t he TS of the facility's license, it is effectively this edition of the standard to which the licensee must comply "to the degree possible" in order i to fulfill the facility's license requirements and meet the requirements of I 10 CFR Part 50, Appendix H. The licensee's commitment with regard to E185-82 i

and the reconstituted capsule is only a commitment to have a fourth capsule in the surveillance program, not to apply that capsule within the verbatim withdrawal schedule in E185-82.

s j The staff has also provided the following evaluation to summarize how each

CNS surveillance capsule is used to meet the requirements of the CNS

surveillance program. Again, the staff has evaluated the program against the withdrawal schedule requirements of E185-73 Case B, since the results from the j first two CNS capsules indicated that the vessel materials would be eypected to experience shifts of greater than 100 *F at E0L. The first capsule 1 withdrawn from the CNS vessel meets the requirements of the first capsule in i the Case B schedule of the E185-73 standard. The second capsule which was i removed in 1991 after nearly the same level of. exposure as the first capsule j is considered to have been a confirmatory test of the results of the first

] capsule. These two capsules are considered by the staff as a basis for the application of Position 2.1 of RG 1.99, Rev. 2 for determining material embrittlement from surveillance data. The third CNS capsule should then be removed in accordance with the requirements given for the "second capsule" and -

the fourth CNS capsule in accordance with the requirements given for the i

" third capsule" in the Case B program of E185-73, to the degree possible. The I caveat "to the degree possible" was a significant inclusion in the TS since I the fourth capsule was installed at an azimuthal position. on the vessel wall where the neutron flux lags the peak vessel flux by approximately a factor of two (Reference 11). Thus, even though the licensee may be permitted to use the peak 1/4T E0L fluence as a reference value within the framework of E185-73, it is still physically impossible for the fourth CNS capsule to achieve an exposure corresponding to 100 to 125 percent of this value as'would  !

be required by E185-73. Likewise, the E185-73 requirements for a " fourth and fifth capsule" in standby are not currently met by the CNS program in the  !

licensee's attempt to comply "to the degree possible".  !

j-  ! The licensee's proposal to defer the removal of the third CNS capsule until i

22 EFPY. (an exposure of about 50% of the peak 1/4T E0L vessel fluence) is

, consistent with the staff's interpretation on E185-73 and the CNS licensing

, basis as denoted above. The staff must conclude that this change does not allow the licensee to take actions "beyond the ambit of the prescriptive i authority granted under the license" (Reference 3) and thus, per Commission Memorandum and Order CL-96-13, does not require a license amendment.

4 Furthermore, though removal of the fourth CNS capsule at 32 EFPY is not

completely consistent with the E185-73, the staff concludes that it is consistent with the CNS license given the "to the degree possible" caveat in

the CNS TS and would be consistent with the requirements of the standard under j which the vessel was purchased, E185-66.

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3.0 CONCLUSION

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Based on the technical and regulatory findings presented in Section 2.0 above, j the staff has determined that the surveillance capsule withdrawal schedule 4 proposed by the licensee meets the requirements of E185-66 and E185-73, to the degree possible. Therefore, the staff has determined that the program i proposed by the licensee is in compliance with the requirements of 10 CFR l Part 50, Appendix H and with the facility's licensing basis. Therefore, the

staff has approved the requested change in the withdrawal schedule without the

[ need for a license amendment. The licensee may defer removal of the third CNS

! surveillance capsule until 22 EFPYs of operation.

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4.0 REFERENCES

[1] Letter from P.D. Graham (NPPD) to USNRC Document Control Desk, Request for Revision of Reactor Vessel Surveillance Capsule Withdrawal Schedule, Cooper Nuclear Station, NRC Docket 50-298, DPR-46, December 13, 1996.

[2] Letter from P.D. Graham (NPPD) to USNRC Document Control Desk, Additional Information Supporting Request for Revision of Reactor Vessel Surveillance Capsule Withdrawal Schedule, Cooper Nuclear Station, NRC Docket 50-298, DPR-46, April 17, 1997.

[3] NRC Memorandum and Order, CLI-96-13, In the Matter of The Cleveland Electric Illumination Company (Perry Nuclear Power Plant, Unit 1),

December 6, 1996.

[4] Letter from G.R. Horn (NPPD) to USNRC Document Control Desk, Proposed Change No. 119 to Technical Specifications Revision of Pressure -

Temperature Limitation Curves, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46, December 10, 1993.

[5] Cooper Nuclear Station Technical Specification 4.6.A.3.

[6] Letter-from G.A. Trevors (NPPD) to USNRC Document Control Desk, Reactor Vessel Material Surveillance Program, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46, July 6, 1987.

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, 4 l [7] Safety Evaluation Report by the Office of Nuclear Reactor Regulation, Related to Amendment No. 120 to facility Operating License No. DPR-46, Nebraska Public Power District, Cooper Nuclear Station, Docket No. 50-298, dated April 26, 1988.

[8] Letter from G.R. Horn (NPPD) to USNRC Document Control Desk, Response to Questions on License Extension to 40 Years from Operating License Issuance, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46, June 7, 1991. 1

[9] Safety Evaluation Report by the Office of Nuclear Reactor Regulation, Related to Amendment No. 143 to Facility Operating License No. DFR-46, l Nebraska Public Power District, Cooper Nuclear Station, Docket l No. 50-298, dated July 5, 1991.

! [10] Letter from G.R. Horn (NPPD) to USNRC Document Control Desk, Proposed Change No. 109 to Technic 4 Specifications, Revision of Pressure -

Temperature Limitation Curves, Cooper Nuclear Station, NRC Docket ho. 50-298, DPR-46, July 28, 1992.

[11] Letter from G.R. Horn (NPPD) to USNRC Document Control Desk, Submittal of Reactor Vessel Surveillance Results, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46, February 25, 1913.

] Principal Contributor: M. Mitchell

Dated

May 7, 1997 s l l

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