ML20217H206

Safety Evaluation Supporting Amend 106 to License DPR-22

ML20217H206
Person / Time
Site:	Monticello
Issue date:	10/12/1999
From:	NRC (Affiliation Not Assigned)
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ML20217H190	List: ML20217H184 ML20217H206
References
NUDOCS 9910220075
Download: ML20217H206 (7)

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• 4 UNITED STATES,

j NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 3068H001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

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RELATED TO AMENDMENT NO.106 To FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated December 31,1998, as supplemented May 17,1999, the Northern States Power Company (the licensee) requested an amendment to the Technical Specifications (TSs) appended to Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant (MNGP). The proposed amendment would revise the TS reactor pressure vessel (RPV) pressure-temperature (P-T) limii curves, delete completed RPV sample surveillance requirements, delete the requirement to withdraw a specimen at next refueling outage, remove the standby liquid, control system relief valve setpoint, and make associated administrative changes. The May 17,1999, letter provided clarifying information that was within the scope of the original Federa/ Register notice and did not change the staff's initial proposed no significant hazards considerations determination.

Following discussion between the NRC and Northern States Power Company (NSP) staffs during telephone conferences on February 18 and March 10,1999, the NRC staff issued a request for additional information dated March 24,1999, to which NSP responded by letter dated May 17,1999. NSP also committed to modify the reactor pressure vessel (RPV) 1 surveillance program in the MNGP Updated Safety Analysis Report (USAR) to withdraw tne next MNGP surveillance capsule in 2003 pending resolution of an initiative by the Boiling Water Reactor Vessel and internals Project.(BWRVIP) to develop an Integrated Surveillance Program for boiling water reactors (BWRs).

These submittals were made in accordance with regulations in Part 50 of Title 10 of the Code of Federa/ Regulations (10 CFR Part 50), Appendices G and H, which govern the development of RPV P-T curves and the RPV surveillance programs, respectively.

2.0 EVALUATION 2.1 Reactor Pressure Vessel Pressure-Temoerature l_imits t

in the early 1980s, NSP withdrew its first RPV surveillance capsule from the MNGP vessel.

i This capsule contained two sets of Charpy V-notch specimens, sufficient to generate two complete Charpy curves. One set was tested and the results were reported in surveillance capsule report BCL-585-84 2,- Revision 1, dated November 1984, along with the results of the capsole's dosimeter wire analysis. The dosimeter wire analysis demonstrated that the capsule 9910220075 991012 PDR ADOCK 05000263 P

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had been subjected to a low lead factor of approximately 0.3 (meaning the capsule acquired about 30 percent of the fluence associated with the peak vessel wall location). NSP decided to -

re-encapsulate the remaining set of Charpy specimens and insert them for additional irradiation in the Prairie Island Nuclear Generating Plant (Prairie Island) RPV. These specimens were withdrawn from PraMe Island in 1996 and tested, and the results were reported in surveillance capsule report SIR-97-033 Revision 2, dated October 1998 (transmitted by letter from M.

{

Hammer (NSP) T NRC dated December 21,1998), along with a reanalysis of the original 1984 i

- Charpy data, i

- NSP r.oncluded that the results from both the original surveillance capsule from MNGP and.

from the supplemental capsule hadiation in Prairie Island were applicable to the evaluation of

)

l the MNGP vessel. A summary of this information is shown in attached Tables 1 and 2 (taken l

from Tables 3-1 and 3-2 of SIR-97-033, Revision 2) for the surveillance plate (heat C2220) and I

the surveillance weld, respectively. Note that the shifts in the 30 ft.lb transition temperature can only be given for the MNGP surveillance plate, since the unirradiated Charpy curve for the l

MNGP surveillance weld material was unavailable. NSP has recently taken action to look for

- and acquire the unirradiated Charpy curve for plate heat C2220, which is based on testing of this material at Oak Ridge National Laboratory (ORNL).

The NRC staff has raised this issue of " lack of unirradiated baseline surveillance data" and its resolution on an industry-wide Ln!s. It was the impetus for the BWRVIP action to develop an l-Integrated Surveillance Prograni for BWRs. The ORNL data was.also used to reevaluate the i

plate's initial nil-ductility reference temperature (IRT ).

l l

In its most recent analysis of the MNGP P-T limit curve methodology (conducted in 1998 to support the MNGP power uprate) prior to its December 31,1998, submittal, the licensee deterrr.!ned that RPV plate 1-15 (heat C2220-2) was the limiting material for the MNGP vessel.

Since the MNGP surveillance plate is representative of RPV plata 1-15, NSP assessed whether l

the new unirradiated and irradiated test data indicated the need to modify the MNGP P-T limit

. curve methodology. Based on the methodology found in Re'gulatory Guide 1.99, Revision 2 (RG 1.99, Rev. 2), for the evaluation of radiation embrittlement in RPV stests, a material's adjusted reference temperature (ART), which for the purpose of the P-T limits evaluation the licensee calculated at a depth 1/4 of the way through the vessel wall (the 1/4T location), can be determined from the following equations:

- ART = IRT. + ART, + Margin where ART, = ti e irradiation induced shift in RT, = CF

• FF with.

CF = the Chemistry Factor, a function of the copper and nickel content of the material or as interpreted from the evaluation of available surveillance data FF = the Fluence Factor, a function of the RPV material's neutron fluence at the 1/4T depth

' and Margin = 34 *F (as assumed by NSP for all MNGP materials)

1 4

l -

.In the NRC ' staff's safety evaluation, dated September 16,1998, of the MNGP power uprate i

amendment, it was documented that the IRT for plate 1-15 was 14 *F, its CF was 125.3 'F (based on a chemistry of 0.17 wt% Cu and 0.58 wt% Ni), and its FF was 0.745. This resulted in an end-of license ART of ART = 14 'F + (125.3 *F

• 0.745) + 34 'F = 141 *F

?

Using the newly acquired unirradiated and irradiated surveillance data, NSP concluded that the IRL value for plate 1-15 was 27 'F, that the chemical composition of the plate was 0.17 wt%

Cu and 0.65 wt% Ni, and that the plate's CF based on an evaluation of the available surveillance data was 130.8 'F. This resulted in (with no change proposed for the margin and a small modification in the FF to 0.730) an end-of-license ART of ART = 27 *F + (130.8 *F

• 0.730) + 34 *F = 156.5 *F Since these results demonstrated that the previous analysis would be nonconservative at end-of-license (and for any fluence level up to end-of-license), the licensee submitted its December 31,1998, application to modify the MNGP license by incorporating figures into the P-T lirnit methodology in the TS that are based on the more conservative material property values above. It should also be noted that since the licensee's evaluation ass;gned the same initial properties, chemistries, and fluences to the other MNGP lower intermediate shell course plate (1-14, heat C2220-1), that plate was calculated to have the same 1/4T end-of license ART as plate I-15 and is therefore an equally limiting material.

NSP also identified passages in MNGP TS 4.6.B.2. and TS 4.6.B.3. which referred only to the historical testing of the first MNGP surveillance capsule. NSP concluded that these passages were no longer relevant for inclusion in the TS and proposed to eliminate them as part of this amendment.

Finally, in a separate action not involving the proposed TS change, NSP committed to incorporate into the MNGP USAR a withdrawal date of 2003 for the next MNGP surveillance capsule. This date may ba changed later pending resolution of the initiative by the BWRVIP to develop an Integrated Surveillance Program for BWRs.

2.2 Staff Evaluation and Conclusions The NRC staff independently evaluated the new unirradiated and irradiated surveillance material test data for plate heat C2220 as it applies to either MNGP plate 1 14 or 1 15.

Regarding the licensee's modification of the IRT value for heat C2220 to 27 *F from 14 *F based on the unirradiated data from ORNL, the staff concurs with NSP's assessment. The staff accepts the use of 27 'F as the IRb. In addition, the use of this measured IRT value permits the margin term to be established as no greater than 34 'F based on the use of RG 1.99, Rev. 2, methodology.

The NRC staff then evaluated the licensee's proposal to use the data from the surveillance material lrradiated in Prairie Island for essessing the MNGP vessel. While the licensee's action to take measures to address the low lead factor problem associated with the first capsule that was withdrawn from the MNGP vessel in the early 1980s is commendable, it also raises certain technical concerns. In report SIR-97-003, Revision 2, NSP noted that the second capsule (from

kN.

. Prairie Island),"saw accelerated fluence (lead factor >10)...." The NRC staff's calculation barmi on the flux and fluence information provided in this report supported this statement and estimated that the lead factor would be about 18.

In the development of surveillance programs for RPVs, the Commission has recognized in its regulations (10 CFR Part 50, Appendix H) the use of American Society for Testing and Materials Standard Practice E185 (ASTM E185). Since at least the 1966 edition of this standard, an emphasis has been placed on maintaining the irradiation history of surveillance capsules such that their irradiation temperature, neutron flux, neutron spectrum, and maximum neutron fiuence are similar to the conditions experienced by the vessel, while still permitting the capsule fluence to lead the vessel so that the testing results are predictive. For the RPV " Wall" capsules required by ASTM E185, the upper limit recommended for the lead factor has ranged from 3 to 5. Such lead factors permit end-of-license fluences to be achieved in a reasonable amount of time while not causing the irradiation history of the capsule to be so different from that of the RPV that the potential for significant flux effects on the irradiation damage response must be evaluated. However, in considering a capsule irradiated with a lead factor of 18, the NRC staff is concemed that some flux effect may occur, particularly since the capsule is essentially being irradiated in an environment (including temperature history and neutron spectrum) ch'aracteristic of a pressurized water reactor rather than that characteristic of a BWR.

i For these reasons, and without additional data from the MNGP surveillance program to verify that the embrittlement behavior projected by the capsule pulled from Prairie Island is appropriate or conservative for assessing the embrittlement of the MNGP RPV, the NRC staff utilized the RG 1.99, Rev. 2, CF Tables in its evaluation. The tables were used in the absence of demonstrated credible surveillance data.

Plate heat C2220 was deterrnined to have a best-estimate copper content of 0.17 wt% and a best-estimate nickel content of 0.65 wt%, which equates to a CF of 128.3 *F. Using the same IRT, margin, and fluence factor used in the licensee's analysis, the NRC staff calculated the end-of-license ART for the limiting MNGP plate as ART = 27 *F + (128.3 *F

• 0.730) + 34 *F = 154.7 *F Th's analysis will also be recordeo !n the NRC's Reactor Vessel Integrity Database (RVID) and the basis for its determination documented there. The data supplied by the licensee based on their testing of the surveillance ca, sule from Prairie Island will also be included in the RVID 9

should the staff's questions regarding its use be addressed in the future.

Note that the results of the staff's analysis show that the CF and end-of-license ART values

. calculated by the licensee are conservative (higher than the staff's value). Therefore, although

~

~ the staff does not concur with the basis of the licensee's analysis (i.e., the use of the Prairie Island surveillance data) the staff finds the results of the licensee's analysis to be acceptable. Furthermore, since the results of the licensee's analysis are acceptable, the TS changes proposed by licensee based on their analysis are also acceptable, since they are equivalent or conservative to the analysis results independently calculated by the staff, and their implementation is consistent with the requirement of 10 CFR Part 50, Appendix G.

. The NRC staff also concurs with the licensee's proposal to remove historical references to the testing activities of the first MNGP surveillance capsule from the TSs. This information has

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. been acqu! red and reported to the NRC as required under 10 CFR Part 50, Appendix H. The historical referer ce in the TSs provides no added value.

Based on the information provided by NSP in its submittals and the results of the NRC staff's independent analysis, the staff concludes that the proposed changes are acceptable. This finding was made since the requirements in the TSs will continue to comply with 10 CFR y

Part 50, Appendix G. The commitment by NSP to incorporate a 2003 withdrawal date for the l

next MNGP surveillance capsule into the facility's USAR is also acceptable to the staff since the licensee continues to comply with 10 CFR Part 50, Appendix H.

2.3 Standbv Liould Control Relief Valve Setooint The licensee proposes to delete TS 4.4.A.2.c, which specifies the Standby Liquid Control (SBLC) relief valve setpoint. The setpoint of the SBLC system relief valves is govemed by the -

provisions of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel -

Code Section XI, as required by TS 3.15. The testing required by TS 4.4.A.2.c is enveloped by the current testing performed by MNGP's inservice Test (IST) Program. The IST Program implements an edition of ASME Code Section XI that has been approved in 10 CFR 50.55a.

1 Any modification sto the setpoint is controlled by the plant's configuration control process, which would ensure the requirements of ASME Code Section XI are invoked as required by TS 3.15.

The IST Program rsquired by TS 4.15 ens ares the SBLC relief valves would be properly tested I

for operability. Therefore, the change is acceptable.

3.0 STATE l CONSULTATION in accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The 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 l

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 (64 FR 6706).' 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.

r-i.

. 5.0 CONCLUSjQN The Commission has concluded, based on the considerations r.:iscussed above, that (1) there is reasonable assurance that the health and safety of the public Nill not be endangered by i

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.

Attachment:

Tables 1 and 2

- Principal Contributors: M. Mitchell F.Lyon Date: October 12, 1999 I

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