ML20215G712
| ML20215G712 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/17/1972 |
| From: | DRESSER INDUSTRIES, INC. |
| To: | |
| Shared Package | |
| ML20215G377 | List:
|
| References | |
| FOIA-84-744, FOIA-87-744 SP-52-166, NUDOCS 8706230360 | |
| Download: ML20215G712 (123) | |
Text
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.e _ m gj g SUDJECT: ULTRASONIC INSPECTION PROCEDURE TOR TORGINGS, BARS, PLATES, BOLTING, PIPE, TUBING 6 PITTINGS - IN,
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- . ACCORDANCE WITH ASME CODE POR PUMPS S VALVES TOR NUCLEAR ~
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.' P_URPOSE
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.U4 1.1 This specification is established to provide a nonder,tructive exa:r.ination procedure by the ultrasonic inspection method in
)3 accordance with ASME Code Pumps & Valves for Nuc1 car Power.
- 2. SCOPE ,
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~.? f 2.1 The precedures are set.forth for the detection of defects by @ the ultrasonic method of examination of plates, forgings, ; % bars, studs and bolts, pipe, tubing and fittings.. The - I f) procedures apply to carbon steel or alley stee1~ compositions. (LQ A ~ ~ " 3. REFERENCES. . I . 3.1 ASTM Suecifications. , ASTM E-317-68 - Evaluating Performance Characteristics - T]el of Pulse-Echo Ultrasonic Testing Systems. u p, j.j ASTM E-114-63 - Ultrcsonic Testing by the Reflection Method M. Using Pulsed Longitudinal Waves Induced 8 ', by Direct Contact. .,7.: . $ . ASTM A-388-67 - Recommende'd Practices for' Ultrasonic g Testing and Inspection of Heavy Steel Torg:Lngs . - c.j , .n 2 ASTM A-435-67 - Method and Spec, fo'r Longitudinal Wave - p Ultrasonic Inspection of Steel Plates j Wj
. for Pressure Vessels. ,
m h ASTM A-578-68 - Spec. for Lorigitud *,nal Wave Ultrasonic l L,] y - Inspection of Plain & Clad Steel Plates ; {'j . for,Special Applications, j
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ASTM A-577-68 - Spec. for Ultrasonic Shear Uave
?l.1 ' Inspection of Steel Plates v - - - ..
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'9) 3.1 (Continued) - ~ ,d . ASTM E-213 Ultraconic Inspection of Meta 1 Pipe 7 & Tubing for Longitudinal Discontinuities. - [;}j ASTM E-214 I:::mersed Ultrasonic Testing by the t1 - Ref.lection Method Using Pulsed Longitudinal 7,!.] Waves. , D;.y . ASTM E-273 Ultrasonic Inspection of Longitudinal S Eh,j ,
. . Spiral Wolds of Welded Pipe & Tubing {
MD { ' y; 3.2 CODES - l ?, ;, . l lt ASME Code Pumps & Valves for Nuclear Power 6 Latest Addenda j $q ,
. ASME Code Section III Huclear Vessels -
4 UN ASME Code Section V Nondestructive Examination -
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J.; 4 QUALIFICATION OF FERSONNEL , 4.1 All ultrasonic, inspection operators shall be performed
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- 5. TIME OF EXAMINATION -
Lp , lp 5.1 Ultrasonic testing, unless othernse specified, shall p!j be performed af cr any heat treatments required by the i materials specification. $n 6 . EQUIPMENT hy,; !, . . Foi 6.1 Ultrasonic excmination with an ultrasonic, pulsed re-p] flection - type system generating frequencies over the .a range of 1MHz to SMHz refer - ASTM E-317-60
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M - 6.2 - Couplants suitable for ultrasonic examination shall W have good wetting characteristics. - }, %g.. . 6.2.1 The coup'lant shall be used betwaen the transducer and
. test surface. .
m,a - @ .] 6.2.2 Couplants shall have no adv.rac affect on the material i, being examined. Por austenitic materials, couplant chall have - a low chloride content. iM@l Nj 6.2.3 Couplantc'used will, depend upon Tccting conditions - oils, light greases, glycerin and water are suitabic. i .d The addition of wetting a;;ents is recommended. 1 A . INST. NO. SP-52-16G
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Fi4CE.3.0.10. PiGES 4 s . r .: To11owing ultrasonic cxamination the couplant .shall be 4 - 6.2.4 (' , thoroughly removed. . &) 7,'. 7. .. EQUIPMEllT CA}IERATI0li m . . $,0 7.1 The proper functioning of the examination equipment shall be checked and the oquipment shall be eclibrated by %g i the uso of the reference specimens as a minimum. . N.$ - 1. At th'e bcg. inning of'ench production run of a R$ given size and thickness of a given material. 1!.j g '
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- 3. At the end of the production run.
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testing equipment is not functioning properly, - T all of.the product that has'been tested since the last valid equipment calibration shail be . .m h , re-examined. -
..y 8 SURFACE PREPARATI0l!
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- )* . 8.1 Surfaces of material .being ultrasonic exam'ined shall be free of gross dirt, scale, or.,other extraneous d surface contamination.
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- M 8.2 Actual material surfaces'shall comply with .the ,
followings yA t .. 'T S.2.1 Torrtings - Torged surfaces shall be free of loose dirt,
.d sEale . Cleaning with a flexible wheci sander to
[% improve the surface is permitted. m 8.2.2 Machined or' Ground Parts - A surface finish of 125 to 8l
- s. - FEIS a.s satisractory.
lr$ 8.2.3 Hot Rolled Surfacas_ - Require the removal of any loose acherent scale or foreign material. <> 3 8.264 Cold Rolled Surfaces - If free from dirt and rust are %[.h suiracle as follco. , 9 METHOD OF TEST _ d>.m r.j\ ., -) 9.1 The following are to be usod as a guide for producing
'j 'well defined interpretable results:
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' - INST. NO.
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..u P/P 9.1.1 . Material'shall be examined by the direct contact method. -
{ ' p . Equipment shall be of the pulse echo type using 'the fi] normal straight beam technique and/or angle beam technique g1., as require,d. . l
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64 9.1.2 To assure complete coverage of the examined material, S. each pass. of the search unit shall overlap a minimum of 15% of the transducer width. Scanning shall be at a rate of 6" per second, max. %(?) _f , 9.1.3 After calibration of the testing. instruments using the Fj reference specimen, the' transducer shall be placed in 94 an essentially defect free area of the material being ' $ examined. The scope shall be checked for icek of or W reduction in back reflection. Should either of these Q indications occu.r, the instrumentation, couplant and . Dp material surface shall be rechecked until satisfactory Il results are obtained. ,
.ty . . -E 9.1.4 The reference specimen shall be of the same' thickness f, and nominal composition as the material being examined.
,'%.O ~10. . PROCED RE . 10.1 Plates *
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$.] 10.1.1 The transducer shall be Vto 'l-1/8 inch diameter or . W 1 inch square.
. approval must be obtained'for use.
If smaller transduc6rs are required, f." / G.; y ' Examination shall be made from either major surface s ,u. V a
- N 10.1.2 Z of the plate. The ' entire .'3C'lR% surf ace shall be / W -
covered by moving the search unit parallel paths .M
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- with not le ss than 15 % over. lap. Scanning shall be iii at the rate of 6 inches per second,i max. -
j ' @n 10.1.3 The surface to be exa' mined shall be suf ficiently j g.d clean and smooth to maintain a reference back reflection als from the opposite side of the plate of at leas.t 50% /'/E of full scale during scanning. q i% n A nominal test frequency of 2-1/4 mhz shall be used i /Q 10.1.4 (y although this' frequency may be adjusted to give a clear, g C A easily interpreted trace pattern. 10.1.5 The examination shall be conducted with a frequency - j7'} and instrument adjustment that will produce a minimum
'D of 50 to a maximum of 75 percent of full scale reference il L j
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- 11. TORGINGS E BARS (OTHER THAN USED POR BOLTING) -- -
1.l . 5,, 11.1 Material is to be ' examined 'throughout its entire volume in accordance with the 'ormc1 straight beam 6.i In addition, ring forgings and all technique. NJ
;G hollow forrings shall be exe ined by'the angic Forgings beam i technique in the circ'umfere tial direction. !
M* . may be examined by alternde methods utilizing distance
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amplitude corrections pr hided acceptance standards ' lg " are equivalent to those' , isted in Paragraph 16
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l ined by the Pulse-Echo Longitudinal !O 11.1.1 Material shall be .exa/ons a't approximately right angles 9 beam from two directi
. to each other with a/ searc$ unit of one inch square or 1 to
'3 1-1/8 inch diameter /and a frequency range of 2-1/4 mhz. H .The instrument shal.1 "?- reflection is 75 15% ofbe screen setisoheight that the when firstthe back scarch unit is placed on an indication free area of the - material being examined. . ,i . 11.2 Hollow Forgings y (a) Examine from circumferential su
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(b) Examinefromonefaceofsurface[pu
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s the axis of the circ 6mferentiri v surface . A Dise Forcings m '11.3 -
" s LM N (a) Examine from one flat surf p. '
') / \ (b) Ex' amine from th circumferential surface.
,' ^ 'h 11.4 Angle beam examination of material *shall be conducted d er.
with { 1: s-y ' # - '- 1 A nominal (45Ef3. frequeBhall D .1!.f_.".Z3h.45* be used. Calibration trans uc c'
f /O Gf Distru6ents shall be made with a squarc notched referenec6 M specimen with the following notch dimensions; O
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l.f (a) Depth equal to the lessor of 3/8 in,ch or 3% .:a of the nominal section thichness.
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ty) 11.5, Surfaces shall be clean and free from 1/ water, N j.l ' loose scalo or any material detrimental t results. .. acdurate u ~ Tlat surface scarch units may be used wi 5 accuracy ' ij 11.6 h 'l DJ for parts r,reater than four (4) inches They may also be used for smaller diame%iar.eter. tere provided f? ' consideraiton is given to the sharp cufvature and M . results .can be interpreted as being accurate. If y this assurcnc4 is not attained, then specially
"* 1 curved search units shall be made available. .i) BOLTS & STUDS (GREATER THAN 2-INCH' DIAMETER UP TO 4-INCH J! 12. .
(.jj D1hM m R.) - 'UI 12.1 All bolts and studs shall be examined prior to . Surface finish shall not excbed 250 RMS. ' ' il ' threading. M ' Material shall be examined' by the' straight beam d 12.2 Z technique using a radial scan.' The size of the search ,'
.>./ g unit shall not exceed one' (1) square inch. If results-V-
- cannot be interpreted as being accurate, then a 1:.!
. specially curved search unit shall be made available. .
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- l;N 12.3 A nomina 5. frequency of 2-1/4 mhm shall be.used. .
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' .. 4 - 12.4' The examination shall be conducted with an instrument /d adjustment that will produce a first back reflection N
. .that is 75 to 90 percent of full screen height. .
Mt " 9, 13. PIPE. TUBING & TITTINGS , u Materiab.shallbeincpectedbyanymethe which will 13.1 produce results which are easily interpre. b1;e. j:, ..
. t Material shall be examin.ed over its full lent.th at W 13.2 Yi- a frequency sufficient to detcet defect indications ;.( equal to or larger than those in the referene,e 0 . spec s.me n . ,
M; g 13.3 Surface's to be examined sh'all be free of oil, d,irt , i etc. , which may produce erroneous results. - ]d . 13.4 13.4.1 Reference Soc'cimen Eference specimen shall be o'f same' nominal MP. . M'
. diameter, thickness and compocition and heat treated condition as the product be'ing ',-
P. , a examined. The separation between standard defectsK
} placed in the cams reference specimen chall be not lesT than twice the icngth of the sensinn .: ,
j unit of the inspection equipment, ll:ST. ND. l
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l %{a a l . - _., 33 13.4.2 Standard defects shall be axial grooves on the outside eami [ly *
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(a) Length not greater than one (1) inch. , i l 1 . n O (b) Width no't greater than '1/16 inch. , d . (c) Depth not greator than the. larger of 0.004 inch! %' or 5 percent of th.e wall thickness.
- f,a 14 EViLUATION AND ACCEPTANCE ,
Kg a i 14.1 Evaluation
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p h} 14.1.1 All defects are to be . evaluated on a re$cetion basis. l J 9 14.1.2 Primary consideration should be given to the two '2 ) . following facts: . ,.m,- ! .
'j (I (a) As' the fre'quency is incre'ased the sensiti ityer of the technique also increases, i.e., sma-defect indications can be located.
d( , .A (b) Ultrasonic waves are 'of low energy capacity . L
' and .as the frequency increases,,the rate of s .i
- attentuation also increases. -
' d. . q 15.' - ACCEPTANCE ,
. ,The following relevant indications are Onacceptable: [(
..u . $ 15.1 Plates d Within a circle whose diameter is 3-inch or oneone
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- ,j 15.1.1 of the plate thickness, whichever is greates /
q 7 or more discontinuities
- which produce a '
total loss of back reflection accompanied by @ continuous indications on the same' plane. JE , ~,4 - a 16. TORGINGS AND BARS (OTHER THAN USED TOR BOLTING) s@} The following relevant indications are unacceptable: 16.1 d u 16.1.1 Upon straight beam examiiution, n or more reflectors ied by a compiere d which produce indications accom;,
- d. loss of back reflection not asso 'ated with or attributable to the geometric cof iguration. Complete
' '. ] Y loss in back reflection is assucied when the back 0, ' refleetion falls below S percent 'of full screen height. ' l l
'l .; ust. a. t SP-52-16G l q
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M 16.2 Upon angle beam examination, one or more reflectors i W which produce indications exceeding in amplitude . A 3.? . the indicction from the calibration notch. - l
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g - 17 BOLTS AND STUDS (GREATER THAN 2.IUCH DIAMETER to 4-1 CH l O LTHiTfio . Q Any discontinuity.which cauces an indication in excess g* , . L 17.1 U; of 200 of the height of the first back reflection or ' any discontinuity which prevents the production of a first
@? .. back reflection of 50% of the .calibrction amplitude is not b '
acceptable. ii 18 PIPE. TUBING S. TITTINGS $d Defects which produce reflections in excess of the dj 18.1 reference standards are unacceptable. g s, sh . 19. - REPAIRS - N., 19.1 Renoval of defects, subsequent repairs and re-examinations shall be made only with specific engineering approval f! J unless otherwise specified on the drawing or appli-cable controlling specification.
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hk [ 2 0.' TEST SYMBOL, REPORT, CERTITICATIO1! - <8,,l Parts having " finished" external surfaces that' are l C) 20.1 non-critical shall be low stress stamped or engraved \,, Ud
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o.4 with the Symbol "UT" when tested and. found to be s 7! - acceptable by this specification.-- . jilt y". 21.- REPORT . :p . d 21.1 The testing' agency shall report in duplicate the l@ essential elements of the Itest asThis conducted and hj LO forward same to the purchaser. shall include: y, id s Location of defects, how repaired and re sults of 21.2 A a) @ final examination of that repair area. Defects which are determined to be within ac'ceptable limits j C1 C need not be reported.
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h 21.3 'A , description of, equipment and procedure shall include: ' ,$8 ~ '. M - (a) Instrument, nams model number. , n s p'd N (b) Type and size transducer. (i Cf (c.) Test frequency. - v ,. . s.y , ' j : I aQ ' ' l S . IllST. NO. E' } SP-52-166ai i _ _ _ . . I .
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Method of test lS:) (d ) .
- U (e ) Couplant * *
(f ), Surfcce" finish O + Y (g ) The surfa'ce or surfaces from which the test shallf - '
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f1 be performed. .
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(h) . Stage of manufacture when tested
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g4 . (i:) . Description of calibration block (Size, y - material, basic calibration reflectors) and d ih calibration method. - Equipment initial calibration date and
- j (j) subcequent rochecks.
s . .g - 9:3 CERTITICATION 22. s .
- )s
- j 22.1 All acceptable material shall be supported by a ,
, (j certified statement, related to a specific part
'h number and/or heat identification symbol, to the .
effect that 'such material has been ultrascrdcally-fmfi X' h d inspected in accordance with thismz;.= 7~Specifi- * (1 , cation and has been found to exhibit no unacceptable
.d;j indication's of defec,ts.
A * . gj 23 RECORDS A 23.1 Records shall be maintained as per IX-225 of ASME, Q Section III, " Nuclear Vessels". [$;' gj ' i 4 Personnel record of. Qualification. - WITNESS 24 . d 24.1 48 hour notification, exclusive of Sundays and . , , Holidays will be given prior to the performance of
' ,~
,y lV tests and inspections requiring witnocs by customer. /. . .a - 3 . 84,
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g Para. 6.2.2 - Added "For custenitic materials , coup ~lant $)jg/17/72 shall have a low chloride coriter.t". Para. 8.2.2 "250 RMS" was "450 RiiS". h 3,,7, .p + e' .i; W r,r . Q' .1 To . g.: ,,,q,)s . a#. ,} , - ,,..=.1: '3 g;., # . ,9... . . . . . e,. a. . >e
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%( ' . . M ,, I i e$b . . g t' ' / hg ENGINEERING DEPARTMENT g 4._,.,_ , . . -. , - , - - - . - s h - ULTRASONIC EXAMINATION QF SAFETY. APPROVED FOR b =
'. VALVE 'D) M
- YOKE Ge b0 C0]jSTRUCTiON C,] 1. PURPOSE fLifaVC).ShS)2fZhS p k
/g;/4. ENGINEERING DEPARTMEN,T/
A og This procedure outlines t e met7re@mo od and etalis . for 0 x< Eejd.pifa ' 9 of Yoke Rods on 3707 RAX 6-21 Safety Valves. \, /
- e
?[, . 2 SCOPE . . g , The ultrasonic procedure is designed to locate surface 'and subsurface * ! .5' discontinuities lying in a longitudinal direction, such as piping, stringers, y cracks and seams.
- j. * '-
depth a'nd length ofThis procedure detected provides the neans of measuring the . , discontinulties. , ,l. ? 3. REFERENCES i (
- 3. T' ASTM E-317-68 Evaluating performance characteristics of Pblse-Echo k
T}1 . . e Ultrasonic Testing Systems. - "j ASTM ' E-114-63 ' Ultrasonic . Testing 'by the reflection meth'od using
- j
,j . Pulsed Longitudinal Waves Induced by direct contact. O i SP-52-166 Ultrasonic inspection procedure for forging, bers, l plates, bolting, pipe, tubing, fittings. In accordance,with ASME code for pumps and valves for Nuclear Power. 1 . j.a .
. DC-663050-27-2 Dresser Industries a ,._ ., .d "
- h. CODES 1 *
'4.1 ISHE Code Pumps and Valves for Nuclear Power an'd Latest Addenda.
Mj .
' ASME Code Section lli Nuclear Vessels. * ,1 41 -
ASME Code Section V Nondestructive Examination. ~ c. n A G y 21 rah t (',a
- 5. QUAllFICATION OF PERSONNEL
. $} 'ey 's M. ;j 5.1 All ultrasonic inspection operations shall be performed by personnel qualified to SNT-TC-IA, its supplements and appendices.
m . d d
- 6. TIME OF. EXAMINATION
- d 6.1- Ultrasonic testing, unless otherwise specIfled, shall be performed (d
p af ter any heat treatments required by the materials specificatlon. o :' m - 1.1 h'; PREPARED BY ED Y. MARTINDAl.E DATE OF ISSUE 12/M m PAGE ' 0F 6 1- DATE OF REY.
.i APPROVED BY' d M ' l i ~**'ew n , . . . , , , , _ _ _ _ , _ , ,_ ,
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. ENGlHEERING .
ggg SPECIFICATION SPEC.N0. l !G. , . , TW 8711 &[ ,
~
ENGlHEERING DEPARTMENT ES n m k *
- i 7. EQUtPMEwr 4 7.1
- Ultrasonic examination with an ultrasonic, pulsed reflection i
- .h! type system generating frequencies over the range of IMHz to
%, . SMHz refer - ASTM E-317-68. Q - '
\
p - 7.2 couplants suitable for ultrasonic e'xamination shall have good M Ph" -
- wetting characteristics.
7.3 The,couplent shall be used between the transducer and test surface. C'.i ,? p 7.4 couplants shall have no adverse affect on.the material being. R examined. For austenttic materials, couplant shall.have a low chloride content. Q]p p.. , .. ,. - .
"? -
7.5 Couplants used will depend upon testing conditions - olts, light - d* .;d grease, glycerin' and water are suitable. The ad'dition of we~tting. '
- agents is recomended. In any event the'couplant used for the test -
y ., shal) be the same as that used for calibration. , c 9 7" 6 Following alt,rasonic examination the couplant shalb be thoroughly .
,d removed. ,
'9 : J
'8 EQUIPMENT CAI.l8 RATION - .') ,8.1 .The proper functioning of the examination equipment shall be checked" i
(. and the equipment shall be calibrated by the'use of the reference f.;,l ' specimens as a minimum. - -- w Crdc a., At the.beginning of each production run of a given size and thick-Q * -
- ness of a given mater'lal, ,
W . M
- b. After each i hour during the production run. *
%' Q D.
M c. At the end of the productio's run. '3 g . , h)M.3' aw- $ .
- f. d. At any time that malfunction is suspected. If during any check 1] -
It is determined that the testing equipment is not functioning Q properly, all of the proc'uct that has been tested since the last f , valid equipment calibration shali ice re-examined. . m k , , P[1 9. SURFACE PREPARATION d - t] 9.1 Surfaces of material being ultrasonic 1y examined shall be free of . 4 gross dirt, scale, or other extraneous surface contamination that 91 ,. will prevent the detection and evaluation of discontinuity Indications _ $( ~~^ described he,iln. . f 'd PREPARED BY rn v wantiunasr DATE OF ISSUE 12/26 m PAGE.1- 0F 6
..i DATE OF REY.
d! ' APPROYED BY ' # I - d q _ _._.. _ ._... _.. .._ _.._. _ _. .. _ ... _ ._. _ ._.. _ .._
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p..pQngg .~ SPECIFICATION SPEC.N0. @g ' gi g , 8711 [, . ' ENGINEERING DEPARTMENT ES __ 0-241 M 10 METHOD OF TEST , ; N g 10.1 The following are to be used as a guide for producing well 4'y defined interpretable results: ' y . 10.1.1
. Material shall be examined by the direct. contact method.
Equipment shall. be of the pulse echo type using the Qg . normal , straight beam technique and/or angle beam ' $y: . technique as required. 10.1.2 To assure, complete coverage of the exam'ined material, y;. - each pass of the search unit shall overlap a minimum . W:: ' ? ' ' '
'aofrate15% of 6" of per thesecond, transducer max. width. ' scanning shall be at ' '
.z . - Kg , * .
, _10.1.3 Aftec calibration of the testing Instruments .using the -
G.s 7 *
, reference specimen, the transducer shall be placed In .
icy ~ P< e ait essentially defect free area of the material beIng examined. The scope shall be checked for lack of or h . reduction in back , reflection. Should.elther of .these N ' Indications occur, the instrumentation, couplant and *
&. . . material surface shall be rechecked until satisfactory "y results are obtained. Once the examiner has obtained .j -
- satisfactory results he shall observe and* note the - .
- 4) .-
difference required to obtain the same reflected signai
- from the test material as the reference block. If this .
fj +
, difference is more than 21 db he shall add or subtract M .
. this difference when making the final evaluation of . . . . .
[!.i , located indications. x] h q
, 10.I.4 'The reference specimen shall be of the same thickness and y,p) -
nominal composition as the material being examined. j 9.b- *11 -PROCEDURE T. ' ,g -i ! j
,1 .
.3
- W .
i mj 1.1.1. The first examination shall be made with,a longitudinal beam. The "1 ! scan0 shall be performed on the entire exposed length of each rod l
@ at 0 and 90'. The first scan at 00 shall be at the option of the !
examiner. He will then make the second scan at 900 from the first. lQ - Z 11.2 The surface to be examined shall be suffletently clean and smooth h y to maintain a, reference back reflection from the opposite side of the rod of at least 100% of full scale during scanning. . 9 11.3 A nominal test frequency of 2.25 mhz shall be used although etir - !.2) f [,,.;. frequency may be' adjusted to give a clear easily interpreted trace pattern. - - 1 c.q k PREPARED BY 12/2W 6 4 ED. Y. MARTIND%L DATE OF ISSUE PAGE t OF
'i -
DATE OF REY. l 1
' APPROYED B'Y ' .t q___ __ _ . ..... _ _ _ . _ _ _ . __ _. _. ... _ - . . . _ . _. .
u ._.. _ . _ . . . . _ , _ - - _ - . _ - - - . . - -- I
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, ENGlHEERING l',,jj ' glgg SPECIFICATION SPEC.NO.
j,d . . . - M 8711
".0 ' . .
l
- m -
ENGINEERING DEPARTMENT . ES o-2ki n * *
- ,\g . -
$.j 11.4 The examination shall be conducted with a frequenc'y that is i.; capable of producing a minimum of 50 to a maximum of 75 percent 1 4 of full scale reflection from the provided. drilled' hole.in the-
, reference specimen (HWK 001 sketch #1) -
? .! . .
. 11.4,1' Reference level is the gain required to obtain a 75% deflection
- ,$ from the drilled holeg l
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- 5'd 11.5 Scanning shall be performed with a gain setting of at least 6 db 31 -
over the reference level setting. y T .
~ .
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- 11.6 Any indication exceeding 50% of full scale will be investigated.
4 to estain the maximum reflected energy. . w, . 11.7 Any Indication exceeding 100% of full scale "a t the scanning gain O ' will be evaluated for possible rejection. The gain'will be re-M,A4 , duced to reference level. Should tWe indieption exceed .20% of h;j reference level for a length greater than 1" It w!!1 be recorded on Form F-66 and F-67 as applicable; and the valve withheld for J P. C. & E. Jisposition. w
- h. 'j 11.8 Rods containing no reportable indications as des'c ribed in t'aragraph 11.7 will be excepted and returned to storage.
%.4+ a
- p. v g.
12 4, I ANGt.E BEAM EXAMINATION I. '
.y ~d ,
12.1 Angle beam examination will be performed us'ing a 49
- 2' shear
$j wave. The transducer shall be of frequency and size to provide b,T .
satisfactory examination results. The transducer will be provided
;h
.d with a lucite shoe ground to fit the conteur of the rod, surface..
)1l,Ij 12.2 The entire exposed length of each rod shall be examined (circum-g ferentially) 100% from two (2) directions clockwise and counter
,t clockwise.
l1
} - -
PREPARED BY EA.Y- WimF DATE OF ISSUE 12n6m 6 PAGE L OF DATE OF REY. APPROVED BY '/> ~/
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@SS3,5i5S35S.NEOTNbEM.NNM$NESE53fd~*NbkNb'N$$$$ bM y , . , " . .: . g gl ,f l; y,; ENGlHEERING ggggggg SPECIFICATION , SPEC.NO. Ni , .
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- j ENGlHEERING DEPARTMENT ES D-24I E.h,,.,.-
5
.g.F) *
- 12.3 Scanning shall be performed with a gain of at leas.t 6db over _ _ ;]
W4 . reference level. 1 j.y
. .~ j h6 12.4 The reference level shall be set such that'a 75% o'f full. scale 1 Q ,
reflection is received from the slot provided in the referen'ce ' g$; . ' speciesn.(Sketch #1). '
. . i 22.5 Any Indlestion excu d!ng 50% of fuit scale wilf be investigated to obtain the maximum ref! acted energy.
[V:{ , ,'
'j m 12.6
~ will be evaluated for possible rejection. The gain will then Any indication exceeding 100% of fult :cale at the scanning gain d .
be re'duced to refererice level. Should th's indication exceed 201 jE ' of reference level for alongth greater than 1" It will be recorded
- 4 on Form F-66 and F-67 as. applicable and the valve withheld for D3 , P. G. .s E. disposition. ,'(see , Appendix A & B for example of forms,)
, ,( 12.7 Rods containing no reportable Indications will be accepted and f.1 '
returned to storage. . b ,
- 13. REFERENCE SPECIMEN .
./ ,.~
p . . N.Q Q - 13.1 deference ' specimen shall be of same nominal diameter,, thickness p! -
. and composition and heat treated condition as the product being* -
' ?j examined. The separation between standard defects placed in the N same reference specimen shall not be less than twice the length fc , of the sensing unit of the inspection equipment. .- &] f.R '13.2 Standard reflectors shall be axial grooves *on the outside of' g 4] - the rod surface. .
}
S
.(/. , a. Length greater than one.(1) inch.
e - Y b. Septh not g' reater than the larger of 0.004 inch or 5% of the wall thickness. &j ' ' /: 14. REPAIRS . J'
- 14.1 Removal of defects, subsequent repairs and're-examinations shall
- n. be made only with specific engineering approval.
y%.4 icy . b
- A 15.
REPORT , 15.! The testing agency shall report in duplicate the essential elements @$d of the test as conducted and forward same to P. G. & E. , l? y - Yi PREPARED BY FD Y . MARTfNBAfF DATE OF ISSUE imsm PAGE 5 0F i DATE OF REY. l APPROVED BY - Q. . . - _ 7
[Mi$$j32@5 N M E.'9dfiD3gj@@.QRQMNRifjyg.M.yWA
- j. 41 '
81 ' '
- ENGINEERING g SPECIFICATibH SPEC.N0.
W 8711 k4 V' . 0
- ES 0-241 w ENGINEERING DEPARTMENT .
$] . ( I.ocation of defects. Defects 'which ar's deterstned within acceptable j f ,e 15.2 Ilmits need not be reported. >.J to:
. .i p't 15.3 A description of equipment and the procedure used shalt include M .
.all Information included on Form 66 (Appendix A) covering but not j9 limited to the following: .
g . .. . .. d.q a. -instrument, name, model, number. Type transducer.. %@ b. .. %1 ' 9,; . ,c. Test frequency. . 9 g. d. Method of, test. c: f e. Coupladt. , , m . . . . .. g . .h __ _ f. Surface finish.
- The surface or surfaces from which the test shall be' performed.
h , g..
) -
- h. ' Stage of manufacture when tested. .,
,w . . . . [j) I. . Description of caf f bration block (Size, Material," hasic call-bration reflectors) and calibration method. , .$3 ' . . . 3:) 'j . Equipment initial calibration date and subsequent rechecks. n, . 3;j f
.'16. RECDRDS
- e. .
~
16.1 A complete file of: records for all insteriais and it' ems examined to @If/ this procedure shall be maintained.in accordance with M. W. Ke11,ctsg . , quality Assurance Program KFP-16. Q.'J d4 b! Vl l" Q* i ':'w ? VW
- 17. !AcK UP TEST
$:.u!! d$ 17.1 Liquid Penetrant or Magnetic Particle will be used on rods which it are suspected of containing discontinuities which may extend into
- .d ,
threaded areas or to support areas where Ultrasonic Tests indicate 7q discontinuities with lengths in excess of 18' long. , F.4 b Y.3 18 FORMS - 18.1 Appendix A Examination Report F-66. - kif r 18.2 Appendix B Rough Drawing Report F-67. b' . ~ 7.ii 18.3 Appendix C Status R'eport Form F-68. . . ] - y .
-PAGE_ 6 op 6
.y PREPARED BY ED Y. MARTINDAI.E DATE OF ISSUE 12/26n3 APPROVED BY m_.....__ = _ _ _.. ._.._ ... ._ _..._ .- _. .... . _ . _ . . _ . _ . . .
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. SME7Y 'VALV6 3707 W (mil -
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. Appendix "3"
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M.V. KELLOGG COMPANY 1,'r
' ' * *. . . , SHEET Appendix C of
[y:: CONTRACT 22-0-8711-0 . j N.I . .
. ULTRASONIC STAT.US REPORT. . . {
2 . yl1 * ,
.. YOKE RODS EXAMINED TO M.V.
K. PROCEDURE
i '.;. 1 ULTRASONIC
' )
. I
*U YOKF, SERIAL MISC. TAG LPT RESULT -{
^
TEST T,} UNIT NO. RCD NO. NO. INF0 NO. RESULT ACC. 1 RE;f. DATE PERFOREF.D BY
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$ name 1s Harold Hudson.
I am submitting this affidavit f y and volunjarily without any threats, inducements, or coercion, to Mr. s Devine, 7
,D who fs identified himself to me as the Legal Director of the Government Account- .O ability Project of the Institute for Policy Studies. I am submitting this state-f,El -1 ment to evidence my concern over a comprehensive quality assurance (QA) breakdown a
for the work of Pullman Power Products at the Diablo Canyon :luclear Power Plant. Dj There is no possible justification for allowing this nuclear power plant to go critical until the i4uclear Regulatory Conrnission (NRC) confinns the full scope of nl> - QA breakdown; identifies the causes; and monitors completion of a corrective \ h p action program, including a full reinspection of safety-related work at the plant. H.i Iny many instances, the reinspection may be the first legitimate quality control ,1 , M caverage the hardware has had. h.> ; p.i ,, i3 s n 3 I base this conclusion on my four and a half years experience at Diablo . i l
,J Canyon in Pullman's quality assurance / quality control (QC) program, including y.,.d m two and a half years, i through 1982, during which I was the Internal Auditor. The j
\
$!3 basic lesson I li!arned is that the conclusions of a Nuclear Service Corporation M' ,l audit of Pullman are more true today than when first published in 1977--the b 3 m program does not meet the requirements of 10C.F.R. 50, Appendix B; and it does @ not have an operative corrective action system. The latter has been demonstrated by the further deterioration in corrective action from 1979-1983. While before, the, system was merely failing to identify and solve problems, now it is actively in . O cog rfing them up. This has been especially true with respect to welding, non- 'ad [p,4;., r destruitive: examination procedures (NDE), and hydrostatic tests--al1}of which I leirned'were consistently uncontrolled, and that some of the procedu'res"for the 1 first two items were not qualified by a testing process which proves the procedures
] actually work as claimed.
. b
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<3 s The system also broke down for vendor quality assurance, where Pacific Gas and
~
fi Electric (PG&E) management ordered Pullman inspectors to stop reporting cracked
- +
i welds found in structural steel restraints supplied by vendors such as Boston. Bergen and American Bridge, jt} fj (s an auditor trying to work within the Pullman site and corpo5te QA system, jy ; h a I leaimed the cause of the QA breakdown and why it has not been corrected. (( Pullman QA Management does not want to know about QA/QC violations. Management's f[f corrective action has been to harass, threaten, and intimidate QA/QC personnel who identify problems, and to, dismiss those who persist. Although I exhaustively hh y reported deficiencies, the major effect of my disclosures was to prompt orders JO ! % from the QA manager to only look where I was told, and his angry threats to "get rid of me!'During one such exchange,he exclaimed Pullman's bottom line: we're not '9 - comitted to building this plant to 10 C.F.R. 50, Appendix B.
~
In that case, I
. :.\
f.; - do not see any legal basis for the NRC to allow this plant to operate.
,c ,
I am not opposed to nuclear power. Rather, I believe in the technology
'si'.i '
3/j enough to insist that it receive the proper respect. I began working in the * <k nuclear power industry in 1974 at the Trojan Plant and have worked at the
'u
;ij Humboldt Bay Plant.With the exception of two months in 1979. I worked at Diablo Inj Canyon for Pullman from September,1978 until Friday the 13th,1984, when I was
. .g uti laid off. The layoff occurred the day after I finished a two-month series of I' y fjij disclosures to the NRC. ,
qJ~ l b;..i For my first three to four months on site, I was a documents reviewer. For O nineteen months I worked as a weld inspector in the pipe rupture restraint j 1 program. In August, 1980, I was promoted to QA Internal Auditor. 21 - 3 :
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i . t?! F Q 1.di"jfN.'!isid@ @ M A i$3.!nEiNi[ M W WIf3,53E;;S.A M285$shFff ' 3 C ,o My responsibility was to evaluate and monitor the entire QA/QC program for compliance with our legal obligations. This is how I learned that Pullman
- does not consider 10 C.F.R. 50 a legal obligation for work at Diabb Canyon.
y "2 :
- 5 g:j , , In Januarys, 1983, I was removed as internal auditor, but rempined in the 2;. ;
j QA phgiy tr' help close out Discrepancy Reports (DR) and Deficient Condition ' g& a f.. Notices (DCN), as well as to complete my pending audits. QA Manager, Harold flq)j h, Karner, restricted me to carrying out his specific assignments. The harass-l ment was so intense that in gid-May I resigned. Through my union, the next h Ab day I return to Diablo Canyon as a pipefitter. There simply had been too many
- g. ' headaches atttempting to work within the corporate system. On my own time, at
,,i home, I finished organizing and summarizing my evidence of QA violations. In
%@) },y November, I completed an initial report. On November 28. I sent it to NRC . . . Commissioner, Victor Gilinsky. On December 6,1983, his office wrote that I t .. y would be contacted by the Office of Investigations (01). Although 01 never
- q. .!
called, on January 6, 9, and 12, I was interviewed extensively by a series of -
,4 Q NRC inspectors from Region V. On January 13, I was laid off.
/(
' ^ll a.,
This statement will summarize the information and list the allegations 0,'d in three written reports already disclosed to the NRC. My affidavit also is yQ 1,3< to submit a written record for allegations which I have only described to the
- k. g.1 Nic in interviews.and identify allegations not yet described to the NRC.
'J f]
tt I. QUALITY ASSURANCE BREAKDOWN FOR WELDING d. J ~
. With a few exceptions, from the onset of construction, the welding Q'; program for structural steel essentially has been uncontrolled--in iolation of
. n legal requirements, as well as contract and design specifications. EThe IM:) ?
- techniques to circumvent quality assurance included unqualified welders; M) p a
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,l, unqualified welding procedures; use of welding procedures so irrelevant for }h the assigned work that, in effect, safety-related welding was widely conducted [ without procedures; reliance upon unqualified inspection procedures to check - the quality of the welds; informal changes of contract specificati ns without @Jj the bquired administrative review or distribution; falsification of records; r ~ m -, kil and liarassment and intimidation of QA personnel who identified and attempted fm. Ma to obtain corrective action against the violations. The abuses occurred both "g Q, during original construction, and during the current modifications due to the w , Vlgd,t , Bechtel/PG&E seismic design review program. Q3 J.d The list below represents a more detailed summary of the allegations M1 and evidence that form the basis for the above conclusions. The list is %.j f. 3, drawn primarily from my November 28, 1983, disclosure and attachments to
'w 9 Comissioner Gilinsky, which are enclosed as Exhibit 1. . 7!
- 1. Weld procedure Code 7/8 for piping and plates has been used w
an ch improperly to weld numerous forms of structural steel on pipe supports. What i ,?.'.- happened is that Pullman substituted American Society of Mechanical Engineers 0:.t (ASME) pipe welding procedures for the American Welding Society (AWS) struc-
;u d tural steel procedures, as implemented. This practice exceeded the legally-ph approved limitations for use of the procedure. The limits were logical, since y ^t N the two types of jobs have little in common. Pipe welding involves working w$a
., around a circumference. In structural steel welding tlie inxis of the weld is' h on a. straight plane (Exhibil 1, at 2). ~ G : Me t
- 2. Code 7/8 has been used improperly to weld tube steel on pipe 0;j l supports. ' Tube steel involves a different type of metal than the P?l material (M covered by ASME procedures. This is significant, because the NRC has identified -l 4
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'! ./
. . , . , _ . _ , , . _ , , _ - . . , , _ . _ _ . . _.__m .
p.E9c.gML / W A M @ @ M.M j y fTQ M K @}E@ g g g y,2.;Q g n @ M Q g I,; e f, ,. , ' use,of the same metals as a precondition to use ASME procedures for AWS work. In fact, tube steel welding is so unique that the AWS Code has a special sec-dl - tionforit(g.,at2-3). ~~ __ i ., b r. "l.i; T
~3. Code 7/8 'was improperly used to weld threaded wet) studs
?.} q ($:q whic bolt plates to civil steel on Class I safety-related pipe supports.- ; fM W"J: The 7 type of welding used for these studs is not listed within Code 7/8, and l T.<:d fM A it bears almost no resemblance to the work legally covered by Code 7/8
- b. (Id., at 2).
ed , jf
- 4. The welding for threaded studs did not even honor the require-
,3 ments of Code 7/8, which calls for the use of a backing bar. Instead, process
- l. ,) sheets operated by the construction department imposed backgrinding, which is s.
3 a totally different operation (H.). i. %j 5. Code 7/8 has been used to welt at least eight pipe support
*d M joint configurations, including flare bevel groove welds, and double bevel '
M g groove welds, not covered by Code 7/8. Each of these configurations repre-Cd sents a unique welding task and legally must have its own approved weld procedure jt.: y specification detailing the joint configuration (H., at 3). ,d M 23 . 6. Process sheets that guide quality control coverage did M gj not consistently call for inspection to verify the fitup of flare bevel groove welds; one of the joint configurations not covered by the 7/8 pro-f: cedure in the first place. That leaves the quality of the ensuing welds hd . doubly unreliable. This uncontrolled work has been occurring as part of g, the current design modification construction work (M.). I have read a Qq s PG&E memor.andum asserting that QC fitup inspections are not required for t ,
$@ flare bevel welds. That memorandum is not sufficient to overrule engineering -
A., .. E q i;p
- r.n ,r n ~ n.-. . ~ . , . .
.xn_..n,.---n--..._--.....- .- - - - - - - -
.D.@ .*Yq.g:&W 5t.M g M g -Q g M ggM gfyg g M (g d Q hTM L W Z., 6 ; c. .
{.; - *
,. s
*'y specification ESD 264, which requires inspections 'of groove welds and full
,; .- penetration welds.
.m..---.
s 8 '/ 41
- 7. Code 7/8 has been improperly used on pipe ruptur.e restraints 1
hlk to weld five types of metal different from the ASME approved P-1 fdaterial. h (eg Thes~ e restraints prevent a pipe ruptured during an earthquake from whipping I ~ j back"and forth, which could damage the rest of the equipment (M., at 4). Dq.f glA 8. Code 7/8 was improperly used to weld two structural steel dw shapes on pipe rupture restraints that are not covered by the procedure--W fi d shapesandtubesteel(M.). $f v3 % 9. Code 7/8 was improperly used for at least 11 joint config-1 p@; , urations not covered by the procedure itself. These joint configurations were
'd not generically prequalified per the AWS Code and were without Procedure J.1 f;! Qualification Records and/or were not detailed on the Weld Procedure Specification n ;
(/ji (1d..,at4-5). J. l )1: 1 . 1
, l} 10. The result of the procedural breakdown was uncontrolled ,I . #9 .
welding. To illustrate, in one example, pipe rupture restraint square groove
.q vn.j ,
welds were conducted without any established or documented procedure that (jb applied to the work in question. In some instances, welds had been completely (,d. removed without any QC record of their disappearance.
, The records reflected h;, QC accepted welds where none existed. For documented repairs, there was only 4,g lg erratic QC coverage due to unexplained procedural changes that deleted the ffi 7;u requ_itement for nondestructive examinations (I,d . Attachment 2).
- n
- t. *1 -
g c.3 . >@ ^
<J,,
t /j ':.r ? M .j)- i.1 b_,__ _ , . . -
-.m,. . ..
g g 3Q gx w.w 3ggp3m.:w.qg M gggqQ Q g:9gggg,y5Qqh955 GL -7
,p s ., . S
- 11. Pullman has recognized the error of applying ASME welding
] procedures to AWS work in an uncontrolled manner and issued Welding Technique Specification No. AWS 1-1, in an attempt to clarify the proper use,of Code '7/8 E on AWS work. But the scope of corrective action was inadequate. t only x.d ' ojd covered the. work in a weld crack repair program on pipe ru6ture '~ 'UJ 21 res[raints (J_d,. d at.5-6). The misuse of Code 7/8 far exceeds the use of d) Q 31 AWS 1.1. The crack repair program only covered about one-fourth of the pipe Q rupture restraints, and none of the pipe supports.
~.I;l '
ij 12. AWS 1-1 failed to fully correct the improper use of Code 7/8 "g M f for welding in the weld crack repair program. The procedure uses a steel not n .. kjj contained in the list of acceptable AWS base metals, without evidence c W that it had been individually qualified to prove its reliability (jd., at 6). Oq p ,. q 13. The above violation was approved on December 20,1979, by
,} V. J. Casey, who signed off as Cognizant Welding Engineer. Sixteen days -
j earlier, however, he had been appointed Pullman's Assistant QA/QC manager, h
.' i;;
according to an interoffice memorandum. To my knowledge, Mr. Casey has never g been listed on the pullman organizational chart as a Cognizant Welding h Engineer. The only way his approval would not represent a false statement is
- o. .e 2,
if he were simultaneously a. construction and QA official. That would be a 6 r,s violation of the NRC's requirement for a QA program independent of construc-
]j tion ( b at 6-7). , J
- 14. I also have serious reservations about Mr. Casey's qualifica-tions, based on his judgment in the field. In 1978, Mr. Casey was my !
"y l g supeIvisor when I began as a welding inspector. Heinstructedmet}omeasure q $j fillet' welds by the throat, when the AWS Code requires the measuremenfs from . l% .) 31 l..' i U 6 e
,_ . ... . _ . . _ , , . , , . - . . . . _ . . . . _ . . . . . . . ~ _ . . _ . . . _ . . , _ . _ _ _._ _ _ _ . _ . . - _
" - ^ ^ .%1Li M :. 10iTi22h_nl1H$2t M h2:.X sd232id W b,
,c the leg of the weld. For approximately two months, ! inspected welds to the ;g s f - . , wrong standard, because Mr. Casey gave me a makeshift gauge not designed to "l measure fillet welds. Other inspectors informed me that Mr. Casey has changed a
- . ' the rules on the spot for equipment anchor modifications in the co$tainment.
4 $m They. stated his instructions were to work to a " relaxed" engineerijg~ specification
? ESD.243. 'M :
1% ' <A.
., :j 15. Through loopholes in its Engineering Specification ESD 223, .@3 Pullman improperly exempted itself from AWS design, fabrication, and erection *d requirements for all structural steel pipe support welding. Writing off the Ih rules in this fashion violated the PG&E contract specifications. '\'i To my know-
- r ledge, there is no documented authorization from PG&E to deviate from the Code requirement, which is still in the contract (M., at 7-9).
q...;
- 16. PG&E contract specifications on welder qualifications were
.)q changed without required review and authorized approval. The rules were M! changed through a cryptic, unexplained note. The changes involved the .
l qualifications standard for all rupture restraint welders before July 10, 1979. The use of ASME qualification standards for welders doing unrelated
).3 AWS work mirrors the breakdown in welding procedures. Again, however, the ,d 7M 1979 corrective action only applied to rupture restraints (M., at 9-12).
, .y 4 a g 17. The PG&E contract requirement for Charpy, or notch impact If.j strength tests, was waived for Code 7/8 and other welding procedures. Charpy
.a ;. i t..
tests are necessary to be sure the welds installed under the procedure can a/9 d meet relevant design and professional code requirements for strength, p
,d Deleting this requirement was a serious step, which should have gon( through
]q 5.a the Con,t aat Specification Change Notice process to assure proper thin,eering review and approval. 1.1
. .t Instead, in January,1974, a P0&E piping superintendent q.;
u
" l
.__m_._. .m.,, _ _ . _ _ _ . _ _ _ , _ _ _ _ . _ . - . .- .- _..._..j
lg21$$$M8!ff$$$ddi1E$fAId2IEI$ESNbbS!.idNEbNUbbb b
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- c.
s 9.n. C; removed this significant QA check with a one-word penciled response, "No". L'i when Pullman asked in a letter if weld procedures for rupture restraints
'E ~..
j,{
'S.
required Charpy impact tests (M., at 12-13). j f.!N! 1- 18. In violation of still unrevised contract specifieltionss f.e Q M specific corrective action commitments on relevant Nonconformance Reports ijf (NCR), and relevant procedures for the weld crack repair program, none of the Kb l full penetration welds less than 9/16 in. thick among rupture
@dj restraints h
pt-were ultrasonically tested.
- This means that the welds in rupture restraints l f 'since July,1979, were not fully covered by quality control tests in a signifi-h cant number of cases. PG8E engineers accepted the-loopholes to Pullman's 1
pl
$} : program in July,1979, again without the required review and approval, and <;ua f4 without revising the relevant contract specification that was being ignored Nl Ee q
(!d., at 13-15).
.N. ,
3.qf. ~ 19. Another weld procedure, Code 88/89 for carbon steel piping, n . 3 has been used to weld pipe support structural steel shapes.and plates during both lf
,., original construction and repair work in the current design modifications.
- ()
3 Structural steel shapes and plates are not covered by Code 88/89 (Id. at 16). N opj Sp 20. In violation of the contract specification, Code 88/89 has F.,
% been used to weld carbon steel plates and structural steel shapes to rupture W .Q restraints with two welding processes, Shielded Metal-Arc Welding (SMAW) and Gas v.a l 3 Tungsten Arc Welding (GTAW).' GTAW is not covered by .the relevant AWS Code'( g J)' /A !w uH 21 . In August,1979, PG&E issued Welding Technique Specification
- h. No. AWS 1-3 to clarify the use of Code 88/89 for AWS welding. Unfdtunately.
W'1 - d the " solution" again repeated the problem. AWS 1-3 covers a weldin'g ifrecess,
$j (GTAW) and a base metal (A-515) not covered by the relevant AWS code provision m , ,fj (M. , at 16-18) . 'v1 .
qu
,y
't.m ?.Lic ,lE W d i:.T W ':G dd M W i[d isis fiE $3fS M!2.EMdh M R RIES M 10 T .., ,
- J , 22. Pullman also substituted welding procedure Code 92/93 for y
m pipe rupture restraints when the process sheets specified that the work i would be done to Code 7/8. The Pullman Assistant QA manager accep,ted the - bl)' 4 switch in an August 15, 1978, memorandum without changing the proc 4ss sheets-- .q . % whici left a record of work to a different procedure than was actullly used. r:id e _i
- ?.J.
, .v (M.j at 18). The only records accurately reflecting the weld procedure used h'h, were the weld rod requisition forms (M., at 21-22). )'. . . .;) @ 23. The informa.1 approval of the welding procedure switch was i j based on a false premise--that both procedures were qualified to unlimited s;.
..h thickness and were technically equivalent. In fact, they only bear a passing s
fjj;j resembl a nce . For example, Code 7/8 does not include a type of welding in
- s. j
- fj Code 92/93 that is only universally approved by the AWS for welds up to 1/4 in.
D
,i thickness. Nor did Code 92/93 have its own procedure qualification test to y
p Q verify its reliability on the welds greater than 1/4 in. thick. In effect, p that welding was uncontrolled and its quality is legally indeterminate. The , h two welding procedures are also different with respect to joint configurations, ,i] (,l1
;i joint details, tacking the joints, weld processes to be used, backing bar JJ requirements, and welding techniques, such as the allowable heat input from Sj y'j AMPS and maximum volts. The controls for clearly distinct special processes gj cannot be legally intermingled through a memorandum (M., at 18-21).
k f c .s
- 24. Contrary to contract specifications, welders qualified to y
..w ASME-based Code 92/93 were used for structural steel welding without being m propefl y qualified to the AWS Code. The switch was accepted on August 15, h' 1978.. Interoffice Correspondence, rather than through an accountable procedure qd, y with review, authorized approval and a Contract Specification ChangI ytice
- k g:1, (H. , at 20-21 ).
e 4
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- 25. An April 14, 1983 Discrepancy Report on 1972 welding in the Spray Ring Piping System for the Unit No.1 containment deme. DR #4713,
?-
failed to identify an organizational breakdown far more significan) than the -
- s. issue it disclosed (variations between the SMAW weld process used a the 3 proc ss reported in the process sheets). DR #4713 also revealed thaft the '
~
bfd 'j proce'ss sheets and rod requisition forms referenced different weld rods m
,G than had, in fact, been used. The response of the QA/QC manager was to 0:c..
y.] accept the violation as is. 1, The DR did not mention one of the most signifi-
.]f cant violations: the production department substituted an unauthorized, 'i./
Q,,. unapproved procedure and process for the procedure which had been properly @,4 . i selected and approved by the QA system and the third party authorized irspector jy from the State of California. m This was done in order to avoid teleys when QA Va
,1 issued the wrong weld rod for Weld Procedure 128. production could not wait J
qi to correct the weld rods, so the foreman just changed the procedure. In Ci
,q other words, the production department's " solution" was to achieve compat- -
ibility by making the procedure as wrong as the weld rod. DR #4713 endorsed
- D
,3 , the procedure switch Qd.. d at 23-25). If production can overrule the QA ., /1
- qC' system so easily on such casual grounds, it means that controlled welding
[1.;j *
.e. g procedures occurred only when tolerated by the construction department.
J '1 1,. 23 Under the circumstances there can be no basis for confidence that the quality y of the welding was controlled. Most significant, in April,1983 Diablo e(a (.;,f j Canyon management was still satisfied with this result. r;
- 26. DR f4713 missed another equally significant violation: QC j@.
pr; inspektors had approved all the welds after visual examination, altf ough the
- e l GTAW ahd SMAW welding procedures do not look the same.
The 1972 failure f7.] raises ' serious questions about the reliability of QC inspections af ttfe - [], . q c1 lf
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time. The failure of DR f4713 to even note the QC inspection failure demon-
- s 4
c #, strates that 11 years later, the acceptance standards have not yet become .
@ realistic. Significantly, before it was issued, this DR was revieSed three .:q !
time by Bechtel and PG&E management, which must assume responsibilf ty for a
)
.,j QA esport that failed to disclose, at all, the most significant QA violations 3 (1,d. .' a t 25-28).
.A e:'.: $j l
- 27. The breakdown in records for the weld rod and weld process 1
T sheets render it impossiblet 'o verify the qualifications of early welders by hdj reconstructing weld rod and process records, as asserted by Pullman in response o,5 g to 1977 Nuclear Services Corporation findings that the qualifications could not $ be established for welders in late 1972. I demonstrated this effect of Nj l^ DR #4713 by applying its findings to a case study on a welder whose qualifica-E h.)i m tions were challenged in the original NSC audit (I,,d., at 28-30). ?; ,j 28. My attempts to perform my audit duties on welding led to N . Lh sustained management hostility, including restrictions on my organizational 9 d ,: freedom, harassment and intimidation, and retaliation through personnel 3O actions. On January 28, 1983, the harassment reached a climax. I had already f 9 been removed as internal auditor on pretextual grounds (infra, at 23-4) f]a ' and was doing research for pending audit reports that I had issued, in this
- 3.
.:1 case Unscheduled Internal Audit #35 on pipe rupture restraings. I was at my hj j,.j desk reviewing the records on three full penetration welds that had been i .*it M tested to the wrong nondestructive examination process. Mr. Karner approached Q ~
f and wanted to know what I was doing. When I told him, he asked if ] had been O y;j direc ed to identify those problems. Because I was completing a pejding audit of whi'ch Mr. Karner disapproved, I accurately answered, "No." He tlien# shouted %] at me that I was no longer the internal auditor and could no longer identify Vi ij q I. r :':{_.
., ,m.,._,m . . . , . . - _ . _ - . - . . . . . _- -- ._
g@hhadNb M$bbE5b b bEbI
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@:'N n; discrepancies unless he specifically ordered me to. At the time I was k; still a quality assurance employee, helping to close out DCN's and DR's. Mr. L.. _l S)) Karner's orders to restrict my inquiries violated the requirement,Jor M l % organizational freedom in 10 C.F.R. 50 Appendix 8. Ji I I pd - h y'q -[ 29. During the January 28, 1983, confrontation, Mr. Karner also g i si . I threatened that if I repeated this type of behavior, he would "get rid of me." @ From his demeanor, I was unsure whether he was referring to my presence on
- .m
\
- h. the job, or my presence--period. Mr. Karner's threats eventually convinced w
E me to resign and to take a pipefitting job. The pervasive atmosphere of
- u
$} M.i intimidation was too counter-productive for an employee to successfully uphold h.1 required QA/QC standards within Pullman's quality assurance program. k!b s 11 t &j 30. Although Pullman has gotten rid of me, the company has kept em the proble.n of unqualified welding procedures. When I left in January,1984, h:;, we were still working to the same welding procedures I had audited. Nothing - % 1 has changed except that after all the notice, it is clear that Pullman and L M,;) PG&E's violations are deliberate. There can be no excuse of ignorance. 3 { $! Corrective action has been nonexistent or ineffective. l fd - There were discussions 3 on-site of attempting to qualify Code 7/8 after the fact, which would have d" 0;;j been ineffective anyway since it was the sponsoring procedure for considerable ( i Jj work that it did not describe. As of my departure, however, even that halfway 2JJ step had not occurred. 3, M. w N. 7 II. QUALITY ASSURANCE BREAKDOWN IN NONDESTRUCTIVE EXAMINATIONS DJ hj y, Nondestructive examinations to test the welds and othed hardware were as unreliable as the procedures to conduct the welding in theSir,st place.
- .q ,
y The indeterminate quality of the testing process leaves the quality of the
- I.
ki i: h j r;a '
/,} - ~ . . .. . .... _ . . .. _,, . .. ._ _ _ _ _ _ _ . _ _ _ _ _ _ .
y*:Lga. -lU ' ?k5N.j$kDEd55E$.{iA55Y$bYbbYSYN$$N ^Y
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hardware in the same status--indeterminate, at best. In some cases, NDE [ . results were compromised due to simple manipulation at manage-ct; ment direction. This phenomenan allegedly occurred when Bechtel ap.d PG&E had B y % the NDE personnel do certain ultrasonic tests (UT) over with a diffprent 3}, , . l appr(ach, after the tests had identified a large number of rejectable welds. h;. A good illustration of the quality assurance breakdown involves s.s y 1972 tests used to measure Seismic Class I valves on the reactor coolant x) pressure boundary for minimum wall thickness in response to an Atomic Energy y Commission (AEC) directive. The UT procedure was not qualified by tests to
.v . ,
A determine its reliability, which was questionable anyway, because the procedure kN f,q did not measure the entire surface of the valves. There is serious question ji ,- whether all relevant valves were examined, in part due to conflicting informa-b tion in the reco rds . Not all the equipment used to measure the valves was
);
[.% traceable and calibrated. The former violction invalidates usage of the equip-hji ment. The latter affects the accuracy of UT results by up to 48 percent, .
'% when the AEC required 98 percent accuracy. Informal changes of contract
'.dr} g specifications, without the required review and approval, again facilitated q
;d the QA violations. To my knowledge, corrective action has not occurred.
M 30 P The unreliability of valve measurements was representative of a
;;2 general QA breakdown for nondestructive examinations. In Internal Audit 101,
\
m l i h I checked 21 such procedures--seven were deficient, representing three forms of nondestructive exams To date, the most significant problem remain. The Li ' basic-flaw was that records were not available to demonstrate that test pro-m] O ceduras were qualified. After I traced the use of one procedure back to the 1
- . l 4 steam generator feedwater nozzle, the QA manager ordered me not to ling out where I
g4 l
!]p a related te.st procedure was u se d.- The response to my disclosure of these
-l
'N problems was to sit on them for over a year. In some instances, there still y ci Y.)
.......y.~, . . . . . . . . . ...
*~__a=y_y___**-*"_*_**."_f.~*__*f_'T7_N_*~i-'_**'"-'*M.*P-'UE
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'n-y;.s
}; . has not been effective corrective action. QA management reneged on solutions to which we had agreed. The situation became so frustrating, that I conducted g .. an audit on corrective action and sent the results to Pullman corporate head-- '-
.g quarters. The response was to reprimand me for breaking ranks, wh)1e the QA .4 4
d vio1(tions continued to be ignored. Below is a more detailed list $ng of related tR _ h
- e. , .
allegitions. e
%n") -
31 . In some instances, the unreliability of nondestructive ge; y-: d examinations is due to manipulation of the test results in order to mask
.g deficiencies. This allegedly occurred in 1982, with respect to tests involving y
{f.) around 230 Unit I full penetration welds--some in the containment--where UT f}. examinations revenied large numbers of rejectable conditions. Witnesses Q.1 h.f.j described the defeits to me as voids, slag, and lack of fusion in the roots i sm h of the welds--whict' raise questions about weld bonding. I was also informed %4; that Bechtel and P3eE management responded by manipulating the UT procedure in a N, a manner that would lower the number of rejected indications. The welds were ?,h$ @ then " accept (ed) as is" Qd., at 15). Ld
- .w hj In other instances, the QA violations are more deeply rooted.
37 MO The case of Engineering Specification ESD 234 for ultrasonic measurement of l G~! valves on the reactor coolant pressure boundary is a microcosm of the break-RI4 gclj down. On January 18, 1982 I initially reported QA violations through Internal
- q 4p Audit #101. I tried again in November, with unscheduled Internal Audit #34
?j On January 2,1984, I finished a report to Commissioner Gilinsky on this still y
h N uncorrected problem, which I have since forwarded to the NRC inspectors at N Diablo Canyon. It is enclosed as Exhibit 2. : g.y 2 Pf4 7 y 32. There is no evidence that the ultrasonic thickness;meesurement Q ., i l4 q p l lN h m,1 s -_~2=_.=---_- .u.1----_-- :2 - - vr * ' - - - - - " ' ~ ~ ~
E i ;y;._2 % sO5 $V$II.NEEN$$A b di2 b Nb N b b b ' }d;M - w.(.y 1
- procedure was qualified through tests to demonstrate the 98 percent level of x-accuracy required by the AEC. The valve measurements were conducted with an
=---
Q, uncontrolled procedure, and therefore cannot be accepted as the basis for
;e g jii conclusions about the quality of the valves. In my audit, I couldreither ww 7
a find'_ evidence of a Procedure Qualification Record (PQR), nor a Procedure.- h Qualffication Test (PQT) (Exhibit 2, at 2-3). h$
- w. -
% 33. There is no evidence of " procedure verification tests," j l
QA Q w required by ESD 236 for the transducers, that take into account the curves. \ l N ridges, and irregularities that exist on every valve and significantly affect
\
h j the measurements (M., at 3). n:M Ri frj i
- 34. Management appears to have conducted the measurements without j N.4 l q
AC any qualification test, despite prior warning that the procedure was too j Q unreliable to sup; ort its findings. An April 17,1973, " Interoffice Corres-
.O ,M pondence" had dist:lo sed:
t/.1 h 3. The transducers available are adequate for flat { f smooth surfaces. There are no adapters, shoes s >.u or wedges available should they become necessary. 4 At this time, it appears the transducers supplied M)
$.' may not be the correct type for thickness readings.
I.4 If this is true, we will have to order new N transducers. M:j
$ 5. The effect of surface contour and roughness must .s be tested prior to making any reportable results.
e;:q 4:? 6. There is no available equipment on the U.T. equip-iud, , ment for review. ap - ad It is doubtful that any meaningful results can be h; gj obtained at this time and it is definite that none can be reported until the above-mentioned j - problems are solved. ; ,
]'
(M., and related attachments) Jy 1 4 ' 4l, _ _ . __ _ _ _ _ . _ _ _ _ . . _ . _ _ - _u
-mm- -
N.S.I,M$ ' NNNM Nbb ' ' $a . g .
.q f.)
u
- 35. Pullman QA manager Harold Karner improperly refused to A take corrective action in January,1982, when I disclosed the lack of pro-
\ 'l cedure qualification records or tests for ESD 236 and ESD 244, the UT Thickness - 5,I 3 i Gauge Procedure. The problem remains uncorrected. His excuse was tihat these
- p. .
4 procedures were only nondestructive measurements rather than nondeskructive ]$
- n. ..<
4
<:4 tests, and therefore did not represent "special processes" whose quality must be controlled (I_d., at 4).
a
.N That semantic distinction is irrelevant. The reason to
$[yll u;.3 require reliable, controlled procedures is to assure the quality of sensitive, W safety-related hardware. Indeed, in 10 C.F.R. 50, Appendix B, Criterion X, (R .. n
>]
the tems " examinations, measurements, or tests" are used interchangeably. gj The safety-related purpose for qualified NDE procedures is magnified for ESD f) y.; 236. ESD 236 was instituted in response to an AEC directive to the nuclear o ... : industry after discovery of valve problems at a series of plents. [d. 4 tl
- 36. Mr. Karner's manipulation of definitions is wrong. UT measure-54 ments constitute a special process which must be qualified.
- c. d They are a special b) process because they are uniquely created to perform a specific quality-Mj related function. Further, pG&E contract specifications and 10 C.F.R. 50, h
A1
% Appendix B, Criteria IX, " Control of Special Processes," identify nondestruc-d ; tive testing as an example of special processes, not as the boundary of the w: concept.
Y.3 v.% d
- e. .- .
- 37. DIA #34 of 254 Valve Wall Thickness Data Reports demonstrated
.b * .y that t'he Data Reports are incomplete and, therefore, are not traceagle, as c
Llg required. : For example, none listed the size, shape, or manufacturerg's yy designation for the transducers that performed the wall thickness. bhe' ESD ,J " f a' y y i
- s. . - . - . - - - -.- - - - - - - = = = - -
M;$IM!R$sMNMNN Nbdb$b i my W.4 . Q u, , 236 Documentatiori Packages do not provide any information on the testing 4 y< equipment beyond the serial numbers. In some cases, there were not even S.- $ serial numbers for the UT machines and the micrometers used as a mechanical-backup measuring device (M., at 5-6). , L] j M . 7 M 4 38. The Data Reports offered unreliable, inconsistentIinformation. %m For instance,19 reports listed two different UT machines as having conducted IM the same valve measurement. Serial numbers for UT thickness equipment and p @k micrometers could not be verified independently. Ten percent of the valves m. h$ y checked physically had seria'l numbers different from those listed in the Data i k Reports . In many Data Reports, original information had been whited-out and ) { k p altered without signature or explanation (Id., at 6). i I b yd 39. 4 Necessary records to demonstrate calibration of the measuring equipment were not consistently available. To demonstrate the potential J[b d.: e : effects, on three UT measurements whose accuracy was tested, the pre- and W jij post-calibration checks showed variations of 10 percent, 48 percent, and 2.6 . n$ percent (M., VIA #34, Attachment 5). The maximum error permitted by the AEC c.; (. Np was 2 percent. kid ,s- - d 40. The AEC acceptance standards were violated when valve fn C measurements from equipment that failed minimum reliability standards (#39, a supra) were used to accept the valves as sufficiently thick (H.). 'f4 m
- 41. Forty-two Data Reports disclosed that the valves were below M.3 3 the minimum thickness, but on the paperwork they were marked as " accepted"
? : ,g without explanation (M.). , 'b P,
[q. . 42. In il cases, the measurements w'ere incomplete. n Thhrfcords
.i simply skip results for required areas of the valve, such as the flat pad at A
3:/ ;.j the bottom (M.).
- l
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'iv.;g ug..jf.F w Ay> ga qw-.q..p..~. g6 w a. h Eif.x.- @ M..pK w. E.a.g }, -M.gc J9)W .w.~ $.a.?,l q,.. M. w?% w.. .M.,~ip.3, a, i j .5 p gw.y: qq--.u: a.:. .. . . ,:- W - wr. ' ,
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- 43. In 14 valve locations, there was no documented evidence that
'the valves had been examined at all (H.).
- k. .
( 44. Therewasnodocumentationtoindicatethatweldppairson . 3
; the valves were controlled, as required by the AEC. To illustrateithe absence of v'erifiable controls, the Data Reports do not have a requirement to list lM whetier valves were weld-repaired, or the weld procedure used (H., at 7).
di r .d d 45. During my research for UIA #34. I discovered that none of the fl lM valves meet AEC and PG&E design requirements. Westinghouse, the manufacturer,
- r. .
$N . r.c had explicitly declared that they "were not designed to meet the minimum wall l thickness requirements of ANSI B16.5"--one of the relevant professional codes f].) W fi, listed by the AEC in 1972. By comparing Westinghouse's communication with dd. pG&E contract' specifications, I learned that the valves also do not meet the qu Qj design requirements in the contract ( M.). a a f.::1
- 46. To my knowledge, there still has not been any corrective -
M action on this problem. If there had been good faith attempts, I should have >:G
.% been contacted as the originator of the audit. I remain available to help
%j follow through.
- q. 9 di V
.s 47. Similar to UT thickness rneasurement procedures, nondestructive M
.w test procedures lacked documentation of procedure Qualification Records or 1
4 Tests. in IA #101, I found this flaw in seven procedures out of 21 examined. y l M Beyond the UT thickness procedures, there were five cases where no evidence Ph N existed that NDE procedures had been qualified. As a result, the quality of 33 Q work examined under those procedures remains indeterminate. Thesekncluded: Y} g.. ?
- 1) ESO 234, for UT Inspection of Groove Welds on pipe rupture restratnts A
g,. d prior to 1979; ESD 241, for UT examination of Safety Yoke Rods on Safety icj pi 6 'l n. e I ~
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$f 4' - Valves; ESD 246, for Magnetic Particle testing, with unknown use; ESD 247, for ps f . Magnetic Particle examination of welds in the crack repair program on Unit #1 l j.7 Steam Generator Feedwater Nozzles; and ESD 270, for Liquid Penetrgnt { kb examinations, with unknown use. On January 12, 1984, I completedhnd delivered W(.x1 [, to htC inspectors, a draft report to Conunissioner Gilinsky on IA lbl. Itis
' j
{ j encipsed as Exhibit 3. 3 ' f8 24 48. The corrective action for procedure ESD 234, consisted S]A 2 45 4 of unreliable, "a f ter-ther fa ct" Procedure Qualification N- .h Tests, whose use was not controlled and accomplished using qualified procedures.
- n Ironically, this is the same flaw the late PQT were supposed to correct.
)
H Further, there is no evidence that management reviewed and approved the y procedures for the PQT (Jd,. d , at 2-3). T,h -
- 49. QA Manager Harold Karner improperly prevented any corrective g
~
1 action for the lack of procedure qualification records on ESD 270. Instead, he t q directed that the Procedure Qualification Records for a similar procedure, J kl ESD 210, should be used for ESD 270. That is unacceptable. If the two pro-V:3 ! i *i cedures have separate numbers, there are at least some dissimilarities. Those bi
- .? unique features of ESD 270 inherently will not have a proven demonstration of d..s .
4 g their ability to identify defects. This QA violation remains ignored. $q1 (,] 50. No. ' investigation was perfonned to determine . 4 4.) where ESD 270 was used. Instead, the QA manager told me to just write up what I had learned already as an audit finding. m : v3
. 51. ESD 241 for UT of the safety valve yoke rods invo{ves the most y rJ significant violations. In addition to the lack of a PQR, the harkware was d tested from December 17-20, 1973, before the UT. procedure it'self was even ~)
6l g' ' issued on December 26, 1973, and prior to approval of the UT procedure "a.i
. ~ ~-=- ~
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e.y g- R g, &q.a, .M;;.4w -Q..-..@~W?fM3b?21.9.$d.id$NM$bb. a-gi I0$2b5N 'q.;R} . - $,1 ' w by PG&E on February 12,:1974. . :,-?. The testing was totally uncontrolled for y's
- y. -
the yoke rods on these valves, which I believe control the release of radiation f fromthecontainment(g.,8at4).
'~~ -
i k 52. s i did ESD 241 was deficient because it violated instruebens from g s Dresser, the vendor for bolts and studs. The Dresser instructions required fM. the Eods to be examined prior to threading. rf At Diablo Canyon, the UT's were
.d 4:4 conducted after the threading. Further, ESD 241 did not use the Dresser 3 instructions to determine the reference point for sensitivity and the criteria m -
to report questionable items (Id., at 4-5). Nd 53. The existing documentation for the tests fails to meet the M %4 A standards both of ESD 241 and the Dresser Instructions. Required information t/J) p
- ad on the testing surface and instrument calibration was not included (Ld.. at 5).
la d 54 Both ESD 241 and the UT inspection records failed to reflect 5,.".j - c a; compliance with a PG&E-imposed requirement for backup inspection "with the i M liquid dye penetrant technique to check the yoke rod ends for indications of $1!j cracking that might extend into the threaded area of the yoke ends" (M., at j$q 5-6). ?i:1 JQ
'N .y 55.
No DR was issued to PG&E on ESD 241, although this corrective );d! action had been agreed to both by Mr. Karner and the NDE supervisor. Mr. Karner Mi
.a improperly reneged on the basis of a memorandum from John Guyler, my successor
$3 as internal auditor. Mr. Guyler dismissed the detailed, documented DR which I 'M h +4 . had proposed with the following assertion: "PPP has accomplished this per
- gj instruction from PG&E.
It is evident that a nonconformance does not exist and Q 63 a DR is not necessary" (id,., at 3-4). Mr. Guyler's response was indequate. : no . h 1 First, the procedure violated PG&E instructions (see #54, supra). Second,
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- .7 even PG&E does not have the authority to validly instruct Pullman to violate
$ 1D C.F.R. 50, Appendix B, Criterion IX- "Special Processes." Third, Mr. R W O,a Guyler did not document his asserted conclusion. y.j y
- 4 e[:(%m 9 56. Overall, Pullman viulated NRC reporting requirtripnts and N PG&E contract specifications by only. reporting the deficiencies for two out ,f of IIhe seven nr.1 destructive procedures to PGM on Discrepancy Reports (id MJ h
l31
- 57. PG&E dispositioned the 'DR for ESD 246 " accept as is", although i t.l .
there is no 'information ilndicating where the nordestructive test was conducted. hh Since the identity of the affected hardware could also impact on thu evaltation "W ,f;' criteria, N&E's acceptance was premature (M., at 7). &1 )hhl
- 58. The reason the location of work tested under ESD 246 could not h{ be identified is that Mr. Karner improperly provented me from looking. After f.)
$su I learned that ESD 247 was used for welds in the crack repair program on feedwater mi q nozzles in the Unit I Steam Generator, he ordered me not to check where ESD 246 - W hadbeenused(~Id.,at6).
${
y l j 59. PG&E improperly dispositioned the DR on ESD 247 " accept as is", 1 @W although the &gnetic Tests in the procedure were referenced to ANSI standards, l K '{' y> :jv rather than the relevant ASl1E Code Section 1; and although the qualifications M.n of the MT personnel conducting the test cannot be verified from the records f:.k t available (M.).
- 2, t 3
$$ 60. e The corrective action .for ESD 246 and 247 involved procedure ij . qualifications after-the-fact (1d Li . at 7). After-the-fact procedury qualifica-4h tions should not excuse PG&E from accountability under NRC rules. 5t best, it @ps ;, means that the damage has been minimized. But it also inherently means that m J.1 M
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17, .. 10 'C.'F.R. 5'0, Appendin L ,was violated, because special processes were con-(% r 9 @.s ducted under uncontrolled conditions. P..
;" 61. Even if it is acceptable to conduct procedure quadification
, y
$fjf tests after the fact, the tardy test must be performed under contrilled cir-s G$p
% j cums ances. In this case, PQT's were conducted with different equipment than I
41 - had been used originally (Jd_.). No documentation was tupplied to support the d@l
$/8 asserted Corrective Action Response that the new equipment made the results more conservative. $w. ,
l$ 3., 62. QA Manager Karner was responsible for the deliberate failure m
'E to provide reasonably prompt corrective action for IA 101. On January 18, 1982, I initially disclosed IA 101; on March 23, 1932, it was finalized after $[ I provided Mr. Karner with additional information which he had requested. On d
e.4 April 6,1982, corrective action for the first finding in the audit' on. lack 7.] of procedure qualification tests was approved. Before implementation, how- -V A ever, he changed his mind. Although the official time limit for corrective . I@ action is ten days, the audit as not closed out for over another year, a despite my repeated memoranda and attempts to formally notify Mr. Karner of p N3 his obligation to address the issue of unqualified NDE procedures (~Id., at 8-11). M pA
., a M
m
- 63. Pallman corporate QA Director A. Eck was notified of the j] failure to take corrective action and improperly refused to help. Instead, he N
N. reprimanded me for bringing the matter to his attention. On June 14, 1982, I
.Si a
49 notified Mr. Eck, through an Interoffice Correspondence, of the overdue 9 correetive action. He did not respond. On July 6,1982, I perfemed and fii Gj submitted Unscheduled Internal Audit #31 to Mr. Eck on the lack of corrective n . e 36 action required by ESD 253 within 10 days. This time I received a kesponse. ,, M., Both Mr. Eck and Mr. Karner reprimanded me for submitting the audit to Mr.
~ @A Eck directly, rather than letting it proceed through the chain of command.
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- This violated ESD 263, they explained. My audit was voided. Both individuals neglected to mention the violation of ESD that I had raised -
d
- d the QA violations were not getting fixed (M., at 9-10), i;.
'i .
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- q 4 1
~
N : 64 In January 1983 I was further punished for Mr. Karner's u.3 $; improprieties. I was removed as internal auditor because only 5,instead
- . a -
lW of the required 18, audits had been closed out. Part of the problem was due r!d b.y t o circumstances - beyond my control. Mr. Karner or supervisors 41 Lj were sitting on some of my audits beyond the required deadline. Mr. Karner El{ ';( . also was loading me down with ancillary assignments.and unscheduled audits were not $! counted. ]A 65 On January 28, 1983, during the meeting in which Mr. Karner E! ( threatened to get rid of me for looking at quality -related issues without t
; being assigned (Supra, Nos. 27-28), I informed Mr. Karner that he had w 'M violated 10 C.F.R. 50, Appendix B. He responded twice that we are not c: ,
yl committed to 10 C.F.R. 50, Appendix B, and that it was "O.K." for him
' b{
ai to violate the Code of Federal Regulations and related contract specifi- "0 / cations. n ,1 Q f% III. t,4 BREAKDOWN IN OUALITY ASSURANCE FOR HYDROSTATIC TESTS. !a M. Hydrostatic testing at Diablo Canyon from 1975 to 1978 does not h .s have the necessary QA documentation to prove the reliability of the tests. lq In hydrostatic tests, water is run through the plant at higher pressures A 3 than normal to see if the piping is reliable. : u c:)~ S}.
- m -
' =
In February.1981, I conducted Intt.nl Audit 86, in-wpch I Q r:n 1 earned that nearly all hydrostatic piping tests for a year, during 1980 gi and 1981 were conducted without required QC. documentation.'-In April 1982 bd j NRC inspection identified f. hat documentation problems identified ' f.l Gj pn...-. .
. . .- -~ - .. -
=--
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.u f.f %- -
in Internal Audit-86. were not ' properly -corrected. I becaine convinced kJ tbat ,. serious problems may exist with the hydrostatic tests. In March 1983
,f- I completed Internal Audit 106, which examined the records for 79 priginal . ^ '
F 4 hydrostatic tests and 118 retests conducted from 1975 onward. w I t arned 3 .~ that the test documentation did not have evidence of required QC oversight, [ QA records, consistent procedures, or controlled test conditions. In short, g .: H there has been a generic breakdown in the QA requirements for hydrostatic JS, v. tests. They must be redone. Internal Audit 106 is enclosed as Exhibit 4. 6.c .
,{j My specific allegations follow.
- 8
,.f
- 66. The procedures for hydrostatic tests conducted before Q
- j Ja nuary 27, 1975 are fundamentally inadequate, due to their failure to q!-
3.- include documentation requirements, and due to lost pages, the inability
})
to even entirely reconstruct the procedure requirement. . 3
- 67. Almost all hydrostatic tests and retests from 197b
]q.. S onward lack required QA documentation. The most significant omission ' .i.s involves QC coverage documented on a piping system closeout - F98 h.m p1 Department Release. This activity is necessary to assure that departments 3 @W performing the test comply with procedure checklists. Unfortunately, q 44 departments only complied sporadically with the requirement to complete .. J . 4]ili and maintain the form whichdanonstrates compliance with the test pro-
.ci cedure. In other cases, there is not necessary backup documentation to A
(i verif.y the conclusions in the release. (Exhibit 4 AAR #1). .?Q W * : n a. -
.1 .
68. $j ., From December 1977 - April 1978, in 28 cases Pullhan, {
;] test requirement forms did not have information necessary under the
-l y,
b a
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- ' t_ _ _ _ _ _ >
}'Mi(:Li;SN! d SSSN225E $$$ $658.511$5 ?S$#$ $ 5 ,- -n .. . M* , <. g.). - (,q., . , procedure ESD 229. Fundamental deta, such as the type of fluid, pressure l$I.' .and temperature, simply is missing (M., AAR #2). E Ge.
%- 69. In 28 cases Pullman's NT procedure data form es J. . ;
1 ljkIf not{matchPG&Erequirements. This form is the guide used to condect the hn ten, .so the distinctions translated into different test conditions that i h; p disqualify the results from Pullman's hydrostatic test. To illustrate, kg in one test Pullman's procedure only had a. pressure of 2485 PSIG when kd { M PG&E's acceptable minimum was 2812 PSIG. (E.., r
.n '. tj N 70. The absence of backup documentation continued after 1978.
l
.9 {
y,y , From March 1978 to April 1980, there were 14 hydrostatic retests without a
;'.Q .
A signed QC field pipe release, dispite the conclusign by Quality Engineering il.0 3 j I gj in the test records that QC had verified the results (M. AAR #3). h.G'" s,I:j ,
- 71. The problems with hydrostatic tests offer another q
4 example of management harassment of QA personnel. During the May 1982 jg NRC inspection, I spoke extensively with. NRC representatives. After the
}]c ..m interview Mr. Karner expressed anger at' the length of the meeting. At a later ,;$[] meeting, during 4his general time frame, he threaten to get rid of me. '3:i i 'y .
4d IV. BREAKDOWN IN VENDOR QUALITY ASSURANCE.
' .1 A% Although I was not as actively involved with vendor QA as (c. '
5,.4 with.special process ahd hydrostatic test procedures, I observed the symptoms of a generic QA breakdown after becoming familiar with two
=
d@l O examples of QA violations involving vendors. One case involved a jendor that calibrates micrometers, a precision measuring device for Pullman - tools and the impact of weld repairs, among other functions. Although 1:
'h. .
the vendor had a clean bill of health and was on the Approved Vendors
'S:
h U.k_ _. .._ _ . _ _ _ . _ - - - . - = - = -
@f?$A!lI5$.W$W$$b$?-- 4; , - y.. .,. 99 ', *
- ,M .
['( n List (AVL) until my October 1981 audit, there was virtually no quality assurance program. Unfortunately, corrective action was solely prospective - to
- i. .4 remove the firm from the AVL. The damage that already has been doce will
-l a: $
h remain. 4 - i M IThe second case involves 1980 and 1982 orders by PG&E for Pullman u u> j,p inspiactors to stop reporting the large number of cracked shops welds found 4,
?
in Boston Bergen and American Bridge.workIhese hardware defects should have M ,y been reported on DR's, but instead were ordered to be ignored because they Vi! ,4 came from a vendor. Specific allegations follow. -t f:R
- 72. The reliability of Pullman's Approved Vendors List t
.h; is indeterminate, due to the inclusion of Microsurface Engineering. This }j firm only had a token quality assurance program, yet had been approved and passed previous vendor audits. My audit demonstrated that Microsurface
'9 did not conduct audits, did not have'a written procedure for calibration, sJ ,
i'l conducted uncontrolled inspections, lacked traceability for use on ';' i . *
'Y -
' Pullman tools, failed to disclose laboratory standards for calibration, and
;.g did not have required documentation for training of laboratory personnel.
8, The violations were so ingrained and pervasive that it is not credible to W 'p} 3. conclude they only sprang up since the vendor passed an audit the previous
\ year.
}d 3 M A 73. Corrective action for the Miscrosurface QA violation
.ss d
}:jj improperly was restricted to the prospective step of removing the firm l:f from Ihe AVL This was inadequate, because the accuracy of measurements h Q made with Microsurface tools is indeterminate. The effects of prevfous vio-Eh lations will remain undisturbed. E* AM.
~
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- 74. In July 1979 Pullman inspectors began finding signifi-
[:ll cant quantities of cracks in welds received from two vendors, Boston hh Sergen and American Bridge. Until 1980 Pullman inspectors wrote 19 '- - Discrepancy Reports on the welds, which displayed a consistent podern of
;$) 1
{
\
Sa linejrindication. The DR's are enclosed as Exhibits 5 24 On ApH1 3.- p .d ;, I eJ 198E however, Mr. Marvin Leppke of PG&E issued a memorandum directing f;- s;u) Pullman to stop issuing Discrepancy Reports on these " shop" welds. The 6,c , memorandum is enclosed as Exhibit 25.
- e yp '
p:. , 75 In 1982 PG&E repeated the improper restrictions on t.,
;;l QA enforcement against the same shop welds. This time PG&E instructed
~
Pullman to delete shop welds from the formal walkdown program that
;. ' represents a final visual check on quality. Relevant supporting documenta-tion is enclosed as Exhibit 26. 'l . ',1 V. '
RECORDS FALSIFICATION r.3
,4 , ;,;] Beyond instances of contradictory and impossible information
- . .i O in the records, in some cases I am sufficiently familiar with the cir-F.d d;,
p cumstances of false records to state that they were intentionally fal sified. Examples involve the qualifications tests for QC inspectors. Nj As a prospective welding inspector I failed one'of-1ny initial test and was x~ .L ;
- j. then given1 copy of the 'tett to study to ' assure. passing on the second attempt.
3 Another inspector was car.tified after taking-e tes't which upon review months 9 later-he:was four<d to have failed. He was retested at that time and passed 76 ,3 with the assistance of coaching. The test was backdated to the ortiginal test p - @.1 date to cover work perfonned # ring the intermin period. The latted example u) ~ > gj occured in 1980. ~ w d M,,
'h I I'd m .:
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C)),2 1 2 1 s s; M 5 E bib db O!d [.355 b N$ $ I N 5)YS N N b b sb E S M b E N b N 48-eM.. 'y . 4) 3 . U - VI. CAUSES OF THE QUALITY ASSURANCE BREAKDOWN. 0 E M q 77. The most significant cause for the QA tireakdown31s the
~
i a environment of repression'and the predictable retaliation againsthA S pers'onnel who diligently try to identify and correct QA violations'. The E s $.! proldem goes well beyond the loss of organizational freedom. Upholding the W Atomic Energy Act at Diablo Canyon can represent professional suicide, iih @ Most significant, the sacrifice is for nothing. The violations remain, q - Lfi uncorrected. My own experience is a case study. Mr. Kamer threatened to ..n f "get rid of" me on three occassions when I persisted in attempts to obtain e$ d corrective action. Mr Karner restricted my freedom as an inspector until IB I could only look at specific problems assigned by him. I was reprimanded,
-j;j verbally and in writing, for communicating with corporate QA management
.? 'q about such a fundamental violation as the failure to take corrective m action against unqualified NDE procedures on safety related work. To add % insult to injury, in January 1983 I was demoted for not finishing enough Aq yj assignments. The demotion was due in part to Mr. Karner's refusal to uw e act on my audits, which made it impossible in some cases for me to' finish A fj sly assignments. - D,2 r$ >=. l't%
- 78. The final act of reprisal against me occurred on January
@l V.' % 13, 1984 I was laid off frod'aiy job' as a'~pipefitter, the ' day after making my third disclosure to the Nuclear Regulatory Commission. NRC inspectors already
,) had fold me that site management had a copy of my first report on welding procedures, and that Bechtel was studying it. On Friday, 50 pipeffiters were y - .j laid off, supposedly due to a lack of parking space. The usual prhttee for these layoffs is to let workers from the local union stay until last. - ~
g.? In this instance 46 out of the 50 employees laid off were " travel cards" l 23 f.} Q_ _ _ _ _ , - _ _ . _ _ _ _ _ _ _ _
[ % $ dbsbh d[J. 1 5N N b bEYI b b- N Nb - n'y C .. 30 fj).c L .; from out -of-town unions. 'Although more travelers were available, four Y' employees from the local were swept out with the travelers. One of the four L.) 1 was having conflicts with his supervisor and one had an absenteeism problem.
-jl
. y f.;{
The other two were y partner and vself. My foreman protested tal the super-4 2 i c.] visor not to lay off g partner and me, and asked for permission to pick- >.A '. j sq someone else. The supervisor referred him to the resident construction manager, who refusa;i the request and told the job steward that we had to be the
^ . ;. ones laid off. My foreman and the job steward recounted these events to me G on the day of the layoff. That day the job steward also informed me of the C
f'.] q perception of site that my layoff was due to " politics" and was decided " higher up". On January 25,1984, the day after retaliation was widely discussed at M} g Congressional hearings, management called me back to work but not my partner. A ,( The pattern represented by my case illustrates why a significant number QA violations ,I have gone unreported, and why the quality of Diablo Canyon is indetermir$ ate. 1 , 1 Those who persist in reporting the violations are dismissed, or harassed i
.q relentlessly until they resign, or give up and stop trying.
D 79. ej Another cause for the QA breakdown is subordination of PG&E's d'j and Pullman's QA department to construction. Until recently, PG&E site QC did not ('N"2 review Pullman Discrepancy Reports. PG&E's Resident Mechanical Engineer, a con-y4 y struction offical, reviewed and approved corrective action to discrepancies. As of
;;J;
,m' N.. May 1983, Pullman Internal Audits were not submitted to PG&E site QC for review but 4 instead submitted to the Resident Mechanical Engineer. ~; Il p;,;
- 80. Another cause for the QA violations was lack of resources. To A{Q! illustrate, from August 1980 to September 1982, Mr. Karner was the(only permanent fD employee in the QA/QC site management. He did not have an assistant QA Manage. ,
11* * "
,E and the QC Supervisor was a temporary employee. _
- 81. The QA breakdown was not due to PG&E ignorance. On i,l4 4
'(
. , .c ,, , m w: +;g,ye.w,;;ry.u .:.. p :;. e. q .n g n 'z g a g . g s w 2..r..ac.
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31 - e repeated occasions, I identified many of' the issues in this affidavit un <- - !
.o. .y 's to a variety of officials within the PG&E supervisory and annagement staff.
y.
'i Althpugh some officials listened and expressed agreement and/or syinpathy,
, t -
IMs norp.of the violations were corrected. I believe that PG&E and Pullman
,.g M m have been gambling that the NRC will not enforce the QA laws, even if m
E.! they are caught. For the sake of the public's health and safety, I hope 4,'f.l c-L 1, that the NRC calls their bluff. 4 l I have read the above 31 page affidavit, and it is true, (L. >] t accurate and complete to the best of my knowledge and belief. r..' .w
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k cuc C kMN d Harold Hudson i
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. .] SUBSCRIBED AND SWORN this b day of January,1984','in 3% . Is .- x ,
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! '/ #L//[/h [ RECEIVING OFFICE Docket Number (if applicable)
' f: < 1. Facility (les) I olved. gj(N
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Description:
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,9 4. Source of Allegation:
F,j (Cheet appropriate bo.) _ contractor employee security guard _ licensee employee news media
.i , _.__ NRC employee private citizen
/
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organization (specify
'. other tspecifyl [m k /I'M
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;J MM DD YY 1 5. Date Allegation Received:
;' i.: I Z. 2. e ( 3
- 6. Name of Individual trirst two initlet and test nemel 8 NM M Receiving Allegation: D.fsMd.fcf /)epyn pur,(,y, 'do#7CT; "ll..j
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- 7. Office: g g b
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- 8. Action Office
Contact:
trir : two initiet and iest namei - SM q
- 9. FTS Telephone Number: yg y y .,,
7 g g Q $ 10. Status:
'j - Open. If followup actions are pending or in progress (Check 'onei Closed, if followup actions are completed k.]"'
MM DD YY
- 11. oare Ciosed: 11.1 Document Nos.
g j ; l N U Remarks: E.ji l l l l l l l l l l l l l l l l l l l l l j jjl l (Limit to 50 cherector.)
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' dis. document lists (or directly references) each allegation or concern brought to the attention of NRC personnel, The purpose of this statement g]')
3 sheet is to assure that all points raised by the alleger are covered. E" ,' If the problem statement is not clear as to who, what, where, when, or why F]- regarding the issue, the commentary section will amplify the statement. The [,h" commentary section will also be used if there is apparent conflicting e information or if there is no or very little original information available 5' which describes the concernTs'). (This can occur if, for example, a line concern was received in an interview). k(Q Problem Statements (use extra sheets as necessary) O.Y (f (j .. A11eastion# Verbatum Statement or Reference a
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!y* m'ean an ALLE2ATION DATA FORM U.S. NUCLEAR RIGULATORY COMW5SION 41W8 s enserveters on eeww side
, . c.) l V A////tr' nsCavlNo omCE .. Docket Number (if applicable)
- 1. Facilitylles) I otved. I I tw more then 3. er w d by., /hI / $ f 8. # # 3 7 [
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v, b1 """" """"
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operations onsite health and safety M scheck esproeriete hostee) 3 _ [':' construction __ offsite health and safety 6 w' - , safeguards emergency preparedness c .. ,w
.j _ other(sp ep yl 1
1 3.
Description:
1P ls.K.lEl IDlo IE.l5l l#lo 171 lFlse 14 k l/ l lC l0 lelf W-l p1 _ t,urna w in esereciere> lyl IWli IrWI I Aldts I lbl ll, il l 1916 lsle k I: li Idlfri os ; ,, j l IP IAIRITl/ IAI). I IFIEINI IWlfl4. Lbl L.sl/ IMEl 141dl N' lbl IfIElE.1sel6lhYl InlAJI IPl/I#1/1 IJlal#lr1715I I
.U lJj 4. Source of Allegation:
;il scheck opp,opriate boal _
contractor employee _, security guard licensee employee news media ['} . NRC employee private citizen d , _ organization tspeettyi j other tspecityl [n # d'M o /
- ;) MM DD YY l I
,' 5. Date Allegation Received:
/ L 2. 9 ( 3
,, /
- 6. Name of Individual trirst two inisteis and seet namel Y N N8
?.1 ' Receiving Allegation: D. fem /A.fch /byn AVGd@. tiesMCC
{f.. ht. I2lC3/f$ gy 7. Office: g y 'fb y ACTION OFFICE
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Contact:
(First two initiale and test namel %, F*A,%e#
- 9. FTS Telephone Number: gg 3 y .
7 g g s .. A 10. Status: Open,if followup actions are pending or in progress
! s ": gen,,g ,,,, _
o.1 j'] Closed,if followup actions are completed Y
- MM DD YY
'i! 11. Date Closed:
l t 11.1 Document Nos. M '
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e., AFFIDAVIT ^ s" .. - - FY ,
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My name is Charles Stokes. I am submitting this affidavit l 4;s *
.a freely and voluntarily, without any threats, inducement or l O coercion, to my counsel, Thomas Devine. This statement ?f p.e. t supplements my Nove=ber 16, 1983, affidavit and is intended to h:l.h j maintain a current record of my allegations to the Nuclear l hj)h, -
Regulatory Commission about the Diablo Canyon nuclear power
.M plant.
- t _
The remainder of my allegations are contained in the N tape of my December 7, 1983, meeting with the Nucles: [.At P Regulatory Commission (NRC) staff and the transcript of a $,?4 January 25, 1984, meeting with the staff. Mr. Knighton of the 1 { iU g. NRC staff premised me a copy of the January 25th transcript. h.i { q' Although the transcript has been prepared, I have not yet
- r W received a copy.
f.; ' M, ORGANIZATIONAL FREELCM @1 1. I
- d. .q .
As of October, 1983, when I left, management's policy 1 g was that we were not supposed to discuss problems with the NRC jyn i or Quality Control (QC). Despite any contrary written policies,
- ?'%
':']
- s. . around the time of internal or government audits, management (gj ns representatives diplomatically would tell us to direct incuiries M
p:y from Quality Assurance (QA)/QC or the NRC to our supe visors
# first, who would decide if we should answer.
4 This was an oral Y. , policy. I never saw it written down. This type of policy led W p to paranoia among the empleyees. We all knew how much the NRC b~ .; ~. n .uculd find if it started looking. To illustrate, in the spring $ ,{ cf 1983 another engineer and I went to the NRC trailer to e _1 W
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c_ u._s_~_:_ r__r 1. _ -_2 : - -~2v'-"' &~ '~2-- ~ ~
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!9.M?.?Bitny 4 2 w-M ?.Q W % C M ffi$$$$ k N$E:.N]lS.$ $ $ b??S $ 0!:!N$$$55bb $
e. Q. n., ,. O.< . m s
!Ifi .v Mi O. - obtain research materials.- We were not going to disclose.
Nfi ._ -i [$9 problems. When we returned, everyone in the trailer was
%.H
;. discussing our visit, wondering if we had "squealid" and what
, would happen to us. We made it explicitly clear 'to everyone, l f
including' management , that we had only gone to pick up copies l Tj}0 of some public documents.
~
7<. :d.3 . 2. I believe that management helped to enforce question-f$ able design practices by hiring aliens on " green cards" who i i
$a .
were afraid to disagree with superiors due to the risk of being ' k3 dismissed and subsequently deported if they could not maintain
- h. their jobs.
I personally know of many Indians brought over y- from the Catawba nuclear plant in South Carolina who felt this q- &gg way. One' Indian who was a friend became so disheartened that g ye . he- just signed off anything, whether it was right or wrong,
;,g.
m That is unfortunate, since he was a good engineer. The Q combination of management intimidation and the large number of x ..,f errors simply were too much, and he lost his spirit. d' Wl QUICK FIX l
" d. 3. Management kept the personnel in the Quick Fix program lt y' largely ignorant of the ground rules. Quick Fix in theory was a $f m a field engineering corrective action progra=. Those merely y .r -
affected by Quick Fix were even more in the dark. Instead, )1/ . management issued informal, uncontrolled documents to only a l hp few of those who needed them. j . The Quick Fix program was later renamad cl}e Pipe l I O'J
% , Support Design Tolerance Clarification (PSDTC) progra=. Tc
$? ..? j
- g. illustrate, on June 16 , 1983, manage =ent issued an informal y
+! ?:.
.r.e.jiv,y.j< g;ft,a g g sy v w+ u..;';3 p3,,p ,.wJ S.p u a.:- .w.,.au 6 . ~;~g X . S. ; g N a a-.aN .l. ~...
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' ' memorandum "to clear up the confusion" about PSDTC.
-:. - r
{u j > (Exhibit 1). Unfortunately, the clarifying meco was only .y 3:j issued to five managers and the file, although there were at t.. i. %y
+
least twenty (20) engineers in the program and many others such A as QC inspectors who needed to be " clear" about how Quick Fix I B i t operated. l a ; d 4. The June 16th Quick Fix memo did net succeed in p
- M clearing up all the confusion, which 6 7 Quick Fix essentially w -
[N q as an uncontrolled, underground engineerirq program. The ! June 16th memorandum missed more of the p. . gram than it l 5h} l covered. To illustrate, it described the proper response "if" , I a:n : i
\
g) a PSDTC is necessary, without describing the preconditions for [hq ,. PSDTC action. That is forgetting the program boundaries, such ' Y as the scope and' extent of a PSDTC engineer's authority a d ,a gaping,loophele. , 'd 5. In the June 16, 1983, memorandum, management decreed
'E n3 that PSDTC's "are to be treated in the same manner as the n:, '
gj original ' approved for construction' drawing" (Exhibit 1) which $ means that an uncontrolled, underground engineering program had m ro \ w .1 the same fercal authority and status of the appreved design.
- 6. Although these Quick Fixes were subj ect to later
- {, ', review, during my employment management told us the findings .c , W ,I
- g. were accepted at a 98% rate. Even that is curious, since nc I T one I spoke with in the design trailers was aware of reviewing l
@ the Quick Fix work. ' g < f 7. The uncontrolled authority of the Quick Fix progr am
- x., -l
%s. was abused. Quick Fix engineers completely redid the cesign cf N 88g-m h
- s. .
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- ,,*,gsff9%$P 9%@@hQ2,${gQ2lM;@.iyi'@hMEMMM$M$sfsMN!59 '
n .. ry 7, ' 4 - - p kf j^
) ,m h[jj.il hangers,-deleted hangers, deleted weld symbols from the - -: .
*l drawings, and took similar actions without the benefit of any E[f _
$m calculations.
The Quick Fix program was a substitute for the formal-(2" 8. h design and construction QA* reporting system throughout the @ program for hangers. At every step, from issuance of the [, j drawing through. final QA review, "PSDTC" could replace the
..- normal reporting system (Exhibit 1). This could mask the 4
deficiencies from NRC review, either through routine disclosure J or audit, unless the NRC audits the Quick Fix sheets
$@.i
$f themselves. A.; j$ 9. Quick Fix substituted an informal repair proSram for m;; i?f,. f controlled corrective action through the formal.nonconformance r i. \ y system. . Quick Fix engineers learned of significant 'E*
- A,.
deficiencies that are recorded on the Quick Fix sheets, instead of being processed properly through the nonconformance reporting { -system. To illustrate, I personally reviewed Quick Fix sheets N that showed ' fixes to problems such as studs welded to plates on i a hanger that fell off after the U-bolt was removed, concrete 8) pji i "shell" anchors that were cut off, and simi".ar defects. These wa y,;j serious hardware deficiencies should have been reported in the )gi ;. nonconformance system for trending and discicsure to the NRC. Jik Q, . SEISMIC DESIGN REVIEW a
. 10. The " final" calculations in the seismic design review re do not include the assumptions sheets, which means that e, -
specific errors cannot be effectively tracked. The c=ission is % cdd, since the assumption sheets were the cnly things the d-
,jg 1
$589ES WIj d gMT'Y?W sp,in9 M .M]Ydg 2 5j d i d @ )hys @ h .3 E sI M syt Q ' Aq lif.!l0 !$'I pg - hh5 . pfl' engineers reviewing our work were supposed to have' looked'at. - Md i) 11. - PG&E's December 28, 1983, letter to th NRC on small t. J bore piping s.nalysis does not contain enough information to h na support a' conclusion. The data is missing or inaccurate.for thermal anchor movements at relevant points maximum design and i operating temperatures: dimensions for the necessary length of pipe to verify txpansion, gaps at the restraint location, and' other examplacs. I will gladly discuss with the NRC the h specific significance thu each of these analytical holes has for the PG&E conclusion that there aren't any proble=s. [n - g 7
- 12. Similarly, PG&E's .'inuary 27, 1984, response to the ,
4 - Ig NRC on expansion anchors relied on data that was incomplete or {%d[('
, of suspect accuracy to support its conclusions. -For' example, I h.] - never saw drawing 054162 referred to on page 2, nor any a
@ calculations concerning the NRC recoc= ended design margins frcm g {Q.g.; - I&E Bulletin 79-02. By contrast, the analysis on page two of . hh the reply.does not mention the more significant NRC Bulletin 79-14,'which should have been relied on to include the f"jl kd , effects of flexibility. I never saw any calculation accounting ! l.'h' 1 for place flexibility. On page 9, the analysis covers Eeating, h'k,; Ventilating and Air Condition dt et supports and electrical %[$1-14 .aceway supports. However,. it skips more significant hardware 80 1- such as small bore and large bore pipe supports. The effect of 4.] the missing information could be to rebut the conclusions that 9 1 % are included. For example, in some instances th'e factors of
~1 l
- QP .. safety are described as approximately "1". However, according -)
W J g to page 9 of the PG&E reply, if the ind case is Hoscri er l Oc j %:. "5~ \ &l} l n .-
f Qgmfy317%%QGL M yQQ)qk;Q j}[h5$Q$$kM WMK[ fy < '- *
- g. .
tw R tid to - ] . Double Design Earthquake (DDE) seismic, the factor of safety is " _l 3 1 y cut in half. This would cause the anchors to fail by PG&E's /n + own analysis. I have thoroughly reviewed the January 27th PG&E j
.W '
N response and will di,scuss my specific analysis with the NRC. l d - 1 EFFECT OF DESIGN AND CONSTRUCTION QA BREAKDOWN ON WELDING y, 13. Because of deficient design drawings for welding, the { various departments were working to drastically different
@ assumptions about the penetration of certain welds. The San ih W Francisco office assumed one set of criteria, site design
'5.I Q engineering another, and QC a third. For example, in designing i 43' - gj bevel and flare bevel welds for tube steel, San Francisco k m assumed that the radius cf the tube steel is 3T for hangers da ( within their scope. I and others in one site design group i [ trailer assumed'2T. Other trailers may have assumed 3T. In i$. sore cases, the radius was only 1.5T in reality. That wculd a {f
-9 .
become the design assumption of a conservative engineering I analysis. In July, 1983, another engineer and I checked a O~ piecc of tube steel that had been cut out at rando= from the stockpile. On at least one visible corner, the radius was 1.5T 1 ' ,I'b when we measured. This means that San Francisco was 100'; off M g . in its assumptien. In my opinien, that is not censervative. i
.,:e a" Drawings depicting the differences between reality and f . assumptions are enclosed as Exhibit 2.
S . 14 The weld procedures and techniques failed to
-:a.
j compensate for the weaknesses in the drawings. It is possible S .
& to partially neutralize the impact of the deficient drawings by .
M ' tg w;y using 3/32" or s= aller weld red. Unfortunately, the techniques d i.';) C 90! %^ a..-..,~....,.._.,..-..-.----- - . . - - - - . - - - - - - --- :- - - -
{Q).M 1M I$'G E N$N$$$[ kb.I.Nb$$!hhNkbN N
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,9 .. 4 L Y, ay by.a , y Wu did not require the use of a rod that. size. A QC inspector "--l lMy told me' that welders often used 1/8" rods instead, and that the L ! : 1 . procedures allowed even more variation. - f.h
- 15. The main relevant Pullman procedure, ESD 223, did not-
- Q .
require the joints to be welded flush for flare and bevel TIb M a-a- welds. As a result, the effective throat of the weld is not accounted for by. procedure., An AWS technical expert agreed-with this analysis when I checked.it with him. If nothing @lf3 P
.$ else, this indicates a QA breakdown. I also believe the NRC Nb . staff was factually mistaken, however, to conclude that the ?: , . !
63 welds were flush in practice. I know this, based on personal
% 4 gj( experience from the Quick Fix program. I am still willing to m
g help NRC inspectors on a plant tour find examples of the welds 3,I described in ey f).R. The NRC did not ask for my help on flare A? - % and flare bevel welds when we took the plant tour in December. V r, U.d I believe that is why their own visual inspection missed the- ! rig ':M problems. dj !
$:y 16. When Pullman did specify an angle in its weld '
k+:Ti procedures and techniques, the specified angle did not catch '
.% h rEf . .a the relevant code requirements in'our design commitments. For
/v.R m:., skewed filler welds, ESD 223 permits an angle down to 15 degrees, Ed C which is not permissible under any code. For grcove welds, until June 23, 1983, the procedures required a 37% degree ry
} .
angle, which is permissible under the American Welding Society yh"" (AWS) code for piping. But it does not satisfy the requirements ! 6
.p .a s.
for prequalified structural steel or tube ;*: eel covered under - !k our design con =1tments to specific A=erican Institute of Steel
- ,djl e .,.
k f
1 6 3.II.dfflEl S $ $NEBi M ON D $ N ech Miri.1 M nh W . , SF M .- . h ye., .
),) ' %.
j Constructions (AISC) or AWS code provisions identified to us by . . . ,
- L ' The 15 degree angle is so far off th,at the AWS
{,j , management. , h code does not even list a way to compensate for luch an error. At our February 3rd meeting, Mr. Kirsch admitted that the NRC
- l. 7 ] .
pj;.3 staff had not examined my allegation #94 in this light. h,9 17. The lack of document control meant that ESD 223 was
..1 9 not applied consistently. This was a generic problem for
- q g.. . u welding procedures. From personal knowledge in Quick Fix, I
)
know that the welders did not have copies of ESD 223. In fact, 9{ C t,, they generally did noc have copies of the welding procedures
\'
k;1 ' h, they were supposed to be following. There was one case where a e d welder's helper did have a bootleg c~opy of a procedure, b'ut he %) ' cold me not to tell anyene because he was afraid of being 5 Q m t' [ fir s d. p[g
- 18. ESD 223 included inaccurate information in the skewed Q
R..y J fillet weld table. It did not adjust to use the sa:e effective M leg length assumed by the San Francisco office. N 19. QC inspectors couldn't read the welding angle and 4t N) , effective throat sy=bols and instructions on the design 71 Q( drawings. That is because the inspectors were not consistently ; %1
' j: qualified.to the AWS code, and none were issued the AWS symbols i s'81
..',0 code. h$
- 20. If QC inspectors refused to sign for welds because f,/;
they couldn't read the symbols, Quick Fir. responded by re=cving h ?.g the symbols from the drawings. In sote cases, these essential
.R,_ s._, variables were deleted without any calculations.
T. 5:J
- 21. Management failed to provide QC with instructicns er 9
N y ,.
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d8[ 44, s . wn..n a _..-,_,,..-.-.n . . . - . = ~ . . - , - . - . . - . a
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n ', f9 .
- g :
8kP y .
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g calibrated tools to measure the visible radii of flare'and * ~ -
-l s
n i 1 flare bevel welds on an as-built basis. As a result, the l
- i j ;
potential mistakes were not caught and reviewed by engineering.- 4 . b./ To my knot 41 edge, this feasible and necessary corrective action g,; - wi p has not yet occurred. ; 1 {;zj- f 22. I am concerned that the'NRC may accept a corrective
]
action program that does not compensate for past mistakes. For v.o)p-g< example, in response to allegation-#93, SSER 21 accepts the q.j explanatien of Pullman QC inspectors that they assumed a ; No - g 45-60 degree angle if the information was not included on the i drawings or relevant procedures to guide their inspections. h] y The NRC staff failed to go the next step, however, to learn (~ when they began this practice. It was not until a June 23, g] . . .
- 4. 1983, Pullman m'emorandum that either installers or QC 4
y inspectors assumed 45-60 degree angles. (Exhibit 3). A b Pullman official issued the instruction after I refused in the h h,o Quick Fix program to accept the 37% degree angle, which I h+ detected frem the information included in other Pull =an welding h t. techniques. (See Exhibit 4 for examples). In other words, the PW facts as described by the NRC only are accurate for new work by th d' Pull =an during the past seven c.cnths. @d fe 23. Corrective action has been inadequate to date on the M<< w bevel and flare bevel velds. The design groups could g h' . recalculate the welds, assuming a radius of 1.5T for all 9.- % tubing. Another solution would be to recalculate by assuming %s 3 ~~' that each joint is a butt weld with a 1/8" effective throat, - instead of twice that as assu=ed by San Francisco. To my b'
,f ff s.
!4 lo f
*___.....s..-.__ _ . . _ . . _ , . . _ _ . . . _ . _ . . _ . _ . _ . _ . . _ . _ _ _ . . _ _ . . _ _ _ . _ . _._.
emg. eg. e,)u.,ws ..W&h%. ~iWW. NWgiz._e' c - u #p.isW.M-%lU .u W M.aww. Ma . 4?W~d Gt.J.sJ a %a.de. (j_ py - aa , su f;; . : , . M:1 q bD .
,- r -
-s
~'! knowledge, neither management nor the NRC has taken this- .m, -.
h necessary step co make the design calculations pi.operly con-G ~
~
,l,d .servative. ;.
., s h+;i u
- 24. Corrective action for skewed fillet and grcove' welds h..., remains inadequate. As Mr. Kirsch pointed out on February 3, 1983, one way to tell would be to drill into the weld, insert a -
@c
- i. re
- F hooked wire and measure how far it extended, with appropriate 4
4 i g deductions for interior conditions such as slag'. In my opinion, ! bn another solution - for groove welds only - would be to re- @?l? : K g calculate, assuming that the welds were butt welds with an h effective throat of 1/8". Another approach would be destructive 4 . hj examinations. To my knowledge, none'of these solutions has
.c., (- been it:p1'emented.
hk I have read'the above ten (10) page Affidavic'and it is pq- : 74 true, accurate, and. complete to the best of my kncwledge and Vf tjaj belief. [$b I & Oh CHARLES STOKE 5 "'Q, I, CHARI.IS STOKES.CERTITY THE AE0VE TO BE A TRUE AND CORRECT STATE!E::T, ~~d!S i Sch DAY OT Teb ruar ,- ,1984 P@3 9 ' / A V CNARLES STOKES k' O , p , NOTARY //
;f , ']gj . I My Co:=ission exrd-es April 2,1984 l ;.9
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hb$'5 .?d$ $2h b $ bY.5 b1$$ $. 'Y k h$, $ h5k,'$0Yh$* pa 4 c' sees qe D 6.. H .,d a, UNITEo STATES "i. * [., 't NUCLE AR REGULATORY COMMisslON h f. [ h WASHINoTON, D. C. 20555
,, h s ,
,/ MAR 15.19M p * *** . .. _
y MEMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing Q Division of Licensing i - p@n FROM: R. Wayne Houston, Assistant Director for Reactor Safety K i,',l[ I:': :'f;ru : e- n':.- O
SUBJECT:
DIABLO CANYON HUCLEAR POWER PLANT,.UM TS 1_AND 2 - i STAFF AFFIDAVITS IN RESPONSE TO THE AFFIDAVIT OF 73
. , JOHN H. COOPER DATED 1/19/84 AND TO ALLEGATION NO. 177 i An affidavit and a response to Allegation No.177 prepared by C. Y. Liang, i and another affidavit prepared by F. Rosa, of my staff are enclosed. These documents have been prepared in response to the subject affidavit and alle- "Q;q gation; the/ are intended to provide input to the staff response to the s;.q Joint intervenor's Motion to Augment or Reopen The Record dated February 14, !
1984. ,,;9, Ei, By copy of this memorandum, the originals of the enclosures are being trans-mi - mitted to J. Rutberg (0 ELD). dl7 .m ;
* ' jp a
hgj ']** v t,/ t. v% f$ R. Wayne Houston, Assistant Director l for Reactor Safety 1.0 Division of Systems Integration l- 1 '; M
Enclosures:
IO As stated i..d
;d.
cc: R. Mattson f4 D. Eisenhut 2j T. Speis FTh G. Knighton' 'M K. Kniel bI R. Capra J. Rutberg (DELD) . .h .!N % L. Chandler (0 ELD) H. Schierling 3 *
. RSB Section Leaders A. Marchese %j ,
O. Parr Q , V. Benaroya W. Jenson j 1;f) Q R. Kendall , o ICSB Section Leaders 5 , 4 jj
~
kl Nanh,RSB OI 0 X24754 f
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Q{ y A . .
~ ]q' , , . & UNITED STATES OF AliERICA NUCLEAR REGULATORY COMMISSION i TJ '
V,v, . .::ld BEFORE THE ATOMIC SAFETY Al:D LICENS1f;G EDL.RD @j ,.
- .c .
Ir. the Matter of ) g %py r
) ~ @ PACIF?C GAS A.ND ELECTP.!C COMPANY ) Docket Nos. 50-275 OL W 50-323 OL- i d -
(Diiblo Canyon Nuclear Power Plant, ! 3
- d Units 1 and 2) l
%:s @ AFF1DAV]T OF CHU-YU LIANG h a .r . REGARDING RESIDUAL HEAT REMOVAL SYSTEM h;!} 1, Chu-yu Liang, being duly sworn, state as follows: I am employed by the U.S. Nuclear Regulatory. Commission as a h.] st 1. Senior Nuclear Engineer n the Reactor Systems Branch, Division of Sys-MITi "f2 tems Integration, Office of Nuclear Reactor Regulation. A copy of my ; z,y j i;. ,, professional qualifications is attached, > 1 have reviewed the Joint Intervenors' Motion to Augment or, If. .:c3 2. pji in the Alternative, To Reopen the Record, dated February 14, 1983, Part E [q and the appended affidiavit of John H. Ccoper of January Ig,1984 con-i.)y, /% cerning perceived deficiencies in the design of the Diablo Canyon Residual yV .- j f.;;, . Heat' Removal" System. _, N Mr. Cooper's affidavit concerning perceived deficiencies in the % 3.
- D b c'esign of the Diablo Canyon Residual Heat Removal System is essentially a ku,f (W reiteration of his technical concerns documented in Allegations No. 37
;g .
Ts thrcugh 45 and 177 with a few new items not previously addressed. My 0 resiense to his affidavit is divided into the following five groups of
. .M w ' 91 techr.ical concerns: <
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b
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,a . ] q (a ) Effectiveness of reatural circulation cooldown and the scope of the natural circulation tests at Diablo Canyon (Issue had not been '. 3 : h previously addressed)
~"
L.? l- I, (b) The changes made by PGLE to the Diablo Canyon operating proiedure ^ I n . c,o. l/ S n [q
~
B-2:11 requiring power removal from the RHR isolation valves after #T jeu l.h - t'hese valves were open conflicts with the PG&E commitment stated in (0 {a",j. the staff SSER No. 7 with regard to achieving cold shutdown from the E control room. (Issue had not been previously addressed) -. i ",] - s p rA t is . q g (c) The R:-;R hotleg suction line should be designed to safety related g($ r,g
,. l ', requirements. The use of RHR system as a part of ECCS during LBLOCA y., ,
and SELOCA. TMI-2 experience with RHR systems. (Portions of these f:.'.I .; ..
"h issues had not been previously addressed) a (d) Effects of inadvertent / spurious closure of RHR hotleg suction isola- c ,b i tion valves. RHR system design relative to GDC 34. Spurious clo- D' '! ' ^ ; "
c / .~ [ , sure of the RHR suction isolation valves is a frequent event. -;,vc D(
-! (Issues have been previously addressed in Allegation No. 37 through .$Iv,..n t , s.. <
? 1 45 and 177)
.]
. ( (e) The Staff responses to Allegation 40, contained in SSER 21, is inad-)
equate and further analyses / review are needed. (Coming from ,
%3 Arb ~ hmc..i- I n W> c.fjab ~' Y C/" .. Exhibit 17b of Mr. Cooper's affidavi )ft At the end of this affida-
' v. - i
),a3 vit, I have summarized my response to the allegations 40, 45, 177 !
I. . ,.."l ,
- q and to the concerns in Mr. Cooper's affidavit. .
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C,% Mj fy ,4. Mr. Cocoer's Ccncerns f3
;;j Mr. Ccoper has expressed en Pages 7 ar.d 9 of his affidavit, l
- N .
f ,goncerns regarding the effectiveness of natural circulation and the scope _
- j of the natur61 circulation tests at Diablo Canyon Plant.
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p - hp- ' Staff Resoonse
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@9 ,Mr. Cooper has expressed concerns regarding the effectiveness of RCS mj C1 cooTing by'natura1 ' circulation and the adequacy of the natural circula-
- m w,
tien tests.to be conducted during the low power test program. His con-f.;., cerns' appear to be based or,1 statements made in draft Reg. Guide 1.139 f that indicate natural circulation cooling is "a poor alternative" for m - [.M core cooling. Draft Regulatory Guide 1.139 has not been issued as a final
.A .w y.p document and does not represent any final staff position. The Staff's-
.nt .- y current judgment differs from the statement from the draft Regulatory .. .n N./ Guide quoted by Mr. Cooper on page 9 of his affidavit.
, . . A significant amount of information exists to support the efficacy 1
-lk of natural circulation in PWRs. Tests have been run in both the LOFT and
- ';j .
9 Semiscale facilities which demonstrate the ability of analytical models to s.a .a ,' predict natural circulation. In addition, tests hcve been run in Westing- >1 % house plants recently licensed (e.g. Sequoyah and North Anna) which have p . demon'strated this ability to remove decay heat by natural circulation. , , , g:d LE l Operational events have alte Ibnwed that natural circulation is a viable,
- q.
y effective means of decay heat removal. N 4;d With respect to natural circulation tests, the detailed procedures g- . T for the natural circulation and boron mixing test for Diablo Canyon plant h $ are currently under the staff review. This test will der.:enstrate whether E 4.; cr nct Diablo Canyon can be brought to the cold shutdown conditi-ons, 1 - i.D Even thouch Diablo Canyon was not reviewed ecainst the B P RSS 5-1 natu- - c;3 44 .
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M '. m4 y ral circulation test requirements, the tests be'ing proposed essentially-2 ., . a .'est the RSB 5-1 position. The low power natural circulation tests that
'ij .
Mr. Cecper referred to on page 7 of his' affidavit are designed for cpera- _~ ~ tor training per the requirements of item I.G.1 of HUREG-0737. 51evia-
- j}
p. bility of the natural circulation cooldown in PWRs was previously addressed M
~
in the NRC steff testimony of W. 1ensen regarding Contentions 10 in the $p
.8 Diablo Canyon full power hearing.
7.y T.h, Following a loss of offsite power, decay heat removal thrcugh secon-hQ m;y
.4 dary system is essential until the primary coolant pressure and tempera-q,g.
ture reaches the conditions which permit RHR system initiation. As long gq as the condensate supplies are available to the auxiliary feedwater w y.) pumps, the decay heat generated from the reactor core and the sensible
'c1 V. heat in the RCS could be continuously removed through steam generators q
7,;l with sufficient natural circulation in RCS. ,4 p 3] . ,;j Staff Conclusion g In summary, it is the Staff's conclusion that cooldown by natural Pj circulation is a viable, effective means of decay heat removal that has - L'j ,s, been extensively demonstrated. Moreover, Mr. Cooper has not identified . i.[.) any specific problems or issues that can be specifically addressed, b-) .u d v.! 5. Mr. Cooper's Concern 1
;$ , On pages 10 and 11 of Mr. Cooper's affidavit, he discusses the a ppa r,ent inconsistency between the PGLE commitment addressed in Staff Id SSER !;o. 7 and the Diablo Ccnyon operating procedure B-2:11. SSER .';o. 7
} . . states that ell operator actions needed to perform plant :ooldown can be $
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?.;y . %j ; * '- m .. [;f accor.plished fror, inside the control room. However, operatinc procedure 96
'E-2:11 reovires an operator to leave the control room to manipulate the
./;,$jj breakers for valves 8701 and 8702. .
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w, : ['Q = 3 P. Peleva t allee!!ien !! umber ff. 4
', V p
lione-pf] . M.1 - f" .d Staff Response N,)s.?
/ At the time the Staff SSER llo. 7 was issued in May 1978, there were 8
V no procedural requirements to remove power from valves 8701 and 8702 in b any mode of plant operation. The applicant intended to always have power f.G Q g! available to valves S701 and 8702. Thus, a plant cooldown could be con-g ducted without any operator actions outside the control room. This de-c.:n f! sign was consistent with BTP RSB 5-1 which had just been implemented by
.a I the Staff. As a result of fire protection review PG&E was requested by 9
t the Staff to remove power from the RHR suction line isolation valves 8701 g and 8702 during power operation. This is discussed in SSER flos. 8 and 9. q n,y ; i This action was intended to ensure that a fire in the vicinity of the RHR. j g'1 u. Q. . isola ~ tion valve control circuitry during normal operation would not cause . f.,i spurious opening of these valves and thus initiate a 1.0CA outside of m.
$ containment. Also as described in the Staff response to allegation 45, PGSE was requested to remove power from these isolation valves after M]' .a p.) being opened. This action was required to prevent spuricus closure of 1
w the Valves for P.HR pump protection and to reduce the possibility of a low w g r-RCS temperature overpressure event. MI
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q, ,) ' jd Thus there are two manipulations of the 870; and E702 motor poner ,3 a supply breakers, both requiring operator actions outsice of the control d.) . '(.,., i room. The first action, re-instating power to the isolation valves in - h;j preparation for valve movement to iriitiate RHR cooling, was founch to be hd W necessary as a result of a fire protection review. The second action, l removal of power from the isolation valves after being opened (during TD <hutdowns) was based on RHR pump and low RCS temperature overpressure r; conce rns .
- It shop 1d also be noted that BTP RSB 5-1 is not considered a re-wat ki quirement, but rather is one acceptable means of meeting the Commission's 1 regulations. Staff experience in implementing BTP RSS 5-1 has shown that Q -
in some instances it is necessary to allow, on a case by case basis, T,, 'l b limited operator actions outside of the control room,to achieve cold
' r:
1 shutdown. 5
.I I
1] 7 1 Staff Conclusion a .,j a, The position stated in SSER No. 7 with respect to the ability to
.g l conduct a plant cooldown frorn the control room and the requirement to remove power from the RHR isolation valves during normal plant operation. .
t .0.); are conflicting requirements. 0; ,S However, the requirements for power removal from the isolation t.. :; . fin v'alves during normal plant operation and shutdown cooling mode are A - 'd es acceptable and the deviation from the BTP RSB 5-1 is acceptable. As dis-c:; p.]) cussed in the response to allegation 45, af:er installation of the low
,;. 4 hj ficw alarms, the Staff is requiring that t'.e rpower remain on the isola-
- 0 -
. . . ion valve during RHR cooling for protecticn against LOCA's outside of l,' 1, s' s
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j.y * * . , ce y e W c ontainmer.t. As discussed in SSER lics. 8 and 9, power removal frorr. tte
'il isolation valves during normal plant operation will continue for fire
- 7 g protection considerations. .
- . 0, :
1 The actions that need to be taken by an operator outside of The
., d fRi ccr.tr:1 ror ar.d the time necessary to take the r have been examined and dj
[q have been determinqd to be acceptable. The applicant has verified that V Q it will take less than five minutes for operators to reach the motor [',g 3., tM control center one level be. low the control room and a short distance away ljp
.J to manipulat,e the breakers. Also the operator will not be exposed to any y;j 4 unacceptable environmental conditions by going to the motor control center.
- .x;,'l 1 ._
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- 6. Mr. Cooper's Concern On page 3, 4 and 5 of Mr. Cooper's affidavit, he states that flow
- 4. .
I; from the RCS hot leg to the RHR system through the single inlet would be I w required for mitigation of a small break LOCA, as was evident during the hi
.i 1 m, sli!-2 accident. Therefore, he asserts that the RHR r.uction line from the Fl RCS hot leg should be redundant.
n Si . [.,; ' . Relevant A11ecations Allegation llos. 40 and 177 -. ld :s Staff Resoonse -)):. As the Staff indicated in'the response to Allegation No. 40, a large O 0
- portion of the RHR system is designed to serve a dual purpose and, as was d
d stat'ed in the Staff response to Allegation 40, the ECCS portions of the d td l').j RER syster. at Diablo Canyon meet the single failure criterion. As a part n .q 1.1 cf the ECCS, the RHR pumps take suction initially frem the R'-l5T and later, f m
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- to durir.g the recirculation mode, from the containment emergency sumps. This e,,
porticn of the RHR system is designed to provide injection or long tern
. recirculation following a large break LOCA. For a small break LOCA, the
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- a .
Ij f RHR system does not provide injection since the RCS pressure is $ormally , F ex well above the shut-off head of the RHR pumps. However, as in the case-of the large break LOCA, the ECCS portion of the RHR system allows long 7q .. - T. term recirculation. Contrary to Mr. Cooper's statements, more than , Ms $... enough coolant would be available in the containment sumps for this moc'e l
- c
-u of operation. If the RWST inventory has been reduced by continued injet- , ;p.
- p... .5
- fon, the fluid lost through the break will be available in the containment W:
y sumps. .i. S.I Mr. Cooper has cited the TMI-2 accident as an example of a SBLOCA
- v. l SJ where the RHR system was used. This is incorrect. The RHR system was ,
k ,g [ never relied on for injection, long term recirculation or decay heat 7.Z e removel. (i.e., suction from the RCS hot leg). Decay heat removal was ]
-b
,$ initially accomplished by using the steam generators and the auxiliary ,.n ,,f feedwater system. E l
.,c e:) .
Staff Conclusion Mi , .. rN The portion of the RHR system relied on for ECCS function have been MI j j] reviewed and approved by the Staff, and are in conformance with 10 C.F.R. 50.46 and Appendix K. The Rns hotleg suction line is not a part of ECCS $l\ p and is not required for mitigation of any size LOCA. The hotleg suction line,is used for plar.t cooldown only. i@d g'a? As was stated in the response to Allegation 40, the Diablo Canyon No 1 % cesign with a single RHR suction line meet the position of SRP 5.4.7 ar.d - 4 , 't
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% Branch Technical Position RS5 5-1 for a Class 2. plant. USI A-45 is per-ts
' forming further assessments of the reliability of various decay heat J/]3
,A ' { ( removal system designs. _~1
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4, N,. Mr. Cocoer's Concerns & Staff Desconses-
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$6 - A number. of other concerns were expressed in the affidavit that l
.1
%.hp .. y have been addressed by the Staff previously. These are listed b'elow. @m (a) On page 2 of the affidavit, Mr. Cooper discusses the possibility 49 of inadvertent closure of the RHR su;(ction isolation valves. The Q h m ,4 Staff has addressed this concern in the response to Allegations 45 [l and 177.
- 2. ,
il (b) On page 3 of the affidavit, Mr. Cooper discusses the concern that y.) '/ the single suction line does not meet GDC 34. ]. ': The Staff addressed Js this concern in the response to Allegation 40. Mj. (c) On page 8 of.the affidavit, Mr. Cooper states the belief that spurious m, closure of the RHR suction line is a " recurring common cause" fault that can cause both safety related RHR pumps to fail. The Staff ! (M.;j response to Allegation 177 specifically addresses this concern.
'Also, the Staff has discussed the necessity for a low flow alarm in. ;
o . H the response to Allegation 45. (..J, I 1 [,a;j 8. Mr. Cooper's Concerns 1 In Exhibit 17b. Mr. Cooper has the following comments to the di g i Staff response en Allegation No. 40. h (a) The RHR system should be iaanalyzed in the light of the TMI
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&w . hy Q (b; Two 1,5ction lines from the RC$ to the RHR are provided -in CE, B&W, "n and the newer Westinghouse designs. 4'--- $, (c) In the proposed Regulatory Guide 1.139, the RilR system is re. quired
~
- p A to be redundant, and withstand any type of an accident, not hu'st a h
W laroe break 1.0CA. k @a (d) ' Accessibility of the RHR isolation valves inside containment during Enj at: radiological conditions. p . 4 1 Relevant Allegation Number g-
% Allegation No. 40 and 177 -
N,,... Staff Resconse y. 'E
. My responses to Mr. Cooper's coments are as follows:
- k) '
, (a) The RHR system was not used during THI-2 post accident operation, jdi-
@ The. lessons learned from THI-2 on the RHR system are irrelevant to Wp) y the subject issue (single failure concern). (b) The early designs of Westinghouse, CE, and B&W plants all have only $ ,, j ; one RHR suction line from RCS hotleg. The current designs are %J M.ij equipped with redundant' suction lines. The NRC Unresolved Safety . El j'd Issue A-45 is assessing the adequacy of the RHR design with respect 3 dl# pu d - to the single suction line and the pressure interlock features on
' )
@s the suction isolation valves. ; (c) Preposed Regulatory Guide 1.139 has not been finalized or issued by W .the Staff. t,q k p (d) PGLE has informed the Staff that the. radiological ccr.ditions in the
- ^J
,.m vicinity of the RHR valves are acceptable for the cperator to eriter - ?
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.the area during normal plant shutdown conditions. During accident- .. o
M - conditions, decay heat should be removed via steam generators and ; Q. git m. r,t j
- the auxiliary feedwater systems or.after a LOCA, by the RHR, system
.in.the recirculation mode. ;
l$ Q] Conclusion' - g .. ff
- As discussed in item above, the RHR system was not used, nor was it
%h '
FA.n - necessary during the THI-2* accident. The single _ suction line design, 1 ls(4 h.
- which is common in most operating reactors, has been reviewed and ap-q.a proved at Diablo Canyon, and is acceptable.
The generic implications of S gw[ RHR system single suction line are 'part of the ongoing. Unresolved Safety l9_
.an .
fM - Issue A-45. y ," . lgj. The Staff's current judgment differs from and does not endorse the fy] lel8 s, statements in the draft Regulatory Guide 1.139, as described in the item fq- above. The Regulatory Guide 1.139 is still in draft form. The positions fg stated in the draft do not reflect current regulatory requirements or N4 .
$'q positions. - .
l bMu im . < M f h
]
Overa'll Summary w,, Mr. Cooper has raised a number of questions and concerns centered LA i; around the adequacy of the Diablo -Canyon RHR system single suction line.
?.@ ^
- 2 Th'e Staff'has addressed each of these concerns in its responses to Alle-P -
y t4 gations 40, 45 and 177, as well as in the discussions provided above. 5j Jn sur.cary, the Staff believes the RHR single suction line design is acceptable, and USI A-45 will assess the overall reliability of decay a
' 13 hRt removal systems.
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m$s u gf The Diablo Canycn design has the capability to remove decay heat
@p withcut the RHR system by relying on the steam generators.and the auxil-hj% { iary f,eedwater system. There are ' adequate water supplies available to - "
h.% ' h the. auxiliary feedwater system, and.'there is the capability to ut!ilize b% . c[ backup water supplies should the safety related condensate storage tank. . _ d E (' CST) supply be depleted.
$p The Staff believes that natural circulation cooling of the PCS is w .
p;.' <. ' viable, and there is sufficient operational experience, experimental data hij . f and analytical calculations to confirm the validity of the p'rocess. [iy Mr. Cooper is incorre'ct 'in his assertion that the RHR system single %rl . suction line must be available for mitigation of certain LOCAs and that lMJ.ln -
- y. this path was used during the Till-2 accident. The Diablo Canyon design t
g$1 places.no reliance 'on this flow path for any LOCA scenario, nor was this
.r t h.
- n1,.
path used in the TMI-2 accident. t ?. lj,j ' Furthermore, the Staff believes that installation of the RHR low pH [q flow alarm will provide positive indication to the operator should either 7N . { I r,$i - RHR suction line isolation-valve inadvertently close while the RHR l M Kj- pump (s) are operating. The installation of this alarm will b'e completed h$ y prior to power operations. Y v M
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Ig: I hereby certify that the answers are true.and correct to the best 9 f.j -
'of my knowledge.
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'w,, /Jtt SU Subscribed.and sworn to bef (WlL,,, .
this /5/Mday of dhe d. ore 1984 me b '. w ..' 5 e:.t / .. q/ - C$ ! .4&. c.c 2 G i.) ' ' 4,', liotary Public
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ppwuwggggggggggsig@gMm&ssMMSiyaBW w.n** 1 ee . [ff p Professional Qualifications - j Qg ; Chu-yu Liano Reactor Systems Sranch- .
.v, g Division of Systems Integration 1 g;
U.S. Nuclear Regulatory Comission . 5 j .
,j 1~ am employed as a Senior Nuc'elar Engineer, Reactor Systems Branch, s.. .
" 1
; Division of Systems Integration, U.S. Hucelar Regulatory Comission, '
s ,.k Washington, D.C. The Reactor Systems Branch is reaponsible for reviewing reactor license applicatier,is and evaluating the design of reactor. l:i
} systems, ir)cluding the residual heat removal and emergency core cooling l systems, of the nuclear power plant with respect to nuclear safety. As j
g .
.part of my duties,-I have been responsible for reviewing the operating h, license applications of several PWR facilities with respect to reactor N
M systems. f From 1965 to 1967, I was employed by Lockwood, Andrews and Newman, h.) l. e i Inc. .(Hour l ton, Texas), where I worked on the design of mechanical systems , i for public buildings including heating, ventilation and air conditioning g:h . {i[j systems, central plant and emergency power systems. l W From 1967 to 1969, I was employed as a mechanical engineer by Avon- $s (Y , dale ' Shipyards, Inc. (New Orleans, Louisiana), where I worked on the N design of marine steam power plants for tankers, destroyers, and cargo 7
, ships.
41 From 1969 to 1974, I was employed as a Senior Engineer in the
- u. -
Department'of Systems Engineering, PWR Systems Division, Westinghouse Electric Corporation (lionroeville, Pennsylvania), where I worked on the f 9,,, design and review of nuclear power plant auxiliary and power conversion %i systams. I served as a lead engineer for 26 Westinghouse PWR plants, ,; l,m W h.k W %_n_ __--- _ -- - -__- -__ - - --__- - - -
M22C?iL2SEME%i10]It$&$5h $ E!LDbSNb?$$$ $EUSW Ob b ya 2 w,1 . . . r 51 providing balance of piant design criteria and 11555 interface recuire- %p.5
' tents' and assisting plant designers (e.g., Architect-Engineers) in the
,f ?: [.)* Areas of auxiliary and power conversion system design. . 1;. g From 1974 to*the present, I was employed by the AEC, in the[ w. Eg Avriliary and Power Cenversion Systems Branch. Divisien of Technical __ "id.i
$ Review; following the reorganization of the AEC, I served as a systems ]N
[ engineer in the Auxiliary Systems Branch, Division of Systems Safety, U.S. fluclear Regulatory Comission. In 1980, I commenced employment with m the Reactor Systems Branch, Division of Systems Integration. 4 ' Y I attended the Cheng-Kung University, Taiwan, and received a B.S. h - ,i Degree in Mechanical Engineering in 1960. I received a Master of Science j%h ' $n Degree in Mechanical Engineering (majoring in steam power plant design)
'A .
i$d from the Oklahoma State University in 1965. I have also attended the na d Graduate School of Engineering at Catholic University, k'ashington, D.C.,
.N y where I took a course in Nuclear Engineering.
- n 2" I am a member of the American Society of Mechanical Engineers.
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EN1217 17b b , SER R E Ed LJ T T d 4 L. 1-10-84
!?G4, I received e ec.py of NUREC- M75, yl;l ,
On. January 7, 3 ')
- S u p p l erne nt No. 21 - the Safety Evaluat son Repcrt re;esee to the - _~*
9 ' c.peration of Diablo Canyon Nuclear po~er Plant, Units ; and 2. [h,, - Thi s coeurnent cont ains the of ficial NRC responses to. the gij " allegations" (NRC$ s t errni nology) which I anece to Mr. Iugene . j Powers of the NRC Of fice of 2 nsceet i c.n a nc Enf orcerne .t cn August
.?
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- Co .. - i ; , a l teos t -o are o i e . . a. . d yea s a;c.
gf-One we.u l e t n : ni+, ; i vs.s. 't h e a rnou r.t c,f I i nne a va i
- a bl e to the Wp NRC t o eval uate any .cc recerns, that the2r engi neers woulc have ce ne a thorc. ugh , lob of eval uat i ng thern. Frc.rn readi ng the appropri at e.
@] '. portions of this supplernent (Allegations G7-45), h e.w e v e r , it be cc.rne s apparent . that4thi s dc.eurnent has been hastily prepared during the last few %eeks in an at t ernpt to inolli fy the growing Q,[; publ i c concern over- the l arge nurnber c f out st anding problerns at p Di a bl o Canyon. The nutober of t yp og ra.ph i c al errors cont ained in ih thi s cceurnent gives sorne indication of the Arnount of t 2 ree spent 47 in it's cre parat i on. E< ut rnore serious, t o ene, is the lacx of g res cons i veness t o rny ori gi nel quest iens, and the lack c. f just i f 2- [9; catsons supp12ec by the NRC for for what 1 cons 2 der t c. De 2nace-quate answers. NRC personnel, in their " responses" t o rny %] i concerns, cont i rtue to ignore the basic facts of rny case, the [Q Jj, opere, ting exper2ences at DiabloLCanyon anc at other nuclear power pl ant s. and t hey ec.nt i nue to insis- that, Cc.de of Federal N Regulations notwithstancing, repeated rnalfunctions 2na safety-M related syst ern are not considered to be a " s 2 ; ni f i c ant safety h concern" unless the systern rnal f unct ions when called upon to act ually peric.rrn it's saf ety f unct ion dur2 ng an eees cent. I arn
- g)
% convinced - ano I t h i nk enest thinking ir,dividuals we.uld agree - tnat t he t i ene to correct problerns, wi t h a safety-related s yst ern is jy bgfgrg it is actually needed to prevent an accident or safely
- shut down the plant.
Ly NRC represent at ives have told sne that i f tne anal f unct ions q* about which 1 aan concerned had happened when there was fuel in L, g tne reactor, or when the systern was called upc.n t o f unct i c.n, then Q m they would be considered a significant safety concern. Since, ho ever, the systern failed ( t wi ce) before fuel was loadec, tnere y was no threat to the health and safety tc. the public, t h e re f c.re l n c. saf et y probl ern. This philosophy is legally, rnorally, ano y%j - l logically bankrupt, and is akin to saying that a hich-speec h $N -
,, aut c reebile bearing de.wn on a pedestrian is no tnreat to nas health and safety until it act ua l ly hi t s hi rn.
W!5 % 7he f o l l e.wi ng paragrephs ec.nt a s n any e r.a
- y s e s c.#. re: ut t e.l s k t c., end ec.inene nt s on the NRC "respenses" ( c.r actually, l e,e x c. f re s p: nses) t o any "ellegations".
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A E ~- AT I ON No. 37 "The s c. ) i d stete pec.teet~ ion systere (SSL3) [g:l M re3eys thet initiate closure of RHR l et cown 2 sol at i on va l ves 6701 a r',d
-[ 6702 perforrn no safety functic.n, reduce tne rea:e-cil2ty of the RHR syst ern, and cause e pot ent i a) for RHR purno ik.i., d arna g e. Therefore, these re:ays 'hou3d s be rerne.ved. " .p -
e y)(.
/. . 'The NRC response to this concern is very disturbing, since W it di'sp2eys a tctai 3ack of understanding of hcw t hi s syst ern
, functions. To ene, this is especially disturbing since t-h i s response i s f rorn t he NRC eng i neet-i ng st af f i n Washi ngton, DC'- U the su; posed " experts" on nuclear po-er plants. h in the first place, the SSPS gses 991, initiate the closure R,
,d c. f <alves. 670* ano 67el. 7ne autornat t e cic s ure c.f. .s. .ne s e 'v a l ve s initiated by the changing of stat.e of a "ec.rnparat c r" enoc u l e
~
2s [
.. } oca.t ed
- i n t he Prot eet i on and Cont ro) Racks in the cab 2 e g spreading roorn (directly below the control reorn) . rora t here, the
.A circuitry passes through four input relays in the SSPS ( I c.c a t e d 3 in a separat e roorn, a d,l a ce n t to the control reorn ) and then back y down t c. ,t,h e cable spreading roorn to the Auxiliary Safeguares Racks, and 'then to the anotor contre.1 centers for the valves. The ESPS ,1p pp ggy arnp2 i fies or changes the signals, or perfortes any y logie fune, tion w:th t h ern - the signals inerely pass tnrough sne 4 rel ays' i n the ESPS. Why PGLE cont i nues to insist s n et "The s o i s c' l .n . state pec t ect ion syst ern e c.'inp 3 e t e s the I c.g i c f unct i c.n aric N generates a larger output signal iarnps. ) which in turn actuates relays in the auxiliary logie cabinet" and tne NRC ecnt2nue to g ,; insist that "Thi s autornat ic iscletion function [is; pe r f ortae c by
% ine West i nghc.use desi gnec SSPS" is t ruly arna:i ng. Cert a:nly soin l
] c. f these organizations have access tc. the circuit cis;rarns for thi s syst ern and the expertise to uncerstanc' thera. One res g it l Q"h .
a linoct conclude that this i s a del i berat e at t eract to taake tnese
] eircuits appear to be a part c f tne engineered saf ety feat ures of
),ys the SSPS when they really aren' t.
- s. Secondly, both PG&E and the NRC cc.nt i nue to insist tnet tne
!(? RHR systern has a ternperat ure/ pressure interlock systera to S autornat ically close valves 8701 and 8702 i f the ternperature or pressure in the reactor coclant syst ern exceed pred e t e rrni ned
@%.; values, thus preventing an "i nt er-syst ern LOCA". E<oth ?GSE anc the NRC are aware that the Diablo Canyc.n Technical S pe c i f i c at i c.ns i require that the pc wer be r erne.ved f rc.ra the actuate.rs for these fj valves during the period when the autornat ie closing act ion woua c Uj be desired. It is a anyst ery t o rne how bc th org ani:at 2 cns ec.nt : nue j Q t o raske this cl a i rn when they knew that the pc wer i s rernc.vec f ec tn the val ve act uat e.rs, prevent ing t hern frorn aute roat ica} }y closing : ,y unless, again, there is a deli berate at t eropt to inislead the public.
'l The NRC asse rt s that " diverse indications and al arrns are
;: rc.v i c e d in the cont rol reorn (inclucan; a RHR syst ern I c.w f3cw e l a rra to be installed during the f2rst refueling outage) to allow D the operator (s) to assess RHR s y s t e rn status and to alert .t h ern to 2 pctentiel s y s t ern cegradation." Gec.rge Or-el l ould be p r c.u d of N :he euther of this sentenee; it is such a fine e w a rn el e of " ne w-I speak". In the sarne sentence, the presence of an alarrn as
e 2 e f med, and the schedu)e fc r insta3 ling it is g : v e ri, pe-naps the - M)Q
, rJ C .vou l d like to explain h c.w it -as pos s2 bl e, -ith all tnese Ny c.i 7, 2 (( k 1k i
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Q "c: ve-se a nc 2 c at a c.ns and a l a rrn s f c.c en 49 p urn; to ce run ! ss -:t hc.u s uet : on, unr.ot i ced by the c perat e-s, for pn; ppg , er,til M' st es cart. aged, 6 urn: nad tc. be replaced.) as happened at M abic. Canyon Jast Octocer, tTne gl4 r Last2y, the NRC states that "this a13egetion does nc.t p involve considerations tnat questi d en pl ant readiness f c.r pc.~e r . f ascension test ing' or full pc.we r c.pe-at i on", and yet goes e.r. t e.
~ say, several pages later, that "during the first cycle.of c,pe rs -
Il7!: tion, pla'nt s operate enore frequently on the RHR systerns testing $ and training requirernent s for a new p2 ant. Thus, the period ~of % vulne-abi2fty to e s:iuri c.us RH: sveticr. CV e}osers reey be k greater snan in suosequent cycl e s. ' T-(ppa re nt l y, ir; these i e si a.: e <r;e m s . tne NRC sees reo h3 : :.. t - e : i ; t i o r.
- f W
%:f e S ALLEGATION No. 36 PG&E is ignoring evidence that the spurious g closure of a rnot o,e operated' valve i s r.ot " i snpos s i b l e " . <d ,3 The NRC$s position, here, was very wel l-t irned. They state: h "The st af f has exarnined in depth the l icensee's act ic.ns in res pc.nse t c. an event involving the spurious initiation of R:-::: h;[{ rne.t or operat ed valve closure as well as the concerns expressed by the alleger regarding the pc t ent i a l for such event ' sic 7, ane plh $: cone
- udec that t i rne) y evaluation and correct ive eneasures ore
$;3 t aken to preclude repetition of such conditions." In the fe h g[ weeks since-this s t at ernent was inade, eniibel' "s;urious i ni t i at ion
- c. f RHR suction valve closure" at Diablo Canyor, causec d e rna ge to 3 an E-4R purnp. This is twice now that the NRC has proclairned that j
'd the pec.bl erns wi t h this syst ern have been " re s e.1 ve d " after raa* i ng l I only paper changes. How rnany rnore t i rne s will the Diablo Canyon @d ha RHR purnps have to suffer carnage bef ore the peop*,e in charge out l
- 's there realite that the problern is i nh e rent in t he circuit ry, not the pec.cedures ? As of today, PG&E and t he NRC coht i nue t c. er.pouse
['r.f the s arne philosophy which has in the past l e d t o purn p c arna g e and
}oss of decay heat r e rnc.v a l capability at Di abic. Canyc.n G.nc at f; rnany other plants around the country.
g y Over three years ago, 2 gave PG&E copies of 16 Licensee Event Reports dc.curnent ing cases of " spurious initiation of RHR , suction valve closure" in various plants around the country. I ' ra sure that rnany rnore cases inust have c.ceurred since that titne
'. j besides the latest one at Di abl o. But PG&E still ce nt i nue s t c.
k} W.; contend that "A failure, such as the s p u r i c.u s closure of a rac t or operated valve. . . has nc.t been ec.ns idered credible." ( 50. R, page 2.1-3), and that " Westinghouse does nc. t consicer sourious Q operat ion of e 3 ectrica31y contro12 ec valves as a crec:ble s:ngle active failure" (FSAR, page l '5. 4 - 6 ) , and that "The probacility of $1 l d *
.* any spurious ve.1ve closure is therefore 2.f4 x 10 t o t he rn i nu s ]
6th pc-e r per valve-hour." ( F S A::, page 6.3-34a). In the face of l (% W t ne over-hel tni ng evidence that spurious valve closures happen cuite regularly,_1 c a n c.n l y interpret the f at lure c f ?G l.2 a r.d tne l h$ NRC to recc.gni:e this evidence (anc e.c t uren it) es yet e ric t '.e c - at em: t o tr.i s 2 e e.c t h e oublic as to sne safety of the A rt ; s y s t e re.. dg Tne N::C states t h e.t "It e c.e s e. pear ther tne licensee as givia; proper at t ent t on t o t he s p u r : c.us c l os u re c.f tne v a,1 ve s an
.i J.; cuest:en". I would lake to point out s nat ane rc f il gpl i e r,. 2 s c.nl y l Im !
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&,,;1.. the faret st ep t c.~are corree 2 n; t ne c;e f: e s e nei es : n t . .e : -:: * @ s y s t ere.. 9g .,1g,. ' i s nec e s s a ry t o a c t u e l '. y' e c.r r e c t t . .e pro:,le. ene Ti"~ prevent it f rern ha ppeni ng again. He.w tr.a ny inc.re t 2 me s wi ' l e r. A.' 1: - (fj' ,
"p u rn p have t c. be damaged at Di abi c. Ce nyor, before tne r.ecessary gd act ion - wiring the interlock circuitry directly from the negan , ,
'f reeks" to the " Auxiliary Safeguarc Reeks", wi t nout it pass:ng -
'q thrbogh the SS?S, and adding the I c.w RHR f i c.w alarm - will be taken? -
...$..g
- y o
V )* e . j pLLEGOTION No. 3? "There is no centrc.1 r-oc.m ennuneiatior, p r c.a. :e: 4 t'o ~ a l e r t the operator (s) when the RHR letdown line hes eeer, j , ,,2.s ol a,J e o c u r i n g .~ oc e s 4, 5, anc 6 (not snuscowr, ec Ie s.n e t -
- y. d e.w n , and refuelin; respectively).
%] h PG&E was instructed, on April 2, 1981, to install a low R.4 R
;j fl ow clarm in the ecntrol room, but was. el lowed t c. wa: t unt i l
[hlMg af ter the first refueling outage to co so. Apparently PG&E i nt end s ,t,c; wai t until the last permissable minute to instell this a l a rrn, everi the. ugh its presence wouac have prevent ed the d e.rna g e A}q. - to the RHR pump wnich c.ecurred t wc. mc.nt n s ago. Tnis e.itituce coes not bc c st. /ny ec.nfidence in PG&E's ec.ramitment to the safe, Qd rel i abl e operat ic.n of Di ablo Car.yc.n. ,M , The NRC i ntencs t o ac'here tc. their original senecule for the 64 installation of this alarm, even in the face of this seconc' $0] incicent, statinga "The staff has concluced t h e.t the ex2stin; cont rol roc.tn indications and pec.ee d u re s at e suf f ici ent tc. essure 7.'] w. Q edequate decay heat rerac.v a l in the i nt e r i ta. "
, .c .
g ALLEGATION No. 40 "The quest i c.n rai sed w e.s witn reg ere t o .< .et he r gh[ or not the single RHR pump suetion line f roin the RC5 hot le; rneet s safety related st andarc s. The newer PWRs are c e si gned N., with redundant RHR pump suet ic.n lines f rc re, the RC5 het le;s." 6 EfP Again, the NRC ' eni ssed the boat on t his one. My ce nt ent ion {.d, was that this system should be reanalyzed in the light of the TM1 {m. accident. PG&E ele.ims that the single RHR suction line i s not pt safety related and is only used during the norrnal ec.c I cown of t he @ plant. I disagree. At TM1, their RHR system was usec t c. rai t i g at e d the consequences of the now f arac.us accident there. I propose that %[.[a this portion of the Diablo Canyon RH2 syst em is inececuete, s:nce qS a single failure in this line we.uld prevent decay hee.t reme. val via tnis syst ern. I offer these facts as evidence: N.. ~ .
- 1. The pec.ven unrel i abi l i t y of the s uet i c.n ve l ves 2n this
.,g line, both at Di ablo Canyon anc' at c.ther pl a nt s.
f 2. That t r e,s se'eiv r e ), f,, t e c s u c t i c.n lanes f ec.m t .e ECE to the RHA are p r c.v i c e d in CE, E &W, e nc' t h e ne-er We s : r.;nc ..se g
.. l c'e s i g ns. Why -c.u l d these e xt r- 22nes be provicec :f . ey e
k . .-
-eren't neecec?
.l m
M 3.. In cegule.tc. y Gusce 1.13 ? , the NA staff sieies te- ne
~
$ resacual Heat A e rne.v a l System is -equ: red to be r e : e nc e r.
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,;.., .J jg g] Apoarently, tne NRC is aware of the deficieneles in snis and
=er.aps other areas in the cesign c.f D 2 a b 2 c. Ce nyc*n, 52nce they.. .
fl.h consicer it a " Class 2 Plant". They go on t o st e.t e that "A single _ [h P.H R suetion line f rorn the RCS hot 2eg is consideced acceptab)e pg for a class 2 p2 ant as long as a single failure could he W4 corrected by enanue.2 actions insice or out si ce of cont ai ntnent., or the plant could be returned tc, hot standby untl2 rnanual a ct i c.ns p .o-
- e n : -m , 6e acco':.isaec.'
-M e :-e c- :ose e :ves ;c.r.
M at t hi s coint. ~%% ' u de- #Ed_io}o.gie..r._). e m:itioJP e ich , would -' q'n:o . . prever.: nunnan ent ry i r.t c. t h.e cont a i ntne nt (sberF,g' were f c.ur.c in E tne cont e.i nenent at TMI), could val ves 6701 or 6701 be opened j'O rnanually or repaired?
- Both of .t nese valves are Joested insice R} ,the cont a i ntnent structures at D1'a .o Canyon. .
.3 <M
- m d ALLEGATION No. 41 The power source of certain relays is nest s h c.w n
% y?q on certain drawings and this caused an operational p ro b l ern, .Q the fai2ure (closure of RHR i sol at i on val ves) " Although the contact s of the ESPS input relays are snown on $}'i
- h. the elee:rical schernat s e of this sys t ern, the power source for tne relay coils is not shown c;p env f22C2veg glagt gtgwigg. The NRC continues to state that the Septernber 1981 ineicent was cue to a
%(f [.V.,,,
" lack of pre-p2anning" i rnp2 yi n; . t ha t if the technician responsible for the incident (and his f orernan ) had just done
]y their job properly, the spurious closure would not have occurred. It seerns t o ine, that if the i nf ortna t i on i s not on any drawings,
#;j then no arnount of pre planning will help, and errors are bound to
<W .y be in a d e . The NRC takes great pride that a technician-drawn p " c ornpo si t e crawing" of this sys t ern has been put t c.g e t h er. I put gj together a sitnilar drawing back in May of 1961, and gave it to the i nst rutnent f orern a n, but that obviously dien' t prevent the
- f. incident 3 rnonths later. Hand drawn, or " bootleg" drawings can (h; get lost, or not be distributed to the persons who need the y
rd infcrmation. That is the whole purpc se of the Drawiing Control Sys t era - to enake sure that eggyrgtet yp te; ggag infgrLngtipp is d readily avellable and accessable to those who need it. Nc.t only Q is this idea just good corntnen sense, but it is part of t he Cc.de ni a of Federal Regulations: .
'u'* r
.J y
h 10CFR53 Appendix B ,' I I I Desi gn Cont re.1 W *
.xs< Measures shell be established to assure that appliecole Ay reg ul a t ory rec ui rernent s and the design basis, as cefinec in I paragreph 50.2 and as specafied in the 2icense appi i eat 2 c n, for g those structures, systems, and cornpc.nent s tc. which th2s apcencix ap;iies are correct l y t re.ns l at ed i nt o specifications, *ewings, J pro ecures, and i nst euct i c.ns. "
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.; L.3 ) nLLEGAT20N No. 42 L i censee sne.ne gegnen was unrespons:ve :( r e e prn- '
?1 - enendat ions t ce prevent spur 2ous e2osure of tne 2sc1e :on
~ :t). ve)ves on the residua 2 heat r e enc.v a l (RHR) sys t er...
y.f Closure c. f
^ the valves disables operation of the RHR systern for cecey l heat reinova ' - -
[ ' The chr ono2 c gy of event s in this case speaks fc.r i;t se l f. At i,j no t isce did PG&E take any effective act i on t o re se.1 ve any o f any Y4 concerns without the int ervention of the NRC. Mc.s t of the the M1 origina2 prob erns 4hich 2 brought first to PGl.E's, ther tne NRC's l a t t ent i on rerna in uncor rect ed even t oc ay, after three years of l
@ unsuccessful at t ernot s on any part , ,N 2 n the NRC as sesstnent of the sa fety signi ficance c.f t hi s - -- -
[',. ) prob 3 ern, an atternpt is being snade to give the i rnpres s i c.n t h e.t l PG&I, L'e st i ng h ou se, and the NRC have all been actively wcrking l for the past three years so3ve these pr obl erns. I find this very )
'h hard to believe, since' no evicence of Any kind has been p rc.d uce o l to support this position. As far as I' rn ec.neerned, the 2ast
$[ L.N t considerati.on that this problera received was i n Nc.vernber of 1951, wnen the Diablo Canyon Onsite Review Group decided to t ake nc. l ! act ion to correct the problern.
, :) .'.
1,
il l ':li ALLEGATION No. 43 The 3oss of the r e s i d u e. ) h e e.: re rne.v a l (RHR) ] >.
syst ern on 9/29/81 due to unplanned elc.sure of the RWR iso-lation valves was an event whien shc.u 2 d have been re;c.-ted to s$ the NRC i n acec.rdance wi t h 10CFR50.71. The 1:censee's failure
'l* to enake such a report was i n vi c l at i c.n cd NRC regule.tions.
Q.q
$.; iOCFR50. 72 Nc.t 1 fi cat ion of significant events i i.'] ta) Each licensee of a nuclear power reactc.r l i ce n s ec' under j.j . paragraph 50.21 or 50,22 of this part shal1 notify the NRC Operations Center as soon as possible and in e.22 cases within one hour by t elephone c.f the occurrence of any of the H[.- following significant events and shall identify that event as 4 being reported pursuant to this section:
[J m. Mg "(6) Personnel error or- procedural i nadec;uacy whi ch, during h norrna l operations, anticipated operational c.eeurren ns, c.c N accident conditions, prevents or eg u,y p r e v e r,.1, by :tse)f, t.kd the f ul fi l 3 <nent of the safety function of t hose st r uct ures, j sy s t ern s , and cornpone nt s i rnport ant to safety that are r.eeded rfj t o. . . ( i ) rernc ve residual heat fo))ow:ng reaetc.r snuscown..." Q ,
.. (any ernphasi s)
Q av ihe NRC cl aiins t h e.t "The loss of res:cual heat rerec.ve.; c e ::e.ei t v i? ch curing a t iene when significant f i s s : c.n p r oc' u c t c ee s.y he a; is h present in the core wou d have safety s g ii f i co :e. - n :n s
% particular 2nstanee, fuel hac r.ot been loacee inte t e D:eelo Ca nyc.n Uni t 1. Therefore, no fission procue: cecey n e e.: -es Q,1 present and l o s s c. f RH:: eaoability had ne. e e: ue.1 safe y s : ; n : f i c e r.ce. " Again, t he NRC coe s.n' ; perceive e r.y s a f e : y p ec.::; e m
[f)1 7 (D ,.} unt: 1 the speeding a u t c.tno b i l e actuei y nits the pe:estria . - e, . ;
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g? O w'.jE G.:T I ON N c. . 44 ine licensee falleC t c. pec.pe*)y proces s a Wk
, Nucleer Plant Pro b l ern Report,
'g";[. g/ I n rny or f g i na l ec.ro p ) a i nt I stated that'the Nuclear Pl ant " - P r o n l ett' Report dveurnenting the first loss of RHR suction at [ Di abio Canyon was "si gned of f as ecinplet e witnout any plant
$ ena na g erne nt review. . . el as si fi ed as ' rec.n-r e pc.r t a b l e' and' wi t hout .% a ny f ol l ow-up act i c.n such as an RHR purnp inspection or ~
f/ Sl; itvestigation ir.te the cause e# the eve.t."
~%e tJ:' e fi t s t et ve *ne a cic.ve ec.ncerns are true (als nougn t wc. rnonths af t er t he w e s e r. . , # ,ot ne ;-: :, ere repo-t -as i n: t : At e - t :- pe r f: rre 4. purg A ~
test), they Just bel i eve that this is an acceptable ay te run a [ p c.w e r plant. In iny di scussi ons wi t h Mr. Jess Cruse c f t he NRC, who interviewed the "princi ples*!. in the handling of thi s probierc f,' report, he stated that "no-one denies that it could have happened .;'; just the way you said 'it Edid3, and I sort of concluded snost likely it did happen Ethat way3". What action did Mr. Cruse take? None. Mr. Cruse also stated that this was not ' repc.rt a bl e' J ,'.[.j
%j 5- because there was no fuel in the reactor at the t i ene. Again, it would seern to rne that the NRC would wish to kreow of probl erns
[t _be#pn there was fuel in the core, but this i s apparent ly not so. M As for the analysis of tne probl era to prevent reoccurrance, this h 4 has not been done by PGJ E even to this cay, as evicenced by the recent (Nov. E3) rep 2 ay of the Se p t ernbe r 1951 incident. PGLE and the NT<C both clairn that " strict procedura2 cc.nt rol s " are adecuate @%] X to prevent reoccurrance, a l t hou g!h tnis rnethod has been proved ingses ',H.t e twice before.
- m, l
.A ALLEGATION No. 45 Section 5.5 of the Diablo Canyon F5AM describes h
9.: the autoelosure i nt eri c ek f o r- the RHR suetson line i sol at i on ) valves (6701 and 6702). Sect i on 3. 4. 9. 3. a of t he Di ablo Y[i; Canyon Technical Speci ficat ions req ui res pc.wer to be rernovec W f rorn these isolation valve oper ation during enc oes 4 0 :.t j I shutdown when RCS cold leg t ernperat ure is less that 323 dh kg degrees F), 5 (cold sh ut down) , and 6 ( re f ue l i r.g ) . This re-q u i rernent defeats the f unct ion c.f aute. closure i nt er Ic.ek for M, s the valves.
- Aq hy In their lengthy analysis of this s i tnp l e a32egation, the NRC adinit s that r ernc.vi ng the power frern these va l ve c.perat e.rs c:e f e at s R
h the autoelosure interlock to the RHR suction va2ves as cescribed 31 in the FSAR. I contend that either the FSAR s h c.u l d be ec.rrect ed $j so it accurately describes.the RHR sys t ern at Diable. Canyon, or the RHR systern should be operated i n conf c.rrnance wi t h the FSAR. Q - ' The Cc.de of Federal R e g u l at i c.ns is c! ear in be.th cases: The NRC b" reust be notified :. f the plant design de es nc t conf c.rrn t o t he criter2a wd bases in the FSAR, a nc t h e F S A R' ra u s t be kept up to case. e y$ 21 , (b,h 10CeR50.55 Conditions of const ruet i c.n p e rini t s w fg te) (1) If t he perini t is for ec.ns t ruct i on of a nuclear pc+e r , jf m i !' plant, the hoicer c. f the perens t s9621 nc t a f y t ne Ce r rii s s a en ! 7 ( l 7To a
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f"* ' N. IOCFR50.71 Maintenance of records, snaking of resort s - g . . . . M (e), each person licensed to operate a nuclear pc.we r reactor @ pursuant- to the provisions of paragrapn 50.21 or 50.22 of % this part Ebtll M2G&it 9tri2GiEallY, as provicied in k'.p , paragraphs (e) (3) &nd (4) if this section, ibt fiDtl Eiff%Y. Yd EDt1Y111 C129C% .iEEE31 originally subrnitted as part of the y applicat, ion for the operating license, to assure that the M i nf orrnat ion included in the FSAR contains the latest anat er i al
- k. developed. This subtnittal shall contain all the changes necetstry to reflect inforene.t i on and analyses s ubeni t t ed to
${ $ the Cornini s sion by t he licensee or prepared by the licensee $y pursuant to Cornrni ssion requirernent si nce t he suer.>.i s s i on of If the original FSAR or, as appropriate, the last upcated F 50 R. Q v;p The updated FSAR shall be revised to incluce the effects of: all changes tnade in the faellity or proceoures e.s cescri eed 6 in the FSAR; all safety evaluations perforrned by t he licersee Yj either in support of conclusions that changes dac not involve gj an unreviewed safety question; and all analyses of new safety
- j. i ssues perforened by or on behalf of the licensee at i V Corarni ssion request. The updated i n f ortnat i on sna21 be appropriately located within the FSAR.
$ (3) (1) A revision of the original FSAR containing those hA original pages that are still applicable plus anew M re p l acernent pages shall be filed within 24 inont hs of eit her h K. July 22, 1960, or the date of issuance of the operat ing license, whichever is later, and shall bring the FSAR up to M.} date as of a enaxirnurn of 6 rnonths prior to the date of filing the revision, y".) yyl The NRC, in it$s a n a'l y s i s , curiously avoids any #nention of the above two regu2ations, but goes on to say that operit ing the d.y . plant with the power retnoved f roan these act uat ors is a violation it 'of their Branch Technical Position ESB 5-1, Pos i t i c n F.1. C. They M.) also state that "There have been tnany oece s ions of sourious ;M; ] suction valve c2osures on ( s i e') operat i ng pl ant s. This has resulted in not only a loss of cecay heat eernoval, but also an @7" overpressure event due to the loss of the 3etco-6 f i c -pe t n. " ~5ey { ce nt i nue that "During the first eyele of operatio., 21 ants oeerete'ecce f requent ly on the RMR s y s t e *n as a resu' . of . ma i nt ena nce, testing and t r e. i ni n; re p u a refne nt s for a new pl a nt . _ $q Mus, the pe'iod r of vulnerability tc e spur:ous :-R svet:en r?.N a - e (m) E _ _ _ . _ __- _ - . . _ ...s - - - -
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' UN1TED STATES OF AMERICA l' NUCLEAR REGULATORY C9MMISSION f
h;g. . BEFORE THE ATOMIC SAFETY AND LICENS1HG APPEAL BOARD h.d.I
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g.v; in the Ma ter of .
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!g PACIFIC' GAS AND ELECTRIC COMPANY ) Docket Nos. 50-274 OL b% ) 50-323 OL
~
hf . (Diablo Canyon Nuclear Power Plant ) -
;% .U.its 1 and 2) )
e,
$q AFFIDAVIT OF FAUST ROSA REGARDING M RESIDUAL HEAT REMOVAL SYSTEM ll" h.
- I, Faust Rosa being duly sw*orn, state as follows: l
;t/o s 1. I am employed by the U.S. Nuclear Regulatory Commission as Chief, In-j)! ,
tj l strumenta. tion and Control Systems Branch, Division of Systems Integra-a g tion, Office of Nuclear Reac~ tor Regulation. ad g{- ,
- 2. I have reviewed the Joint Intervenor's Motion to Augment or, in the alter-m; Q
- native. .to reopen the record, dated February 14, 1984, and John H. Cooper's
&[M:a affidavit of January 19, 1984 attached thereto, concerning perceived de-Ea. ficiencies in the design of the Diablo Canyon Residual Heat Removal System.
- 3. Mr. Cooper's affidavit concerning perceived deficiencies in the design I of the Diablo Canyon Residual Heat Removal System is essentially a re-(c] iteratinn of his concerns documented in Allegations No. 37 through 45
- t. r
"$j and 177 with a few new items not previously addressed. My technical G5 gej j
, evaluation of his affidavit is limited to the following three areas y
involving the instrumentation, controls and electric power design of the
&,k Residual' Heat Removal System:
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[j . * . i; yh, (a) The use of relays and power supplies in the solid state pro- ,( ; blf tection system (SSPS) to provide the automatic closure fea- ' cp + pg ture for. the residual heat removal (P.HR) system isolation .x
-l 59 p valves whenever the ' reactor coolant system (RCS) pressur'e.
ik pg
. 1 '\ r t exceeas 7 pre-determined setpoint. -
4.% 'D [..ct"' hf 4b) Non-conformance of the design to the recommendation of Regu 5 latory Guide (RG) 1.139 in regard to failure of a power DO D (( }@S M;i supply causing a change in valve posit' ion. q? th A,n., (c) The lack of control room annunciation.or alarm of loss of RHR system flo,<. q None of the foregoing matter:. oh a concern regarding design yg1 e.
- ] quality assurance. "
Til J2 . Mr. Cooper's Concern a '4'. ' Pages 1, 6 and 121 (Pg. 2 of Exhibit '17B) of Mr. Cooper's affidavit $f ( reflect his view that the use of relays and power supplies in the M g . SSPS to effect' automatic closure of the RHR isolation valves whenever A.a RCS pressure exceeds a pre-determined setpoint is unnecessary and . .s
- d should be eliminated; the design is such that. loss of the SSPS power Eh 3 1 supply will cause an unwanted automatic closure of an isolation valve
+ '
,)d with consequent eventual RHR pump demge assuming no operator action. t - h I The valve's referred to are motor operated valves (MOV) 8701 and 8702. $hi' e,q. 7 Pi ' %}t i( . 8
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- gj nelevant Allecation Number l@j ,
VQ ~ Aliegation No. 37.
$1,o ,]
Staff Response - P . 1
,4,;7.4 The original staff response to Allegation No. 37 was provided in Supple- %v ment 21 of the Diablo Canyon Safety Evaluation Report (NUREG-0675). This Ji affidavit is intended'tisuppTement the original response to this allega-jl tion. .e 6l rey:
- It is my. understanding that the automatic closure feature and the prevent
',l.y
- a. opening interlock for the RHR isolation valves are as described in Amend-eg ,
h ment 4'of the Diablo Canyon FSAR, Section 7.6.2, Residual Heat Removal M 43 Isolation Valves. This section of the FSAR is provided as Attachment 1 c.m . f;f to this affidavit. Mr. Cooper's concern is with the detailed implementa-33 tion of the automatic closure feature.
'M M
y1 As described by Mr. Cooper, the initiating signals for automatic closure
,..- originate in RCS pressure instrument bistable modules (also referred to .~
g as signal comparators). Thus, for each valve, one of these signals is Mc input to the SSPS where it energizes an input relay; a contact from this h relay is used to energize, using an SSPS power source, an auxiliary relay g$n i located in a engineered safeguards cabinet; and a contact from this aux-iliary relay is in turn used to initiate the closure circuitry in the n,. . . i motor controller of the isolation valve.
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y , k h snould be noted that the diverse automatic closure signal (pres. I If surizer steam space temperature, for one valve only, see Attachment 1-) ,c 4f - - G., u is' incorporated into the signal from an RCS pressure bistable before rg M this signal leaves the instrumentation cabinet. Therefore, this aspect
/. .,
c' -t e t . c. .a t i c ci e s . e c'e s i;r is r.c r ei s . a r. te M . Oce;er's cor; err.. m 9 As stated in the staff response to Allegation No. 37, the automatic l M closure circuit is designed to " fail safe". on loss of power, i.e., p . 2t to initiate*
- closure of its associated isolation valve should loss of
%l ,
4.1 control power occur. This is required by General Design Criterion Q (GDC) 23, Protection System Failure Modes; which states:
- i. 'l J
3,.1 i f "The protection system shall be designed to fail into a na safe state or into a state demonstrated to be acceptable FM on some other defined basis if conditions such as discon-
.y ' nection of the system, loss of energy (e.g., electric
/1 power, instrument air) or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, jPd and radiation) are experienced."
n.
@ The principal safety function of the RHR isolation valves is to protect !$1 the RHR system from overpressure and possible consequent LOCA outside ;d q containment. The principal safe state for these valves, therefore, is q
closed. Thus, the loss of any one of three control power sources: (1) M) ? 4 y, the SSPS power used to energize the auxiliary relay, (2) the power feed
;)
- 3. to the RCS pressure instrument, or (3) the power feed to the pressurizer c:.1 steam space temperature instrument (for one valve), will automatically
-i ,
M., close an isolation valve. Also, in order to meet the channel / train in-4 f dependence requirements of IEEE Std. 279-1971, for each valve, all tnree
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y,p power feeds should originate from an independent inverter supplied vital - W il .1 l
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- . 8' cM instrument bus. It is my understanding tha't the design has been imple-h.. mented in this man,ner. Thus, the loss of either one of two vital instru-e ,
k raent buses would automatically close' an isolation valve. As'_long as the t B{ fail safe. feature is retained, this automatic closure would occur on loss ; M
~~
ot.,rci psatr ra;&rdlass ci it.5 so.erce. !,hl f$ It should be noted that the RHR design does not provide automatic closure "[q. of the isolation valves on loss of actuation power. This is because MOV's M *
- p inherently fail "as is" on loss of actuation power. Thus, they will remain M ,
closed, i .e.., in the safe position, if actuation power is lost when they k[N y.4
.[.3 are ac'tu* ally performing their principal protection function. The fact L
that the isolation valves will remain open if the po. car failure occurs n, 4 - during the cooling mode is also acceptable for the following reasons: (1) t a RCS pressure transient requiring closure of the isolation valves con-
' d, y current with or immediately following 2 loss of . actuation power is a 4 -
@ very unlikely event, (2) redundant sources of actuation power (offs % - h @g and onsite emergency power) are available for each valve, and (3) the ?$ b W control and instrument power for each valve is an independent battery [ backed inverter so that, given the loss of both the offsite and onsite G.m 3 actuation power for one valve, the other valve would have its independent 'd g .t onsite actuation power available and its independent automatic closure f! 4 circuitry available to close the valve if this was needed for protection e, v . K against an RCS pressure transient. Therefore, in my judgement, the over-v d all instrumentation, control and actua: ion power design of the RHR isolation g - 6 4
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fV , m.n h valves is'in full conformance with the requirenents of GDC 23, provtces 1 3% M:? suf ficient assurance of RHR decay heat removal capability and RER ^-syste .. y,{, overpressure protection, and is, therefore, acceptable. S?t, I have no direct knowledge of the specific considerations involved in th @b A cecistor. to use SSFS inpet ra.ays and poner scarces to imp'iemar.t tr.e au 4g.4 It obviously could have been implemented dif fer-g] matic closure function. e ently. It is noted, howeve,r, that'the SSPS is the point where the tran-e;:d p n! sition from protection instrumentation channel (s) output to protection M M ' system train (s) input normally occurs, and from which the train o d
;;$i protection actuation signals normally originate, for most safety-rela
$) . :t This is true for the reactor trip and all engineered safe-functions.
$j The automatic RHR isolation valve closure d ouards actuation functions. l circuits are safety-related and redunt'.nt, and include instrument chan l.;i Therefore, I believe
- I inputs and train oriented actuation signal outputs.
jy d na that the SSPS was used in order to meet the channel / train separatio M. ;T independence requirements in a manner consistent with the gen v 2 .: t @jQ used for implementing these requirements for essentially all the pro ec pj id tion system functions. l.Tj so y Conclusion td. ,. . The design of the automatic closure circuitry for the RHR isolation %j . .% valves meets the applicable regulatory requirements for safety-re-g;;:, , lated systems; these include the " fail safe" feature required by $[ w . E u. k
.]
1 I
- * ~ ~ - - . . . _ , _ , , , , _ -
,2 , . i $ h i$lL ? $ ? $ M $ $ ki$ 0$ b? $ $ $ $ Y b Y Y Y Y $mr6 . g*, w 7 %a tW GD;-23, Protection System Failure. Modes, and the requirements for w g . protectio.n channel / train independence and separation of IEEE Std. 2,.; '- Wj . 279-1971, Criteria for Protection Systems for Nuclear Power Gener - 4 .@ ating Station (10 CFR 50.55h). Therefore,Ifindthedesignaccepts d o . able. ViY . 3g. . . . . ( ('clj 5. Mr. Cooper's Concern h l fj Page 8 of Mr. Cooper's affidavit cites the lack of conformance of the ac l gj design 'of,the Diablo Canyon RHR system to the guidance provided by 7.i . j] Regulatory Guide 1.139, Guidance For Residual Heat Removal. t et In the l i h!] area of instrumentation and controls, he states that "The- Diablo Can- ' 11 'J e yon RHR system does not meet the criterion that " Failure of a power
. c.
@3
'd supply should not cause any valve to change position"."
. ... ~ Relevant Allecation i C. ! F%. None. d.h OL
'1 Staff Response s{
}j] The above cited " criterion" is taken from Position 2.a of proposed 44 R.G. 1.139 dated May 1978. This version of the guide was issued C for public comment. Subsequently, a draft Revision 1 (dated June Q.. , '~3 5: 1980) of this proposed guide was prepared by the staff. In this re- ) y . vision the guidance regarding power supply failure reads as follows: h- "Upon loss of actuatino power, (emphasis added) isolation valves should fQ g not change position unless movement is to a position that provides W N w . - t
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rs .,. ., . m .; eater safety." Neither version is ' consistent with GDC-23 wnich re- [,i M - quires a fail ' afe s design on loss of pcwer without~ qualification as S;y . . ,4 o "whether it is control or actuation power that is lost (See the staff ~ " ~ s
- response'in item 4 of this affidavit.); the original version of the 1
ll '
,O guide does not specify a fail. safe design, while Revision 1 speciiies a W ;
flD fail' safe design only for loss of actuation power.- id iD . - . . - [f0! However, neither the original version or Revision 1 of this guide was offi-' l g@ . cially issued by the NRC. Therefore, this guide does not reflect Commis-sion pblicy or guidance and is not used by the staff in the review process. (h a i Furthec development of this guide is now deferred pending completion of I P@/ Unresolved Safety Issue, TAP A-45, Shutdown Decay Heat Removal Require-
%( l ments. Diablo Canyon will be subject to any new requirements relating to
\
l 3. r.s instrumentation, control and elect'ric power that may result from the work
)) ,:0 of TAP A-45.
I l 1
-m ..;} . The acceptability of. the existing RHR system design in this area for as-u-
d3 suring plant safety is discussed in Items 4 and 6 of this affidavit. i W . d. N3 Conclusion 9 1[ers The regulatory guide cited by Mr. Cooper has not been officially ;
)( issued by the NRC and is, therefore, not applicable to the evalua-h 8 .
tion of the design of the RHR system. The principal criteria used (J r , by the staff for this purpose in the area of instrumentation ar.d con-2 2 3 trols are GDC-23 and IEEE Std. 279, as stated in Item 4 of this af-y
@d fidavit.
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- 5. Mr. Cooper's Concern
' .l (dh - Pages 6'and 123 (Pg. 4 of Exhibit 178) of Mr. Cooper's affidavit re- ~
m@.1 flect his view that an RHR system loss of flow alarm should be pro- [ hp l N vided ,in the control room immediately. i EU . h
.!( ,
Relevant A11ecation G
',$ Allegation No. 39. , I d )
1' '
;jj Staff Res'ponse f j As stated in the staff response to Allegation No. 39, the licensee was i . ..
- h. required to install a loss of RHR system flow alarm in the control room 1 5 j jl, during the first refueling. The staff found this acceptable based on i M. the following considerati6ns which in aggregate provide a high degree of M
j assurance' of decay heat removal. capability: (1) the presently available A['N,
,c, control room indications of loss of decay heat removal and RHR system !;g M status, (2) the time available for the operator to take corrective
.y$ 9)1 action following a spurious RHR system isolation, (3) the alternate gj
- r. . y, q means available for decay heat removal in event the RHR system is in-
.a d ' operable, and (4) the technical specification requirements that provide $m., assurance of sufficient decay heat removal capability. ! ] .e $$f. . isf 1 N n ! a m -H 9 ,h l fka 1 !sj .
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ihe licensee has since committed to install a RHR low flow alarm prior [q
~
T to entry into Mode 1 operation (PG and E Letter No. DCL-84-057 to' G. W.
$is] ,
(d Knighton (NRC) dated February 15, 1984). The licensee has also identi-
@Wlfi fied and described the administrative controls and procedures which are pg in effect and wf11ch govern the removal of power from the RHR isolation
$j i
valves (MOV's 8701 and 8702) by opening the associated breakers; this-
?@O.
s.a will be done with the yalves close'd in operating Modes 1 through 3 and M ; 3 We have established that E with the valves open in Modes 4 through 6. ! Q h
,, c opening these breakers will not deenergize the valve position indica-tion lights _in the control room.
L
?h The removal of power from these valves has been evaluated from the
.u9-qj standpoint of operator ability to c'onduct a plant cooldown from the k^
d control room and found acceptable; this evaluation is provided in Item N l'.j 5 of the affidavit filed by Mr. Chu-Yu Liang of the NRC staff. The s 71 accelerated installation of the low flow alann and the removal of
)
.4 jd power from MOV's 8701 and 8702 during RHR cooling should effect a sub-I stantial reduction in the vulnerability of the RHR pumps to damage due !
Q@4 to spurious closure of the isolation valves.
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6 %2CL W.L 2L?$db5$2EN1 SSY$ SW~ Y' f,; . . - e .. 11 cv4 . t.H1 i,. m e ,1) It is noted + hat opening the isolation valve breakers during RHR cool-gy - tQ ,. ing defeats the automatic closure overpressure protection for the RHR.
~
t ii .,- g.g system. However, af ter insta11ation' of the RHR low flow alarm, the
..,9 jp staff will require that these breakers remain closed, thus, the auto-
- .atic ciosure feature wi'i oe reinstatto. In tne interim, Ine RdA M a .
safety relief valves, and the plant and RHR system status indications and ala'rms available in the control room coupled with the administra-
??
JJ tive controls in effect' during RHR cooling; provide sufficient assur-j@g:j ance that overpressurization of the RHR system will not occur. N[: !$ Conclus'io'n b In the interim before installation of the RHR system low flow alarm in f:,; f,i , the control room prior to initial operation in Mode 1, the existing 5 @A c control room status indications and alarms and the existing procedures IM are sufficient to assure adequate' decay heat removal capability, and in 1 3l)0 nq V conjunction with the RHR safety relief valves, will also provide ade-Q quate protection against overpressurization of the RHR system. ll! - y.$; 7. Overall Summary 1 Mr. Cooper has raised a number of concerns regarding the adequacy of a 1h the instrumentation and controls design for the Diablo Canyon RHR f system. The Staff has addressed these concerns in its response to t9
- fl Allegations No. 37 and No. 39, and in the discussions provided above 0]9 k . in this affidavit.
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..j&;;; In summary, the Staff concludes that:. (1) the design of the automatic l i .d HT4 6q closure circuitry for the RHR isolation. valves 'is acceptable, based
- , m.. .;
gj on its conformance to the applicable regulatory criteria (GDC .23 and 3.,1 . lj IEEE Std. 279); (2) the citation of R.G.1.139 by Mr. Cooper to sup- {h fo
;; r: t 's p: ! '- ' o . :.at l os s O f ::r. r:'. ;:,<!i - s r.:.'. i r.: . re s & i r, i s:,- 1 'j.
hN1'r lation valve closure is not valid because this proposed guide.was not nc . y,}j formally issued and does not refle,ct official Commission policy or
.o. .l I:/.,i guidance; and.(3) the existing design, procedures, and control room in-k F dicatio'ns and alarms provide sufficient assurance of decay heat re-9 jj '
moval capability and RHR system overpressure protection during the p. a [t.') interim until the RHR low flow alarm is installed prior to initial $h f.h entry into Mode 1 operation. 9 Q .u., The' above statements and opinions are true and correct to the best ,:.::1 p of my knowledge and belief. . h C 0 9,a J3
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$p 41' Ifj! Subscribed and sworn to before me N] this 67h day of(/- ) kid /1984 1% MA M yJ
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,'7.6.2 7.ISIDt'AL F. EAT REMOVAL ISOLATIol; v/1VIS
{,, Q ' M - Descriotion kl ?M
- n. -
( There are ,two motor operated gate valves 'in series in the inlet line from the h Reactor Coolant System to the Residual Heat Removal System. They are normally 1
.:r. '
closed and are only opened for residual heat removal af ter system pressure is S.s 4 y refute!. .bt".sv apprezimate'y 400 psii, and syste. te.=;,trtture has 'seen reduced t-g',. approximately 350*F. (See Chapter 5 for details of the Residual Heat Re=cval
.$$ System). They are the same type of valve and notor operator as those used for g.. accu =ulator isolation, but they differ in their controls and indications in the
' j following respect:
$[p ' l w 1. One isolation valve, that nearest the Reactor Coolant System, is inter..- @(n y ' locked with a pressure signal to prevent its being opened whenever the {j system p/ essure is greater than 425 psis. The valve vill also be closed automatically whenever the system pressure increases above approximately 600 psig. This interlock and automatic clostag action $s derived from',,
" one process Control Channel, h@d t
n g 2. The other valve, that nearest the Residual Heat Removal System, Ls j similgly interlocked and automatically controlled. Conttol signals are h- " derived fr'om a second process control channel. In order to ce= ply with IEE-279 and to provide diversity, this valve vill also be prevented frgs 4
% opening when the pressurizer vapor space temperature exceeds approrimately 455'T and automaticAl[y closed when the pressurizer vapor space te=perature I k@..a J
exceeds approminately 490*F. This te=perature control signal is .4erived M . from one process instrumentatidu protection channel. - m}
- n Analvsis
;]
hy Eased on the scope definitions presented in Reference 2 (IEEE-279),1971) and
$gg Reference 3 (1EIE-338, 1971), these criteria do not apply to the residual heat re= oval isolatien valve interlocks; hevever, in order to neet AEC requirements
((et and because of the possible severity of tbc crnsec.uences of loss of function, - Md the require ents of IEEE-279 vill be applied with the following co =ents. C 1 pj
'fl 7.6-3 Amend.ent 4 (February 1974) 3.d H%]
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n' ., ' e'g,gr .ed l . .. 2 7er the purpose ' of- applying IEEE-279,1971, to this circuit, the ' h following definitions vill be used.- s
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- a. Proreetion System -
1
-Tne'two valves in series in each line and all components of i ;
~
l $m;y .t. heir interlo:hing and closure circuits.'
- 4
. .J Protective Action b. g gd Tne. automatic initia, tion and maintenance of Residual Heat Removal . System isolation from the Reactor Coolant Systen f . . d). g,, pressures above redidual heat removal design pressure.
~
M , n i
!i IEEE-279, Paragraph 4.10: Tne requirement for on-line test and 8.yA. 2.
calibration capability is applicable only to the actuation signal g' . and not to the isolation valves, which are required to remain n? closed during power operation.
- M r.v g c:
ve IEIE-279, Paragraph 4.15: Tnis requirenent does not apply,.as jy 3. 4 the 'setpoints are independent of mode of operation a.nd are not Bj changed . i.h A 4 N Enviro =. mental qualification of the valves and viring are discussed in M
$ Section 3.11.
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% 7.6.3 M rovide i-
. ilectrical interlocks (L. . , limit switches) a during fue1Aiandling cing the possibili of damage to the f mary ceans oper ations . M nanical stops are rovided as tb .
/ ets of the m
.jg preventin uel handlingj eti, dents. For ey (=ple, safety as
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M ' N M WPDSffFCWF9p p g g geg g ~ g @j . . . ' ., rg . FAUST ROSA ? . $l! d PROFESSIONAL OUALIFICATIO!!S R{f A;q i INSTRUMENTATION AHO CONTROL- SYSTEMS BPANCH r,$g .
'DIVIS10H OF SYSTEMS .1HTEGRATION
~
M $ 1 nave.onen empioyco oy Ine hucitar Reguiatory Comnission since January 197i. W From January 1977 through 1980 I served as Chief,'PBker Systems Branch, and g a since January 1981 as Chief, Instrumentation and Control Systems Branch, both g , Prior to these i h branches being in the Division of Systems Integration. p%,6 5 assignments,' I served as a Section Chief in the Electrical, Instrumentation M,1
?N .
and Control Syptems Branch, Division of Systems Safety, and in the Plant Systems h 9 Branch, Division of Operating Reactors. I have participated in the revies of @} instrumentation, control and electrical systems of numerous nuclear power 1 ..:ij stations and in the formulation of related standards and Regulatory Guides. t it M ' The Instrumentation and Control Systems Branch performs an in-depth technical Q W1 review of the design and operation of nuclear power plant instrumentation and g-m control systems important to safety including: , protection systems, engineered ,jy 41 safety feature control systems, safe shutdown systems, information systems, 11
@ interipck systems, plant control systems and essential auxiliary supporting gj a1 This review includes a comprehensive assessment of these systems systems.
[ q@h for all power reactors for adherence to appropriate codes and standards and g .
. encompasses complete evaluation of applicant's safety analysis reports, h Further, the k
d}- generic reports, and other related system design information. Eranch develops the bases for Regulatory acceptance criterk for instrumentatio l }Q .I @ and control systems designs; evaluates experience obtained during the cons V
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- m[A ' and operation of nuclear power ' plants and relates this infomation to futur'e '
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evaluations and acceptance criteria; and participates in the development of M -
- j Regulatory Guides and regulations pertaining to instrumentation and con 6ol
- l .
,] . systems important to safety.
2 .. l The Power Systems Branch performs comparable reviews, evaluations and criteria
, development functions primarily in the area of electric power systems i important to saf'ety.
g b qq I hold a Bachelor of Electrical Engineering degree ~from the University of
- f Ap Pittsburgh, Pittsburgh, Pennsylvania. In addition, I have taken courses i
fd in Mathematics, Theoretical Physics, Nuclear Physics and Engineering, and @$ P.adiation Shielding at the University of Pittsburgh and at the Reactor M! . C{;.! ' School of the Bettis Atomic Power Laboratory, Westinghouse Electric Corporation. - ,
- L l
7 fiy nuclear engineering experience background derives from my employment at il bj the Bettis Atomic Power Laboratory of Westinghouse Electric Corporation, 4 West Mifflin, Pennsylvania, from May 1955 to September 1962; and from my $N h m employment at the Bechte1' Corporation, Vernon, California, from September 1969 I I f' to January 1971. At Bettis Laboratory I was a lead enginaer in the nuclear I submarine power plant group with technical responsibility for nuclear instru-gij J mentation, rod control, and reactor protection systems. k*ork involved com-
*d 3 penent and system design, installation, testing, modification and documenta-
- j.[ tion. I also 'serYed as Bettis representatiYe during full-Scale tests con- [
r M m ducted by the Navy. At Bechtel I conducted engineering studies and prepared
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w,. . l Ml . reports and specifications relating to the design'and constrw. tion of the Rancho Seco Nuclear Power Station. This work was primarily.in the; areas o.f , ., ,. . _ safety-related electrical power, instrumentation and control systems. f 1 - 90 b '
-Mi n:r.-nui'!aar engineerin5 b6ckground derives primarily from my ecpioyment in
{pg the Construction Engineering Department of the National Tube Company, United States Stee1 ' Corporation, Lorain, Ohio, from June 1947 to April 1955; and from 3 ray employment at the Rocketdyne Division,of North American Rockwell Corpor-I ation, Canog'a Park, California, from October 1962 to March 1968. At National g$yj Tube I served 4s a Senior Engineer engaged in de' sign and development of pQ(f electrical power and control systems for new pipe mills from conceptual design r fh through detail design, procurement, installation, and initial operation. i hy j This work extended through completion of two major pipe mill construction b projects. At Rocketdyne I was a Research Specialist engaged in design and development of controls and instrumentation for a dual turbo-pump liquid , 4 l l I hydrogen feed system for a nuclear rocket engine. My primary responsibility was for control system integration extending from conceptual design through t procurement, installation, and completion of the test program. 4 Md I am a member of the Institute of Electrical and Electronic Engineers and have hl served on its Standards Board. I have participated in the nuclear standards m development work of this organization since 1972, te hv I am a registered Electrical Engineer in the State of Ohio, Registration lN M. E-020165. f ~~ 4 N m m d.- . . . _ .. . ____ _
IPy E $. .a K 7,pa.n!. f+ F.~.BPN 2mawua : : .. W_ MM M W7W W$.Rmi.WP~@.n.,:RS.an:.ME.wpi;M.f4..u.5W.w&.u.,d_by'h. . = 4W u- ... M ywyWW#MW. n i R::*A ri$ . W '
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h D Cherecterization M f% ggj .. - The allegation relates to the RHR pump common suction line valve control and e p#f % potential damage to RHR pumps due to loss of suction as a result of a single f failure. , Bjjj' . @:e f% Related A11ecations: 37, 39, 40, 45 (previously discussed in SSER 21) f% ,:n '.y di Implied Sionificance to. Plant Desien, Construction or Operation i {
- id. ,,', l g The RHR suction line from the RCS hot leg in the Diablo Canyon design contains g
7, Q w two isolation valves (8701 and 8702) in series that are normally closed during M. power operation and hot standby condition (Modes 1, 2 and 3) W The RHR suction line from the RCS hot leg.is only used during Mode 4 (hot shut-down with RCS . N) cold leg temperature less than 323 *F), Mode 5(cold shutdown) and Mode 6 44 l i .,. 7 (refueling). A postulated inadvertent closure of either isolation valve (8701 l {s or 8702) in the RHR suction line during plant shutdown could c;use potential WQ 7g damage to both RHR pumps. w, a Yt 2
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M ~~ -6 (p; C >.'. : a y a N A.ble Canyon SSER 22 A.4-177.1 J. t sh , - - - m,--- strnremer m_w --_- --m-m--s-z-m,-~r------ 2" e % ~e = =- - = * ' > = - - + " *""#"" " '"-'- J
p.4.,94d @ '[j %.x > MQ. _4 ..'l'@a.%i#e, g:x M ?.eM2MgP@i-QAMp. :#:lW# 79_,.#_inc g.,.Wf;!pM. b4WM(, u c _ w . . . ld e 61 . g < . 1 g . b ..J y M " M NssessmentofSafetySionificance g w - Udj n d A This allegation overlaps concerns previously expressed in Alleaations 40 and dq, if v' i:5 '.g . t t s er a id E s i s i r' y 2.s s ts.f ~ i r. Dit t *. : C r.r.y:t !! E r !!:. . I *. . re h This concern also has been discussed by the staff at an ACRS meetinij qn
%v.! !
df: February 10, 1984 Wl. x h *
'52 The potential, damage of both RHR pumps due to loss of suction as a result of a Ah "i single failure is prevented by the following provisions: >f 4
y$pg 1. In response to the staff reovirement in SSER 21 regarding Allegation 45,
&yh PG&E hes committed, in a letter dated February 15, 1984, to install the f.4 d.f .w RHR low flow alarm prior to entry into power operation (i.e. Mode 1 with % 7:.
associated decay heat generation). The low finw alarm will be set so that
.Rf
,Qg.- : sufficient time would be available to alert the operators to trip the RHR
@;#f pumps before pump damage occurs.
W w; N M 2. The current Technical Specifications and operating procedures for Diablo p$
$ Canyon Unit 1 preclude the inadvertent closure of either of the two RHR 'd j
'!M pump suction line isolation valves (8701 and 8702) by maintaining the valves hl s;.j *
- in an open position with power removed for the valve operators during nn O Modes a, 5 and 6.
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The apolicant stated at the ACRS meetinc on February 10, 1984 that RHR pump D . g -damage could occur in 10 to 15 minutes following loss of suction flow. ~ -i I,h Operating experience from the Calvert Cliffs Nuclear Power Plant showed that bm 9 t' . e ?.' * :'.: .: set 1r ve e f t .! pet a:: :x' .ite'y M -ir.utes af te- 1.!! c' U::t':r y;{. gi fl ow. The failure of both RHR pumps is an event beyond the design basis and h,:9 its occurence is highly unlikely based on the plant specific design and O administrative controls discussed above. Howevsr if failure of both RHR bK( a g;j pumps shoulil< occur.during plant shutdown, the following steps could be taken to maintain ,a, safe shutdown condition: c
.s
{X%y &.u),f MC4 Agh
- 1. If both RHR pumps failed during the period when the decay heat level is hEk] still relatively high, then the plant' conditions would permit decay heat dMS Jn to be removed by the steam generator (s). Condensate supplied from the Qab Q(aE,- condensate storage tank, raw water reservior, and the auxiliary salt water g system (unlimited supply) via temporary connections could provide a lon'p term source of auxiliary feedwater for decay heat removal. j kk33
- 2. If the steam generator (s) were not available, and the decay heat is
%,p.gj p @j relatively low, one RHR pump is generally used to remove decay heat ..; g hi 7tQ:*, , with one pump in standby, in accordance with the requirements o' Technical .. ,,o . d Specifications 3.9.8.2. In case the operating RHR cump is damaged due to
- 3. ; .
j closure of a suction valve, the standby RHR pump could be used to continue l j
$ the decay heat rerioval function af ter the clnsed sucticn isniatic.n valve (s)
K]t'4 J., 4 is r.anually oper.ed by an o:erator. 1.nalyses indicate that i# asi decay Q *i 1 k!{ . s o., A.4-177.3 Q.'9 - c .; .
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$j.f ;. heat removal capability were lost at the time of reactor trip, at least Qj , ;g .2 hqurs would be available for the o'perators to restore decay heat removal ;v g,y capability before core uncovery. I' decay heat removal capability were
- st vie or D/'? h:9;, ::. tide- 81.s r e t he t'!- 2 h;u
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! I . 1 v.uld be a Ph tk availab'le for operator action to correct the situation. Mi w Ms %p;
- 3. If both RHR pumps were Samaged l while the steam generators were.open for w
maintenance (or during any other period in which all steam genrators were x- unavailable), the charging pumps or safety injection pumps could be used h!.f to injsct water into the RCS for core cooling. If the manways on the +;f j[1. stean generator primary side were open for maintenance, water would flow %J lM out the manways and onto the floor of the containment. The containment h si? spray system and the fan coolers, which are independent from the RHR A system, could be used to remove decay heat inside containment to the $ ultimate heat sink via the component cooling water or the essential service )N) 'W a water system. ,sj h . ii 4 Diablo Canyon Operating Procedure flo. E0P-17 addresses the emergency 24I procedure under the condition that both RHR pumps are damaged during W l plant shutdown. W
- .0 in summary, the staff recocnizes that closure of either of the two isoir. tion
)d valves in series in the RHR hot leo suction line would prevent the RMR system 4.] h...; #ror perforninc its decay heat removal furict'.cn and could result in damace to h the EHR purps if.not corrected. Our evaluatien has concluded that: $y?. T A.4-177 A n'i.g f\ -
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PACIFIC GAS AND E LE C T RIC C O M PANY n statt stacct, saw rnancisco, c4uronnu mos M;.] gg , Tetcewowc om 7ei mi -
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mj . a-Q , 't February 29, 1984 d gfj PN PGandE Letter No: DCL-84-085 L. . q O h.] Mr. John B. Martin, Rpgfonal Administrator % U. S. Nuclear Regulatory Commission, Region V E4 1450 Maria Lane, Suite 210 !.h Walnut Creek, CA 94596-5368 ou d Re: Docket No. 50-275, OL-DPR-76 2.j Diablo Canyon Unit 1
'h Board Notification No. 84-014 Minimum Wall Thickness Measurements, Hudson Allegations i]1 ,
j
Dear Mr. Martin:
t
'] Enclosed is PGandE's response to the issues identified in Board Notification No. 84-014, dated January 24, 1984, concerning valve minimum wall thickness 't measurements originally performed in 1973 and 1974. From a review of this g* allegation, PGandE has identified twenty-two issues or concerns which have ;
been grouped into four categories. Each of the twenty-two issues is addressed I
'j ,
in the enclosure.
%, Additionally, in the introduction of the enclosure, PGandE addresses an issue
, raised by H. O. Hudson in a letter to V. Gilinsky dated January 12, 1984 -
,j. ;
Subject:
Report #3. a:
%,' Kindly acknowledge receipt of this material on the enclosed copy of this i letter and return it in the enclosed addressed envelope.
ml Sincerely, l
.9 J. B. Hoch d, for J. O. Schuyler '4 ,
M Enclosure 1 O h cc: T.W.Bisho)l
,;(? D. G. Eisen1ut c;j H. E. Schier11ng
;u Service List
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ll:l PGandE Letter No. DCL-84-085 h:p V y ' M ENCLOSURE !ll + - . <j 1 MINIMUM WALL THICKNESS E ASUREMENTS . m. $y$ % 1. INTRODUCTION 27 k 2. INADEQUATE QUALIFICATION AND VERIFICATION OF PULLMAN UT PROCEDURES q? - A. Issue 1. Qualification of Ultrasonic Measurement Procedure !.h.: B. Issue 2. Qualification of Ultrasonic Measurement Technique
% C. Issue 3. Procedure Verification Tests g D. Issue 4. PQRs for ESD 236 and ESD 244 mf 3.
[: DATA REPORT OMMISSIONS OR INACCURACIES
- h. A. Issue 5. Transducer Identification
[;,' B. Issue 6. Transducer Serial Number n C. Issue 7. UT Thickness Tester Serial Number
;{ D. Issue 8. Transducer Frequency ,
Calibration Information N E. Issue 9. F. Issue 10. Certification of Equi > ment Calibration r@) , G.' Issue 11. Micrometer Serial Num)ers <1 H. Issue 12. Micrometer Serial Numbers
? I. Issue 13.
J. Issue 14. Step-Wedge Blocks Pre- and Post-Operational Calibration
,; K. Issue 15. Minimum Valves Marked Acceptable
'y' L. Issue 18. UT Examinations on 14 Valves not Documented h M. Issue 19. Serial Numbers for 20 Valves Y d 4. INADEQUATE DOCUMENTATION AND DISPOSITION OF VALVES UNDER MINIMUM WALL ,;j THICKNESS ,3 M A. Issue 20. Issue 21. Follow-up Documentation on Yalves Below tiinimum Wall Documentation of Weld Repair b! B. g v, . A 5. MISCELLANE0US ITEMS RELATING TO WALL THICKNESS PROGRAtt A 6 gj A. B. Issue 16. Issue 17. Clerical Issues Measurement Inspection 7, J - C. Issue 22. Reactor Coolant Pressure Boundary Valves 75) iQ $g . NM -
\ ? sa ;.d g]-
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ymwwga ggggggggggggpMW#58NgngM@gggggW w - E g y W . VALVE MINIMUM WALL THICKNESS EASUREMENTS f,, l 41 w INTRODUCTION J
~1 E.[ On January 9, 1984. H. 0. Hudson identified to the NRC a number of issues b regarding minimum wall thickness requirements. The basis for these issues g% stem from two Pullman audits Nos. 34 and 101. These audits 7eport on i W
4 possible problems with quality assurance records for ultrasonic thickness measurements for Unit 1 and 2 reactor coolant pressure boundary valves. /,; PGandE has thoroughly examined the issue of valve minimum wall measurements A. and can find no instance where valve measurement integrity is suspect, d 7,j PGandE has reviewed the second Hudson allegation (Letter from H. Hudson to F V. Gilinsky dated January 2, 1984 -
Subject:
Report #2) concerning quality WP assurance records and has grouped responses to the issues into four areas: M (1) Inadequate Qualification and Verification of Pullman UT Procedures, d hq (2) Data Report 0 missions or Inaccuracies, (3) Inadequate Documentation and-Disposition of Valves Under Minimum Walt ; Thickness, and '! 6. (4) Miscellaneous Items Relating to Wall Thickness Program. W.) $ Also taken into consideration were the latest H. O. Hudson allegations, f. X contained in " Report #3 - Quality Assurance'Discrepaucies Associated With Pullman Power Products Internal Audit #101 At The Diablo Canyon Nuclear N- Plant," dated January 12, 1984. A review of these al!agations revealed that, : L although the specific instances or components were different, the basic issues remained the same as those presented in Hudson's January 9, 1984, letter to 4 .;s the NRC. The primary issue of Procedure Qualification Records has been V p d adequately responded to in Section 1 of this enclosure. The specific allegation that procedure ESD-241 was used to UT valve yokes prior to the W procedure's issuance is correct. From a review of the test documentation and E< confirmation from the Level III technician who performed the test and authored 1 ESD-241, it is apparent that the essentials of ESD-241 were addressed during %] the yoke testing with the aid of a copy of Dresser's UT procedure. The Data d Reports show that the parameters of ESD-241 were satisfied, including ?:] transducer frequency, shear wave angle, couslant and calibration b requirements. The only data missing were tie dimensions of the calibration
- ,; standard; however, the calibration block from 1973 was located and it was confirmed that the block's dimensions were in compliance with ESD-241 G.)! requirements. In support of this last item, a comparative test of a typical j valve yoke in Unit #2 verified acoustic similarity between the calibration h blocks and yokes. This dimension and acoustic data will be added to the Data
@} Report packages. r.M Q1 (lj 1. INADEQUATE QUALIFICATION AND VERIFICATION OF PULLMAN UT PROCEDURES @w A. ISSUE 1: "There is no evidence of a Procedure Qualification Record K (PQR) documenting that the ultrasonic measurement procedure (ESD 236) G is qualified by a proven demonstration (procedure qualification test) f3 of valve wall thiccness measurement. This is a nonconformance to y% 1 10CFR50 Appendix B IX and XVII and PGandE Contract Specification .!
#8711.4.3.23 and 4.3.29 (referenced in ESD 236)."
r" { 1 it ) m p; _ 0295d/0005K ,,< . _ _ , . . . . _ , _ ~ . , - .,_ _.. _ _ ,. ~ , _ _ _ .
gny%.pwg,g g g g gg g.m g g g gg g g g g g g g gggq W Q ;, . Q e p .. M ~ RESPONSE: Criteria IX and XVII of Appendix 8 require qualified s, f. procedures for special rocesses and the collection and maintenance of '9 quality records. The a>ove-referenced sections of PGandE Contract , a Specification 8711 also requires special processes to be performed using qualified procedure and requires calibration of measurement ~ ..._., J.: 1 %i equipment. . e The UT thickness procedure ESD-235 was qualified and cattbrated with d each application using step-block reference standards. ESD-236 h.] requires this pre- and post-operational calibration for each valve-
.s
% tested and the results are recorded on the Data Reports which are % maintained at the site. This calibration verifies the accuracy of the procedure when applied to a specific material, in addition to P operational adequacy of the ultrasonic test unit and transducer. ' Separate Procedure Qualification Records (PQRs) are not required or g applicable to thjs process (see response to Issue 2). W: Pullman Procedure ESD-236 for the valve wall measurement program used h;f.7 a thickness measuring procedure which provided the following:
- m. .
9 (1) Calibration blocks of material acoustica11y similar to the ! material of the valve being measured. Four blocks were used, $ types 304 and 316 of both cast and forged, matched to the valve j z#g 1 material. I t,s , Oj (2) Transducers and ultrasonic thickness testers that formed a system l r.j capable of measuring the specific material and thickness of the ui valves in the program scope and were calibrated and qualified as a gg system with each application. - 4 J d,s (3) Preoperational, post-operational, and in-process calibration of
? the system to ensure correct measurements with mechanical J reference checks to micrometer readings of action valve sections, I
d when practical. j d (4) Recording of the data from items (1), (2), and (3) above on each l d j measurement Data Report which is maintained as a Quality Assurance i d record, c( O Prior to measuring a valve, the ultrasonic thickness tester was 3 calibrated on a calibration block of known material and thickness. j,1 When possible, the calibration was repeated on the valve itself. This is a verification of the procedure, repeated for each valve tested and M documented on the Data Report. m] 8. ISSUE 2: 'There is no evidence of a Procedure Qualification Record A documenting that the ultrasonic measurement technique (ESD 236 M Procedure) is demonstrated to have a maximum error in repeatability 0i ' of the wall thickness as required by It is alleged of andaccurac5etterofnot more than6-20-72 (25page 2,~ paragraph 2). h the A.E.C. that PG&E/M.W. Kellogg (Pullman Power Products) did not perform a
- Procedure Qualification Test that demonstrated that ESD 236 had a 3 - maximum error in repeatability and accuracy of not more than 2% of the ,
wall thickness. There is no documentation (PQR) in the ESD 236 9 Documentation Packages that records the Procedure Qualification Test. y, l M' 0295d/0005K W l
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g %./tg h glyf;g g g.p gyNer g pM n W g igpfj)3g g iygjgM EfM F1 h H .* &* Without the PQR, PG&E/M.W. Kellogg (Pullman Power Products) does not i - W . comply with the A.E.C. requirement for documentation."
~
RESPONSE: For ultrasonic thickness measurement, the procedure is Eq verified each time the preoperational calibration is made. Good. ultrasonic technique requires calibration / verification of the~ ^- .E % procedure each time it is used, periodically during the performance of
- 3 the test and after completion of the tests. This assurjes accuracy, !
within two percent, under testing conditions and allows for % adjustments to be made for temperature, battery conditions, individual $(q -equipment components, etc. Review of the actual data reports shows UT measurements to be well within the +25 to10rance when compared to the respective calibration block in the range applied, ]1 w s] C. ISSUE 3: "There is no documented evidence of.' procedure verification tests' as required by ESD 236.6.7 (see attached U.I.A. f34, Quality Audit Checklists page 15) to determine that transducers will be of %]k
?
suitable frequency, size, and adapted with shoes, wedges, or saddles as each valve measurement requires.. It is alleged that PG&E/M.W. M d Kellogg (Pullman Power Products) did not perform ' procedure 7 verification tests' to determine transducer requirements. Without
' procedure verification test' documentation there is no assurance that e the transducers used were of a suitable frequency, size, and adapted ;
Q
% with shoes, wedges, or saddles as each valve measurement requires. !
O See Audit Action Request (A.A.R.) #1, UIAf34. A " Complicating the issue is an M.W. Kellogg Interoffice Correspondence,
'M M
dated 4-17-73 (see attached U.I.A. f34, attachment #6A), that states:
. '3. The transducers available are adequate for flat smooth surfaces.
@1 There are no adaptors, shoes or wedges available should they be i necessary'; '4. At this time, it appears the transducers supplied may i not be the correct type for thickness readings. If this is true, we ,a will have to order new. transducers'; '5. The effect of surface '} contour and roughness must be tested prior to making any reportable q .v results.' s4 "The absence of documented ' procedure verification test' to determine j L the proper transducers to be used, and the IOC report of the absence of adaptors, shoes or wedges, and that transducers supplied may not be h the correct type, raise serious QA questions about the transducers f.; . used to perform the UT measurement. m N 'There is no documented evidence of the testing of surface contour and R roughness for effect as referred to in the 100. This testing should M have been a part of the ' procedure verification tests' for which there l;;} are no records. V "The 100 concludes 'It is doubtful that any meaningful results can be sf obtained at this time and it is definite that none can be reported a.i T until the above mentioned problems are solved'. Without documented records of ' Procedure Qualification Tests' and/or ' procedure Q verification tests', there is no assurance that these problems weie W resolved."- - Jy - s d M-0295d/0005K Nt m ..,4
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, frequency, size, and adapted with shoes, wedges, or saddles as each m'> valve measurement requires, as determined from procedure verification
>d tests." q: ~. 9 The accuracy and suitability of the procedure and transducer are verified against a known standard of like material with each [.);[ application (see response to issue 1). This verification is
- a. documented on each Data Report. Thickness teeasurements were conducted Pid at a frequency capable of resolving the thickness range to be y: measured, as evidenced on all data report. (Re: current ASME Section V, Article 5, T-561) In addition, because of the transducer
%.y sizes utilized, valve contact was adequate and shoes, wedges, or N 1.c; saddles were not required and, therefore, were not used or verified. h Surface roughness of manufactured valves would have no affect on a i U thickness test since the back reflection from the inside wall of the valve is quite strong and would override any surface-induced static on l l d a CRT screen. Additionally, the use of couplants compensates for .) 40
?
variation in surface roughness. The referenced IOC of April 17, 1973, was an internal Pullman memorandum written prior to full receipt of UT I Ji equipment.and procedure development. M D. ISSUE 4: There were no formal PQR's for Procedures ESD-236 and i Y+Q ESD-244. RESPONSE: These procedures control UT measurement technique and, as j "y 2 discussed in the second issue above, do not require formal Procedure ! Qualification Records.
' .1 j !
.i As an additional item, Mr. Hudson's allegations of January 12, 1984,
.y discuss the lack of Procedure Qualification Records (PQRs) for q various Pullman ultrasonic examination procedures, including ESD 1 241-UT, Examination of Safety Yoke Rods on 3707 RAX 6-21 Safety ;
Q Yalves. Though special processes require qualified procedures, the a 1 qualification approach is provided by the governing standard. No i jd applicable standard has been found specifying written PQRs for j pl utrasonic testing. For example ASME Section III at this time only h;U requires PQRs for liquid penetrant and visual examination )rocedures. M UT procedures are qualified by equipment calibration and tie use of j u. ultrasonic reference standard blocks. .k 4
.Jq 2. DATA REPORT OMISSIONS OR INACCURACIES
.i
[,'d There were a number of concerns expressed in the allegation regarding ij omissions and errors in entries on the UT measurement data sheets. A review 9 of all the subject data sheets and documentation related to the audit ih h j."' referenced in the allegation (Unannounced Audit No. 34) shows that the omissions and errors had no impact on the credibility of the resulting measurements. hd., As discussed above, the accuracy of the UT machine is verified by use of - @N calibration blocks over the range of wall thicknesses before and after each j valve is measured. There were four calibration hiocks, one for each valve material (i.e., one each of Type 304 and 316 - cast and forged). As a
?j reference check, a micrometer was used for spot verification of step thickness X~~
0295d/0005K _ _- -._ _ _ - . _ _ _ _. _
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W + , l I D and actual UT measurements on valve bodies. Because the UT machine and !
.7 .i transducer were essentially calibrated for each valve measured, the essential I kt data requirements are the wall thickness and calibration measurements and their traceability to the respective valve and calibration block. In all v k,7 cases the essential' data was available and the errors and omissions, though
~" -
% deviations from Procedure ESD-236, do not compromise the measurement integrity. .A 4 Although during this review calibration certifications were not found for the Iyj calibration blocks, PGandE has since confirmed calibration of the blocks to
- . traceable standards. '
M I Qp A. ISSUE 5: "None of the 254 Data Reports audited listed the size, I kj sha e, or type designation (manufacturer's designation or description)- l M- of he transducer used to rform the valve wal thickness measurement 1 q as required by ASTM E114-6 .6.1.2 (referenced in ESD 236.3.2 and 5.1). See A.A.R. #1." { j{: J M RESPONSE: The transducer size, shape or type is not'a required entry on the Data Report by procedure. ESD 236 referenced ASTM E114-63 Mi M because it was used in developing the procedure. ASTM E114 Test Data Record requirments are delineated for use when lon itudinal waves are M- utilized for locating discontinuities and not thic ness dimensions g (reference 1.1 in Scope of ASTM E114-63). I A B. ISSUE 6: "Most of the Data Reports do list a transducer serial d number but ESD 236 Documentation Packages do not provide any information or description for transducers by serial number." .. i RESPONSE: The only meaningful information for a single-element Vl transducer, of the type used, might be frequency. This information is N,j entered, with only 19 exceptions, on 259 Data Reports. In any case, . y the lack of specific information on an individual transducer has no y affect on the performance of the measurement since the accuracy of the transducer and test unit is verified as a system during pre- and ]a post-operational calibration and documented on each Data Report. M C. ISSUE 7: "Seven Data Reports do not list a serial number of the UT thickness tester used to make the measurements. See A.A.R. #2." ,i RESPONSE: The serial number should have been recorded on the data
- 3) report; however, since there were only three UT-units (Branson 1 Sonoray Models 301, 303B and Sonic Model Mark !) used for the testing '
e !- program, each of a different manufacturer or model, the model of the ah UT-units used can be determined from the Data Reports which provide $; adequate traceability. N Q D. ISSUE 8: " Nineteen Data Reports did not list the testing frequency or nominal frequency of the transducers used to make the inspection. q( See A.A.R. fl." N RESPONSE: The frequency of the transducer used should have been h.5 recorded. However, the appropriateness of the transducer frequency chosen for the thickness of valve wall to be measured is verified w during preo rational equipment calibration. As responded to in id Issue 3 th ekness measurements were conducted at a frequency capable 4 of resolving the thickness range to be measured as evidenced on all data reports. (Re: ASME Section V, Article 5. T-561) t.
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- 3..U . E. ISSUE 9: ' Nineteen Data Reports list both the Nortec NDT-120, SNf 12224, and the Branson Sonoray Model 303B, SN# 18060, as the UT thickness tester used to make the valve measurements. But there is e,. only one set of calibration information and valve body wall _'~
g measurement results for each Data Report. The actual UT equipment used to make the valve measurements cannot be de.terstined for the v.]3 e, . purpose of traceability. See A.A.R. #2." ,
~ .,, 4 ..
ra RESPONSE: The technique used during this time period was to either P line through the unit not used or circle the unit used. All of the units were capable of producing accurate measurements when calibrated f% v in accordance with the procedure. The Nortec NDT-120 was used during L ' procedure development, but was not used to collect actual wall
? thickness data; therefore, the Branson unit would apply to the cases 7q; identified in the audit report. , % F. ISSUE 10: "Two hundred and seven Data Reports referenced serial b numbers for UT thickness tester equipment that could not be traced 'p to documentation for certification of equipment calibration from either the manufacturer or any other calibration organization as QH required by ESD 236.4.2. See A.A.R. f2."
y i:] RESPONSE: Procedure ESD-236, paragraph 4.2, requiras that the d manufacturer's cortiffation of UT equipment calibration be maintained i on site. Though there is evidence of calibration certification being originally supplied, it has yet to be located in the files. The i ,.i search for the certification will continue. i a ; Ff It should be edded that certification of factory calibration is of 1 little practical use for units used for thickness measurement. ' Certification only means the machine operates and that the horizontal f]l "i and vertical gains are linear within factory tolerances. The q:' preoperational calibration is the controlling verification for
.i assuring accuracy for the measurements, y
d G. ISSUE 11: " Fourteen Data Reports do not list serial numbers for the micrometers used to check the calibration accuracy on the valves by a ' @y M mechanical measns (sic). The micrometers used to make the valve 1 M measurements cannot be determined for the purpose of traceability. ) N u See A.A.R. f3." N
~
8tESPONSE: The serial numbers for the micrometers should have been d entered on the Data Report form. In some cases the entry was marked a' as not available, indicating that the micrometer did not have a serial Ei number. A h.j Since crAibration blocks were fabricated of material identical to the "b, type icing measured, and each step of the block was marked with the thickner,t, of each step, the use of micrometers was a reference check. M The calibration blocks were uted as the controlling verification, as kj described by procedon and as the<n on Data Reports. A complete Ip review of the Data Reports showed no case uhere the UT equipment was - rea4 usted as a result of the nechanical check performed by the
@j micrometers.
0295d/030bK ) u.___._.-._ _. _ . _ _ _ _ _ _ _ _ _ J
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%. ISSUE 12: "Eighty four Data Reports referenced serial numbers for zicrometers that could not be traced to documentation'for s
[ ^ certificatfon of equipment calibration from either the manufacturer or any other calibration organization as required by ESD 236.4.2. See *~- - d dn'i y A.A.R.'f3."- , % ltESPONSE: See the response to the previous ites. ? p ' - M I., ISSUE 13: "Six Data Reports do not list any information concerning the step-wedge blocks used for calibrating the UT tester equipment. hi . See^A.A.R. f4." M 't RES?0NSE: Four specially fabricated calibration blocks.were supplied ?- to M.W. Kellog. These blocks were the only step-wedge / blocks available.for use in this time period. The record of preoperational d.$4 : d' e ' calibration'gives evidence that the thickness of the steps corresponds to the correct calibration block. The thickness of each (}[,jiE d ;! step of the blocks was marked on the block and each block has unique step dimensions. Therefore, by taking the step dimensions from the N v.a preoperational calibration, one knows exactly which block was used. Q J. ISSUE 14: " Eleven Data Reports do not list pre or post operation calibration information. See A.A.R. f4." %y, $9 RESPONSE: ?reoperational and post-operational calibrations were performed using the calibration blocks in all cases. In all cases, this calibration is documented on the data reports. Any preo>erational or post-operational mechanical. calibration performed .- % on tia cetual valve was supplemental to the block calibration and was
;; accomplished when contact could be established, which was a mejority .:,a"j of cases.
1-f)i K. ISSUE 15: " Forty two Data Reports indicated the valves as below the d minimum allowed wall thickness but the Data Report forms were signed d in the item #7 space that indicated the valves were physically marked 2 m cs acceptable. See A.A.R. f6." i;1 RESPONSE: All valves were tagged either " accept" or " reject," as wd appropriate, in addition to marking the accepted valves. Since there. Or was no space to indicate that rejected valves were tagged, the P,1 technician inadvertantly initialed the space for marking accepted 9 valves as a method to document that he in fact performed his I3 thickness determinations. PGandE has recently reviewed all data e sheets and confirmed that all valves identified as under minimum wall jg were replaced, repaired, or accepted through engineering evaluation. a i? L. ISSUE 18: " Fourteen valve locations, listed by Westinghouse Letter N #PG&E2080 to be measured, had no documented evidence (Data Reports) [/@;j of being UT examined. See A.A.R. f7." ms 9 RESPONSE: The fourteen valves identified were either excluded from
- d valve wall thickness as a result of their not being the pressure
7e containing item or were deleted at a later date by an aumendment to - M the original list by Westinghouse. These valves are: 0.I 1-8010 A, B. & C Valve wall measurement was not j 2-8010 A, B, & C required in accordance with the d exception described in Westinghouse N Letter PGandE 2080. N A_ _ _ _ 0295d/0005K fQ P Q & iM{ $ $,Q$k%?W W Q W 2l.jf[ W $MQ& @ Q$ & $ $$ h h5' W g * ' g : . 1-8368 A, B, C & D These were not primary pressure @ 2-8368 A, B C. & D boundary valves and were deleted a from measurement program by rtf Engineering per modification of the R . or'ginal list by Westinghouse. - M Valve identities were changed by an aumendment 1ssued March 14, 1973 Al n (Westinghouse Letter PGandE 2273). k.f M. ISSUE 19: "Two of the 20 valves physically checked had serial b numbers that did not match the Data Report serial numbers. See Y A.A.R. #8." pr h .
- d. RESPONSE: The two valves have been physically checked and the serial numbers do indeed match the Data Re> ort serial numbers. However, it Q
hj- should be noted that the serial num>ers were vibratooled on the cast gr gj surface and sighting may have been difficult. Nevertheless, adequate indication was recorded to provide traceability between the test g
- en reports and valves.
p- 3. INADECUATE DOCUMENTATION AND DISPOSITION 0F VALVES UNDER MINIMUM WALL 4 r.c THICKP E55 k A. ISSUE 20: "There are 47 Data Reports that indicate that valves were r; below minimum wall requirement. U.I.A. f34, A.A.R. #8 (see attached) identified two valves (Locations i 2-PCV-455A and f 2-PCV-4558). that ?d; were weld repaired to meet minimum wall requirements. But the M, . ESD 236 Documentation Packages do not specify which of the valves were weld repaired. It is an item of concern, that the NRC should [y fnvestigate, as to which valves were weld repaired or replaced." ..:;n j RESPONSE: A total of 33 valves for both uq1ts were found to be under 3 minimum wall. (Mr. Hudson's number of 47 data reports indicating rejection probably evolved from some val n: receiving multiple tests % and rejections.) It was not specified by ESD-t36 or intended that fa the data reports indicate the final disposition of rejected valves.
- .9 -
Ij The rejected valves were processed through the deviation reporting N system and the information is readily available. The final M P i disposition of the 33 valves is as follows: j J .1 -15 valves replaced with new valves ij -9 valves accepted "as-is" through Westinghouse calculations g -9 valves received weld build-up by vendors mi f B. ISSUE 21: "The ESD 236 Documentation Packages do not provide any information as to the weld M Locations #2-PCV-455A and # procedure used to weld repair valves in2 M have been weld repaired. There is no documentation available that assures that the A.E.C. reviewed and approved the weld procedures y g used or the description of techniques used to verify the q acceptability of the repaired valves. The NRC should determine what i g valves were weld repaired; if the weld procedure used was acceptable; and if the technique used to verify acceptability was adequate." ,d ., q RESPONSE: As discussed in the previous iter, it was not required b that the ESD-236 documentation packages, which are for thickness d I,Q 0295d/0005K j1
% g n i d d Q C;M M M M @ g i b M 4 Q Q M M M js[$ M i sjdd M M M j M if f (N - - 7 p p ' N ' measurements, would include documentation on valve repair. All
$ ", valves that were found to have unacceptable wall dimensions were M returned to the vendor for repair or replacement. These valves, when
% returned to the site, were rameasured and Data Reports for these M replacement. valves are included in the documentation packages for ~" - ' 9;n. ESD-236. _
* ~ ; Repair procedures used by the valve vendor were supp1ied to the M Commission, but this information was not given to the site contractors and was, therefore, not available for Mr. Hudson's Q
g review. The weld procedures were submitted to the AEC by Mr. Searls'- aj 1etter of July 23, 1974, to Mr. Engelken of Region V.and PGandE is W reviewing the files further to identify any additional correspondence to the AEC on repair procedures. %n 5 4. MISCELLANEDUS ITEttS RELATING TO WALL THICKNESS PROGR%M~~ i.4 3'- A. ISSUE 16: "Many Data Reports were found to have original information d$ whited out and new information inserted. There are no signatures of persons making the changes or explanctions for the changes. See 8 A.A.R. #6." tt y?1 RESF0NSE: During the time period of the testing of valve wall R thickness, the use of " white out" was the practice for making iy corrections. The use of " white out" on quality documents was subsequently prohibited. , W !
.l1 B. ISSUE 17: " Eleven Data Reports did not have a complete measurement oj* inspection of all areas of the valves as requiredl>y the procedure.
There is no documentation authorizing the incomplete measurements. Q See A.A.R. #7." l RESPONSE: The eleven valves in question were new valves that the M supplier (Westinghouse) had shipped to replace originals that had d been returned for unacceptable wall thickness. These new valves i Qd required performance of full ultrasonic thickness measurements prior .cd to shipment to the site. On receipt of these valves at the site, it was noted that certain areas were identified in the suppliers N,3 J documentation as having questionable wall thickness. PGandE elected b to perform additional thickness measurements on the valve areas in question. M.W. Kellogg Company (Pullman) performed these tests and tf",1 found the valve wall thickness to be acceptable. This testing was an y added precautionary measure to assure data validity. The actions p; taken above can be evidenced in a PGandE letter to file from H.E. f Petersen, " Wall Thickness of 10C482 Check Valves Documented in h Westinghouse letter PGandE 2479", dated December 5, 1973. h A C. ISSUE 22: "The Westinghouse supplied reactor coolant pressure boundary valves (pressurizer safety valves) were designed to meet the WC requirements of Article 9 of the ASME Boiler and Pressure Vessel W Code, Section III (1968 Edition). This code is not referenced in the D A.E.C. Letter which specifies relevant codes and standards. d Westinghouse states the valves were not designed to meet the minimum wall thickness requirements of ANSI B16.5 (USAS 816.5) which is referenced in the A.E.C. Letter as one of the relevant codes. This Q) raises the question of whether the Diablo Canyon reactor coolant a $ 1 0295d/0005K q
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pressure boundary valves (pressurizer safety valves) meet the A.E.C. (NRC) code requirements. The NRC should investigate this issue to assure that the Diablo Canyon reactor coolant pressure boundary 1 valves (pressurizer safety valves) comply with the relevant codes and standards as established by the A.E.C. (NRC) for valves within the __ aj reactor coolant pressure boundary, as defined in subsection 50.55 (a)
' ~
.e 4 (codes and standards) of 10CFR 50. 9 'The Westinghouse supplied ' pressurizer safety valves' (reactor a h; coolant pressure boundary valves) do not comply with PG&E C.S. 18711 V Section 2.2.1 Code requirement to be designed, manufactured, and d M., fabricated to ANS B31.1." RESPONSE: The statement that the Westinghouse-supplied reactor a coolant pressure boundary valves (pressurizer safety valves) were g designed to meet the requirements of Article 9 of the ASME Boiler and yq Pressure Yessel Code Section III (1968 edition) is not completely g accurate. The subject valves were actually designed and fabricated el to the requirements of USAS B16.5 using the stress criteria of the g!.g ASME BPVC Section III as the basis for establishing stress levels. ASME Section III, Article 9 was used only for operational design
- 'Rj requirements.
@g H The requirements defined in subsection 50.55(a) of 10 CFR 50 specifies the use of ASA B31.1 or USAS B31.1.0. USAS B31.1, Chapter (1j IV, " Dimensional Requirements," in paragraph 126 directs the use of USAS B16.5 for design and fabrication of valves and is the basis for its being used as the criteria for the pressurizer safety valves.
- q
-1 Therefore, the requirements of USAS B31.1 were met. As defined by Westinghouse in letter PGE-2080, the bodies of the valves in question j are not the pressure containing item and, therefore, were not
'?7 designed to meet the minimum wall thickness requirements of USAS B16.5. n W For clarification, it is noted that the reactor coolant pressure >fj boundary valves were not supplied in accordance with Specification O 8711. They were supplied by Westinghouse as part of the Nuclear Steam Supply System in accordance with Specification 8700. j Y, j G N Nl 9 m M - ,j ~ Wr
.} 0295d/0005K
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