ML20090H438

Forwards Analysis of Vessel Wall Neutron Dosimeter from Browns Ferry Units 1,2 & 3 Pressure Vessel, in Response to NRC Request for Addl Info to Support Util 830922 Request to Change Tech Specs TS 191

ML20090H438
Person / Time
Site:	Browns Ferry
Issue date:	10/19/1983
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML18025C005	List: ML20090H438 ML20090H445 ML20090H451
References
NUDOCS 8310280148
Download: ML20090H438 (17)

Text

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. TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374ot 400 Chestnut Street Tower II October 19, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

In the Matter of the ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 By my letter to you dated September 22, 1983 we submitted a request to change the technical specifications of the Browns Ferry Nuclear Plant (TVA BFNP TS 191). The requested specifications enclosed with-the letter revise the pressure-temperature limit curves for hydrostatic pressure test, reactor vessel heatup and cool down, and nuclear core operation.

The enclosure to this letter provides additional information in support of TS 191 requested by your staff.

The enclosure includes information regarding chemical composition of reactor vessel material, end-of-life neutron fluence, and Charpy impact testing results. Please get in touch with us through the Browns Ferry Project Manager if additional information is needed.

Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Ma ager Nuclear Licensing Subscribed a sworn to beror me this. '/ dayof[O68[g, 1983 h0Atl/$ Y $ N Notary Public

,My Commission Expires _

Enclosure e oc: See page 2 8310280148 831019 PDet ADOCK 05000259 00l p PDR An Equal Opportunity Employer 8(3 a

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,T Mr. Harold R. Denton October 19, 1983

.cc (Enclosure):

Mr. Charles R.. Christopher-Chairman, Limestone County Commission P.O. Box 188

-Athens, Alabama ~ 35611 Dr. Ira L. Myers State Health Officer State Department of Public Health State Office Building Montgomery, Alabama 36104 Mr. R. J. Clark U.S. Nuclear Regulatory Commission Browns Ferry Project Manager 7920 Norfolk' Avenue Bethesda, Maryland, 20814 _

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ENCLOSURE ADDITIONAL INFORMATION REGARDING REACTOR-VESSEL MATERIAL TOUGHNESS AND NEUTRON FLUENCE IN SUPPORT ~OF AMENDMENT REQUEST TVA BFNP TS-191 BROWNS FERRY NUCLEAR PLANT (DOCKET NOS.. 50-259, -260,_-296)

Attachment 1 - Reactor Vessel Beltline Material Toughness Data

~

Attachment'2 - GE letter dated February 9, 1971 - Information Regarding Neutron Fluence Attachment 3 - Browns Ferry unit 1.-' Analysis of Vessel Wall Neutron Dosimeter'- (report by Southwest Research Institute)

Attachment 4 - Browns Ferry units 2 and 3 - Analysis of Vessel Wall Neutron Dosimeter (two reports by Southwest Research Institute)

Attachment 5 - Fabrication and Testing of Reactor Vessel Material (report by TVA's Singleton Materials Engineering Laboratory).

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Table 1.. ,

Keactor Vessel.Deltline Material Toughness D.its ,

Browns Ferry Nuclear. Plant Units 1, 2, and 3 Shift in Finill RTNDT for RTNDT for-Minimum Initial' 40.F.FPY:

40EgPY.

-Heat Material Cu ' - NDTTs 50 ft }b/35 mil temp ( F) , RTNDT I 1/4T - .at:1/4T:

Component Unit No. Crade (5)1 (oF) ,PMWD 4 ,o. NMWDS,o (or) '(oF) (oF).

Dase Hetal 1 C-2884 SA 302 GRB 0.12 **

02 +34 - +543 0' 32 32 -

HAZ 1 .- Weld. - -03! +60 +803' +20 32 ~ .52 0.10 WELD 1- - Weld . +133 -

+38 +13 26 39-Pase Metal 2 A-0981 SA 302 CRB 0.14 -102 - -14 '+143 -10 39 ' 29

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Weld - 03 -40 -283 0 39 39.

WELD 2 -- Weld 0.20 03 - +18 0. ~ 58 - 58 ,

Base Metal 3 B-7267' SA 302 cRB 0.10 - 202. -36 .-183 ,

-20 ' 26 '6 IIAZ 3 - ' Weld - 03 +20 +473- 0' ,26' 26' WELD 3 - Weld 0.11 03 - ,-22 0 z 29 29 10ee letter fron T. F. Henry, Combustion Engineering, Inc., to' John Fox dated April 6,1993 20eneril Electric Company' letter from R. B.' Beers to H.'M. Bankus, " Reactor Vessel Surveillance Program",-dated February 9, 1971 (ecpy attached).

3 Estimated based on Branch Technical Position HTEB 5-2 (copy attached) and minimum valves from Charpy impact data prov'ded i by~

Singleton Laboratory (see attached memorandum from F. Van Meter to T. F. Ziegler dated August 11, 1983--L21 830812 112) kP arallel to rijor working direction Snormal to major workinc direction

• 6Determined by using minimum values from Charpy impact data provided by Singleton t.aboratory.

7The greater of NDTT or yC (50 ft Ib/35 mils) - 60 for NMWD 8Estimated using 1 above, Regulatory Guide 199 Curves (copy attached) and neutron fluence determined from flux wire testing (see,_

C16 790223 019 and L26 800318 883--copy attached).

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Interoffice Correspondence

.- POWER SYSTEMS

. April 6, 1983 Mr. John Fox Tennessee Valley Authority

Subject:

_ Materials Analyses .

Project No. 900051 -

Job No. 98120216 ~

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-Unit 1, Unit 1, Unit 2, Unit 3, Sampic: 2868 2884 E6B Unit 2, Unit 3 Unit 1, 65E EBM 67J-(P23079) (P23080) (P23031) Weld Plate (P23082) (D39368) (D39369) (D39370) ~

C .22 .27 .20 .18 .15 .16 .17 Mn 1.39 1.33 1.35 1.27 1.49 P .007 .008 1.50 1.45

.007 .007 .010 .011 S .009 .015 .011 .012

.007 .011 .012 .013 Si .23 .21 .19 .22 Ni .52 .09 .10 .09

.52 .55 .51 .33 Cr .07 .09 .28 .30

.11 .08 .08 .08 Mo .48 .46 .49 .06

.47 .49 .49 .48 V .002 .003 .002 .002 .003 .003 .003 Co. .009 .010 .012 .010 .011

.09 .010 .010 Cu .12 .14 .10 -

.20 Al .032 .11 .10

.035 .025 .015 .007 B <.001 <.001

.006 .010' e.001 s.001 <.001 <.001 W <.01 < 01 .01 <.001- "

<.01 <.01 c.01 <.01 As .006 .009 .015 .010 .013 .008 .009 Sn .006 .008 .013 .008 .009 .005 .006 Zr <. 001 <.001 s.001 <.001 <.001 <.001 <.001

'N .006 .007 .007 2 .005 .007 .006 .006 .

We certify that the information recorded above is an accurate copy of the data obtained from tests performed at the Metallur-gical and Materials Laboratory.

COMBUSTI N ENGINEERING, INC.

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\ ATOMIC POWEn cc TENT GENERAL' CLcCTnlO . COMPANY. 175 CUnTNCn AVE , - SAN JOSc. CAL.lf. 95125 Phone (108) 297 3000. TWX. NO. 910 336-011G ~ ~ ~ " '" M EW ;- - - - . ,

ncuvro FEB 1 2 1971

. tttr. R.11.1Dunham Letter No. BP 3'G93 etsien utenianc.6 .

~C hicC !!cchar.ical Engineer February 11, 1971 n 6 twe i t.m ser,s -

Tennessco Valley Authority *

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General Electric Company t iunai_ _, um i .

1301 llannah Avenue H. W.

Knoxville, Tennessee 37921

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Subject:

" Reactor Vessel Surve.illance Program .__.i i ._.J.'^ '

Reference:

TVA to GE Letter 6599 and 3-3336 T 6,

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. We have attached numbered. Our answers a marked-up refer to the copy above of TVA Letter 6599 with the questions '( #

nuir.bers: e JF

1) shell

2) - According to our QC records, the following is the data on the f4K !!58 f,c,'. -

plate material:

6) lleat fio. Plate'No. NDTT ( F) I TVA I C-2884 6-139-19 0 C-2868 C-2753 6-139-20 6-139-21 0

-20 h-.'te.W-

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TVA II o A-0981 6-127-6 -10 c'/,

D g to C-2467 6-127-10 -10 C-2849 6-127-20 -10 7 c./ /) /M TVA III C-3201 C-3188 6-145-1 6-145-2

-30

-30

/UM Ab. -

B-7267 6-145-6 -20

- 3) B!JI cites the second sentence of Par. 3.3.1 ("The Buyer can furnish a plan which the Seller may use as a guide.") as permission to use Appendix B and Drawings 11781549 and ll7B1500.in lieu of preparing the required document. G&W stated that they followed these documents in preparing the surveillance specimen. 8t.U states that the surveillance samples will be taken from either MK !!57 or 14K #58 plate mat ^-i.4 w.- - . . . . . .

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GEN ER Al @ ELECTRIC Mr.'H._M. Bankus February 9, 1971

'4) Further correspondence with BT,W has failed to obtain any further docu-

5) mentation or information beyond that described above. We are continuing to pursue this matter.

7) No special handling tools are provided for removal of the surveillance

8) specimen holders from the reactor vessel. The normal pool tool accessori.2 (J-hook, L-hook, . actuating pole, etc.), plus. the in-core detector cutter

_ are sufficient to remove, handle and prepare for shipment. Shipment is necessary because testing of the' individual specimen requires the services of a radiation-laboratory. It is left up to each customer to select the laboratory and contract for the particular services they desire.

Shipment is most easily accomplished by separating the specimen basket (intact with capsules containing-the individual specimen) from the sur-veillance holder pole. This is done with use of the in-core detector cutter. The basket is deposited in a small lead shipping cask and sent -

to the laboratory for processing. The remaining pole and handle assembly may be cut up and discarded as rad-waste.

9) GE would be willing to provide testing and evaluation service on irradiate

10) surveillance samples. Actual contract conditions will have to be negotiated through our Chattanooga office.

11) The maximum calculated integrated neutron dose (1 mov) on the vessel wal'.

12 is 3.80 x 1017 NVT. This exposure is expected to occur at an elevation 13 'three feet above the bottom of the active core at azimuthal positions 14 opposite the corner elements. The azimuthal positions are approximately 25*, 65*, 115 , 155 . 205*, 245*, 295 and 335 .

The calculated dose on the surveillance baskets at azimuthal locations c

. 30 , 120*, and 200 are equal and equivalent to

• 3.07 x 10 8 neutrons 2

or 3.9 x 1017.NVT cm - sec at. rated power for 40 years. These values are based on the same calculati type as the maximum fluence base. The axial variation is assumed to be proportional to Figure 1.

The above data applies to all three units.

15) As stated in NED0-10ll5, Para. 5.7: "Because the boiling water reactor is a constant-temperature device, no special temperature monitoring devices are required." This has been confirmed by temperature monitot ing devices which had been installed in earlier plants.

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. l'- _ G E N ERkt (h E LE CTRIC Mr. H. M. Banlius - -

.3- February 9,1971

_16 ) Two copics of NED0-10115 are enclosed.

This' completes ~our response to TVA 1.ctter 6599 and 3-3336.

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B. B ,eers/l .h if -

_ Project Manager TVA Project JRO:kn Enclosure -

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General Electric ,,q CUSTONER ORDER.HO:- 205-55577- D A TE: _

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6.1.0.- 017 7- 51 Uni t # 1 TYPE OF MATEEIAL: Mn-Mo-Ni Cod - Case 1374 13 HO: 58A SPEclTICATION: u in?, <o - -A '

_ COMPON EN T S ER I A L nun'3ER:__.6-139-19 II'NN lie A T NO. i?. lC KH ESS: 6 1/8 Min .

• C l 1.in P l S Si Lukens C2884-2 .22- 1.33 .015 .016 Cr Ni no

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+10 50, 44, 50' .036, .028, 35, 20, 25 Height - 3' .027 l Weight 100; Lbs. 240 Ft.-Lb. Energy. Load S PE C il!EN CODE HEAT HO. TENSILE Sil:EliCTH YlELD P0lHi ELOHCATION P.S.I. REDUCTION OF P.S.I. -- lH 2 INCHES AREA 5 BEN EST "3 - GRA'IN Silf' G-139-19T 88,250 69,500 26.6 68.2 Good' Ho:nogeneity 6 and C2884-2 J. - 13 9- 19B 92,250 69,750 OK Finer 24.2 69.3 f

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(Cf.XOf /510-019'h cil fini t //1 TTPE OF SATERIAL; Mn-Mo-Ni Code R^

Case 1 73 %

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SPECIFICATIONi ~3- O'>

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TEST I'~ID3 GRAIN. Sill i5-139-20T 86,500 68,000 27.3 69 9 Good Homogeneity 6 and.- .'

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D16 790223 U19 J. G. Dewease, Power Plant Superintendent, Browns Ferry ant Nuclear Pl H. J. Green, Chief, Nuclear Generation Branch, 727 EB-C FEB 2 81979 WALL NEUTADN DOS 4MTER FRett BR0 Attached are four copies of the completed report on s sthe analy i of the Browns Ferry unit I vessel wall dosimeter.

This analysis was performed to comply with the requirements of Browns Ferry Surveillance i (SI) 4.6.A.4. nstruction The results of the analysis do not meet the acceptance criteria of the above mentioned 51 Alch states that the experiment determined neutron-fluence. neutron fluence shall be less than ae the maximum The experimentally-determined neutron fluence for the end 10 I7 cn -2 of core cycle 1 at the inside diameter of the vessel was 5 64 x .

cm

-2

, while the calculated value of neutron fluence was 1.37 X II 10 Ve are therefore required to perform an analysis to determin e the ef fect of increased fluence on vessel-well embrlttlecunt HDT*

and RT Regulatory Guido 1.99 stipulates that the fluence value theatI/l.T (thickness) vessel wall location be used theshiftinRTbOT. Finure I e mas n ngthe basis for d f the same document presents a means to determine the shift in RT ET based on known fluence levels and material copper content.

In a letter from A. L. Vest, of General Electric Company, to me datud Decanbar 19, 1977, the curves of Figure I were extrapolated downward to allow determination of transitionperature tem shif ts at .the lower generating flux values for bolling eactors. water r A letter dated August 23, 1977, fmm J. E. Cllleland to A

. S chwence r, o f HRC, established an upper limit for plate and weld metal copper content of 0 30 weight percent.

Utillring this copper content, the projected 1/4T fluence value of 2.62 X 1016 c ,-1 per EFPY (derived frcro the '

product of 0.67 X 3 91 X 1016 c,-2

/EFPY taken frors the Swnl analysis),

and the extrapolated Figure 1 of Regulatory Culde , the 199 following chart l's generated. .

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J.' G. Dewease VALL. NEUTRON DOS! HETER FROM BROW EFPY 1/4T Fluence (ca~ ) Sh!ft In RT NDT F) 1 2.62 x 10 I 0 2

5.24 x 10 16 20 3 16 7.86 x 10 25 4

1.05 x loI7 29 5

1 31 x 1o17 33 6

1.57 x loI7 36 7

1.83 x loI7 38 8

2.10 x 1017 gi 9

2 34 x 10I7 43 to

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2.62 x 10 I7 45 11 2.88 x 10I7 48 12 3.1% X 10 7 50 13 3.41 x 1037 52 14 3.67 x 10I7 54 15 3 93 x 10I7 56 16 4.19 x 1037 58 17 4.46 x 1017 60 13 4.72 x 10I7 61 19 4.98 x 1017 63 20 5.24 x loI7 65

'l 5 50 x loI7 66 22 5 76 x 1037 68 l 23 -

6.o3 x loI7 69 24 6.29 x 10I7 71 25 6.55 x 10I7 73 26 6.81 x loI7 74 27 7.07 x loI7 75 28 7 34 x loI7 29 76 7.60 x loI7 78

30 7.86 x 1037 79 l

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J.'G. Dewease VALL NEUTRON DOSIMETER FROM SR EL SEL Effl.

1/47 Fluence (ca' )

Shlft in RTNDT ( F) 31 8.12 X 10'7 80 32 8.38 X 10'7 81 33 8.65 X 10II 82 34 8.91 X 10I7 OS 83 9.17 x 1037 36 85 9.43 x 1017 66 37' 9.69 x 1017 g7 38 9.96 x 1037 68

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39 30 1.02 X 10 40 89 1.05 x 10 18 90 lho operator's curvos for unit I will not be af fo,cted by the above-roforenced bottline curve shift in RTMDT unti the alghth effective full tow r yea r o f ope ra t i on s i nce t he non be l t l i ne reg ion cu rve i s dcrainan t until the beltlino curve shifts by 40* F.

The predicted end of-core life adjusted reference temperatura e at th l/4T position determines the specimen capsule renovel schedule in accordance with 10 CFR 50, Appendix H.

l Page 22 of the attached report contains a table which depicts two different specimen capsule removal schedules based on the initial RT RT f r,the unit NOT tM M tline region plates. N HDT I reactor vessel beltline region is obtained as follows.

A letter dated February 9,1971, from R. 8. Seers , of GE. to H. M.

cankus, of GE, specifies that the highest NDTT value for unit I reactos vessel beltline region plates is O' F.

l This value la dartved frora drop-

• might tests performed by the vessel manufacturer. Copies of the test l certificates for the llaiting heats (C2868 and C2884) show the drop- L I

welght tost, as well as the singla-temperature Charpy notchVImpact O

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. SOUTHWEST RESEARCH INSTITUTE - FINAL RE70RT ON VALL NEUTRON 00SIMETER FROM BROMIS FERRY UtilT 1 PRES test results (longi tudinally-orlanted specimens) .

Section 8.1.1. (4) of

' Branch Technical Pcsition HTES 5-2, " Fracture Toughness ements Requir for Older Plants," specifles how to determloe the RT WT c rresponding to the plate material if limited Charpy V notch tests were performed et a single-temperature on specimens oriented in the longitudinal direction It states that .

the RT if the minimum absorbed energy Is less than !+5 ft/lbs .,

t4DT is estimated as 20' F above the test temperature RT . Therefore, HDT - 1 F (test temperature) + 20' F = 30' F.

Since the above value of RT NDT Is greater than 12* F, the end-of-core life adjusted reference tersperatura at the 100* F.

1/4T position will exceed This dictates a specimen capsule withdrawal schedule conforming to Section ll.C.).b, 10 CFR 50, Appendix H.

RT SInce the 50* F shift In NT ccurs at approximately the !2 EFPY point, which exceeds thn one-fourth servico life value of 10 EFPY, the first specimen capsule removal should occur at the 10 EFPY point.

(Hota that the table on pego 22 of the attached report i s incorrect for this estirmte.) Note however that the results of the specimen tests are not utilized to prodlet in RT WT, but to vertfy, the shift If needed, Figure 3.6-2 of the Drowns Ferry unit I tc,chnical specifications.

Cnple , of applienble sections of all referenced docunents ached. are att H. J. Green BOD CAM:RWC:08 Attachments ces ARMS PP, 823 EB-C

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