2CAN018605, Forwards Projected Values & Method of Development of Resistor Temp for Pressurized Thermal Shock,In Accordance w/10CFR50.61

Forwards Projected Values & Method of Development of Resistor Temp for Pressurized Thermal Shock,In Accordance w/10CFR50.61

ML20205J240
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/22/1986
From:	Enos J ARKANSAS POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 2CAN018605, 2CAN18605, NUDOCS 8601300131
Download: ML20205J240 (5)

Text

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK, ARKANSAS 72203 (501) 3716 January 22, 1986 l

l 2CAN018605 l

l Mr. George W. Knighton, Director

{

i PWR Project Directorate No. 7  ;

1 Division of PWR Licensing - B U. S. Nuclear Regulatory Commission {

Washington, DC 20555

SUBJECT:

Arkansas Nuclear One - Unit 2 l

Docket No. 50-368 License No. NPF-6 l Pressurized Thermal Shock Submittal I Required By 10CFR50.61 i

Dear Mr. Knighton:

10CFR50.61 requires that projected values of RT p be submitted for each PWR with an operating license. TherequiredvaluesIbdadescriptionofthe method of development are attached. The most limiting material is one of the intermediate shell plates with a current RT of 173Fupontheexpirationofthe.operatinglicebIh.of129FandaRT[kals Nofurthersubm$

are required by 10CFR50.61 as the screening criteria of 270 F is not exceeded.

1 Very truly yours, l M

. Ted Enos, Manager (

Nuclear Engineering and Licensing JTE/MCS/sg Attachment

.E W J N TON

[ Rsa (71toMAS)

EICSR (PARR)

"' ~

g a g y %gg2ge p P

MEMBEA MtOOLE SOUTH UTIUTIES GYSTEM

DETERMINATION OF RT FOR PTS ARKANSAS NUCLEAR ONE - UNIT 2 PER'10CFR50.61 INTRODUCTION This report describes the determination of RT PTS f r Arkansas Nuclear One - Unit 2 (ANO-2) as prescribed in 10CFR50.61 (50 Fed Reg 29937, July 23, 1985).

BASIS OF INPUT DATA FLUENCE ESTIMATES The ANO-2 FSAR in Section 5.2.4.2 states "... for reactor operation at the maximum expected output, 2,900'MWT, and an 80 percent plant capacity factor, the vessel fluence (E >-1 MeV) will not exceed 3.47 x 1019 nyt over the 40 year design life of.the vessel. This is an upper limit based on the most unfavorable combination of reactor vessel diameter, core power distribution and a 10 percent uncertainty in the fluence calculation."

On May 1, 1984, Batelle Columbus Laboratories issued a report on the examination, testing and evaluation of the ANO-2 97-degree surveillance capsule. Based on data from the flux monitors in that' capsule and transport calculations Lsing D0T-4.3, they projected a fluence (E > 1 MeV) of 4.21 x 1019 nyt assuming a 40 year lifetime with an 80% capacity factor.

In September 1985, EPRI issued a report, NP-4238, of work done by the Un'iversity of Missouri-Rolla involving transport calculations'of the neutron energy spectrum in ANO-2 using the Codes XSDRNPM, ANISN, and DOT-1V.2. The calculations are based on an R-0 model of ANO-2 set up for the transport Code D0T-V.2. Using results of an R-Z calculation from the same code and a

~

1-D calculation from ANISN, a 3-D flux. synthesis was accomplished. The flux thus obtained was used to obtain reaction rates for several threshold foils irradiated at in vessel and ex-vessel locations. For the fuel loading existing in Cycle 2 of ANO-2 (not low-leakage), they projected tb; fluence 1

(E > 1 MeV) at the maximum point on.the vessel inside surface to be 5.54 x 1019 nyt ' assuming a 40 year lifetime with an 80% capacity factor.

A 40 year lifetime with an 80% capacity factor is 32 EFPY. Using the highest projected fluence described above, a fluence rate of 1.73 x 1018/EFPY was derived. As of December 3, 1985, ANO-2 had experienced 3.7 EFPY.

A June 13, 1978 letter from Daniel H. Williams-to J.F._Stolz was filed on Docket 50-368 giving, among other things, maximum-EOL fluences from the work reported in the FSAR for individual welds and plates. Except for the Upper to Intermediate Shell Girth Weld all fluences were reported as 3.47 x 1019 nyt. The Upper'to ~nttrmediate Shell Girth Weld maximum'E0L fluence was given as 1 x 1018 nyt or .03 times the other locations.

All fluence estimates used in this report are based on the following:

1.73 x 1018 (3.7 + B) where:

B = EFPY after 12/3/85 = (.8) x years after 12/3/85 except for the Upper to Intermediate Shell Girth Weld which is based on:

4.99 x 1018 (3.7 + B)

The ANO-2 operating license currently expires on December 6, 2012.

CHEMICAL COMPOSITIONS The bases of the chemical composition of the materials-in the beltline region of the reactor vessel is the June 13, 1978 letter identiffed above which is consistent with the data supplied in Table 5.2-4 of tria ANO-2 FSAR.

The June 13, 1978 letter did not provide Ni. data for welds. These were taken from FSAR Table 5.2-4.

2

MATERIAL PROPERTIES

~

In~the June 13, 1978 letter identified above, the weld material properties were estimated by the application of MTEB Position 5-2, " Fracture Toughness Requirements for Older Plants." Since a measured value is not available, the generic mean values provided in 10CFR50.61 were used for the welds.

With the exception' of one intermediate shell plate (Heat C8182-2)' for which transverse data was generated by the testing of the surveillance program baseline specimens, the only fracture toughness data available for plate material was from longitudinally. oriented specimens. From this longitudinal data, RT was estimated using MTEB 5-2. Since no generic mean value for NDT SA533 Gr.B plate was provided in 10CFR50.61, these estimated values were used except for the one plate for which transverse data was generated. For that plate the measured value was used.

CALCULATIONS Using the input data derived as described above, RT was calculated using PTS Equation 1 of 10CFR50.61(b)(2) since Equation 1 yields a lower value than Equation 2 for each material . considered.

RESULTS The results are presented in the attached table. They indicate values well' below the screening criteria not only on the expiration date of the license, but also at 32 EFPY, a value relavantEto any future modifications to the license to extend it to 40 years from the date of issuance of the. operating license.

l I

1 3 l 1

10CFR50.61 EVALUATION OF ARKANSAS NUCLEAR ONE - UNIT 2 Inside Material Description Chemical Constan'ts for Surface Fluence, nvt Calculated RT PTS Reactor Vessel Heat Composition, w/o RTore Calculations. *F License License

. Beltline Region Location Number Type Copper Nickel Initidl RT g, Margin Expire 32 EFPY Expire 32 EFPY Intermediate Shell 10120 Linde 0091 .05 .18 -56 59 4.38x1088 5.54x108' 28 -30 Longitudinal Weld-

. Lower Shell .

10120 . Linde 0091 .04 .18 -56 59 4.38x102* 5.54x10l* 20 21 4 Longitudinal Weld Intermediate to Lower 83650 Linde 0091 .05 .18 -56 59 4.38x10l* 5.54x1088 28 30 Shell Girth Weld Upper to Intermediate 10137 Linde 0091 .23 .18 -56 59 1.26x10ts 1.59x10:s 67 .72 Shell Girth Weld

, Intermediate Shell C8161-3 SA533 Gr.B .12 .63 5 59 4.38x101S 5.54x1018 173 180 -

Plates C8161-1 SA533 Gr.B .08 .59 10 59 4.38x10l* 5.54x10t* 135 139 C8182-2 SA533.Gr.B .08 .60 0 48 4.38x10t* 5.54x1019 114 119

^

Lower Shell Plates C8161-2 SA533 Gr.B .08 '. 59 10 . 59 4.38x108' 5.54x1028 135 ~ 139 B2545-1 SA533 Gr.B .07 .66 -20 59 4.38x1018 5.54x1088 97 101 B2345-2 SA533 Gr.B .07 .65 -20 j59 4.38x1019 5.54x1018 97 101 1

O' e~

O

_ y ________ __.__. _ _J