ML20042G711

Responds to NRC Questions Re leak-before-break Licensing Submittal for Stainless Steel Piping.Kerotest Valves in Rh Sys Will Be Replaced in Byron Unit 2 During Next Refueling Outage Scheduled to Begin on 900901

ML20042G711
Person / Time
Site:	Byron, Braidwood
Issue date:	05/07/1990
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
0966T, 966T, NUDOCS 9005150325
Download: ML20042G711 (27)

Text

{{#Wiki_filter:- 1 ,s -m o r :6 y \\ Commonialth Edison i . 61400 Opus Place \\ ~h/ Downers Grove, Illinois 60515 { w May 7. 1990 Dr.' Thomas E. Murley, Director Office of Nuclear. Reactor. Regulation .U.S. Nuclear Regulatory Commission Washington, D.C. 2055'; Attn: Document Control Desk. l j

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Leak-Before-Break Evaluation i HRC_DorknLNoadf-454/455 and 50-456/457

Reference:

(a) May 16, 1989 R.A. Chrzanoweki letter to T.E. Murley. ] Dear Dr. Murley -i Reference (a) provided for NRC staff review the Commonwealth Edison }

(Edison) Leak-Before-Break Evaluation for Stainless Steel Piping for the Byron l

and Braidwood Stations.

j

-{ The purpose of this letter is to provide the Edison response to the fourj(4) NRC questions. relative to this Leak-Before-Break (LBB) licensing

submittal'(SL-4518). (These questions were forwarded to Edison via telecopy on l

February 28, 1990.) The response was discussed in an-April 30, 1990 ) teleconference with Messrs. P. Shemanski and S. Lee of the NPC and Edison i believes.these address the NRC staff's concerns. The response is provided in 1 Attachment A. 1 As was statediduring the teleconference, the formal incorporation of l these changes into the licensing. submittal will be deferred until the receipt. of the NRC Safety Evaluation Report (SER). At that time,. SL-4518 will be y revised to incorporate the remaining changes, outstanding and submitted.to the ?NRC staff. 1 The next refueling outage for Byron Unit 2(B2R02) is scheduled to 1l e start on; September 1, 1990.- During that outage, the plan to replace Kerotest .j valves In the RH system is scheduled to be implemented. Also, approximately j 20 to 30 enubbers are scheduled to be removed. The Leak-Defore-Break i

evaluation supports-the basis for doing both ef forts; In the case of-the i

Kerotest valve replacement proc am, having the Leak-Before-Break evaluation approved by the NRC will allow the replacement valves to be sized without .] Lhaving.to' consider the faulted loads. In the case of the snubber reduction. the Leak-Before-Break evaluation directly justilles the removal of the j snubbers,= selected. In general, the Leak-Before-Break Evaluation is relevent l to_the on-going evaluations regarding stratification and the need for a more t flexible plant design to better absorb unexpected thermal transients. l l. P gh u a

b . Similar. planning is in place for Braidwood Station. As a result,. this same justification applies to the same activities planned for the next Braidwood Unit i refueling outage (AIR 02), scheduled to start in March, 1991. Therefore, Edison requests that the NRC staff complete their. reviews by August 1, 1990, to support this~ocheduling. Accordingly, Edison is available to meet with the NRC staff, as deemed necessary, to facilitate the ' review process. Please direct any questions regarding this submittal to this of fice. Very truly.yours, b ) ~ S.C. Ilunsader Nuclear Licensing Administrator cc: P. Shemanski - Project Manager, NRR S. Sands - Project Manager, NRR ~i W. Shafer - RIII BY Resident Inspector ~ BW Resident Inspector /Imw: Scil 0966T ' p -- i

{ t [ f hd b M t Vi I g i t ' Response for Question 1: Subsequent to submitting the Leak-Before-Break (LBB) Evaluation report for NRC review, a revision to the PIFRAC database, PIFRAC2, was received. This issue included additional heats of thermally aged cast stainless steels, SA351 CF3, and CF8. This new material test data has been incor-porated in the LBB evaluation and a revised J-Tearing evaluation was-performed. Table 3-5 has been revised to include the additional cast stainless steel heats of SA351 CF8A, CF8 and CF3 material specification. Table 3-6 has been amended.to include the room temperature tensile properties for the additional heats which have not been thermally aged. A comparison of the installed material tensile properties in Table 3-3 to the unaged room temperature tensile properties shows the PIFRAC2 test specimens to be bounding. Also, an additional table was generated comparing the chemi-cal composition of the installed cast stainless steel to the PIFRAC2 test specimens.. This comparison demonstrates that the PIFRAC2 CF8, Heat Nos.. 63-66, weight percentage of carbon, chrominum, silicon and ferrite percentage closely matches or exceeds the weight percentage of the installed material.. In particular, the ferrite content is significantly higher and the. chromium content is greater in the PIFRAC2 test material. -Based on the bounding ten-J sile properties and chemical composition comparisons, it is concluded that the PIFRAC2 CF8 heat numbers listed in Table 3-5 conservatively represent: the installed' cast stainless steel. Using the operating temperature stress-nrain data plotted in Figure 3-2 and the tensile properties added to Table 3-7 for the aged and unaged test .y specimen, the mean'and lower bound stress-strain data are identified. The lower bound stress-strain data was determined to come from test specimen CF8A-7C of Heat No. 6, which is not thermally aged. The mean stress-strain data, chosen from only the thermally aged test specimen, was determined to come from test specimen Il-29L of Heat No. 8. The mean stress-strain data is not changed from the previous' evaluation and, therefore, the Leakage. Size Crack Evaluation is not revised. Figure 3-3 was amended to include the J da test results at 550*F for both the thermally aged and unaged test specimen. From this-figure, test speci- ^ men 691-3T of Heat No. 70 was determined to have the lower bound toughness properties. This test specimen is SA351 CF3 material which was thermally. aged at 842*F for 3000 hours. Although the CF8 test material was determined. s to conservatively represent the installed material, the CF3 test specimen was chosen for added conservatism. Page 1 of 19

.t 4 - The chosen lower bound stress-strain data for test speciman CF8A-7C was curve fit using the Ramberg-Osgood relation. The curve fit constants are presented in revised Table 3-9, and the Ramberg-Osgood relation plotted with the test data in Figure 3-6. The chosen lower bound toughness data for test specimen 691-3T of Heat No. 70 was curve fit using the bifunc-tional requirements of FLET. The curve fit constants are presented in revised Table 3-10, and the curve fit. plotted the test data in Figure 3-7. The results of the revised J-Tearing evaluation, using the lower bound tensile properties of an unaged test specimen coupled with the lower bound -toughness properties of a thermally aged test specimen, are presented in' revised Table 7-3. The results demonstrate the required crack size 4argin of 2 and load margin of 1.4 are met. Additional

References:

Tensile and J-R Curve Characterization of Thermally Aged Cast Stainless Steels, NUREG/CR-5024, September 1988 Long-Term Embrittlement of Cast Duplex Stainless Steels in LWR System, NUREG/CR-4744, Vol.1 Semiannual Report October 1985-March 1986, September 1986; Vol. 2 Semiannual Report April-September 1987, August 1989 i Page 2 of 19

em ep~ )g;i.;','1 1 y w, u ,? .s.,- s s u s. ,,.c. o w,,, 1 1 x o - a-m'y ~uw,-

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',I ' ::;p ~ ;;, : 7.; m .nu. y - .,,.;-.r - m .s .s N 1., 4 i; e- "if*6 J. ; ,1 Materia 1LTest Data in. I ~ ,, co a-l < M i:Tablel3-5 '(Continued)i PIFRAC2-O.,,* v,. ~ i?. J s 1 Material ' i Heat [: a;4 '

SpecificationJ No. -

a y a J 5 A. . i 3 ~ y@f. . Cast Metal:D SA351 CF8A' -6' SA351 CF3 - ' 7 :~ SA351 CF3

8 AGED-c

~

SA351 CF8 -

63: 1 ,y.. e /SA351? CF8 ' 64' . AGED: gp po 2 SA351;CF8 65-AGED -SA351LCF8= 66: ' AGED 4 ,s A SA351,CF3 - ' 67; 4 3,. - SA351.- CF3

-68 c

' AGED =" n* SA351 CF3-

69 iAGED' m

< AGED: .o SA351 CF3 - 70' F.h ,.k2 { p5.' N r 4 + 4 I y y L, y ,%z )* 4s G b \\ 1 p ,l f; y;;; y j g -:t s;s -a i 3 o,0j 4 ( .y -y q 4

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^ t q w j ns ' Page 3 of'..19 ' wa O / (- ( -y jjjl4;[., r l

~- e -PIFRAC2 Base Metal Tensile Table 346 (Continued) Properties at Room Temperature Cast Stainless Steel: Heat No. Spec. ID UTS (psi) YS (psi) 6 CF8A-6C 89659 44250-6 CF8A-6L 80943 38840 63 683-40 76868 39754 63 683-41 75011 40522 67 693-40 87890 -40435 67 683-41 84641 39652 Note: 1) Heats'64, 65, 66 are the same as heat 63, material SA 351 CF8, but have been thermally aged. 2) Heats 68, 69, 70 are the same as-heat 67, material SA 351 CF3, but-have been thermally aged. 3) Heat 6: SA 351 CF8A e4) Heats 7, 8': SA 351 CF3 5) Heat 21: SA 351 CF8 was water quenched from 1025 C. Since'the tensile properties of heat 21 do' not meet the minimum yield and ultimate tensile-strength required by the ASME Section II material specification, this heat will not be 4 considered in the material selection. l Page 4 of 19

g-, I c s s PIFRAC2 Cast Stainless Steel Tensile ' Table 3.7 (Continued) Properties at Operating Temperature i '~ Cast Stainless Steel: - Heat No. Spec. ID UTS (psi) YS (psi) Flow (psi) 6 CF8A-2L 62567-21711 42139 6 CF8A-7C 64772 21856~ -43314 l 6 CF8A-8C 68035 21900 44967 63 683-42 57622 .23597-40609 63 684-40 59797-22683 41240 3 64 - 682-27 65613 26846 46229 65 682-18 67165 23394 45279 66 681-6 70602 24612 47607 66 682 73068 27788 50428 67 693-42 61044 -27672 -44358 67 694-40 60479 25671 43075 68 692-27 65453 25105 45279-69 692-18 64511 23771 44141 70 691-6 69572 25772 47672 I 70 692-9 69195 25482 47338-Page 5 of 19

n l'r 'e:i ^ FLET Tensile Material-

Table'3-9 (Revised).

Properties-l i Lower Bound Cast Stainless Steel Test Specimen: CF8A-7C Heat No. 6.SA-351CF8A Unaged. o = 21856 psi g ff=0.000789 a = 4.850431 I n-=-6.586406 6 ) E = 27.67 x 10 ] a = 64772 psi u u= 0.3 l l j i i 1 I \\ ic i f Page 6 of 19

$ ff -- : g. a FLET Fracture Toughness Table 3-10 (Revised) Material Properties Lower Bound Cast Stainless Steel from PIFRAC2: Test Specimen: 691-3T Heat No. 70 SA-351 CF3 Aged 2 1858 in-lb/in J = ic 6315 c = 0.3296 m = a) 3794 = -2445 a = 2 0.09452 a = 3 d a transition = 0.06336 Page 7 of 19

$,h pt ' "J' c i v.+

:,V n {C

. ' av '* ~ c l-av 1 &v' ] i' c- <tw A:l Cast Stainless Steel SA-351 e n,' ip< Table 3-11 ( Added) CHEMICAL COMPOSITION COMPARISON r

. i -:

'i Ferrite s" Co N 4%- GRADE. C Mn P S Si Ni Cr Mc sw i ... INSTALLED CF8A ' Low .04 .45 .021 0.005 1.11 8.69 19.54 12.4. .09-High .6 .80.033 0.010 1.3 9.36 20.35 0 .11 17.0; 'i t.l b? A,t. LP! FRAC 2 CF3 Low .02 .48.01 .010 .83 8.43 20.23.45 .03 17.7: i y .03' 23.6' ) High .02 .69. 03 .010 1.20 8.71 20.49 r,'"~ : a.. Ip41f i g 's 4 1 . 06 : 23.4 CF8-Low- .05 .67 .02 .010 1.13 8.08 20.85

High 405

.67 .02 .010 1.13 8.08 20.85 06 23.4 y d 17.4' CF8A Low .06* .66.02 .02 1.17 8.58 20.42 . 0.07-0.07 ' - 17.4 J] - High .06 .66' .02 .02: 1.17 8.58 20.42 EPRI CF8M .045.76 22 .001 1.16 10.1 :20.7-2.59 .038 20.0- .i L. vs D t JA 1 ) f 4 i l-l E a m 1 u Page 8 of 19 l a; t ,<o 4 i;.

.w i AMENDED FIGURE 3-2 AGED ~ CAST STAINLESS STEEL TRUE STRESS STRAIN CURVES SA-351 TYPE CF3 550 F 60000 - n Cn O_ v ~ 40000 - u W W W~. 1 ~_ l W 20000 - !Ll 1.ABEL REF. SPECJD. LABEL REF. SPECJD. 3 $ f A 8 11-28L E 69 692-t8 / B 8 Il-29L F 70 891-6 I C 8 12-20L G 70 692-9 D 68 692-27 o 0.0000 0.0050 0.0100 0.0150 0.0200 ~~RU E S-~RAI N Page 9 of 19

= AMENDED FIGURE 3-2 AGED CAST STAINLESS. STEEL TRUE STRESS STRAIN CURVES SA-351 TYPE CF8 550 F 60000 - n (f) l Q_ ~ v 40000_ g W ~ LtJ T l W 20000 - i L1J uaB REF. SPEC.lD. J [ A 64 682-27 e es Sea-is C 66 681-6 l D 66 682-9 ~ 0 0.0000 0.0050 0.0100

• 0.0150 0.0200

-RUi S-RAN Page 10 of 19 . _. - ~, .... ~..

+ .. ~ e ~ l l l WENDED i FIGURE 3-2 UNAGED CAST STAINLESS STEEL TRUE STRESS STRAIN CURVES l SA -351 TYPE CF3 550 F 60000 - l 1 n (0 O_ v t 40000_ 1 g W [LJ M CE ~ F._ W t 20000 - [_LJ LABEL REF. SPEC.lD. I ~ } A' 7 12-3L ~ 8 7 12-6L E C 7 13C-2L ~ -i ~ 0 0.0000 0.0 050-0.0100 0.0150 0.0200- - RU E. S~ RAIN e 9e n er '9

~ 2; .~ AMENDED ~ FIGURE 3-2 UNAGED CAST STAINLESS STEEL TRUE STRESS STRAIN CURVES SA-351 TYPE :CF3, CF8A AND CF8 550 F 6oooo - A ~ cn Q-v ~ 40000 (/) g Ld c 20000 - LU veo. acr. secC.e. uset ner. secC.o. ] ~ A 63 683-42 E 6 CF8A-2L ~ B 63 684-40 F 6 CF8A-7C C 67 093-42 G 6 CF8A-8C l D 67 694-40 1 r t ~ o ..........i......... $o 0.0200 O. coco 0.0050 0.0100 0.01 h Page 12 of 19 ~

u \\ L i AMENDED FIGURE 3-3 AGED l CAST STAINLESS STEEL J-R PIFRAC ^~ 12000-J l t 8000-n N C i N 4 in _O o l 4000 - c g ._v t ~ USEL REF. SPECJD. AGED A 8 11-SLC-IO YES y 7 B 8 12-4LC-OO YES C 68 601-6T YES D 69 091-ST YES E 70 991-3T YES ~ I L .0 ......i......... l 0.0150 0.1150 0.2150 O.3150 0.4150 Je a C.in Page 13 of 19 =. = =

L MENDED ~ FIGURE 3-3 AGED CAST STAINLESS STEEL J-R PlFRAC SA-351 TYPE CF8 550 F 12000 - 1 i i 2 8000 m eq C z A \\ to / ~ O [~ 4000 - C v i usa. ner. secc.c. no l A 64 881-48 YES V l 8 65 081-ST YES c se e81-37 YES t 0 0.0150 0.1150 0.2150 0.3150 0.4150 (in) Jel a .Page 14 of 19 =.

l-l A'iENDED FIGURE 3-3 UNAGED CAST STAINLESS STEEL J-R PlFRAC SA-351 TYPE CF3, CF8A AND CF8 550 F 12000 - i t I W 2 ~ ^ 8000 A 1 CN C i \\ w _o I t 4000 - c ~ LABEL REF. SPEC 10. AGED LABEL REF. SPEC.lD. AGED A 6 CFhA-2CL NO E 6 CF84-6CL NO 4 B 6 CF8A-2CR NO F 6 CF8A-6LC NO V C 6 CF8A-2LR NO G 63 683-7T NO 4 _) D 6 CF8A-5LC. NO H 67 693-7T NO j 4 0 .........i ..........i..........i........ O.0150 0.1150 0.2150 0.3150 0.4150 i 3e -O (In) Page 15 of 19 -, _ _ ~. _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ - _ - _ _ _ _ _ _. _ _. _ - _. _

l

1. iENDED FIGURE 3-3' UNAGED l

CAST STAINLESS STEEL J-R PIFRAC l SA-351 TYPE CF3 550 F l 12000 - l l l ~ 2 8000 n eq c \\ <n i _O I l 4000 - c U t.ABEL REF. SPEC.lD. AGED ~ A 7 11-2LC-ID NO i B 7 12-2LC-CD NO T J ~ ~ 0' ..........i.. .......i.......... o.0150 0.1150 0.2150-0.3150 0.4150 Je a in Page 16 of 19 - m- = - -

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Stainless Steel PIFRAC2 1 , T'able 7-3 '(Revised) - P4 il s

-?...

? ce t f y[x - g t s t-. m Assessment Instabiliti' ] , s in 1b in-1b ini -g i n';- T 1 T in ' ?.: < a w. . Crack 512e Nargin of 2: 3000.0 60.0-3370.5 - - 66.084 l n i yr p r. -y -f ;: {h y .. l I 1 It '

Load Margin of 1.4

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[A. l l Response for Question'2: The pressure loading was included in the crack stability evaluations, l ~ however. it was not reported in Tables 4-1 thrcugh 4-3.. These tables will be revised to add the pressure loads which were used in the crack stability evaluations. The pressure terms to be added are pre-sented below. f ~ Table 4-1 f RHR 12" tn = 1.125" Pressure 2287 psia Table 4-2 SI 10" tn = 1.0" Pressure 2332 psia SI 10" tn = 0.365" Pressure 700-psia i Table 4-3 RC Bypass 8"' tn = 0.906" Pressure 2332 psia -These pressure loads were included in' the master curve and J-Tearing crack stability evalustions presented in Chapter 7. The Ma term presented in Tacles 4-1.through 4-3 is the torsional moment. Less than 4% reduction in the resultant moment occurs when the torsional component is removed from the resultant moment. Since this change is very small, it will not significantly change the crack stsbility margins which are conservative by including the'slightly higher resultant moments. There-fore, the master curve and J-Tearing crack stability evaluations have not been revised to remove the torsional component. i e Page 1 of 1 I (

o14 q

l Response for Question 3: After a thorough review of the critical crack length calculations, it was determined that a discrepancy in the Z-factor for weld metal occurrcJ hy using the piping nominal diameter instead of the outside j diameter. i i The critical. crack lengths and margins for circumferential crack size and load margins have been recalculated using the modified Z-factors for weld metal. The change in critical crack. lengths and. margin factors is less than 1%.for all weld metal evaluations and the required margins of 2 on crack size, as well as the required margin of 1.4 on load, have been met. Tables 7-1-and 7-2.have been revised to incorporate the new .l critical crack lengths and. margins. j j 1 I 1 I i 1 t t Page 1 of 3, s m

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c

~ Margin [.,0 System ASI 'CCL' LSC-( CCL/ LSC.>2.00)- h f ( .l RHR . ~. i .sf. N,,e. !.

Weld, 26,197 12.584.

5.652

2.23?

Base'Hetal .17,455 13.990 -5.652- ' 2.48 ; N-gl . g. g i. a I

SI "t=1. 000"-

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.RC Bypass.

gi li [q 15.342 10'006 3.6991 2.71 ~v Base Metal

Weld 30,021 7.589.

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1 System' ASI-CCL-

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S!"t=0.365" j

r Base _ Metal-18,738 12.771 5.142 2.48-i ; t Weld 29,163 10.952' 5;142 2.13 h 1 5 RC Bypass- 'y + b si b '? ~ Base' Metal -21.479 8.734 3.699 2.36: ? 1 SAW 42,030 5.499 3.699

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aoo O May 3, 1990 Res,ponse for Question 4: The' Applicant complies with the regulatory position in Regulatory i Guide 1 446 with the following clarifications keyed to paragraph number! the regulatory position: 1. identified leak sources are piped to either the RC drain tank or a miscel)aneous drain tank which is utilized for this purpose only. Temperature of selected drain lines is monitored to identify leaks. Tank inventories are monitored. Temperature monitoring is more sensitive to small leaks than flow rate. monitoring specified in the Position. The containment floor drain and reactor cavity leak detection measure-ment is recorded and an alarn is provided with a setpoint at 1 gpm. 2. Unidentified leak sources are monitored to as accurate an equivalent flow rate as is practicable. Accuracy of the containment floor drain and reactor cavity leak detection is approximately + 0.095 gpm and + 0.063 gpm, respectively. The sensitivity of the containment atmosphere particulate and the gaseous radioactivity monitors are given in 5. below. These leak detection systems will respond to a 1 gpm leak, or equivalent, in one hour or less. 3. The following leak detection systems have been provided: Identified Sources u a. RC drain tank level indication and temperature indication of selacted inlet lines, or b. miscellaneous drain tank level indication and temperature indication of selected inlet lines. Unidentified Sources Containment floor drain and reactor cavity flow and a. sump level, b. containment atmosphere particulate radioactivity monitoring, and containment gaseous radioactivity monitoring. c. 4. Intersystem leakage between primary and secondary plant is monitored via air ejector off-gas radiation monitors. Also, pressurizer and makeup tank levels are monitored to yield total roactor coolant leakage. Page 1 of 2

~ . < av > o / s, i Page 2 R$sponse for Question 4: ) f 4 5. Leak detector sensitivity is designed as low as practicable. The containment floor drain and the reactor cavity leak detection measurements have an accuracy of approximately + 0.63%. These.loak detection systems will respond to and alarm a 1 gpm leak in one hour or less. The containment atmosphere particulate and gaseous J radioactivity monitors are located in a low back-ground area and are provided with background particulatechannelis10ggsitivityfortheair subtraction. Instrument s

1. Ci/cc.

The instru-ment sensitivity of the gaseous radioactivity channel is 10'6 # Ci/cc. The sensitivity of these two channels should enable detecting a leakage rate, or its equivalent, of 1 9pm in less than 1 hour. The containment floor drain and the reactor cavity 6. leak detection system are designed to remain functional following a seismic (SSE) event. The airborne particulate radioactivity monitoring system is scismically supported. 7. Convorsions to common leakage equivalent are supplied to operators wherever possible. Conversions to a common leakage equivalent are not possible in all In these cases, the system is inter.ded cases. primarily for localization or identification of a leak with no quantitative implications. Leakage is recorded in the control room, and a high leakage alarm with a setpoint of 1 gpm is also provided. 8. Comply. 9. Comply. -....m. ,,,,,,}}