ML20085J716

Forwards BAW-2140, Analysis of Capsule Wiep Point Beach Nuclear Plant,Unit 2 - Reactor Vessel Matl Surveillance Program

ML20085J716
Person / Time
Site:	Point Beach
Issue date:	10/15/1991
From:	Zach J WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20085J719	List: ML20085J716 ML20085J733
References
CON-NRC-91-121 VPNPD-91-364, NUDOCS 9110290318
Download: ML20085J716 (7)

Text

. . -- -. --

-Wisconsin 1Electnc POWER COMPANY m w uenom po b ?on um w na guj m ms VPNPD 364 10 CFR 50.61 NRC-91 12) . 10 CFR-50, Appendix H October 15, 1991 Docume.it Control Desk U. S. NUCLEAR REGULATORY COMMISSION

-Mail Station P1-137 Washington,:DC 20555 Gentlemen:

QQCKETS 50-266 and 50-301 REACTOR VESSEL SURVEII 4LANCE CAPSULE TEST REPORT

-AND RT pts SUBMITTAL POINT BEACH NUCLEAR PLANT. UNIT 2 AND CHAREy UPPER-SHELF-ENERGY STATUS.

POINT BEACH NUCLEAR PLANT. UNITS 1 AND_4 Surveillance Capsule S was removed from Point Beach Nuclear Plant Unit 2 on October-24, 1990. The capsule testing results are submitted-herewith in B&W-Nuclear Service Company (BWNS) Report .

BAW-2140, dated August 1991, as. required by 10 CFR 50 Appendix H.  ;

A summary of the capsule test results-is provided'in Attachment A.

~ Three copies of this repor t -are enclosed for your information.

The surveillance capsule test results support continued use of the current Point Beach Technical Specification operating pressure-temperature 11mit curves. The results-indicate a temperature shift less than,that predicted by Regulatory Guide-1.99 Revision 2, which >

was used to calculate the curves. Additionally, in calculating the heatup and cooldown curves, no credit was taken for the neutron flux-reductions implemented in 1989. - Therefore, the actual vessel-

' fluence will be less than predicted through the expiration date of -

the current curves.

NRC rule 10-CFR 50.61 requires that-licensees submit projected values of RT 3 for reacto:: vessel beltline materials.- Because-PointBeachboesnotprojectthevalue1oranymaterialinthe beltline region to exceed the. PTS screening criteria, the assessment must be submitted with- the next update of the- pressure-temperature l!mits,_or the next reactor vessel' material surveillance report, or-five years from the effective date of this-rula,1whichever comes first. Please note that this submittal also fulfills the RT p73 evaluation requirement for Point Beach Unit 2 (See Tab 1c 7.6 of BAW-2140).

k 9110290318 911015 ' h

- PDR ADOCM 05 COO 266 P- PDR I\

i

u. v v i os-d subs 44vrofHisvwn Duns nwyrm, tin 1

t Document Control' Desk October 15, 1991 Page 2 Point Beach Unit l's PTS submittal will be forLhcoming once additional, accurate, fluence projection data Js available from our cavity dosimetry program. Preliminary calculadions indicate that the RT pts values for all materials in the Unit 1 beltline region will remain below the applicable screening criteria. The Unit 1 submittal will occur within the time requirements specified in 10 CFR 50.61.

On September 30, 1991, Mr. Barry Elliot of NRR's technical staff, Mr. Robert Samworth, NRC Project Manager for the Point Beach Nuclear Plant, and members of the Wisconsin Electric staff participated in a conference call regarding our projections of Charpy-upper-shelf energy for the reactor vessel materials in both units. Mr. Elliot requested that we include a discussion of that issue with this letter. The Unit 2 upper-shelf energy projection is discussed in both BAW-2140 and Attachnent A. That discussion is generally applicable to Unit 1 riso. Based on current end-of-life fluence projections, best estimate material chemistries, surveillance data, and the prediction techniques described in BAW-1803, Revision 1, " Correlations for Predicting the Effects of Neutron Irradiation on Linde 80 Submerged-Arc Welds," we predict that the mean value for the controlling vessel weld metal upper-shelf energy will not decrease below 50 ft-lbs during the vessel design life for either unit. BAW-1803 was developed specifically to addreus the need for an estimating method for the Automatic Submerged-Arc Mn-Mo-Ni Wire /Linde 80 Flux class of weld metals.

BAW-1803 Revision 1 was transmitted to Mr. Barry Elliot directly from BWNS on October 4, 1991. Attachment B presents a listing of our overall reactor vessel integrity program.

Please contact us if additional information is required.

Very truly yours, c- ,,,- x. e J., J. Zach Vick President Nuclear Power Attachments Copy to: NRC Resident Inspector NRC Regional Administrator l

' Attachment A

SUMMARY

OF CAPSUIJ_6 REACTOJR VESSEL MATERJALS TESTING POINT BEAgli_t[QQlEAR PLANT, UNIT 2 BAW-2140: ANALYSIS OF CAPjnE, S WISCONEU1 ELECTBIC POWEB COMPANY POINT BfhCH 1[QCLEAR PLAliT_Uli1T NO. 2 Unit 2 was shut down on October 6, 1990 for its sixteenth refueling shutdown. Capsule S was removed from the reactor vessel on October 24, 1990. It had resided in the reactor vessel for approximately 14.8 effective full power years of operation.

This capsule represents approximately 119%. of peak (inside surface) reactor vessel fluence for estimated lifetime radiation embrittlement considerations.

Capsule S received 19 2 an average fast neutron fluence (E > 1.0 Mov) of 3.47 x 10 n/cm . The predicted peak fast fluence for the reactor vessel T/4 location at the end of the sixteenth cycle 19 2 (14.8 EFPY) is 1.06 x 10 n/cm (E > 1 Mev). Based on the calculated fast flux at the vessel wall, an 80% load factor, and the continued use of the current fuel management techniques, the projected peak fast fluence that the Unit 2 reactor pressure vessel inside surface will receive in 40 calendar years of operation is 2.92 x 10 19 n/cm 2 (E > 1 Mev)19and the 2 corresponding T/4 fluence is calculated to be 1.93 x 10 n/cm (E > 1 Mov).

(See Table Below)

Peak Fast Neutron Exposure for Point Beach Unit 2 (E > 1.0 MeV) n/cm2 Eoc 16 Insido Surface T/4 Wall Loce. tion 14.8 EFPY 32.0 EFPY 32.0 EFPY circumferential Weld (SA-1484) 1.59E+19 2.56E+19 1.69E+19 Intermediate Shell Forging (123V500) 1.60E+19 2.92E+19 1.93E+19 Lower Shell Forging (122W195) 1.59E+19 2.66E+19 1.76E+19 The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure.

The behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility. The weld metal enibited greater sensitivity to neutron irradiation than the base metal. However, the difference in tensile properties are insignificant relative to the conservative analysis of the reactor vessel materials at this time period in the reactor vessel service life.

The Charpy impact data results for the base metal forging materials, the weld metal, and the correlation material exhibited the characteristic shift to higher temperature for the 30 ft-lb transition temperature and a decrease in upper-shelf energy (USE). The 30 ft-lb transition temperature shift is in

relatively good agreement with the values predicted in Regulatory Guide 1.99, Rev. 2, and the predicted value is conservative when the margin is added. The maximum end-of-life RT yp7 is 280'F for the inside surface intermediate shell to lower shell circumferential wc1d (SA-1484). The most limiting end-of-life RT p7s is 283 F, also for SA-1484. This RT p73 value is below the screening criteria of 300'F listed in 10 CFR 50.61.

The decrease in Charpy USE with irradiation showed good agreement with predicted values for the base metals. However, the weld metal decrease in Charpy USE was less than predicted. This is probably due to the lack of data available for developing the estimating curves for material with similar copper contents as Point Beach reactor vessel surveillance materials. BWNS has developed a method to evaluate the radiation induced decrease in upper-shelf energy for the Automatic Submerged-Arc, Mn-Mo-Ni Wire /Linde 80 Flux. We believe this approach is more accurate than Regulatory Guide 1.99 Revision 2 for vessels similar to ours. BWNS approach is reported in BAW-1803, Revision 1. Based on surveillance capsule data and the prediction techniques presented in BAW-1803, it is predicted that the controlling vessel weld metal upper-shelf energy will not decrease below 50 ft-lbs during the vessel design life.

j

. -- . . _ _ - , - - . _ . - . -. . .- - . - -..~. - .- .-

I

^

Attachment B POINT BEACU NUCLEAR PLANT REACTOR VESSEL INTEGRITY PROGRAM  ;

1984 - PRESENT l Proiect Date Completq

1. Neutron exposure evaluation of Point Beach reactor vessels. December 1984

'3. Tested Unit 1 Surveillance capsule T. December 1984

3. 10 CFR 50.61 - Pressurized Thermal Shock (PTS)

Submittal. January 1986 Correction to PTS submittal. March 1986 Safety evaluation report received from NRC. July 1986

4. Reactor iessel Life Extension Study, i

Initiated study in May 1986.

Evaluation of fuel management techniques and internals modifications (shielding) to meet flux reduction goals. September 1987 Identification of critical components in NSSS, I including the reactor vessel, and compilation of transient data associated with these components. October 1987 Comprehensive scoping risk assessment to examine Point Beach specific concerns and the propriety of the flux reduction goals. December 1987 Developed bases and specifications for a plantwide  :

on-line fatigue monitoring system. December 1987 5.- Inservice Inspection a.- Second Unit 1 Reactor Vessel Ten-Year Exam:

LPerformed ASME Code exam utilizing SWRI standard data acquisition system, including 50/70 tandem-near surface search units. May 1987 Performed exam using NES/Dynacon Ultrasonic Data Recording and Processing System (UDRPS) concurrent with ASME Code exam above. May 1987

b. Second Unit 2 Reactor Vessel Ten-year Exam: t SWRI Enhanced Data Acquisition System (EDAS) was utilized. October 1989

\

6. Joined Babcock and Wilcox Owner's Group (BWOG)

Materials Committee. August 1988 Full participant in BWOG Reactor Vessel Integrity Program (RVIP). August 1988 Participant in BWOG Reactor Vessel Life Extension Surveillance Program (RVSP). 1989 Developing master integrated reactor vessel nurveillance program to include Westinghouse utilities with Linde 80 wolds in their reactor vessels. (BAW-1543) March 1989 Submitted BAW-1543 Revision 3 to NRC. October 1989 Safety evaluation report received from NRC for BAW-1543. June 1991

7. Inr,tallation of excore neutron dosimetry (radiometric monitors and solid state track recorders) over one octant of each unit's reactor vessel. Analysis of sensor setn and correlation of cavity measurements with transport calculations will be performed after each fuel cycle for first three sets. Thereafter, a three year interval will be used until sufficient data is obtained to increase the interval.

Install mounting hardware and first set of dosimetry in Unit 2. November 1988 Install mounting hardware and first set of dosimetry in Unit 1. May 1989 First sensor set analyzed for Unit 2. November 1990 First sensor set analyzed for Unit 1. Decemuar 1990 Second sensor set analyzed for Unit 2. October 1991

8. Pilot project: On-line fatigue monitoring of Unit 2 pressurizer surge nozzle (related to reactor vessel life extension study fatigue evaluation). November 1988

9. Implement super Low Leakage Loading Pattern (L4P) cores and axially-zoned hafnium inserts in the guide tubes of peripheral assemblies.

Unit 1 May 1989 Unit 2 November 1989 l

-2 - 1 i

j

( 10. Performed image enhancement of selocted radiographs i of important reactor coolant system components I

(reactor vossols, piping, steam generators, etc.)

and retained radiograph image on media more permancnt than original media. 1989

11. Submit revised heatup and cooldown curves using the guidance of Regulatory Guido 1.99, Revision 2. August 1989 Technical Specification change approved by NRC. January 1990

12. Tested Unit 2 Surveillance Capsule S. August 1991 B

_ _ _ - - __ _ _ _ _ _ - _ _ _ _ - _ - _ _ _ __ ____