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. dTJ G-s Inspection Report Dates Inspector Hours Area of Review 84-01 1/84-2/84 Nicholas 20 Preoperational test Young results 84-07 2/84-3/84 Conte 30 Verification of NUREG Young 1019 commitments
- 84-11 4/84-5/84 Conte 10 NUREG 1019 open item Young' closecut TN-lY :rli..-slefrv v.Mdli.s P9 73 q ACI l C4 9 P&
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V It is understood that GPU/ TMI #1 dumiruer desires concurrance from NRC to repair _the steam generators by expanding the tubes (which have EC indications above the midthickness point of the tubesheet) for 2 inches in the bottom half of the tubesheet. This is obviously a new way to fix a steam generator and raises a few questions: j 1 The change in attachment method from a nulti- i pass fillet weld with a secondary hard 1" long roll to a 2" ro p st be considered '
/
an engineering change in design. SCIII does not specifically require a tube weld, but industry standards do require a tube weld attachment. c 2. She tube weld joint selected by B & W appears to be one where the tube attachment acts as a " stay" supporting the tubesheet. The design assumptions associated with the tube attachment must be considered. 3 The B & W design utilizes a pre-stressed tube weld to account for differential expansion. i' Obviously, the " rolled joints" not only can-not be pre-stressed, but will represent tubes which must change the environment for the " pre-stressed tubes around the unstressed tube attachment. 4 The rolled joints made in oxidized holes must not only be lea'Et'ightV-but must also be able 1 l
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e ._ . . _ . _ _ . _ - . - _ _ . _ . . _ _ _ . _ l y - * , . , p ~~ -
-- . f_- to remain leak' tight when ( 'coef. o'f expansion). thermal stresses are applied in heating and cooling cycles.
Strain = = (7.1-6) x 10-61n/in/F x 14 00F
= 1440 in/in C
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.OTSG 110T TEST ]
DATA SilElir 110. 2
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. 1:'1T I_D 'ls THI UNIT 1 Effective Page 0 ' {O' ." y[C,.lkW * #
TP 600/2 Page 122
. Enclosure 8, g3E P/'UN
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- Initials
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co i Vacuus Pump : Flow Rate
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sy i' - [ P;ge 2 of 3- ,
-L ., Activity. Respon-_ Commitment _ Revised sibility Date Date
-11)- S.T.P. - Operation Procedure _ . Site Eng 1/17/83-
, a) Component & System
. . Operating Limits and.
Precautions TF 1/7/83 1/7/83 b) Starttp & Test
-Instrument Calibration, Hydro, Punp Check-out SU&T 1/16/83 C) Endmill Tubes to be Plugged
- 1) ' Tooling Operation Procedure B&W V5/83 Draft 1/7/83
'2) FCA, .
B&W V7/83 ~1/7/83
- 3) DF - P.R./P.O. to do B&W work TF V12/83 '
-Safety Evaluation .
Fire Hazard Analysis
- 4) Tooling B&W/M&C V15/83 1/20/83
- 5) Recommendation on Endmill Depth B&W 1/7/83 D) B&W Stabilizer Removal
- 1) Weld Joint Efficiency TF 12/23/82
- 2) Proposal B&W 12/20/82 Issued
- 3) Mini . Tech Spec TF 12/23/82 V7/83
. a) DRF Safety. Evaluation Fire Hazard Analysis
- 4) B&W Operation Procedure B&W V5/83 1/7/83
- 5) FCA ' B&W 1/7/83 ,
;6) Installation Procedure M&C V15/83 71 Job Order- M&C V W83
- 8) P.O./P.R. TF V3/83 V7/83
- 9) Tooling ,
B&W V17/83 , T E) Tube Plugging and Stabilization ! (Inclusive of Welded Caps and Ex
- 1) plosive Plugging Plugging) and Stabilization
- Spec TF 12/23/82 Issued
- 2) DRF TF V12/83 V12/83 a) Safety Evaluation b) Fire Hazard Analysis
- 3) B&W Operation Procedure B&W 1/5/83 Issued
- 4) FCA B&W V7/83 Issuad 1 5) Installation Procedure SiteEng/M&C V15/83
- 6) Job Order M&C VW83
! 7) Hardware - Materials B&W 1/15/83 a) Stabilizer Junp Packs
,and Crinpers V15/83 i 8 OTSG V7/83
. Remaining 1/15/83
.o .
Page 3 of 3
? . Activity Rnspon- Commitment Revised sibility Date Date b) ~ Explosive Plugs ( E-3) 1/15/83 c) ~ Practice. Segments 1/15/83 d) Explosive Plug Delivery 1/15/83~
c) Practice Weld Caps
. B OTSG 12/24/82 V7/83 A OTSG 1/7/83 1/7/83
- 8) DF - Materials - P.R. TF 1/12/83 9)' Blast Plan B&W/M&C V3/83 210) Welder QJalification M&C ' Issued Procedure a) Lab SLpport TO 1/7 to.1/25/83
- 3) Endmill Remaining Tubes A) -Revise Spec TS-120012 . TF 12/30/82 -
002 Rev. 3 Total Remote Endmilling Procedure to Reflect Semi-Automatic Process B) B&W Endmilling Operation Procedure B&W 1/13/83 C) FCA :B&W V13/83 D) B&W Proposal B&W 12/17/82 Issued
- 4) System Flush.
A) P . R./P .O. TF 12/30/82 B&W Proposal B&W 12/22/82 B) Acceptance Criteria-01em Analysis Cleanliness TF TED DRF ,
- 5) Westinghouse Plugging ,
A) Westinghouse Installation, FTF 12/3V82 Procedure B) Installation Spec. TF V7/83
- 1) DRF V7/83 i Safety Evaluation Fire Hazard Analysis
- 6) Pre-Service Testing A) Eddy Current Testing Nuclear Inspection Procedure Assurance TBD Scope and Objectives '
TF 1/7/83
- 7) OTSG Freepath A) B&W Proposal B&W 1/10/.83 B) .Possible Low Pressure Felt Plug Blowing -
J. P. Hawkins M&C Scheduling X4001 d
, s.k t y Toa y TMI-1 OTSG Tubes Critical Crack Sizes and Operating Leakrate TUBE LOCATION: CORE PERIPHERY Tube. Load @ 100% Power (Lbs.) ' 200 (Tension) 500 (Tension)
TRANSIENTS: ',
.1 - MSLB Transient Tube Load (Lbs.) 1408 3140 (Tension) (Tension)
Critical Crack Size (Inches) 1'.28 0.52 .,$ Leakrate (GPH) ad- @ 100% Power Operation 14 6 e 4
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m
. 4 .: ~
no. ac.t ' 'hF: , 100*F/Hr. Cecidown; 140*F A T Limit "7 w/280# Tube Pre-load 0 600 - - O
*/4 Press. & Temp. vs. Time [j s.u 500 - -
2500 , 1. u
'400 8 4 94 -
2000g -
. u. us N '
E a300 - - 1500 m ..
>Wy .
200 RkS Press. Limits 1000 E }g 13 100 - Op - 500 0 . . < . 0 1800 - Tube Loads vs. Time .- 1600 - Perip6eial Tube's #'I v Y
.o &
d 1400 -
/ r/1 o
O Av . Tubes J \ .
;;; _120.0 -
'n h 4 ,
.E1000 -
H::: .< Core Tub 800 -
-c e
600
.- a
;5 400 -
, , i i t 1 2 3 4 5 g
. Cooldown Time (Hrs.) -
cf@^
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n.. an.2 y g" CecidSwn; 70*F A T Limit - w/280# Tube Pre-load [ 20.
.l. .
600 - -- Press. & Temp. vs. Time Y 500 - 4,,,Aug, Pe - 2500 s 4 400 's s4,4p g - 2000 E' o s CD ca
.i'M o
us , L g300 og- 1500} j.
.f.
u 200 n m , 1000 2 n. sRCS Press. Limits 100 - '
. 500
, OTSG press.
0 m.c 0 -
* :2,
?:."
1200 - -?ti
?M m
Peripheral Tubesk- E
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Tube Loads vs Time 46 N m,,.1000 - . ' " ~ ~ ~~ , .
.Q h. .
a
- ::.. ;.:,,% >, T ^ N., ]e 800 -
- k. ' * /
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- .s. '
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.M!.Coie Tulies[
l -O 200 -
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b A 1 2 3 4 5
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Cooldown Time (Hrs.) WM
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TMI-1 OTSG- . .
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PEAK TUBE AXlAL LOADS DURING PLANT C00LDOWN ( TUBE LOAD (Les) CENTER PERIPHE8AL 0 C00LDOWN W/140 FAT 800 1965-
~
l (280 LBS PRE-LOAD) , 0 f C00LDOWN W/70 F AT 510 1200 (280 LBS PRE-LOAD) NOTE: PCAK LOADS ARE BASED ON ASSUMED MAXIMUM RCS PRESSURE OF 2250 PSIG DURING COOLDOWN. 4 6 4 Lt -
TABLE NO. 8-1 OTSG Tube Leakrates During Plant Cooldown Tube Location ' Core Periphery Crack Size Basis Generic TMI-1 Generic TMI-1 ~ MSLB Specific' MSLB Specific MSLB MSLB Crack S'ze (inches) 1.25 1.5 0.5 0.63 een en % \ stu Cooldown Tube Load A B A B A B A B Basis (seenotebelow) 0 Cool'down w/140 F AT Limit ' CooldownTubeLoad(Lbs) 600 800 600 800 1500 1750 1500 1750 Leakrate 9 cooldown (GPH) 73 110 96 144 34 42 74 92 0 Cooldown w/70 F AT Limit Cooldown Tube Load (Lbs) 150 400 150 400 800 1050 800 1050 Leakrate 9 Cooldown (GPH) 8 40 11 52 12 19 29 44 Note: A = Tubes which have jumped down in the upper tubesheet. B = Tubes with pre-load O E e 6
-w- - ---- , .,- r, -~ --n-.-.._._,.v. _ -
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,, fu.uc.s TMI-1 OTSG
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Tube Leakrate During Plant Cooldown From MSLB Pre-Critical Crack in One Tuba
' at Various Cooldown Conditions n
. 2 L_;.J AREA A TMI-1 SPECIFIC MSLB l l AREA B GENERIC MSLB ..
7 s
. CURVE
- .::.:.. .. Ng* BASIS 140 -
- A ::. .
~
. 1- NO TUBE PRE-LOAD:
130 - , C00LDOWN W/70*F
~
.. SHELLTO TUBE A T LIMIT.
1,2 0 - .
',$.4.A ,
. . . 2- 280# TUBE PRE-LOAD:
110 - .,..ll ., C00LDOWN W/70'F -
?. . .. . ?. '. .
SHELLTO TUBE A T
.i LlMIT.
100 - _.: :.a:: 'lii'si:*i! : 3- NO TUBE PRE-LOAD: ....;.. .. hf'o[gs
. g, _
90 - s E .s.
.S
.; C00LDOWN AT 100*F/HR il.~{'jp/i3'A
. Oc -
o l# .
-: 'c? -
.' W/140*F SHELL TO TUBE A T LIMIT
- itw a . 80 - .~
= 'D
\ 4- 280# TUBE PRE-LOAD:
>4 o 70 - .. :!. -
.. s .: g o -
- !! !- .: i:!- C00LDOWN AT 100 F/HR EE.U ~! ;!
; 2 A'-
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1 W/140*F SHELL TO TUBE (g cn ,e 60 -
,. : 'i. ,
g I g A T LIMIT ! Ju , s .y,," :- S-5- W 50 - !s \~ .4B 3&W GENERIC O$ l f. ; \ C00LDOWN TUBE LOAOS.
)
i 40 - !! O
.. .: .i 3B DOES NOT INCLUDE RCS &
- ! r .
I: . : 5B s / OTSG PRESSURE LOADS OR TuaE PRE-LOAD
;j;' ' g*:;, -
2B: - 20 - /j"1 B 1 y A V
/
! ' 10 - / 6 GPH l
. /
l - 0 i i. i ! O 500 1000 1500 2000 l TUBE AXIAL LOAD (LBS.) (During Cooldo'wn) l
,-v - - - - - - - , . , . - esvrwyv ----wv-ww------,.-+..--e-- *<------+---,-w _-- =- ---_
- m. _, _ _ ,
TMI-1 OTSG - RESIDUAL PRE-CRITICAL CRACK COD AFTER A C00LDOWN
' TUBE LOCATION PERIPHERY CORE CRACK ARC LENGTH (lNCHES) O.5 . .1. 25 ~ ~
.-m..,.
1 REQUIRED RESIDUAL CRACK OPENING'(MILS) ' O.45 0 . 2 . ... . J.
'C00LDOWN TUBE LOAD MAX / MIN (LBS) 1050 800 k0b kb ." -
C0D 0 C00LDOWN (MILS) - 1 13 0.9 1.0 0.32
; . . . . ~ .
- 1. %o RE0'D RESIDUAL 0PENING TO C0D 8 'W<$,$.7
. . w. .
C00LDOWN (%) 40 50 20 60; . NOTES:' 1 . TUBE LOADS ARE BASED'ON MINIMUM RCS PRESSURE DURING COOL-P DOWN. 2- MAXIMUM TUBE LOADS ASSUME 280 LBS TUBE PRE-LOAD. MINIMUM TUBE LOADS SSUME ZERO TUBE PRE-LOAD. e 9
x .
,,, , ice, .* ws-t. 6ews***5 OTSG Loading Cycle for Tube Mechanical Evaluation -
p _ _ _ _ _ _ - - - _ .- - - - - - - - , + ' FIV - 2.4 X 10s CYCLES /YR. e'
#% I
.HEATUP ' '
g et** N i f \
/ m C00LDOWN
_ / _f 100* TS LIMil g, NORMALOP;
,. -g g
/ \ \
/ \
\
\
g f g \
/ s \
N \ TIME -*- l *
- FIV ALTERNATING LOAO, REGION I ~ 554 P:li, .003" MAX. DISP.
e HEATUP/COOLDOWN, E CYCLES /YR. e 40 YEAR LIFE -
- MAXIMUM AMPLITUDES >
0 fic $
i l l l Figure 5.3-1. Calculated Tube Natural frequency ! Vs Axial Load ; 70 1 60 i E -
. 50 ',
U
=.
g - 45 i
. i E -
hpss4 ed
- 40
\ '"*8 Y
30 . COMPRESSION TENSION l f I I I I
-600 -400 -200 0 200 400 600 AXtAL LOAO, L8 i coese()mo apper sh ny a hNe b- ^i=C- -
O e
. 4
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~
TMI-2 FIV INSTRUMENTATION RES'ULTS - STEADY ' - STATE TANGENTIAL DISPLACEMENT l 1.1 1
- i. - -
E ~
- a. -
37% POWER l8,7 - l e ..
=
E 8 .5 - 75% POWER - gW 8.4 - 5 m s.3 - 8.2 -- s'. t. - 40% POWER 1 t ' ' OTse 1s 30 sg 3 a 3 g n 3 n 4s . 3 se 40 35' 33 23k119 4 1
,, LANE TUBE LOCATION,glNCHES (ARROWS SHOW ACTUAL TUBE LOCATIONS) o STEA0Y CfLS.
STATE DEFLECTION FOR FRACTURE MECHANICS ANALYSIS = 3 x MAN RMS VALUE = 3 o 0~E CAN SAY WITM A CONFIDENCE LEVEL OF 8t% THAT FOR A G~AUSSIAN DISTRIBUTION THE
,CAXIMUM AMPtlTUOE WILL NOT EXCEEG'THREE TIME 8 THE RMS.
res s ,
) 1
/
-.-,-v--,--y.- ~-
d da/dn vs Ak for INCO 600 - l 10-4 , A 75% JOURNAL OF ENGINEERING MATERIALS CURVE .
- O 600', JOURNAL OF ENGINEERING MATERIALS CURVE G 77' MIT CURVE ;
- A .
O 554* MIT CURVE ,
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l JN J 10-5 - E -
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A en09eS 508' P0lNT BORATED WATER g.. ; 10-7
' ' ' ' ' ' " l I ' ' ' 'l'I '
1
+- Ti lt 100 M ..
4
^
AK Kst JTi (5r4 mh.cc %) '
t - . TDR No. 388 Rev. 1 Page 10 of 25 i Table 1 .
- The stress intensity factor ratio K1/6min the tube containing an inner .
circumferential semi-elliptic crack, and subjected to a uniform axial > membrane stress 6 ,, OD = 0.625 in. h = 0.034 in. ; !
- n 7 !
Lo/h +
- f. ,
c/h 0.1 0.2 'O.3 0.4 0.5 0.6 0.7 0.8 0.9
)
i
- 1. 0 , .0.115 0.161 0.193 0.215 0.231 0.243 0.249 0.255 0.293 {
2.0 0.118 0.173 0.219 0.258 0.290 0.319 0.336 0.3 54 0.428 3.0 0.119 0.177 0.230 0.279 0.323 0.363 0.392 0.421 0.526 4.0 0.119 0.180 0.236 0.291 0.343 0.393 0.433 0.473 0.608 5.0 0.119 0.181 0.240 0.299 0.357 0.414 0.46'3 0.514 0.678 6.0 0.119 0.182 0.243 0.304 0.367 0.430 0.487 0.548 0.740 i 7.0 0.120 0.182 0.244 0.308 0.374 0.443 0.507 0.577 0.794 8.0 0.120 0.182 0.246 0.311 0.380 0.453 0.523 0.602 0.542 g g.g gy supf Te THlCK WM #888'#I"Y (M $6 :P eoMt gen #M 4e, y gg tgg T% 4
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MSLB +\ ' 5 ECT -+- (1408 lbs) \- ,
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MSLB - 1.50 - LINE M - '
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(3140 lbs) '\. . s. O \ j 1.25 - 'g E ECT + , E \* g 1.00 - m. D \. . I s. r - N m
's
[ .75 -
- ' ow N O w eb *# N.
< m # .* N s'{ l 20 -
DAX IFITIAL CRACK SIZE TG Ai. LOW (2 YRS. 0F STABLE f STABLE CRACK PROP 0GATION. wd\ AKg=4.0 % d CRACK GROW 1
.2G -
p
. Wp t i l 1 0 20 40 SO 80 100 DEFECT DEPTH IN % WALL THICKNESS v.c. x 3
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.OTSG Leak Rate as a Function of Crack Length & Tube Tensile Load 100 l
So,
- 1107 # tension peripheral tube load @ 100*F/Hr ,
se - Lcooldown _ l i 70 - l
\
2a. I S so - 500 # tension l
- w. = !
Q' . peripheral tube load
=
@ Full Power i 5
w 50 - i a w a .
- w 40 -
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- o .
30 -
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l l L 289 # tensies : l ! 29 .- - I l 3 core tube leed
@ Full Power- l i
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> i i i i l i 1 l 1.2 1.3 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 ,
l TUBE CRACK OD ARC' LENGTH (INCHES)
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THI-1 STEAM GENERATOR PROBLEM o November 21, 1981, discovered primary to secondary leak while RCS pressure was.45 psid. By nitrogen bubble test, found $150 tubes leaking. o Initially found 3-10,000 tubes with defects by ECT. All but %200
.were within upper tubesheet (UTS) . Defects are sulfur induced
~
stress assisted IGSCC. Later found many more tubes had defects in
-HAZ of upper seal weld.
o Sulfur was primarily thiosulfate from Reactor Building Spray system. Failure scenario is: Sodiun thiosulfate leaked by valves to get into spray pump suction line, spray pumps tested by taking a suction
. from and discharging to Borated Water Storage Tank (BWST). BWST water entered RCS during periodic injection system testing and cooldown from HFT in September 1981.
o IGSCC caused during or after cooldown from HFT when "right" conditions of high stress, susceptible material and aggressive environment existed.
'o RCS inspected in April 1982 by visual (remote camera ), UT, RT, PT and destructive examination to look for indications of damage - none found. Inspection conrantrated on susceptible materials such as
'IN x 750, IN 600, SS 304, but examined virtually all materials used in RCS.
Repair Program o' GPU is performing a 17" kinetic expansion on all tubes. A qualification program was conducted to verify that a 6" joint free of defects would handle load carrying requirements (3140 # for MSLB) and be leak limiting (design goal if/hr for all tubes) o Qualification program involved axial load testing & leak testing before and after thermal cycling; residual stress measurements; adjacent shot effects; induced strain effects etc. o Proof test conducted on >100 tubes ist Mount Vernon, IN, witnessed by NRC. o Independent third party review concluded that it ~was acceptable to proceed with repair. o Demonstration test conducted in TMI-l OTSGs. 9 new ECT indications -
. found out of 151 tubes in "B" OTSG using 8 x 1 prob.e. 3 new ECT indications out of 285 tubes in "A" OTSG.
o New indications appear to be defects below ECT threshold which
" fish mouthed" open during expansion. Defects did not propagate or. tear during expansion.
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o After completion of expansions, GPU will conduct a baseline ECT and follow normal ISI inspection program if additional defects are identified. c GPU must justify for defects in qualified zone that these defects do not affect joint integrity. ECT Program o Nearly all defects are 90-100% throughwall . There are %76 tubes with defects below the UTS which ECT says are <40% throughwall which GPU proposes to leave in service. o - About 1200 tubes total require plugging based on a 100% full length ECT program (about 250 previously plugged). Post Repair Testing and Operation o' Consists of. baseline ECT, cold leak testing including a N2 bubble test and hydrostatic test. o Hot testing will include cooldown transients to place tubes under test followed by return to cold shutdown. o GPU-proposed 90 days of full power operation prior to further ECT. o Corrosion test program underway to model plant operation, o Desulfurization will probably be conducted using H22 0* Schedule Estimate o Repairs and cold testing complete by March,1983. o HFT - March, ready for operation - April t
u -A a aa a m-_. k n-4m+ 4 .A
, 3: -
di THRFF MILE ISLAND UNIT 1 4
-e RECENT LEAK TESTING REVEALED 86 LEAKING TUBES IN S.G.
3- "A"; 38 LEAKING TUBES IN S.G. "B". e ECT EXPECTED TO BEGIN DECEMBER 6, 1981. ! e MODE OF FAILURE IS UNDER INVESTIGATION. LICENSEE PLANS + TO REMOVE SECTIONS OF FAILED TUBES FOR LABORATORY ANALYSIS. ,
- - e STEAM GENERATORS ARE BEING PLACED IN WET LAYUP CONDITION l IN ATTEMPT TO MINIMIZE FURTHER TUBE DEGRADATION.
e STEAM GENERATORS HAVE BEEN -IN A PARTIALLY DRAINED WITH N2 BLANKET CONDITION OR FULL WET. LAY-UP SINCE MARCH 1979. e IHE OCCURRENCE OF TUBE DEGRADATION DURING PERIODS OF WET OR
-PARTI ALLY WET LAYUP CONDITIONS IS NOT UNIQUE TO IMI-1, ALTHOUGH IMI-l IS THE MOST EXTREME EXAMPLE.
e STAFF WILL NOT PERMIT RESTART UNTIL SATISFIED ON INTEGRITY OF S.G. s 9 *, f 4 DEISENHUT
! X27672 e
-o ,
MCGUIRE NUCIFAR STATION - UNIT 1 INSPECTION RESULTS e 11/15/81 - SHUTDOWN FROM 50% POWERJ NO INDICATIONS FROM ECT OF "A" S.G.; EXTERNAL MONITORS PUT IN ALL S.G. e 11/24/81 - RESTART AND OPERATION TO 75% POWER; MONITORS INDICATING TUBE RESONANCE PEAKS AT 63 AND 68% POWER e '12/2/81 - SHUTDOWN FOR ECT OF "A" S.G.
- ' INDICATIONS > 20%, CONTINUE OPERATION
- INDICATIONS < 20%, ECT ALL S.G'.
INSPECTION AND OPERATING PLAN e MID-DEC - RESTARTJ POWER - ASCENSION -TO 90% (l itK) e LATE-DEC - EVALUATE 50-90% POWER DATA
- OPERATE AT 100 POWER {1DAh)
- CONTINUE OPERATION AT APPROPRI ATE LEVEL N
e DEC/JAN SHUTDOWN TO ECT ALL SGS
- INSTALL INTERNAL MONITORS (1 S.G.)
- EVALUATE DATA
- RESTART AND OPERATE AT APPROPRI ATE LEVEL
- PENDING PERMANENT FIX i
- DEISENHUT X27672
- . . =. .
[. ,. t OCONFF UNITS 1, 2. AND 3
- 1. OPERATING EXPERIENCE O CIRCUMFERENTI AL FATIGUE CRACKS AT OR NEAR OPEN INSPECTION LANE 01 EROSION / CORROSION AT SUPPORT PLATES IN REGIONS OF DEBRIS DEPOSITS -
MOST ADVANCED AT UNIT 1 UNIT 1 UNIT 2 UNIT 3 NO. (%) 0F TUBES PLUGGED 311(2'D 30(<1%) 101(<1%) NO. 0F LEAKERS 11 3 5 MOST RECENT LEAK 2/81 9/81 5/80
- 2. -CORRECTIVE ACTIONS 10 PREVIOUS CORRECTIVE ACTIONS TO ADDR5SS FATIGUE CRACKING HAVE NOT ELIMINATED THE PROBLEM. THESE INCLUDED:
ORIFICE PLATE ROTATION AUXILI ARY FEEDWATER N0ZZLE BLOCKAGE , l' REVISED TURBINE STOP VALVE TEST PROCEDURES t STRICTER ADHERENCE TO FEEDWATER CHEMISTRY SPECIFICATIONS j O . B&W IS DEVELOPING A BLOCKING DEVICE TO REDUCE THE FLOW PASSING
- l. UP .THE OPlti, INSPECTION LANE .
O CHEMICAL CLEANING IS UNDER EVALUATION BY B&W AND THE LICENSEE TO ELIMINATE DEBRIS DEPOSITS
- 3. STATUS THE STAFF WILL CONTINUE TO MONITOR THE CONDITION OF THESE UNITS.
DEISENHUT X27672
H.B. ROBINSON UNIT'2
- - ; -1. OPERATINa EXPERIENCE 0 ACTIVE PHOSPHATE-THINNING ON THE HOT AND COLD LEGS AND IN THE.
U-BENDS-i O IGA AND SCC IN TUBESHEET CREVICE REGION ON HOT LEG ' SIDE 0 - SCC IN SLUDGE PILE REGION ABOVE TUBESHEET 4 0 PLANT WAS SHUTDOWN IN 8/81 AND 11/81 AS A RESULT OF PRIMARY TO SECONDARY LEAKAGE O 1068 TUBES (11% OF TOTAL) HAVE BEEN: PLUGGED TO DATE I 2. CORRECTIVE ACTIONS O HOT AND COLD WATER FLUSHING. O REDUCED POWER OPERATION O SLEEVING BEING CONSIDERED FOR FUTURE USE IN LIEU OF PLUGGING
- 3. STAFF DISPOSITION THE STAFF. IMPOSED THE FOLLOWING LICENSE' CONDITIONS IN AUGUST 1981:
0 1900 PSID PRIMARY TO SECONDARY HhDROTESTS TO BE PERFORMED AT 24 > EFPD INTERVALS PRIOR TO NEXT REFUELING OUTAGE I t - O MORE RESQICTIVE LIMITS ON PRIMARY TO SECONDARY LEAKAGE i . i
- DEISENHUT X27672 l-r
~
e i 4 SAN ONOFRE UNIT 1 _ 1. OPERATING EXPERIENCE O INTERANULAR ATTACK (IGA) IDENTIFIED AT TOP OF TUBESHEET ELEVATION FOLLOWING PLANT SHUTDOWN IN APRIL 1980 . 2. CONNECTIVE ACTIONS BY LICENSEE O SLEEVE AND PLUGGING REPAIRS PERFORMED ON ALL. TUBES (ABOUT 7000) IN CENTRAL BUNDLE REGION WHERE ATTACK IS MORE ADVANCED 0 HOT AND COLD WATER FLUSH OF SLUDGE REGION O REDUCED , TEMPERATURE / POWER OPERATION j 0 MORE STRINGENT CONTROL OF SECONDARY WATER CHEMISTRY i
- 3. STAFF DISPOSITION l 0 SLEEVE ' REPAIRS APPROVED AS ACCEPTABLE ALTERNATIVE TO PLUGGING
~
O APPROVED PLANT RESTART IN JUNE 1981 SUBJECT TO THE FOLLOWINS CONDITIONS: l 1800 PSID PRIMARY TO SECONDARY AND 800 PSID SECONDARY TO PRIMARY HYDROTEST PRIOR TO RESTART MORE RESTRICTIVE LIMITS ON ALLOWABLE PRIMARY TO SF.CONDARY LEAKAGE < - PLANT WILL BE SHUTDOWN FOR ITS NEXT STEAM GENERATOR INSPECTION WITHIN SIX EFPM AFTER RESTART
+
PLANT RESTART FOLLOWING THE SIX MONTH INSPECTION OUTAGE WILL BE SUBdECT TO NRC APPROVAL
- 4. STATUS O PLANT.HAS OPERATED FOR ABOUT TWO EFPM SINCE RESTART WITH NO STEAM GENERATOR PROBLEMS REPORTED i
i DEISENHUT X27672 e
INDIAN POINT UNIT 2 1
- 1. OPERATING EXPERIENCE O . EXTENSIVE AND ACTIVE DENTING.
- 0 SMALL LEdKS IN APRIL AND AUGUST 1981 0 477 TUBES (3.'7% OF TOTAL) PLUGGED TO DkTE ,
- 2. STAFF DISPOSITION O INSPECTION PROGRAM MUST BE SUBMITTED FOR NRC REVIEW AND APPROVAL
-60 DAYS PRIOR TO EACH SCHEDULED INSPECTION OUTAGE 0 OPERATING INTERVALS IN EXCESS OF EIGHT MONTHS BETWEEN STEAM GENERATOR INSPECTIONS MUST BE APPROVED BY THE NRC l
O INSPECTION RESULTS MUST BE REPORTED FOR INFORMATION WITHIN 45 DAYS - RESULTS INDICATING AN INCREASE IN RATE OF DENTING MUST BE REPORTED IMMEDIATELY l ! DEISENHUT l X27672 t i i 9 9
- l I
l
, INDIAN POINT UNIT 3
- 1. OPERATING EXPERIENCE O EXTENSIVE AND ACTIVE DENTING, BUT NO DENT RELATED LEAKS SINCE 12/78 O EXTENSIVE PITTING ATTACK FIRST OBSERVED ON COLD LEG SIDE FOLLOWING LEAK OCCURRENCE IN AUGUST 1981 0 801 TUBES (6.1% OF TOTAL) HAVE BEEN PLUGGED TO DATE - THIS INCLUDES
' 329 TUBES PLUGGED AS RESULT bF PITTINd IN EXCESS OF 65% THRU WALL-0 762 TUBES CONTAIN PITTING BETWEEN 40 AND G5% THRU WALL i 2. CORRECTIVE ACTIONS O BORIC ACID TREATMENT,,TO RETARD DENTING HAS BEEN DISCONTINUED PENDING EVALUATION OF PITTING OCCURRENCE 0- SLUDGE LANCING, FLUSHING, AND STRICTER ADHERENCE TO SECONDARY WATER .
i CHEMISTRY HAS BEEN IMPLEMENTED TO RETARD FURTHER PITTING ATTACK 0 LICENSEE PLAN 51TO PROPOSE SLEEVING REPAIRS OF TUBES SUBJECT TO PITTING
- 3. STAFF DISPOSITION O APPROVED INCREASED PLUGGING LIMIT FROM 40% TO 65% THRU WALL FOR PITTED TUS,ES 0 STAFF CONSi~ D ERS THIS TO BE AN, INTERIM PLUGGING LIMIT PRIOR TO THE NEXT REFUELING OUTAGE IN MARCH 1982 0 RESTART FROM THE MARCH 1982 REFUELING AND INSPECTION OUTAGE IS
, SUBJECT TO NRC APPROVAL 4
DEISENHUT X27672 T
._ _ , _ . . _ . . . ~ . . - _ . _ . . . _ _ - . _ _ . _ . _ . . . _ . . - . . _ .
l POINT BEACH UNIT 1 1, OPERATING EXPERIENCE O HAS EXPERIENCED WIDESPREAD IGA AND SCC IN TUBESHEET CREVICES 0 627 TUBES (13% OF TOTALI HAVE BEEN PLUGGED TO DATE .
- 2. CORRECTIVE A'TIONS C
O HOT AND COLD WATES, FLUSHING AND . REDUCED TEMPERATURE OPERATION SINCE NOVEMBER .3/3 HAVE . APPARENTLY- BEEN. SUCCESSFUL IN REDUCING THE RATE OF FURTHER ATTACK IN THE CREVICE REGION O TWELVE SLEEVES WERE INSTALLED DURING THE OCTOBER 1981 OUTAGE AS A DEMONSTRATION PROGRAM 0 LARGE SCALE SLEEVING IS PLANNED - THIS WILL INCLUDE THE SLEEVING OF PREVIOUSLY PLUGGED TUBES AFTER THE PLUGS HAVE BEEN REMOVED
- 3. STAFF DISPOSITION O THE UNIT G@TTNy[S TO OPERATR UtlpgE PORT-ION. CR ORDERS DATED ,
NOVEMBER x, .9/9 AND APRIL 4, J3W INCLUDING: ' MORE RESTRICTIVE LIMITS ON PRIMARh TO SECbNDARY LEAKAGE 4 PRIMARY TO SECONDARY AND SECONDARY TO PRIMARY HYDROTESTS PRIOR TO RESTART FROM EACH INSPECTION OUTAGE
/
ADDITIONAL REPORTING REQUIREMENTS REGARDING STEAM GENER'ATOR
' INSPECTION RESULTS O THE INSTALLATION OF 12 SLEEVES DURING THE OCTdBER 1981 AS A DEMONSTRATION PROGRAM WAS REVIEWED AND APPROVED BY THE STAFF DEIS5NHUT X27672 4
I SAIFM UNIT 2 -
- 1. OPERATING EXPERIENCE O NO ADVERSE OPERATING. EXPERIENCE WITH THE SALEM UNIT 2 STEAM GENERATORS HAS BEEN REPORTED TO DATE ,
M 0 l e i . . . t l DEISENHUT X27672 l l l
3
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o ,1
- - 7-
- j. .
. i OTSG POST WELD HEAT TREATMENT
. HEATUP TIME ": TIME AT C00LDOWN TIME-OTSG- 200 F TO 1100 F. G.T. 1100 F/850 F . 1100 F TO 200 F !-
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i ~ e TUBE FAILURES ARE ASSOCIATED WITH SPECIFIC F LOCATIONS IN THE GENERATOR NOT HEAT RELATIONSHIPS. t e THE DEFECT PATTERNS IN THE TWO GENERATORS APPEAR TO BE DIFFERENT AND THIS WILL NEED TO BE EXPLAINED BY A PARAMETER OTHER THAN HEAT NUMBER.' l e HEATS OF MATERIAL EXIST WHICH HAVE HIGH DEFECT [ . FREQUENCIES IN BAD AREAS AND THE SAME HEATS WILL HAVE LOW DEFECT FREQUENCIES IN GOOD AREAS. [ - b I l - i b t 5 .
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- 1. ' /;
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/ .o. LOWER'END GENERATOR.ALWAYS SUBMERGED-(WETTED), UPPER END ALTERNATE -
);
. WET AND DRY WITil AIR (0XYGEN) INTERFACE i -
.. .o WATER LEVEL IN THE PRIMARY SI,DE.0F OTSG WAS IN UTS FOR BETWEEN.31' 1 AND 2f13 DAYS S ,t I'
o SOME DIFFERENCES IN AMOUNT OF FLOW SINCE FEB '79 TOTAL PUMP HOURS 0TSG - A = 681 HRS . TOTAL PUMP HOURS OTSG - B = 393 HRS . L - BACK FLOW IN OTSG - B FOR 10 HRS ;.. I
. DURING SEPTEMBER '81 C00LDOWN -
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f . ,. . i . OPERATING HISTORY OBSERVATIONSu 2 , /[r o POTENTIAL SULFUR SOURCES PRESENT SOME OIL INfRODUCED :INTO RCS IN MAR '79 . O ,; - SULFURIC ACID ADDED.T0 RCS IN OCT '79 , S0DIUM THIOSULFATE ADDED TO RCS AT VARIOUS TIMES OVER LIFE.0F PLANT i 5!. o S0DIUM. THIOSULFATE THOUGHT TO BE PRIMARY CONTRIBUTOR ACCUMULATED IN BUILDING SPRAY PIPING - 1979 AS A RESULT OF j
,j 4
VALVE LEAKAGE
- JUN, AUG, SEP '81 OPERATION OF SPRAY PUMPS ADDED SOLUTION TO BWST .[
INJECTION INTO RCS 0CCURRED DURING SEP '81 C00LDOWN il
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. KEY ELEMENTS IN EXP REVIEW -
+ . F e IGSCC 0F I-600 OBSERVED AT-5750F IN SULPHATE.CONTAINING WATER; UNLIKELY TO OCCUR I UNDER PWR PRIMARY SYSTEM REDUCING ENVIRONMENT $ - NOT ASSOCIATED WITH DEGREE OF SENSITIZATION n; ' ) e IGSCC 0F I-600 OBSERVED AT 75-2250F IN SULFUR OXYANION (E.G., THIOSULPHATESL CONTAINING L WATER; MORE LIKELY TO OCCUR IN PWR.PRIf1ARY SYSTEM
, - SU CE BIL Y PENDS ON SENSITIZATION', PH, TEMPERATURE, AND ELECTROCHEMICAL j POTENTIAL .i 1 !-
e PLANT AND MODEL B0ILER EXPERIENCE IS ENTIRELY RELATED TO SECONDARY SIDE PROBLEMS { -
;. e NONE OF PRIMARY SIDE INDUSTRY EXPERIENCE IGSCC 0F I-600 ATTRIBUTED TO-ATTACK BY !
SULFUR SPECIES b l ., l i i l y 1 J i ! 4 l i A
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_ AGGRESSIVE ENVIRONMENT r, . . SO4 -~ AND S23 0 -~ CONTAMINATION PROBABLY PRESENT ' i
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- HOT FUNCTIONAL -- DIFF.ICULT .
s TO PREDICT SPECIES PRESENT AFTERWARDS ,l
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O SUMARY OF FAILURE ANALYSIS
.o .ALL CRACKS ARE STRESS ASSISTED INTERGRANULAR COR -
lNITIATIO( ON THE ID SURFACE o EDDY CURRE!!T EXAMINATION HAS BEEN A RELIABLE INDIC CRACK LOCATION o . IllCIPIENT CRACKS HAVE NOT BEEN DETECTED Ili CLEAN SEC (H0 E.C. I!!DICAT10iiS) DF TUBIHG BY VISUAL A!!D DESTRUC
, EXAMIllATION o
CARBON' IN THE FORM OF A HYDROCARB0t! APPEARS AS THE C0ilTAMIllAliT Cil FRACTURE SURFACES. SULFUR Ai!D CHLORINE ARE PRESENT AS SECONDARY C0i!TAMIi!A!iTS o RESIDUAL STRESS MEASUREMEllTS' lii ROLL AND ROLL TRA . SHOW N0 STRESS PEAKS BUT RATHER A UNIFORM DISTRIBUTIO o CHROMIUM LEVELS IN THE GRAlli BOU!!DARIES VARY FROM 8 TO 20 WT. % -
,~ o THE INC0!1EL MICROSTRUCTURE APPEARS TYPICAL FOR ST TUBING WITH DISCRETE CHR0!ilUM CARBIDE PARTICLES IN BOUNDARIES ,
. o SMALL AREAS OF INTERGRAilULAR CORROSION SEVERAL GRAIN HAVE BEEll OBSERVED ON THE ID-AliD OD SURFACES AT RA
. o NO RELATIC:! SHIP HAS BEET! ESTABLISHED BETWEEN MATERIA .
1 AND DEFECTIVE TUBIliG o MECHANICAL TESTING OF Ui! CRACKED TUBES SHOW THAT TH EXCEEDS MINIMUM SPECIEICATI0i! REQUIREMEllTS
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g e . o PRELIMINARY CORROSION TEST RESULTS s CORROSION TESTS IN ACTUAL PRIMARY COOLANT INDICATE IT IS CURREllTLY INNOCU0US e REDUCED SULFUR SPECIES CAN REPRODUCE THE. TYPE OF CRACKING OBSERVED IN STEAM GENERATOR TUBES
- e THE DEGREE OF SENSITIZATION (I.E., PRIOR HEAT TREATMENT) IS A KEY PARAMETER IN DEFINING THE MATERIALS SUSCEPTIBILITY TO IGSCC e CRACK-INITIATION' APPEARS TO BE THE RATE CONTROLLING PALAMETER e CRACK GROWTH' RATE IS VERY RAPID ON THE ORDER OF IMM/ DAY e CRACKING APPEARS TO BE A LOW TEMPERATURE OCCURRENCE eCRACKINGTENDEllCYISREDUCEDBYRAISi""~4EPH ,
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SUMMARY
TUBE STRESSES - 1000F/HR C00LDOWN
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(TUBE: INNER SURFACEL S '
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TUBESHEET TUBE RESIDUAL FABRICATION STRESSES - PSI AXIAL. CIRCUMFERENTIAL
@ 22000 -
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-@ 26000 22000
@ 0 - 26000 0 - 22000 000LDOWN TRANSIENT STRESSES - PSI
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o 1004.7 Revision 7 IMPORTANT TO SAFETY p .4c l , 7 ' E' ' NON-ENVIRONMENTAL IMPACT RELATED THREE MILE ISLAND NUCLEAR STATION UNIT NO. 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE 1004.7 0FFSITE/0NSITE DOSE PROJECTIONS Table of Effective Pages Page Revision Page Revision Page Revision Page Revision 1.0 7 31.0 7 61.0 7 2.0 7 32.0 7 62.0 7 3.0 7 33.0 7 63.0 7 4.0 6 34.0 7 64.0 . . 7 5.0 6 35.0 7 65.0 7 6.0 6 36.0 7 66.0 7 7.0 6 37.0 7 67.0 7 8.0 6 38.0 7 68.0 7 9.0 6 39.0 7 69.0 7 10.0 6 40.0 7 11.0 6 41.0 7 12.0 7 42.0 7 13.0 7 43.0 7 14.0 7 44.0 7 15.0 7 45.0 7 16.0 7 46.0 7 17.0 7 47.0 7 18.0 7 48.0 7 19.0 7 49.0 7 20.0 7 50.0 7 21.0 7 51.0 7 22.0 7 52.0 7 23.0 7 53.0 7 24.0
- 7 54.0 7 25.0 7 55.0 7 26.0 7 56.0 7 27.0 7 57.0 7 28.0 7- 58.0 7 29.0 7 59.0 7 30.0 7 60.0 7 Signature Date Signature Date Document ID: 0047W l
DRAFT
/d
1004.7 DRAFT "- " ' " 7 THREE MILE ISLAND NUCLEAR STATION UNIT NO. 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE 1004.7 0FFSITE/0NSITE DOSE PROJECTIONS 1.0 PURPOSE The purpose of the procedure is to provide:
- a. Techniques and methods for calculating projected doses (whole body, and thyroid dose equivalent which might result from monitored releases of radioactive materials from TMI Unit 1.
- b. Techniques and methoos for predicting the downstream concentrations of radioactive liquids resulting from a major accidental release of radioactive liquids to the Susquehanna Valley.
- c. Contingency methods for estimating projected doses if monitors are out of service or off-scale high.
The Radiological Assessmeat Coordinator is responsible for imple-menting this procedure. 2.0 ATTACHMENTS 2.1 Attachment I Dose Assessment Sheet 2.2 Attachment II Meteorological Data 2.3 Attachment III Calculation of the Source Term and Onsite/Offsite Dose Projections 2.4 Attachment IV Contingency Calculations 2.5 Attachment V Liquid Release Calculation 2.6 Attachment VI Protective Action Guides 2.7 Attachment VII Field Monitoring Nomograph 2.8 Attachment VIII Computerized Dose Calculations 2.9 Attachment IX High Range RMS Dose Calculations 2.10 Attachment X Dose Conversion Factor Calculation 1.0 DRAFT
- v. +.
1004.7-Revision 7 2.11- Attacnment XI Hydrogen Purge Calculation 2.12 Attachment XII Thumbrules 3.0 EMERGENCY ACTION LEVELS 3.1 ksrequiredbyanEmergencyPlan-ImplementingProcedure. 3.2 As directed by the Emergency Director or his designee.
- 4.0 EMERGENCY ACTIONS INITIALS
___ --- __ = __-____--__----=-
- _=__= _-__--__________
- NOTE: The TRS-80 minicomputer may be used in lieu of :
- written hand calculations to determine dose projec- :
- .tlons. Utilize Attachment VIII " Computerized Dose :
- Calculations" to operate the minicomputer. -
h
- NOTE: Perform steos in order: :
- If the release is radioactive' materials to the :
- atmosphere, perform Steps 4.1 - 4.5. :
If release is of radioactive liquids to the :
- Susquehanna River perform Steps 4.6 - 4.8. :
- NOTE: Refer to EPIP 1004.6, Additional Assistance and :
Notification. Attachment III (pg. 10.0) for :
- back-up sources of meteorological information. :
4.1 Complete the Meteorological section of the Oose Assessment Sheet by completing Attachment II. 4.2 Complete the Release section, Scurce Term and Dose Projection section of the Dose Assessment Sheet by completing forms on Attachment III. If High Range RMS is to be utilized then refer to Attachment IX. Use Attachment X if a DCF is to be calculated. Use Attachment XI for a Hydrogen Purge calculation. Use Attachment XII for'a Thumbrule calculation. 2.0 t
4 1004.7 . Revision 7 ; e . l 4.3 Utilize' Attachments'VI and VII'to evaluate Field-Monitoring data and recommend Protect 1've' Action. g - 4.4 -Util'ize Attachment IV to project dose based upon contingency-calculations. v:- 4.5 AlwaysJreportfdose rate,-dose, time used,'and basis for the' time estimate to the Emergency Director, or his designee. .
^ ~
-~ 4. 6 Compile the. expected downstream concentrations by performing the !
~
^
steps and completing the forms 1n Attachment V.
~
4.7 Compile the time for the flume to: reach downstream users and a 24
' hour average concentration by completing the remaining steps-in Attachment V.
. . 4.8 Report"results to the Emergency Director or his designee.
=
i s i s
'E 3.0
y- 1004.7 Revision 7 ATTACHMENT.IV. CONTINGENCY SOURCE TERM CALCULATION
-------------------------- Instructions for'Using Attachment Four ------ --- =- ---
--l
- 1. Select a release pathway from the posted menu:
A. . Case'I Secondary Side Release Includes: OTSG-tube rupture. Loss of electric load Loss of power Direct steam release-
--- GO TO CASE I --
B. Case II Reactor Building Release Includes: Loss of coolant accident (LOCA) Maximum hypothetical accident (MHA) Rod ejection accident Spent fuel accident in the RB
-- GO TO CASE II --
C. Case III Auxiliary and Fuel Handling Building Release Includes: Spent fuel handling accident in the FHB Fuel cask drop during transfer Op Waste decay tank rupture
-- GO TO CASE III --
- 2. For the selected release pathway follow the logic diagram and calculate the noble gas and radiolodine source terms (S1 and S2 respectively).
- 3. Enter the following items on the dose assessment worksheet (Attachment I Section 4.0) 51 = Noble gas source term (CI/sec)
S2 = Radiolodine source term (CI/sec) S3 = Whole body DCF (MREM /HR/uci/cc) 54 = Thyroid DCF (MREM /HR/uct/cc)
- 4. -Attach to the dose assessment worksheet (Attachment I) a completed h Worksheet A.
s-12.0
1004.7-Revision.7 CASE I: SECONDARY SIDE RELEASE
.Section A - Determine the Reactor _ Coolant Activity by following the flow diagram starting in the' upper left hand corner then continue to Section B.
Section B - Deter.nine the OTSG tube rupture leak Rate by following the flow
~
diagram starting in the upper left hand corner. Then continue to Section C. Section C . Determine the transport fractio.is by following'the flow diagram starting in the upper left-hand corner. Then continue to Section D.
.Section 0 - to do the dose assessment daciculation use the answers from Sections A,.B and C. Fill in the appropriate blanks and calculate Si and S2 then proceed to Section E.
Section E - Follow directions as indicated at top of page. 13.0
a CASE I: SECONDARY SIDE RELEASE A .' Reactor Coolant Activity _ l Is the RML1 (RC Letdown Monitor) -- YES -- Enter the RML1 High Reading = A1 = CPM High Channel Reading in-cpm available? (Y/N) Calculate the RCS Activity ut in el - ( A1 + 22) = D1 (Enter Item D1 in Section D) NO Fill in the blank and check the appropriate item in Section E C1 = "RML1 High Channel Reading of CPM indicating 'uc" A D1 ml.
---- GO TO SECTION B ------------. -- -- --- -------------
Is the RML1 (RC Letdown Monitor) -- YES -- Enter the RML1 Low Reading = A1 = __ __ CPM Low Channel Reading in cpm available? (Y/N) Calculate the RCS Activity ut in ml - ( Al + 1220) - D1 (Enter Item D1 in Section D) NO Fill in the blank and check the appropriate item in~Section E C1 "RML1 Low Channel Reading of _____ __ CPM indicating'__._ _ uc" Al D1 ml-
-- GO TO SECTION B - - - ====------ -- -- ----
Is the most Recent RCS Sample Gross - .YES -- RCS Activity in uc = (Enter Item D1 in Section D) Beta Gamma Activity in uc/ml ml D1 available? (Y/N) Fill in the blank and check the appropriate item in Section E uC NO C1 "Most Recent RCS Sample of mi" D1
------ -- =- GO TO SECTION B ---- --- - ----- ==- =
RCS Activity e uc 360 in mT - - D1 (Enter Item D1 in Section D) C1 - FSAR Assumption of 11 FF and 360 uc Gross Beta Gamma activity. (Fill in the blank and check the ml appropriate item in Section E.) GG TO SECTION B
q' 1004.7 Revision 7
- CASE I: SECONDARY _ SIDE _ RELEASE OTSG Tube Rupture Leakrate B.
__L_ _ . _ _ _Is the Identified RCS Leakrate -- YES -- Enter the identified Leakrate - GPM D2 (Enter Item D2 in Section D)' in GPM _available? (Y/N) GPM" C2 " Identified RCS Leakrate of 02 NO (Fill in the blank and check the. appropriate item in Section E.)
----- -- GO TO SECTION C --------------------------------------------------
Is the Enter the unidentified Leakrate = _ GPM Unidentified RCS Leak- -- YES -- '(Enter Item D2 in Section D) D2 rate in GPM available? (Y/N) GPM" C2 = " Unidentified RCS Leakrate of
~2~
D NO (Fill in the blank and check the appropriate item in Section E.)
-- GO TO SECTION C --
Leakrate = 400 GPM = D2 (Enter item D2 in Section D) C2 - FSAR Assumption of 400 GPM leakrate (Check the appropriate item in Section E) GO TO SECTION C -- - ------ 15.0
p "
- i. 1004.7
^
Revision 7 6 -
' CASE I: SECONDARY SIDE RELEASE C. Transport Fractions l
Is there a direct release of - NO - Radiolodine Transport steam to the atmosphere (Y/N) Fraction .0075 - D3 (Enter Item D3 in Section D) (Check the appropriate item in Section E)
'YES~ C3 - Condenser Off-Gas Release
- - --- GO TO SECTION D -- ==- - ---------
.Is-E A fraction of total steam - NO - Radioiodine Transport flow through the condenser Fraction 'l - D3 hotwells? (Y/N) (Enter Item D3 in Section D)
(Check the appropriate item in Section E) C3 - Direct-Release of Steam YES to the Atmosphere I
--------- GO TO SECTION D -----------------------
Enter the fraction of steam flow directed to the condenser hotwell as (A) (See Table i " Steam Discharge Flowrates" Attachment IX) Radiolodine Transport Fraction - ( (A) *
.0075) + [(1 ~~ (A) ) *1]= +
= = D3 (Enter Item D3 in Section D)
.C3 - Combined Release-of Steam to the Condenser and Directly to Atmosphere (Check the appropriate item in Section E)
-- GO TO SECTION D --
I 16.0
y 1004.7-Revision 7'
~.
CASE I: SECONDARY' SIDE RELEASE
------------------ - = ------ WORKSHEET A ------ == ---- ---- ===-
D. Dose Assessment Calculation
~DI - Reactor Coolant Activity in (uc/ml) from Section A 02 = Primary to Secondary Leakrate in (GPM) from Section B D3 - Radiolodine' Transport Fraction from Section C Noble Gas Source Term CI in (sec) = 01- . D2 . SE-5 S1 Radioiodine
. Source Term CI in-(sec) - D1 . D2 . D3 . 2.5E-6 S2
-- GO TO SECTION E --
53 - 4E5 MREM
.HR uct CC
-S4 = 1.6E9 MREM HR UCI cc Enter S1,'S2, S3 and S4 onto the
- Dose Assessment Sheet, Attachment 1.
v - I a 0 17.0
2 1004.7 Revision 7
---------------- ==---======-------- WORKSHEET A --------------------------- =- !,i E. Dose Assessment Assumptions.(check appropriate entry and fill in the blank) a ~
uc C1 - -l[ll "RML1 High Channel Reading of CPM indicating iiiT" Al D1 uc
l[l
' "RML1 Low Channel Reading of Al CPM indicattnq D1 mI"
~uc l[l "Most Recent RCS Sampl'e of mI" DI.
- . uc
][l: FSAR Assumption of la FF and 360 iiiT C2=l(( " Identified RCS Leakrate of 02 GPM" l[l " Unidentified RCS~Leakrate of GPM"
,D2 l-_l C2 - FSAR Assumption of 400 GPM Leakrate
'C3 -l [l C3 - Conden er Off-Gas Release l[l C3 - Direct Release of Steam to the Atmosphere
. l[l C3 = Combined Release of Steam to the Condenser Off-Gas and Directly to Atmosphere-r 18.0 EE
1004.7 Revision 7 CASE II: REACTOR BUILDING RELEASE Section A - Determine Accident selection by following flow diagram starting at the upper left hand corner. =Then continue to Section B. Section B - Determine the_ Reactor Coolant Activity by following the flow diagram starting in the upper left hand corner (2 pages). Then continue to Section C. Section C - To make the calculation of Reactor Building Radionuclide concentrations, answer the question in the box then do the necessary calculation, then proceed to section D. Section 0 - To do the calculation of Reactor Building Leakrate follow the flow diagram starting in the upper left hand corner. Then continue to Section E. Section E .To do the calculation of Reactor Building source terms use the answers from Sections B, C and D then continue to Section F. Section F - Follow directions given in prerequisites.
. CASE II: REACTOR BUILDING RELEASE.
A. Accident Selection 1 Is the release associated with a spent fuel handling -- NO = Go to Section B accident? (Y/N) C1 - Spent Fuel Handling Accident in the RB (Check the appropriate item in Section F YES
-- NO - -
C2 = Number of damaged fuel rods 1s-FSAR postulated 208 Is the number of
. damaged fuel rods (Check the appropriate item in Section F) available? (Y/N)
Noble Gas Concentration YES ugl ( cc) = 1.7 = El (Enter Item El in Section E) Radiolodine Concentration uct
~
( EE) = 1.5E-3 = E2 (Enter Item E2 in Section E)
-- GO TO SECTION D --
Enter the number of damaged fuel rods - A1 - C2 - Actual number of damaged fuel rods (Check the appropriate item in Section F) Reactor Building uci Noble gas concentration ( cc) = (1.7 . ____ ) + 208 Al uct
= ci = El (Enter Item El in Section E)
Reactor Building uct __) + 208
~
Radiolodine Concentration ( EE) = (1.5E-3 . Al uqi
- cc = E2 (Enter Item E2 in Section E) -- GO TO SECTION D --
r
- 1004.'7 Revision'7 CASE II: REACTOR BUILDING RELEASE B. Reactor Coolant Activity, Cont'd Is the Enter the RML1 high reading =-A1.= CPM-RML1 (RC Letdown Monitor) . -- YES --
High Channel Reading in cpm Calculate the RCS Activity _Available? (Y/N) UC.
= A2 (Enter item A2 in Section C)
In ml - ( Al + 22) = NO C1 = "RML1 High Channel Reading of ._ CPM" Al-(Check appropriate item in Section F)
-- GO TO SECTION C --
Is the Enter the RML1 low reading = A1 =.___ _ CPM RML1 (RC Letdown Monitor) -- YES -- Low Channel Reading in cpm Calculate the RCS Activity _Aval.lable?__(Y/N) llc in ml = ( Al 1220) = = A2 (Enter item A2 in Section C) NO C1 = "RML1 Low Channel Reading of ._ ___._ CPM" A1 (Check appropriate item in Section F)
-- G0.T0 SECTION C --
1
4,4> o
-9 + IMAGE EVALUATION
///// / p M
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+ r.
l.0 5EnILE - ['ENas i,i j,m En 1.8 1.25 1.4 1.6 4 150mm >
< 6" >
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- b. , ;
r , 1904.7 Revision 7 l
.I CASE II: REACTOR BUILDING RELEASE B. Reactor Coolant Activity. Cont'd ,.
Based upon in-core instru- . . mentation does the ED -- YES -- Reactor Building noble gas concentration. susnect fuel meltina? l' E . E1 . (cc) = 5.7E3 ( cc) = El . (Enter item El in Section E). Reactor Building radiciodine concentration NO UC WC1 . (cc) = SE2 ( cc) = E2 . (Enter item E2 in Section E) C1.= Fuel melting as indicated by in-core instrtsnentation j (Check appropriate item in Section F)
-- GO TO SECTION D --
Based upon in-core instru-mentation does the ED sus- -- YES ------------- Reactor Butiding noble gas concentration sect fuel c1;tddine *ce? UC1 UC1 ( cc) = 160-( cc) = El Reactor Building radioiodine concentration El WC.i ( cc) = 13 ( cc) = E2 ~. NO C1 = Fuel cladding damage as indicated by in-core instrumentation f (Check appropriate item in Section F)
-- GO TO SECTION D --
ut Enter RCS Activity in e1 = = A2 (Enter item A2 in Section C) Most Recent RCS Sample Gross - YES -- Beta r- Activity in uc/m1 C1 = "Most Recent RCS Sample" (Circle appropriate item in Section F)
-- GO TO SECTION C --
NO -- -
- ===- :----- Reactor Building noble gas concentration MC (al) = 160 = E1 (Fnter item El in Section E)
Radiciodine concentration E (al) = 13 = E2 (Enter item E2 in Section E) C1 = "FSAR assumed fuel cladding damage"
.(Check appropriate item in Section F.
-- GO TO SECTION D --
q 1004.7. Revision'7 ; q CASE II: REACTOR BUILDING RELEASE C. Calculation of Reactor Building Radionuclide Concentrations I C1 - FSAR assumed fuel cladding damage Is the total number of . (Check-the appropriate item in Section'F) gallons of RCS leakage into -- NO --- the-Reactor Building available? (YlN) Reactor Building noble gas concentration uti (Enter item El in Section E)
~
( Ed) = 160.= El , Radiolodine concentration 1 uC1
~
( 2c) = 13 - E2 (Enter item E2 in Section E) YES
-- GO TO SECTION D --
Enter the total number of gallons = __ = A3 (Enter item A3 below) l C2 - Actual number of gallons of RCS leakage into the Reactor Building (Check the appropriate item in Section F) uct
~
Calculate the Reactor Building Noble Gas Concentration ( 2c) = ( A2~~ . A3 . 2950) + 5.6E10' = _. - El (Enter item El in Section E) uct
~
Reactor Building Radiolodine Concentration ( Ec) = . ( A2'~ . A3 . 56) + 5.6E10 -
= 'E2 (Enter item E2 in Section E) -- GO TO SECTION D --
ki 1004.7" Revision 7-
.c CASE II: . REACTOR BUILDING RELEASE D. Calculation of Reactor Building Leakrate I
A4 - 50.6 psig. Is the actual Reactor C3 = FSAR postulated RB pressure of'50.6 psig Building internal pressure -- NO -- indicated on: PT-291? - (Y/N) . (Check the appropriate item in Section F) Enter item A4 below and perform the calculation
~
YES Enter the actual pressure = = A4 C3 = Actual RB Internal pressure (Check the appropriate item in Section F) , cc --- Calculate the actual RB leakrate (sec) I cc
= (656 * -A4 2) = (sec) - E3 (Enter item E3 in Section E)
( 50.6 )
-- GO TO SECTION E --
HORKSHEET A CASE II: REACTOR BUILDING RELEASE E. Calculation of Reactor Building Source Terms CI Noble gas source term (sec) -Enter Items S1, S2,.53 and 54 onto the. dose assessment sheet, Attachment 1.
-( El . E3 ) + IE6 - S3 =.4E5 MREM S1 HR CI Radiolodine source term (sec)
-( ~ E2~ . ~'3'~
E ) + IE6 - S4 - 1.6E9 MREM __ S2. HR Tcl
~ Tc1 F. Dose Assessment Assumptions (Check the box and fill in the blank items if applicable) l _l C1 - Spent fuel handling accident in the Reactor Building l ~l C2 - Number of damaged fuel rods is FSAR assumed 208 l' l C2 - Actual number of damaged fuel rods is _ __.
Al l[l C1 - RML1 high channel reading of __ CPM Al l[l C1 - RML1 low channel reading cf _ CPM Al l[l C1 - Fuel melting as indicated by in-core instrumentation l[l C1 - Fuel cladding damage.as indicated by in-core instrumentation l[l C1 - Most recent RCS sample l[l C1 - FSAR assumed fuel cladding damage C2 - Actual number of gallons of RCS leakage into the Reactor Building l_l lll C3 - FSAR postulated RB pressure of 50.6 psig l l C3 - Actual RB pressure of . . psig A4
1004.7 Revision 17 l CASE III: AUXILIARY AND FUEL HANDLING BUILDING RELEASE ; Section A - Follow the accident selection flow diagram and then continue to
- the section indicated by the answer, (yes or no).
2 Sections-B,'C,.0,~ and E . In these sections do'the necessary calculations to get the answers S1, S2, and Cl, then continue to Section F. Section F . Fill out sheet completely.
1004.7 Revision 7
-CASE III: AUXILIARY.AND FUEL HANDLING BUILDING RELEASE A. Accident Selection
'l Has a grab sample-been .
I obtained of the affected_ . -- YES -- GO TO SECTION B area.and analyzed ~on a geli spectrophotometer? (Y/N) NO-Does the accident involve. fuel assembly damage in the - YES -- GO T0'SECTION C spent fuel pool?- (Y/N) NO Does the accident-involve fuel cask drop during trans- -- YES -- GO TO SECTION D. fer operation? (Y/N) NO Does the accident involve -- YES -- GO TO SECTION E a waste gas release? (Y/N) NO --------- = GO TO SECTION F --
1004.7 Revision 7 CASE III: AUXILIARY AND FUEL HANDLING BUILDING RELEASE B. Source Term Generatica Based Upon a Grab Sample CI LCalculate the noble gas source term in (sic) where
-uC
.B1 - total i i of noble gas isotopes as indicated in sample B2 - ventilation flowrate from affected building in CFM Noble gas source term - x x 4.7E CI Si m B1 B2 sec CI
- Calculate the radiolodine source term in .(sic) where uc B3 - total EE of radiolodine isotopes as indicated in sample B2 - ventilation flowrate from affected building in CFM CI Radiolodine source term - x x 4.7E (sec)
B3 B2 S2 Enter B2, S1, and 52 in Section F Enter the Time /Date of the sample in the blank below C1 - Grab sample analyzed on a geli spectrophotometer at Time /Date-(Check the appropriate item in Section F and fill in the blanks)
-- GO TO SECTION F -- ,
g- - 1004.7- -l Revision 7 j CASE-III: AUXILIARY AND FUEL HANDLING BUILDING RELEASE
- C. Fuel Assembly Damage in the Spent Fuel Pool CI 4.2 Noble gas source term _(sR) - S1 (Enter Sl and S2 in Section F)
CI- 7.5E-4 Radiolodine-source term (sR) - S2 C1 - FSAR postulated fuel assembly damage in the spent fuel pool (Check the appropriate item in Section F)
-- GO TO SECTION F --
D. Fuel Cask Drop During Transfer Operation (Enter S1 and S2 in Section F) CI 1.2E-3 Noble gas source term (s R) = S1 CI 4.5E-4 Radiotodine source term (sR) - 52 C1 - FSAR postulated fuel cask drop during transfer operation (Check the appropriate item in Section F)
-- GO TO SECTION F --
1004.7 Revision 7 CASE _III: AUXILIARY AND FUEL HANDLING BUILDING RELEASE E. Haste Gas Decay Tank Rupture CI .!6 Noble gas source term-(sec)'- il (Enter Si and S2 in Section F) CI .004 _Radiolodine source term (sec) - S2
-Cl--.FSAR postulated waste gas decay tank rupture (Check appropriate item in Section F)
-- GO TO SECTION F --
l F
1004.7 Revision 7 CASE III: AUXILIARY AND FUEL HANDLING BUILDING RELEASE I F. Dose assessment assumptions (Check the box and fill in applicable items)
-l((lCl-Grabsampleanalyzedonagelispectrophotometerat Time /Date l((lC2=Ventilationflowrateof CFM B2 l((lC1'=FSARpostulatedfuelassemblydamageinthespentfuelpool l((lC1-FSARpostulatedfuelcaskdropduringtransferoperation , -l((lC1-FSARpostulatedwastegasdecaytankrupture S1 - (CI/sec)
S2 - -(CI/sec) S3 - 4E5 (MREM /HR/uct/cc) 54 = 1.6E9 (MREM /HR/uct/cc) Enter Items SI, S2, S3, and S4 onto the Dose Assessment Sheet, Attachment 1. t
1004.7 Revision 7 . ATTACHMENT V LIQUID RELEASE CALCULATION
- 1. . Estimate quantity,of radioactive 11guld released or the release. rate of the liquid being released gallons or gpm.
(la) (ib)
- 2. From recorded information or sample analysis determine the activity level (in uC1/ml) of the released liquid:
pC1/ml. (2)
- 3. - Obtain the river-level by calling the River Forecast Center in Harrisburg at phone number 782-2256 or 782-3488 and record the reading: ft.
(3)
- 4. Find +he river flow corresponding to the river level No. 3 above, in Table-I, and record: CFS.
(4)
- 5. Calculate the average and maximum downstream concentrations of radio-active material as follows:
Maximum
.uC1 .: cfs pCl**
ml x gpm x 2.33 x 10 gpm + cfs - ml (2) (ib) (4) (5)
- NOTE:
** If the average or maximum downstream concentration :
- is > 1 x 10-' pC1/ml, notify downstream users to :
- curtail intake. :
r 2 e, + %- y+-,m--,, ,yy - - y, , , , - , , , , , , _ _ , , _ , _ _ , _ , , , , , _ , ,_,_,, -
1004.7 Revision 7 ATTACHMENT V (Cont'd) Time for Flume to Reach Downstream Users
- 6. Downstream Points (Table II)
- 7. -Distance to Point in miles (9) (9) (9) (9) (9)
-(Table II)
- 8. River velocity,in mph cor-(10) (10) (10) (10) (10) responding to river flow
-from (4) above (Table 1)
- 9. Calculate a time in hours for the fiume to reach selected point: Step 7 Step 8 a
24 Hour Average Concentration in Unrestricted Areas
- 10. Record the duration of the release in minutes: min.
- 11. Calculate a 24 hour average concentration in unrestricted areas:
uC1 4 pCl ml x min x 6.95 x 10 - ml (Sa) (1) (1)
1004.7 Revision 7 ATTACHMENT V (Cont'd)
- 12. Determine the estimated fraction of MPC:*
uCi mi + MPC.** - Fraction of MPC (13) (12)
==_
- NOTE: ' * 'If the ratio obtained in'(14) of Attachment is >S00, :
- notification of NRC is required with 24 hours per :
- 10CFR20.403. .If the ratio obtained.in >5,000, :
- immediate notification is required per 10CFR20.403. :
_________________________________________________________________- =
- NOTE:
** MPC is the weighted MPC for the isotopes released. :
- If unknown. use 3 x 10'* pC1/ml. :
5 L \ L i_______________. _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
C 1004.7 72- l Revision.7_ ATTACHMENT V (Cont'd) l TABLE I RIVER FLOW VS. RIVER LEVEL
, .A- B .C .D L
- Gauge Reading River Elevation Market Street at TMI. _ River Flow River Bridge, Hbg. .(Feet Above (Cubic Feet Velocity (Feet) Sea Level) per Second) (MPH)
> 4.3 278.7 20,000 . '. 9 -
5.3 279.5 40,000 1.4
-6.2 280.1 60,000 .7 7.1 280.7 80,000 2.0 8.1 .281.3 100,000 2.3 10.4 282.5 150,000 2.6
_12.5 283.6 200,000 3.1 14.3 -284.9 250,000 3.3 16.1 285.8 300,000 3.5 17.9 287.0 350,000 3.7
. 19.5 288.1 400,000 3.9 21.2 289.7 450,000 4.1 22.7 291.0- 500,000 4.3 24.3 292.6 550,000 4.5 25.6 294.0 600,000 4.7 26.9 295.2 650,000 4.9 28.1 296.1 700,000 5.1 29.3 297.1 750,000.- 5.3 l30.4 298.1 800,000 5.5 31.3 299.2 850,000 5.7 32.0 .300.1 900,000 5.9 32.6 301.1 950,000 6.1 33.1 302.0 1,000,000 6.3
- -NOTE: River elevations 302.0 feet at water intake struc- :
- ture TMI requires initiation of EPIP 1004.2 ALERT. :
^
1004.7 Revis18n 7 ATTACHMENT V (Cont'd) TABLE II
-DOWNSTREAM POINTS Downstream Distance To Water Users User (6) (miles) (7)
Brunner Island Steam Electric Station 5.0 Wrightsville Water Supply Ccepany 16.25 Borough of Columbia 16.75 City of Lan: aster 16.75 Safe Harbor Water and Power Corp. 27.25 Holtwood Reservoir 34.75 Chester Water Authority 43 City of Baltimore 49
g.;-_ 1004.7 Revision 7 ATTACHMENT VI Protective' Action Guides / Protective Action Recommendation Protective Action : Actual or Projected Exclusion Guide (PAG'S) : Area Dose (rem)
- Whole Body : Thyroid
' Lower Limit-(PAG) : I : 5
-Upper Limit (PAG)- : 5 : 25
.g.
- h. .
O l [
( 1004.7 Revision 7 LOGIC DIAGRAM DEVELOPENT OF PROTECTIVE ACTION REC 0f9ENDAT10NS (PAR) . I SITE OR GENERAL EMERGENCY CONSIDER SHELTERING DECLARED FOR A SIGNIFICANT EXCLUSION AREA PRO- """" ~ ~ ~~~""" f""* JECTED DOSE RESULTING l FROM A PUFF RELEASE EPA - PAG'S (LOWER LIMIT) YES _ EPA
- PAG S (UPPER LIMIT)
I NO EXCEEDED OR PROJECTED 1 EXCEEDED 04 PROJECTED To BE EXCEEDED 7 TO BE EXCEEDED 7 l I YII UNCERTAIN l LARGE FISSION PRODUCT UNCERTAINTY AS TO C $1 DER SHELT MILE RAD W
\
==
l Y 1, THE RELIABILITY Op ' INVENTORY IN CONTAINMENT? 5 nlLE y lND PLANT FUNCTIONS TO # (SUFFIClENT PAG S LOWERToLIMIT EXCEED) EPA
- PR0 ECT THE P IC SEE NOTE YES/- ISRkEASE g
RATION IXPECTED g TODELyG SUBSTANTIAL CORE I
\
(DAMAGE PROJECTED )
> 20% FUEL DAMAGE NO No VES
'r A \
- C EVAC (EATI RECb4 MEND SHELTE NG
\,
j SE LISHE l PRI TO P OF AL I ARRIVAL 7 POTENTIA I l* CONTINUg NO CONTAINMENT INTEGRITY N NOTE 2 ]D ASSESSMENT FAILED R ECTED ,f 5 \g YES VES \ g RECOMMEND CAN EVACUATION SE N IVACUATION Ytt ACCOMPLISHED PRIOR TO PLUM 4i 4RR IVAL
.\'
/
'\ N ./ OF AFF(CTED AREA (S)
/ (SEE NOTExt)
- NO x '\ m w ,?
Rteo mi EVACU4TtoM ("'
/
\ \ \ ,/ '
0FAREA AFF(C{f (Si , ', RfC 70 ND Shti.TERING /0R AREA (S) THA CANNOT SEsEVACU ED Pal 04 uME ARRIVAL AND COMMEND
% REA(S)
,/
y ' N.gg\, \' E CU4T ON OF 0T
' SEE NOTEj Cs \\ s f' NOTE M CONSIDttAtl0N SHELTERfsVICE EV CUAlls.
SMOULR N BE GIVLN TO THEMECTE0 EXPOSURE TO BE REC ( h AND AS$UML A PROT CTION (ASE FACT DURAfl0N. THE0dNOF PATHWAY 2 FOR UP TO OF LEAST THE FIRST EXPOSURE SHOULD Z HOURS OF St CH0$tN. PF OF A F0 >2H R$ Of t 2:' v21LEVACUAfl0N TIT , ESTIMATES LUwtR (hours) UPPER (HOURS) BES t$f MAft (N GMT) . TYPI L WEtKDAY/ NORMAL) . 1 . ADVER$twtAyt wf t
- D STATE OF eM(RGtNCY READINESS (SLOW SCENARIO U tR LACK OF ADt0UATE P4tPARAfl0N TIME (FAST SCENAtlD}
suCGMENT AS TO THE NetD FOR PROTtCTlve Actl0N RECOMMtNDATIONS, ANY Nott 3: IN a FRCISING TH UNCtRTAINTY CONCERNING THE STATUS OF PLANT FUNCTIONS NttDt0 FOR PROTE PUBLIC, THE LENGTH OF TIME THE UNCERTAINTY EXISIS, THE PROSPECTS FOR E ARLY RESOLUTION OF AM81GulTIES, AND THE POTENTI AL DeGRADAtl04 0F THE PLANT FUNCTIONS Nf tDtD FO PROTECTION OF PUBLIC $N00LD At CONSIDFREDJ lete, RELI ABILITY OF PLANT FUNCTIONS 70 PROTRCT THE PUBLIC ENTENDING SR PROTICTIVE ACTION RECOMMENDAtl0N TO SHELTER PERIOD IS A SUFFICIENT RAlls FOR MAKING MILtl DOWNWIND. CONTINUE PLANT AlltBSMENT. NITHIN A 2 MILE RADIUS OF THE PLANT AND O e
ATTACHMENT VII 1004.7 Revision 7 ,
- 3. AIRBORNE IODINE SAMPLE NOMDCRAFR (Cro:s CFM-Bkg. Crt0 Note: This nomograph is to be used for Iodine 131 10- air samples. counted with a SAM II. This nomograph assumes an ave. ' counter factor of 16000 for SAM II's.
1 Airbo e Activity A1 3 ample S Volume l 20 (al or ec) (uci/ad'!-5 (Ft ). -
*9
- 6 40 -
s
.4
~
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- E-6 '
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- g\.
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. 2.0; _734 ,
7a; 4.# 55 ,,',E 7 7M -2E5\ 1000 : N \ 3
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7000j f ~ ' ss \, - \' (N._s k, .'
>E-8 10000 '/
/ ,7-s, \\ g ( \ s//
3 j
~
20000 , s fastructions: Draw a line through Net CFM (A)
~ ndAirSageVolume(5)usingastraightedge -!
40000 and read 1 Airborne Activity (C) on the line. "*
-s
~'
'""N .
1014 E .\ OMOCRA?- -
*E4 100000 : -:
J J
. - - - __.......___.,7- _._m _ , _ _ _ , _ . , ,,_._-__.~,,.~m_ ,___,.,_.-,m,_
1004.7 Reviston 7 ATTACHMENT VIII COMPUTERIZED DOSE CALCULATIONS
- 1. Ensure computer components are connected as pictured in Attachment 1A.
- 2. Energize the system components in the following order: ,
- a. Quick Printer II
- b. Video Display
- c. Keyboard Terminal
- d. Expansion Interface
- 3. Computer will respond with the following message:
MEMORY SIZE - Strike the ' ENTER' Key
- 4. Computer will respond with:
RADIO SHACK LEVEL II BASIC READY
- NOTE: For loading Unit II programs go to Step 7. :
- 5. For airborne release:
Place cassette labeled ' Program "D" Altborne Dose Calculations' in recorder and ensure cassette is rewound. Depress the PLAY button, set volume level to '4'. i
- 6. For liquid release:
Place cassette labeled ' Program "L" Liquid Release Calculations' in recorder and ensure cassette is rewound. Depress the PLAY button, set volume level to '4'. 1 P
1004.7 Revision 7 ATTACHMENT VIII (Cont'd)
- 7. For Unit II airborne release with RMS system in. service and on-scale:
Place cassette labeled "EMERG2" Unit II Emergency Dose Calculations' in recorder and ensure cassette is rewound. Depress the PLAY button, set volume level to "4".
- 8. For Unit II airborne release with RMS system out-of-service and/or off-scale:
Place cassette labeled " Emergency Contingency Calculations" in recorder and ensure cassette is rewound. Depress the PLAY button, set volume level to "4".
- 9. Enter the following command from the keyboard:
CLOAD "0" for Unit I airborne; CLOAD "L" for Unit I liquid; CLOAD "EMERG2" for Unit II Emergency Dose Calculations; CLOAD " CONT 2" for Unit II Emergency Contingency Calculations and strike the ' ENTER' key. At this time the cassette will begin loading the program into the computer memory. Program loading will take approximately 2 1/2 or 3 minutes. One steady and one blinking star will appear in the upper right corner of the video display to signify program loading is in progress.
- NOTE: If both stars appear, with neither blinking; 1.e. :
- both steady replace cassette with new copy and start :
- over at step S. :
- 10. When program loading is completed, the computer will respond with:
READY Depress stop button, rewind the cassette and remove it from the recorder.
1004.7 Revision 7 ATTACHMENT VIII (Cont'd)
- 11. To begin program execution, enter the following command from the keyboard:
RUN and strike the ' ENTER' key.
- 12. General notes on program operation:
- a. All responses must be followed by striking the ' ENTER' key.
-b. Numbers in scientific notation should be entered using the following formats:
9.2 x 10' = 9.2E3 4.0 x 10'* = 4E-4
- c. All responses requiring a yes or no, are to be answered with a Y or N.
I r
1004.7 Revision 7 ATTACHENT VIII COMPUTER CONNECTIONS AND COLOR CODES 110VAC 110 VAC 11') VAC l' OV a, o o n
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E F 1004.7 Revision 7 ATTACHMENT IX HIGH RANGE RMS DOSE CALCULATIONS Section A - System Description The High Range RMS is categorized into three distinct subsystems (See Schematic):
- 1. Radiolodine Processor Stations
- 2. Containment Air Sampling
- 3. High Range Noble Gas Channels Subsystem (1):
Radiolodine Processor Stations Three stations allow samples to be obtained independent of radiation monitors RM-AS, A8 and A9. The stations are controIIed by solenoid valves which actuate flow through one or more of the (3) parallel filter cartridges per station. The sampilng times for each filter cartridge are adjustable on each local control panel. The filter cartridges must be manually removed for analysis. Subsystem (2): Containment Air Sampling The post accident RB atmospheric sampIlng station is located at the 322' level of the intermediate building, one floor above radiation monitor RM-A2. Three-way ball valves are Installed in the RM-A2 sampling lines downstream of containment isolation valves. The sampling lines are connected downstream of CH-V1, CH-V2, CH-V3 and CH-V4 at P-108.
1004.7 Revisicn 7 ATTACHMENT IX HIGH RANGE RMS DOSE CALCULATIONS SUBSYSTEM 3: HIGH RANGE NOBLE GAS CHANNELS
- DETECTOR : RANGE : CONVERSION : FLOWRATE MONITOR : EFFLUENT TYPE pC)/cc : FACTOR CPM /pci/cc : CFM RECORDER DESIGNATION : PATHWAY RM-A8G High : Aux and FHB : GM Tube : IE-2/IE-3 : IE3 : FR-151 RM-A9G High : RB Purge : GM Tube : IE-3/1E2 : 2.6E3 : FR-148
- RB Purge : Ion Chamber : IEl-1ES : 9.5 : FR-148 RM-G24 RM-A5 High : Condenser : GH Tube : IE-3/IE2 : SE3 : See Table 1
- Off-Gas : :
RM-G25 : Condenser : Ion Chamber : IE1/IES : 1.5 : See Table 1
- Off-Gas : : :
RM-G26 : A, B Main : Scintillation : 1E-2/IE3 : 1020 : See Table 1
- Steam Lines : : :
RM-G27 : C, O Main : Scintillation : IE-2/IE3 : 1056 : See Table 1
- Steam Lines : :
(MR/HR/ pct /cc
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4a5 5g 55 ae m$9 - ne dB EB E 'I / 1004.7 Revision 7 ATTACHMENT IX HIGH RANGE RMS DOSE CALCULATIONS Section B - Source Term Calculations 1.0 Calculation of the Radiolodine Source Term utilizing the Radiolodine Processor Station. 1.1 Enter the radiotodine concentration in microcuries/cc as determined per EPIP 1004.31 from the stiver zeolite cartridge: 1131 pel/cc I132 pci/cc I133 pcl/cc I134 pci/cc I135 pcl/cc Total pci/cc (A) 1.3 Enter the release flowrate in cubic feet per minute (CFM) as determined from the Table below:
- Release Pathway : Release Flowrate (CFM) :
- Station Vent : FR-151 :
- RB Purge Duct : FR-148 :
- Condenser Off-Gas : See Table 1 :
Release Flowrate (CFM) (B) 1.4 Calculate the Radiolodine Release Source Term uttilzing the following equation: Radiolodine Release concentration (pCl/cc) x Flowrate (CFM) (A) (B) x Curie Conversion x flowrate Conversion Radiolodine Source Factor IE-6 CI Factor 472 cc . Term CI pit- seq iii CFM 1.5 Go to Attachment 1, Section 4.0 " Dose Assessment Sheet" 2
1004.7 r Revision 7 TABLE 1 Steam Discharge Flow Rates (1) Steam Generator "A", "B" Valve Tag No. Steam Flow #/hr. Press. PSIG
'MS-V17A, MS-V178, C and D 792,610 1050 MS-V18A, MS-V188, C and 0 799,990 1060 MS-V19A, MS-V198, C and D 814,955 1080 MS-V20A, MS-V203, C and D 824,265 1092 MS-V21A, MS-V218 194,900 1040 MS-V22A 70,212 200 MS-V22B 76,793' 200 (2) Steam discharged from, steam Generator B similar for Valve MS-V17C, O MS-V18C, D MS-V19C, D, MS-V20C, 0, MS-V218. (MS-V228 is 76,793 #/hr at 200 PSIG)
(3) Steam Dump to Atmosphere MS-V4A and B
% Valve opening demand Steam flow #/hr.
20 1.77 x 10' 40 3.6 x 10' 60 5.09 x 10' 80 5.61 x 10' 100 5.767 x 10'
.(4) Condenser Vacuum Pump Olscharge Path To be read from flowmeter on the pumps, or if unknown, use 20 SCFM.
(5) In the event of a direct release of steam to the atmosphere utilizing RMG-26 or MRG-27 to monitor the source term; the following term shall be included with the release flowrate: Total Steam Flowrote RMG-26 or RMG-27 Steam Flow Past
g- . . Q.r 1004.7 f)
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i 1004.7 Revision 7 I ATTACHMENT IX (Cont'd) 2.0 Calculation of-the Radiolodine and Noble Gas Source Terms utilizing the Containment Air Sampling Station. h 2.1 Enter the Containment Air Sampling Bomb Radionuclide Concentrations in microcuries per cubic centimeter (pci/cc) as determined per EPIP-1004.31: Noble Gas Nuclides Radioiodine Nuclides
- KR85 pc1/cc Il31 pci/cc KR85m pci/cc I132 pci/cc KR87 pci/cc 1133 pci/cc l
- . KR88 pc1/cc I134 pci/cc l
- -XE133 pc1/cc 1135 pct /cc XE133m pci/cc Total Radiolodine pci/cc XE135 __ pci/cc (A2)
XE135m pci/cc L Total Noble Gas pc1/cc i (Ai) L 2.2 Enter the Reactor Building release flowrate as determined from the I Table below: Release Flowrate CFM (B) h I k : Purge Valves Open FR-148 :
- Purge Valves Closed See Table 2 :
2.3 Calculate the Noble Gas Release. Source Term utilizing the.following equation: L Total Noble Gas x Release Flow x Flowrate Conversion Concentration uti Rate CFM Factor 472 cc (A1) cc (B) sec CFM Curie Conversion x Factor IE-6 Ci - Noble Gas pci Source Term Ci Sec
7 [ 1004.7 Revision 7 ATTACHMENT IX (Cont'd) 2.4 Calculate the Radiolodine Release Source Term utilzing the following equation: Total Radiciodine Release Flow Flowrate Conversion Concentration pc1 x Rate CFM x Factor 472 cc (A2) cc (B) Sec CFM Curie Conversion x Factor IE-6 C1 - Radioiodine pc1 Source Term C1 Sec 2.5 Go to Attachment 1, Section 4.0 " Dose Assessment Sheet" 3.0 Calculation of the Noble Gas and Radioiodine Source Term utilizing the High Range Noble Gas Channels. 3.1 Enter the Noble Gas Channel reading in CPM: CPM (A) 3.2 Enter the meter conversion factor as identified in Section A: CPM or MR (B) HR UCi cc Enter the postulated mixture conversion factor as identified in Table 3: pc1 Pos. Mix (B1) cc pci Cal. Isotope CC Enter the nuclide class fraction as identified in Table 4: pci Noble Gas (B2) cc
, uti Pos. Mix CC 3.3 Enter the Release Flowrate in CFM as identified in Section A:
CFM (C)
1004.7 Revision 7 3.4 Calculate the Noble Gas Source-Term in curies per second (CI/SEC) utilizing the equation below: Noble Gas Channel + Meter Conversion Reading (CPM) X Factor CPM or MR X pc1 Pos. Mix (A) (8) HR (81) cc uct uci Cal. Isotope cc Cal. Isotope cc x pc1 Noble Gas x Release Flow x Release Flowrate (B2) cc Rate CFM Conversion 472 cc uci Pos. Mix (C) sec cc CFM X Curie Conversion Noble Gas lE-6 CI = Source Term pc CI (D) sec 3.5 Calculate the Radiciodine Source term in curies per second (CI/SEC) as follows: 3.5.1- Enter the Noble Gas Source term as calculated in Step 3.4: CI (D) sec 3.5.2 Enter the fraction of Radiolodine as determined from Table 4: Radiolodine (E) Fraction-3.5.3 Enter the fraction of Noble Gas as determined from-Table 4: Noble Gas (F) Fraction 3.5.4 Determine the Radiolodine source term utilizing the equation below: Radiolodine Noble CI Noble Gas X Fraction + Gas (D) sec (E) (F) Fraction
- Radiolodine Source Term CI (G) sec < 3.6 Go to Attachment 1, Section 4.0 " Dose Assessment Sheet" l
r: 1004.7 Revision 7 ATTACHMENT IX-(Cont'd) TABLE 3 Postulated Mixture Conversion factor (uci Calibration Isotope to uCi Pos. Mixture) CC CC
- MONITOR : EFFLUENT : CALIBRATION : CONVERSION DESIGNATION : PATHWAY : ISOTOPE : FACTOR CPM /uct/cc RM-A8G-High : Aux and FHB : XE133 : 0.7 RM-A9G High : RB Purge : XE133 : 0.7 RM-G24 : RB Purge : KR85 : 0.01 RM-A5 High : Condenser : XE133 : 0.6
- Off-Gas : :
RM-G25 : Condenser : XE133 : 0.6
- Off-Gas : :
RM-G26 : A, 8 Main : KR85 : 0.007
- Steam Lines : : 0.005 RM-G27 : C, D Main. : KR85 : 0.007
- Steam Lines : : 0.005 Iodine spike assumption based upon plant transient.
l l l l u_.~. l
1004.7 Revision 7 1 ATTACHMENT IX (Cont'd) TABLE 4 Nucilde Class Fraction Of Postulated Mixture (uCi Pos. Mixture to uCi NuClide Class) CC CC
' MONITOR : NOBLE GAS : RADI0 IODINE DESIGNATION : FRACTION : FRACTION RM-A8G High : 0.94 : 0.07 RM-A9G High : 0.94- : 0.07 RM-G24- : 0.94 : 0.07
.RM-A5 High : 1.00 : 0.0008 RM-G25 : 1.00 : 0.0008 RM-G26 : 0.80 : 0.06 0.67 : 0.24 RM-G27 : 0.80 : 0.06 0.67 :
0.24
- Iodine spike assumption based upon plant transient.
1004.7 Revision 7 ATTACHMENT X. DOSE CONVERSION FACTOR CALCULATION
----- ==--------- =- Instructions for Using Attachment X - ------------------
tl. Select'a DCF calculation from the posted menu: A. Whole body DCF calculation based upon gamma spectrum analysis.
-- GO TO SECTION A --
B. Thyroid DCF calculation based upon a gamma spectrum analysis.
-- GO TO SECTION B --
C. Whole bo'y d DCF decay correction (assumes I hr elapsed time from original sample analysis).
-- GO TO SECTION C --
D. Thyroid DCF decay correction (assumes I hr elapsed time from original sample analysis).
-- GO TO SECTION D --
E. Default DCF (DDCF) calculations
-- GO TO SECTION E --
- 2. For the selected DCF calculation determine the whole body or thyroid DCF, S3 and S4 respectively.
- 3. Enter the following items on the dose assessment worksheet (Attachment I, Section 4.0).
S3 - Whole body DCF (MREM /HR/uct/cc)
-S4 - Radiolodine DCF (MREM /HR/uci/cc)
- 4. Attach to the dose assessment worksheet (Attachment I) a completed Worksheet B.
- - - - - - - - - ,,.--,,..-,n , y ,
1004.7 Revision 7-A. WHOLE BODY-DOSE CONVERSION FACTOR (WBDCF) CALCULATION
-------------------------------Utilizing the Attached Worksheet=- ---- --
- 1. Enter the Date/ Time of the sample analysis.
ILc
- 2. Enter Concentrations in (cc) for the listed nuclides in column 2.
- 3. Multiply the concentration of the listed nucildes (column 2) by the photon energy (column 3) to obtain.the photon contribution (column 4).
Enter the Photon Contribution in column 4.
- 4. Determine the~ total concentration of the listed nuclides by adding items a -m. of column 2. Enter the Total IConcentration as item Al.
- 5. -Determine the total photon contribution of the listed nuclides by adding items a.-m. of column 4. Enter the Total Photon Contribution as item A2.
~6.- Enter. items Al and A2 in equation A-1. : Calculate the WBDCF on the worksheet as item S3.
Il-1004.7 Revision 7
---------------------------------------- WORKSHEET B -------------- - -=---- ==- - ===-
WBDCF Calculation
- Sample Date/ Time COL 1 COL 2 COL 3 COL 4 Nuclide Concentration- Photon Energy Photon Contribution
. a.) KR 85M X .18 -
b.) KR 85 X .0022 - c.) KR 87 X .79 - d.) KR 88 X 2.2 - e.) XE 133M X .02 - f.) XE 133- X .03 - g.) XE 135M X .53 - h.) XE 135 X .26 - 1.) I 131 X .39 - j.) I 132 X 2.2 - k.) I 133 X .6 - 1.) I 134 X 2.6 - c.) I 135 X 1.5 =
.Tctal Conc. - Total Photon Cont. -
Al A2 Equation A-1 WBDCF Calculation: WSOCF in MREM HR uct CC Sample Date/ Time 2.1 x 10 5
.- ( A2 + Al ) . 9ES - S3 L
1004.7 Revision 7 B. THYROID DOSE CONVERSION FACTOR (TDCF) CALCULATION .----------------------- === Utilizing the Attached Worksheet - =-- =-- ---- '1. Enter the Date/ Time of the sample analysis.
- 2. Enter the Sample Concentrations in pci/cc for the listed nuclides in column 2.
- 3. Multiply the nuclide concentrations in column 2 by the isotope DCF in column 3 to obtain the isotope contributions (column 4). Enter the Isotope Contributions in column 4.
- 4. Determine the total concentration for the listed nuclides by adding items a.-e. of column 2. Enter the Total Concentration as item A1.
- 5. Determine the total isotope contribution for the listad nuclides by adding items a.-e.
of column 4. Enter the Total Isotope Contribution as item A2.
- 6. Enter items Al and A2 in equation 8-1.
- 7. Calculate the TDCF utilizing equation B-1. Enter the TDCF as item S4 on the worksheet.
b '
.joo4,7
. Revision 7
.______ - ...___________. _...--------- WORKSHEET B ------ --- -- --
Thyroid DCF Calculation
- Sample Date/ Time COL 1~ COL 2- COL 3 COL 4 Isotope Nuclides Concentration Isotope DCF Contributions
- a. I 131 1.6E9
- b. I 132 7.9E7
- c. I-133 5.4E8
- d. I 134. 4E7
- e. I 135 1.6E8 Tstal. Concentration - Total Isotope Contributions -
Al A2
. Equation B-1 TDCF. Calculation:
TDCF in; MREM HR uci cc
- (-
~--
A2 + Al ) .- S4
1004.7 Revision 7 C. WHOLE BODY DOSE CONVERSION FACTOR (WBDCF) DECAY CORRECTION
---- -- -------- =- --- -- Utilizing the Attached Worksheet == =- - =--- - - ==- --
- 1. Enter the Date/ Time of the original sample.
- 2. Enter the Original Concentrations in pcl/cc for the listed nuclides in column 2.
- 3. Multiply the original nuclide concentrations (column 2) by the remaining fraction (column 3) to obtain the present concentration (column 4). Enter the Present
' Concentrations in column 4. Factors in column 3 account for 1 hour decay.
- 4. Multiply the Present Concentration (column 4) by the photon energy (column 5) for the listed nuclides to obtain the photon contribution (column 6). Enter the Photon Contributions in column 6.
- 5. Determine the total concentrations for the listed nuclides by adding items a.-m. of column 4. Enter the Total Present Concentration as item A1.
- 6. Determine the total photon contribution for the listed nuclides by adding items a.-m.
of column 6. Enter the Total Photon Contribution as item A2.
- 7. Enter items Al and A2 in equation C-1.
- 8. Calculate the WBDCF utilizing equation C-1. Enter the WBDCF on the worksheet as item S3.
- 9. .The elapsed time between the original concentrations and calculated WBDCF is 60 minutes. Determine the date/ time of the calculated WBDCF by adding 60 minutes to the original sample time. Enter the date/ time on the worksheet.
1004.7 Revision-7
- - - - - - - = = - - - - - = - - - - - - - . - - - - - - --- ---- HORKSHEET B -----=-- - =
(WBDCF Decay Correction) Original Concentration Date/ Time COL. 1 COL. 2 COL. 3 COL. 4 COL. 5 COL. 6 Original Remaining Present Photon Photon Nuclides- Concentrations Fraction Concentrations Energy Contributions a.' KR 85M 0.86 .18
- b. KR 85 1.00 .0022
.c. KR 87 0.58 .79
- d. KR 88 0.78 2.2 e..XE 133M 0.987 .02
- f. XE-133 0.995 .03
- g. XE 135M 0.08 .53
- h. XE-135 0.93 .26
- 1. I 131 0.996 .39
'j. I 132 0.74 2.2
- k. I 133 0.97 .6
- 1. I 134 0.45 2.6
- c. I 135 0.90 1.5 Total Present Total Photon Concentration - Contribution -
Al A2 Ecuation C-1 WBOCF Calculation: WBOCF in MREM HR uct Date/ Time cc of decay corrected DCF
=( A2 + Al ) . 9ES - S3
1004.7 Revision 7 D. THYROID DOSE CONVERSION FACTOR (TDCF) DECAY CORRECTION
-- =----- - =- Utilizing the Attached Worksheet -- =----- - ==
- 1. Enter the Date/ Time of the original concentrations.
- 2. Enter the Original Concentrations in pct /cc for the listed nuclides in column 2.
- 3. Multiply the original nuclide concentrations (column 2) by the remaining fraction (column 3) to obtain the present concentration (column 4). Enter the Present Concentrations in column 4. Fac, tors in column 3 account for 1 hour decay.
- 4. Multiply the present concentration (column 4) by the isotope DCF (column 5) for the listed nuclides to o'otain the isotope contributions (column 6). Enter the Isotope Contributions in column 6.
- 5. Determine the total concentrations for the listed nuclides by adding items a.-e. of column 4. Enter the total concentration as item Al.
- 6. Determine the total isotope contribution for the listed nuclides by adding items a.-e.
of column 6. Enter the total isotope contribution as item A2.
- 7. Enter items Al and A2 in equation D-1.
- 8. Calculate the TDCF utilizing equation D-1. Enter the TDCF as item S4 on the worksheet.
- 9. The elapsed time between the original concentrations and calculated TDCF is 60 minutes. Determine the date/ time of the calculated TDCF by adding 60 minutes to the time of the original sample. Enter the date/ time on the worksheet.
1004.7 Revision 7'
~
---------------------------------------- WORKSHEET B ----------------------------------------
. THYROID DCF DECAY CORRECTION Original Concentration Date/ Time COL.'1 COL. 2 COL. 3 COL. 4 COL. 5 COL. 6 Original _ Remaining Present Isotope . Isotope Nuclides Concentrations Fraction Concentrations TDCF Contributions
- a. I 131 0.996 1.6E9
- b. I-132~ 0.74 7.9E7
- c. I-133 0.97 5.4E8
- d. I 134 0.45 4E7
- e. I 135 0.90 1.6E8 Total Total Isotope Concentration - Contribution -
Al A2
~ Equation D-1 TDCF Calculation:
TDCF in MREM HR uct cc
= (. A2 + Al ) -- S4 Date/ Time l
I
- - - . . . . _ . . . ~ , . ~ . - _ , , - - - - - . ~ . - , . - - , , , - . - , - - . - - - - . . . . - , --- - ,,. , - - - - - . . . - , - - - - . . , , - - .
1004.7 Revision 7 E-. DEFAULT DOSE CONVERSION FACTOR (DDCF) CALCULATION
Utilizing the Attached Worksheet -----------------------------
- 1. Enter the Date/ Time of reactor shutdown.
- 2. Enter the Date/ Time of the requested (DDCF) calculation.
I
- 3. Determine the time since reactor shutdown by substracting the time (Item 2) from the time (Item 1).
- 4. Select the proper accident classification from column 1. Circle this item.
- 5. Select the proper " time after Rx S/D" from column 2-10. Circle this item,
- 6. Enter the whole body dose conversion factor (WBDCF) and thyroid dose conversion factor (TDCF) as items S3 and 54 on the worksheet.
1004.7 Revision 7
---------------------------------------- WORKSHEET B -----------------------------------------
DDCF Calculation
- 1. Reactor shutdown Date Time
- 2. DDCF calculation Date Time
- 3. Time (hrs.) from Rx SD Hrs.
Column 1 Col 2 l. Col 3 l Col 4 l Col 5 l Col 6 l Col 7 l Col 8 l Col 9 l Col 10 Accident TIME FROM REACTOR SHUT DOWN IN HOURS 8 10 12 24 Classification 0 1 2 4 6 OTSG Tube Rupture WBDCF 2.1E5 1.8E5 1.5E5 1.0E5 8.0E4 6.4E4 5.5E4 4.8E4 3.3E4 TDCF 5.8E8 6.1E8 6.4E8 7.1E8 7.7E8 8.4E8 9.1E8 9.8E8 1.3E9 Fuel Handling (RB) WBDCF 2.7E4 2.7E4 2.7E4 2.7E4 2.7E4 2.7E4 2.7E4 2.7E4 2.7E4 TDCF 1.2E9 1.3E9 1.3E9 1.4E9 1.4E9 1.4E9 1.4E9 1.4E9 1.5E9 Fuel Handling (FHB) WBDCF 2.4E4 2.4E4 2.4E4 2.4E4 2.4E4 2.4E4 2.4E4 2.4E4 2.4E4 TDCF 1.6E9 1.6E9 1.6E9 1.6E9 1.6E9 1.6E9 1.6E9 1.6E9 1.6E9 Rod Ejection WBDCF 1.4E5 1.2E5 1.1E5 9.9E4 9.0E4 8.5E4 8.1E4 7.8E4 7.2E4 TDCF 1.2E9 1.2E9 1.3E9 1.3E9 1.4E9 1.4E9 1.4E9 1.4E9 1.5E9 Waste Gas WBDCF 5.5E4 4.9E4 4.5E4 3.9E4 3.5E4 3.2E4 3.1E4 2.9E4 2.7E4 TDCF 5.4E8 6.1E8 6.6E8 7.7E8 8.5E8 9.2E8 9.7E8 1.0E9 1.2E9 Others WBDCF 8.2E5 7.0E5 6.0E5 4.6E5 3.7E5 3.0E5 2.6ES 2.2E5 1.3E5 TDCF 3.8E8 4.6E8 5.3E8 6.2E8 6.9E8 7.4E8 7.9E8 8.3E8 1.0E9 WBDCF TDCF <
l- ATTACHMENT XI. HYDROGEN PURGE CALCULATION The RAC shall complete this attachment should the Emergency Director (ED) decide that a hydrogen purge of the reactor building (RB) is necessary in compliance with EPIP 1004.4 Item 3.1.2.b. The purpose of this procedure is to provide the (ED) with guidelines for the reactor building ventillation flowrate.
- 1. Date Time
- 2. Obtain and analyze a reactor building post-accident sample in accordance with EPIP 1004.31 Item 4.8. Determine the noble gas and radiolodine airborne concentrations in accordance with EPIP 1004-7 Attachment IX Item 2.1. List the noble gas airborne concentration (Item A1) and the radioiodine airborne concentration (Item A2) oelow.
Noble gas airborne concentration (AI) uCl ( CC) Radiolodine airborne concentration (A2) uCl ( CC)
- 3. Determine the dispersion factor (X/Q) at the exclusion area (EA) in accordance with Attachment II. List the (EA) dispersion factor (as Item A3) below.
Exclusion area (EA) dispersion factor (A3) sec ( 3) meter
- 4. Determine the whole body and thyroid dose conversion factors (DCF) in accordance with Attachment X. List the whole body DCF (HBDCF) as item A4 below. List the thyroid DCF (TDCF) as item A5 below.
Whole body DCF MREM (A4) HR uC1 cc Thyroid DCF MREM (AS) HR UCI CC
- 5. Calculate the (RB) ventillation flowrate that corresponds to 1000 MREM whole body dose rate as shown below. HR 2.2 x 10' + ( (A1) x (A3) x (A4) ) = (A6) CFM
- 6. Calculate the (RS) ventillation flowrate that corresponds to 5000 MREM thyroid dose committment as shown below. Hr 1.1 x 10' + ( (A2) x (A3) x (AS) ) - (A7) CFM
- 7. Compare calculated (RB) venttilation flowrates (ltems A6 and A7). Choose the most limiting of items A6 and A7. Explain to the ED that this flodrate would yield exclusion area dose rates consistent with EPIP 1004.4 criteria. Also, that continuation of the purge for one hour would yleId dose rates consistent with the EPA lower limit PAG's.
1004.7 Revision 7 ATTACHMENT XII THUMBRULES
----------------------------- Instructions for Utilizing Annex I ----------------------------
- 1. Identify the release pathway. Select an effluent monitor.
- 2. Select the appropriate thumbrule form Table I or 2.
- 3. Complete the worksheet by calculating the ratto of actual to assumed conditions for the listed parameters. As an example; RM-A5 is reading IE5 CPM (Enter in Col. B)
Condenser off-gas flowrate is 10 SCFM (Enter in Col. C) The dispersion factor at the exclusion are la IE-5 sec (Enter in Col. D) M3 Thumbrule No. I from Table 1 shold be utilized. The worksheet should completed as follows: Affected Ratio Ratio. Ratio Correction Uncorrected Corrected Monitor Col. B Col. C D Factor Dose Rate Dose Rate RM-A5 1E5 .1 10 .5 1E-5 .1 .1x 5x.1 .01 MR .01 x .005 - IE6 20 lE-4 = .005 HR SE-5 MR HR The correction factor is the product of the individual parameter ratios. The corrected dose rate is the product of the correction factor and the uncorrected dose rate.
- 4. Enter the corrected dose rate on the dose assessment worksheet (Attachment VI). Attach to the dose assessment worksheet a completed Worksheet D. Enter "thumbrule" in items S1, S2, S3, S4 and S5.
I
1004.7 Revision 7 ATTACHMENT XII - TABLE 1
~ Low Range RMS Thumbrules for Dose Protection
- G = Gaseous Channel RI - Radiolodine Channel Column A Column B Column C Column D Column E
. Monitor Reading Ventilation Dispersion Dose Rate Flowrate Factor MREM /HR CPM CFM sec/M3 WB
- 1. - RM-A5G IE6 20 1E-4 .01
- 2. RM-A8G IE6 1ES 1E-4 50
- 3. RM-A8RI IE4
- IE5 IE-4 100 **
- 4. RM-A9G IE6 SE4 IE-4 25
- 5. RM-A9RI IE4
- SE4 1E-4 50 ** *
- CPM MIN
** MREM Thyroid Dose Committment HR
----------------------------------- NORKSHEET D ------------------------------------
Affected Ratio Ratio Ratio Correction Uncorrected Corrected Monitor Col. B Col. C Col. D Factor Dose Rate Dose Rate l
1004.7 Revisicn 7 ATTACHMENT XII - TABLE 2 High Range RMS Thumbrules for Dose Protection G - Gaseous Channel-Column A Column B Column C Column 0 Column E Monitor Reading Ventilation Dispersion Dose Rate (CPM) Flowrate (CFM) Factor (sec) HREM/HR (WB) M3
- 1. RM-A5 High (G) 1E6 20 1E-4 25
'2. RM-G25 (G) 3E3
- 20 1E-4 1
- 3. RM-G26 and 27*** IE3 5.6E6 ** IE-4 5
- 4. RM-A8 High (G) 1E2 1ES IE-4 450
- 5. RM-A9 High (G) 1E3 IE4 IE-4 20
- 6. RM-G24 (G) 1E2
- IE4 IE-4 10
- MR EE
** lb Gi
*** Release via the condenser off-gas
----------------------------------- HORKSHEET D ------------------------------------
Affected Ratio Ratio Ratio Correction Uncorrected Corrected Monitor Col. B Col. C Col. D Factor Dose Rate Dose Rate i
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N ews Relea.... :.:.. s. .o . ..;' ..,;,. . gy:. ..- . p .1 '. ' ' Three Wille Islan[ ' Nuclear station Post Office Box 480 f (.< Nudew..'. ' : k,. Middletcwn. PA 17057 Public Information Serv. ices , , . 717 94s ats7 .. For Further information contact: John Fidler . Date: January 25', 1982 i For Release: Imediately ' if6-82N . STEAM GENERATOR REPAIRS FOR THI UNIT 1 . . Middletown. PA -- Officials of GPU Huclear Corporation said today that repairs e to steam generator tubes at Three Mile Island Nuclear Station's Unit I reactor prob-ably will result in at least a six month delay in the readiness of the reactor for
. restart. .
Recent testing of steam generator tubes indicates the repairs will be sub-stantially more extensive than initially anticipated, Company officials e.xplained. The Company had expected the unit to be ready for restart by the end of
..=.
[. s
. February, subject to permission from the"U.S. Nuclear Regulatory Comission.
Additional testing and evaluation will be required before the full extent of the steam generator problem is known. Company officials met with members of the NRC staff today to brief them on the information developed to date. 1 l . Company officiais said it was uncertain whether the timing for restart of .
, Unit 1 would be' controlled by the steam generator problem or by'a recent order of t
' the U.S. hurt of Appeals, if allowed to stand. The Court has directed the RRC to
- l make an environmental assessment of psychological stress that might result from Il return of linik 1 to operation. .
The Unit I reactor was shut down for refueling at the time of the March 28, 1979 accident that damaged TMI Unit 2. THI Unit 1 remained closed for modifications, t and for restart hearings before the NRC's Atomic Safety 1.icensing Board. Inspections and tests during the 21-years after the accident did not reveal 1,' any abnormal conditions Company officials said. Indeed, the hot functional tests is
~ '
January 25, 1982 .. '8*#
#6-82H
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r- . It was not until a . - - L- conducted in late sumer ,showed no evidence of the problem. repressurization of the system in November that small leaks were discovere ,,
'the tubes in both Unit 1 steam generators.
The tubes normally carry hot, pressurized radioactive water from the reactor. , This This, water causes non-radioactive water outside the tubes to turn to steam. steam turns the turbine, which in turn spins the generator to make ele . The tube leaks have resulted in minor additions to the routi radioactive releases from the plant. All releases have been well within federal . . environmental technical specifications. , , The two steam generators are about 70 feet high, and each contains appro mately 15,500 tubes, which are 52 feet long and five-eights of an inch in diam Their walls
.The tubes are made of inconel, an alloy of iron, chromium and nickel.
three hundredths of an inch thick. .'(,are e -t-In a pressurized water reactor, such as Unit 1, the steam generators where the plant's radioactive " primary system" and its non-radioactive " second - The steam generat' ors adjoin the reactor system" pass each other to exchange heat. Nonna11y, there is no radioactivity in the. in the Unit 1 containment building. .
" secondary system."
The Atomic Safety and Licensing Board has concluded extensive hearings the restart of TH1 *., and has advised the NRC that the management of GPU N ' Corporation, technical modifications'to the plant, and plans for emerge -
- ness are sufficient to assure that Unit 1 can be restarted without endanger .
health and safety of the public. A special inquiry into cheating on NRC operator examinations has a concluded, and a final report from the special hearing master,is expected f> m f
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t GENERAL INFORMATION B&W FACILITY REHGNN OT.1'd F'ast.upI Mer/4 N . t . r. . . c.
- 1. $lessroomet.ocation
....- ..n N PG.h The m will be 66.; at the B&W N). in Lynchburg, Virginia (see enclosed map
- 2. Motel Accomodations 7
', Two suggested places to stay are:
Harvey's e Highway 29 Business Lynchburg, VA Single Room Rate: $24.00 plus tax Phone: 804/239-2611 Sheraton Route 29 & Oddfellows Road -- Lynchburg, VA . e Ok/ -
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Please call them ~diVEWif 'you wish to make a reservation. g k 3. Transport _ation v . . ( Automobile transportation while in'at'tendance at the course will be l'" necessary to consnute to and from the Training Cegre
- 4. Lenqth of Cours_e_
The course will begin at 7:30 AM on Monday, Cecember 7,1981. See ' the attached Classroom Schedule for , full details. .
- 5. Course Manuals Course Manuals will be supplied by B&W at the Trqi,ning. Center.
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{ APR 711982 O ls/, : MEMORANDUM FOR: S. D. Ebneter, Chief EPB fIlip THRU: S. D. Ebneter, Acting 01 dor M & PS, EPB FROM: S. D. Reynolds, Jr., Reactor Engineering Inspector, M.& PS
SUBJECT:
Comments on Young to Keimig Memo dated 4/8/82 I have reviewed Young to Keimig Memorandum dated 4/8/82 and have the following comments: 1 Paragraph 419 4 (a) (8) which eliminates the portion of the tube within the tubesheet from the EC examination requirements is contrary to the 4.19 Objective paragraph. The Objective paragraph '". . .is to provide assurance of continued integrity of the tube n. Dortion...of the OTSG.... ,To insure integrity of the tube, the entire tube must be examined.
- 2. The portion of the tube that forms the boundary between the primary and secondary coolant systems is the entire tube.
I fully concur that the licensee's Technical Specifications are unacceptable,as currently written. This problem appears to be a generic problem. Besides t( II # Ifurous corrodant ID to OD pro-blems which can occur in the tube to tube hole area,
@ o +han. p\MS) there is general IGA in crevice areas on the OD of the tubes.
Failure of a tube within the tubesheet should not
- result in a large leak similar to Ginna, Prairie Island, h
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c -2 o Doel, or Surry because the leakage should be metered by the unrolled tube / tube hole annulus. Failure in
-=this restricted area should be easily d,etected by a change in the radioactivity of the secondary water.
The entire tube length should be examined by EC-procedures. Special calibration standards and coils should be ' utilized for those areas where differential coils do not result in a sensitive examination. GPU should not be immune to EC testing require-ments based on their ambiguously written Tech Spec. S 9 e e 9 L
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Subsect RWP WAIVER Service To MANAGER / DEPUTY MANAGER acatiert TMI Nuclear Station RADIOLOGICAL CONTROLS Middictown,'PA 17057 . o, $ It is requested that the person (s) indicated below have the requirement for General /1Wf)(Circle one) Rad Con training prior to entering the Con-trolled Area waived: Waiver Requested BY: F ycotX , WAIVER FOR Name: E @RA') Date of Entry: - h8Yf 7-Company: O SO(2.6. Reason for Entry:_ "TFOSPECTICA) 0 Nhb'$ hf. .. Name: Date of Entry:_ Company: Retraining Scheduled: Reason for Entry: The above personnel will be briefed regarding the radiological risks that may exist in areas they will visit. Requirements for protective clothing will be discussed if applicable. The personnel will be escorted by RWP trained person-nel at all times. - - - Training Provided By: 8 k8t7dM ., Escort (Name): E 6/M M / , l Waiver Expire - _ /IdM - Approved: ' Date: h
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cc: 2 control Points Unit I i Security Requestor Training Dept., % Daughn Silar File: Original . RP/R 2 1 5/18/81 GPU Servee Coroo at on s a s .r.1 e. : ' Metal Pubhc U. tas Corooration i
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News Release Three Mlle Island m e.,..d.- E"p"l Nuclear Middletown. PA 17057 ' 717 ses.ais7 ~ Public Information Services Fur Further Information Contece Doug Bedell l., For Meioase: Immediately - Date: February 10, 1982 *
. . #9-82N
_5TATL;S OF STEAM CENERATORS AT THREE MIt.E ISLAND UNIT 1 MICOLETOWN, PA - Three Mile Island Unit 1 is switching from preparations for restart. to a substar.tial job of diagnosing and repairing wd.ly discovered corrosion problems in the two Unit 1 steam generators Robert C.
- Mrnold, President of the GPU Nuclear Corporation, said today.
l(
, Oepending on the outcome of studies still underway, the repair process
.r',
w'ill take at least six months and could take longer.
- Results of two months of extensive tests of the steam generators have led to projections that on the order of 8,000 to 10,000 of the 31,000 heat-exchange tubes in the steam generators may recpire repair because of corrosion which has caused tube wall cracks. . Cracks can vary from a depth of a few thousandths of an inch to all the way through the well. Only 134 of the ,
tubes leaked during pressure tests and the projections of more extensive - ' e repairs are preliminary results from as yet incomplete examinations of the y tubes' by a process called "addy current testing".
.,z. .
;!.1 Virtually all of the corrosion appears to be located in the uppermost E portion of the steam generator tubes, in the 24-inch thick upper tube sheet
', region where the upper ends of the tightly spaced tubes are held in place.
There are 15,531 tubes in each steam generstor. Each tube is 56 feet long. l
==4
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Februarg'10,1
- - Mo. 9-43N The bulk of the corrosion appears to have occurred at or just belos a I .i region known as the roll transition zone. 1his is the point at which tho'
' j ', ! five eighths of an inch basic diameter of the ttees is expanded for the last inch to seal the tubes into the drilled holes in the tube s.W t. The tube,
)f vs11s are about three-hundredths of an inch thick.
Corrosion in steam generator tubes is a common probism in pressurized water reactors like the one at Three Mile Island Unit 1. The THI-1 protdem. however, is somewhat different in that:
- - - The number of affected tubes is unusually large and the protdas '
I became manifest in a relatively brief period. The first leaking tubes ware *
# discovered during low pressure testing of the Unit I reactor system in late ,
, November. Three months earlier, when the plant was brought to full T operational pressure and temperatura during hot functional testing of the -
systes, there were no tube leaks. In 3Jne,1960, addy current testing revealed no probless of the nature which now exist. .
- - - In instances of steen generator corrosion at other nuclear plar.
the corrosion has typically started on the outside - or secondary side - of the tubes and workad inward. At TMI-1, however, the corrosion has procese y from the inside - or primary side - of the tubes and worked outward, ; Iq., indicating that a corrosive agent or agents were in the primary reactor
,* l- . coolant water. -
.;f;
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na m sea of the apparent nature of the corrosion, officials of (FU
!' Nucleer'are extending the investigation to materials in the Unit i reactor
- t. * ,
- itself. Present plans are to remove the head of the reactor and at least partially defuel so that internal components of the reactor can be inspected.This will be done to determine if the corrosion that attacked the lk o
- more -
.. _ _ ,_ __ _ i , _ _ _ _Ti
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r3- February 10,1982 ' No. S 42N > alloy tubes in the steam generators also acted on surfaces of materials in the ! reactor. The steam generator tubes are made of inconel, an alloy of nickel, ; .
~s chromium and iron. The Unit I reactor includes w,-Js of inconal es well !
. 1
, . as stainless steel, which is metallurgically similar to inconel. ! ;
1 The cause of the corrosion has not yet been determined. The search for
,, the corrosive mechanism is a primary focus of shwitas being made by (FU - -
Sw 1==v Corporation and several teams of contractor consultants and laboratories.. Nineteen tubs sections have beer. removed from the steam
, -A generators for laboratory analyses. *:
The corrosion is described as a condition involving chemically assisted I.
- intergranular attack (ICA). The corrosion attacked the boundaries of grains ~
of the metal in the tthe walls. d (5 Three possible agents or conditions are preliminary suspects in the , ?.* .. , i
+
Unit I corrosion: *
?, ,
- - - Sulphur has been found in cracks in affected tubes. It is not .
s . clear whether the sulphur was the initiating cause of the cracking or became concentrated in cracks initiated by other processes. Sulphur is known to ,. s
*4- .
> - accelerate IGA cerrosion. .
,, - - - Chlorine is being investigated as a possible initiator, or -
- omplicating factor. , k
- - - stzes=== that. Ausult from assemoly of the steam generators, of '-
/ , *:
, , , , from normal plar,t operations may also have contributed to the problem. '
',' y ^ -
' I,
,,; sulphur and chlorine would only be in the reactor coolant water as an,
~M'.:
] impurity and chemistry control procedurus are directed toward minimizing their presence..
s ;$ i
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, - more ==
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~ ~ ~ ?
8
- q. q
. ,' =4- FM=7 Id, No. 9-83N As we do more investigation, we expect to narrow down the elements i caused the corrosion so that we can be confident of the extent of the prob'
! and keep it from recurring", Mr. Arnold said.
In pressurized water . reactor systems, the steam generators are where
~~
plant's radianctive " primary system" and 1.ts non-radioactive " secondary system" pass each other to exchange heat. Hot radioactive water from the' ; rw1 mar reactor passes through the steam generator tubes and causes water c i .- . - the s 4=y side of the tubes to flash to steam. That steam turns the plant's turbine-generator to generate electricity.. t Normally, there is no radioactive water on the h4=ry side of the. heat-exchange tubes. When the first Unit 1 tube leaks were discovered duri
/, low pressure testing of the reactor system in late November, there was some '
leakage of radioactive water from the primary side to the secondary side of
,.,the steam generators. Leakage to the s w 4=ry side has not continued sinc-s Unit I was returned shortly thereafter to a non-pressurized condition.
J The tube leaks were located by a process known as " bubble testing" it:
?
, which the ttbes were filled with water and subjected to nitrogen gas pressu ,
from the secondary side, so that bubbling could be spotted at the top of th i leaking tubes. p , As soon as equipment and engineers could be mobilized, an extensive O. ? round of cophisticated testing was started on all 31,000 tubes in the two' i,'. , steen generators. This process is known as " eddy current testing". Probes e insertad down each tube sense and signal back electrical characteristics of the tube. Fluctuations in the electrical characteristics must then be
'!g evolusted to determine if the tube is degraded or if the electrical signal '
varying for other reasons, i
= soZe = .~ .,
c - 3 '
~
February 10, 1982
't 1 ..
No. 9-83N *-
*a
, t t, .
The TMI-1 eddy current tests have been on-going since early December. .
, [ The initial phase started in early December,1981 and was completed 'in: ea
's February,1982. It consisted of at least partially inspecting all 31,062 3' .
t mes with the type of probe used during prior inspections. The nature of the defects is such that this awa=ination is a v=1mah1= indicator of where ! additional examination and analysis are required. The next phase is uM14fng.- J: e new addy current probe which was developed and fabricated during neremh*T i r
. and January. Although this device still does not provide all of the '
examination results necessary, it is much more r=14*1e about indicating the i !'* existence of defects in the specific region of the tubes where the corrosign' ' i* A appears to be concentrated.. This phase probably will not be complete until i l' sometime in April. - '. Testing is still not complete and the data collected to date have not been fully analyzed. [A In parallel with the additional. testing, the Company . i-will be doing the inspection of the internal components of the reactor vessel ' and plugging steam generator tubes that are known to leak. - Four possibilities for repair are being considered. Final h i=ians on -I
, \
'how to proceed must await completion of studies by the several task forces' '
.." ?
that have been organized by the GPU Nuclear Technical Functions Division. : i The possibilities for repair are: ,
'.. i
.. y
,, - - - Plugging tubes. Tubes that do not meet operational criteria would -
4' be plugged, and the. plant would restart, probably at less than rated power. ,
.. This process would'take approximately six months.
y-1
., - - - Plugging and sleeving tubes. If too many tubes need to be '
4 .
,, iplugged, a combination of plugging and inserting sleeves inside the defective ,,
) .., tubes might be used. This approach could take perhaps a year to complete. *
, 5- -
- more - ,
_ l
,m , . '4.*.
6- .._, _ February 10,1932
. No. 9-82N ,
- - - Replacing all the tubes. ,
e
- - - Replacing the steam generators.
o - The time required to complete either of the last'two approaches is not .- [.y,e
.i known yet, but other plants have required outages of about. one year for work -
. of similar scope. Depending on engineering re w irements and material .- l 9
symf1shility, these two approaches could require longer then one year for , l TMI-1. s . itistaver repair options are chosen, the repair work will be costly,,sith,. full tube or generator replacement being the most costly. . At this point, costs have not been determined. j A number of outside organizations are working on the steen generator, l
- - task forces that have been organized to analyze the steen generator corrosion l M problem. A wide array of technical and laboratory discipl.ines, Irw bmig ,
. chemistry ath metallurgy, are represented in the effort to find the osuus of Q, ;i the wsesion and timeiria on appropriate repairs. The: outside otgenizations .
l
. *. I include Babcock & Wilcox, of Lynchburg, VA; Satte11st Laboratories of Oslumbus, Chio; the Westinghouse Electric Corporation of Pittsburgh; and the FIsetric l l
. - Power Research Institute of Palo Alto, California. Consultants from Oak Ridge j *. .
i )
;. National Laboratory are also being utilized. . ,
- t. {
. '~
The Nuclear Regulatory Comission is being kept advised of the stamm* ; j
'i ' '
generator studies. l
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a v.coermott company c([ Deder M, M81 $$ 12$'t R GPU-81-153 Lynet. burg. Virgir.ia M35 (co4)3c&5111 Mr. D. G. Slear TMI-2 Project Engineering Manager GPU Service Corporation 100 Interpace Parkway Parsippany, NJ 07054
Subject:
GPUNC(TMI-1) Master Services Contract Effective Date June l',1977 Reference Nos. : B&W 582-7105, GPU M77120 Task 141 - B&W Engineering Assistance - Response to Task Force Questions of 12/6/81 Attentioni Mr. J. J. Colitz
DearM!.Slear:
On Sunday, December 6,1981 the. Task Force transmitted a series of twelve (12) questions to B&W 'for their attention and response. Although we have provided verbal response to those questions, this letter will document the. responses. _All questions asked except the question dealing # yith qualification of the tuce/tude sneet roi s are ouuresseo in tne - attachment. In connection with our previous verbal responses, we transmitted an engineering sketch which was labeled LDD-SK 12781 Rev. I and dated 12/7/81. It is imperative that GPU understand this was an engineering sketch and is not representative of manufacturing or as built dimentions; rather it contains typical dimentions used in our response. Very truly vours,
/
L. J. Stanek Engineering Product Manager LJS/mlw . cc: w/ attachments J. J. Colitz D. W. Dcmers J. G. Herbein H. D. Hukill R. A. Knief J. A. Mahn R. J. Toole
7 _ ; ,, ;- _ O {_ ' _ Engineering Input to Task Force Questions (December 6,1981)
- 1. Question:
How'small a section of tube can we remove from above the rolled section in the upper.tubesheet? ,
-Response: ,
7 There is no unrolled section above the roll expanded portion of the
, tube. .
- 2. 'Ouestion:
Can we cut the tube just below the rolled section and remove the rolled section intact? Response: - I . Present tooling and techniques will not accomplish this. The standard tube pull drills out the rolled section. A method for cutting the tube within the tubesheet has not been developed. However, tooling to perform this task can be developed. gg 9 ,
- 3. Questi'n:
o , Is there an unrolled section of tube above the rolled tube section but - below the top of the upper tubesheet? How much? This relates to Question 1.
Response
,. The tube'is rolled to the top face.- There is no unrolled portion above the rolled portion.
' 4. Question:
Can the tubes be rerolled (expanded) at the rolled section or just below the present rolled section?
Response
B&W does not recommend cold working Inconel unless a subsequent stress relief is nerformed, Cold worked Inconel is more susceptible to stress
- corrosion cracking. During manufacturer of the OTSG a stress relief is performed on the OTSG assembly after the tubes have been roll expanded.
- 5. Question:
We are planning to do a bubble test and vary primary level (possible) . or resposition a stopper in a leaking tube to determine when bubbles stop (level of crack, defect). These plugs must be made. Send sketches
. of what Crystal River used ASAP and information on what CR did to .
determine location of leak with stoppers. , v nh.nye- w:-- -mg,- ,- * - - - * "
, ----,-------m we e w.
- -- e ,_.m__ _ . . - . . _ . _, _ _
____.m. . p
. . . Response:
B&W did not develop any stoppers which were used to determine leak elevation. The leaks at CR-3 were BPP.A damaged related and were found at the top of the tube ends in the weld area. 6.' Ouestion: .
-During(manufacture
-loads tension) or.is there of the OTSG's a pattern to how they are are all stressed?
tubes installed What is under the it?.
-Response:
Manufacturing procedures have been established which places tension
.in the tube at the time they are rolled. This ensures that when the tube is welded it is not in a compressed condition.
~7. Question:
What are the design tube stresses expected when the generator is hot (operation and cold? Is there a pattern dependent on location (center / periphery))within.the generator? *
Response
(See Attachment 1).
- 8. Question: 3 .
Tube manufacturing heat numbers. Is there any relationship between
- heats and heat tube location that would relate to periphery vs center tubes?
Response
At.this time there is no evidence to support that there is a relationship _ between heats periphe,ry vs center of tube bundle. The manufacturing process results in the tubes (see attached sketch LDD-12781 Rev. O attachment 2) being installed in a layered fashion not an inside/outside pattern. 9 .' Ouestion: How are the tubes installed.in the OTSG during manufacture i.e. OTSG l on its side, orientation of axis (i.e. W-Y vertical Z-X herizontal) the ' objective is to correlate tube heat numbers with location in the generator to determine if heat numbers (tubes from specific heats) can ! be correlated to the leaking areas. Do we know pattern and heat numbers? I
Response
See response t'o qu stion'#8.
- aye q-.e wmmm e.c . re ma ~..a . ~..-..as.. . - , . . .------.-a~. .- .-
s_ --~ . ___ _ _ . . v .
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- 10. Question:
' Crystal River - was there a pattern to the location of leakers at Crystal River - possibly not due to damage from loose parts.
Response
All of the tubes found leaking at Crystal River had been damaged by the loose parts and the leaks were in the area of ,the damaged welds.
- 11. 'Ouestion:
To aid in designing a stopper / plug that can be inserted in a tube for bubble testing provide inside tube dimensionsvs length of tube. (Especially in rolled section).
Response
The typical tube is a 0.625 "[ $ 0.D. x 0.034" min. wall.
- 4 M I
(.. -
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Calculated Tube End' Loading
}
~
Center of: , Outermost . 1 Tubesheet Tu'oe . 9 -(Load,lbs) (Load,lbs) , t O I neat-up.
-670 -773 i
*i;
- j. .
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.- 649 , 1107 ll 1
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-100 100-8% power- ,
-216 t
i 1 l b 3
. T 1 jf Tube end load changes linear -across tubesheet radius i : i ,
Reference:
5 I Topical BAW-1588, Determination of Minimum Required Tube Wall Thickness for 177-FA Once Through Steam Generators ..; - i, s l . . .., (
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* . SCHEDULED REGION I SPEAKING ENGAGEENTS ;
.AS OF; JANUARY 14,"1982 DATE ORGANIZATION LOCATION SU8 JECT SPEAKER-5/16-19/82 Conference on Welding .Gatlinburg, TN " Effects of Joint 'S. Reynolds
- Technology for Energy Design and Application Fabrication on Tube to Tube Sheet Welding" 2/3/82 John F. Kennedy Medical Edison, NJ NRC Inspection ,L.~ Friedman Center, School of Nuclear Process T-Medicine
)
,. ji _
, j lPRIORITYATTENTIONREQUIRED DAILY: REPORT'- REGION'I PRIORITY ATTENTION, REQUIRED 11/14/82.
TO: Leonard I. Cobb, Director, TAS FROM: Ronald C.;Haynes, Region I Fccility - Notification Item or Event - Regional Action: DIVISION OF RESIDENT AND PROJECT INSPECTI
'N N Three Mile 1/14 RRI . Fax - Update ~on Once Through Steam Generator (OTSG) Per MC 2515.
Island Unit 1-
. Tube Degradation. Eddy Current Testing is con-tinuing in both OTSG's, with about 22,000 of. .
.Y (DN 50-289) 31,000 tubes tested. Preliminary evaluation in ' 9 dicated a failure rate of 11 percent. .The. initial W7 Tube Removal Program (removal of 4 tubes in the -
"B" OTSG) has been completed. Two tubes were a shipped to B&W in Lynchburg, VA and two to Battelle_
in Columbus, Ohio. Based on initial Battelle
- [gv s e o analysis, the failure mechanism may be inter .
granular attack from the inner face of the tubes. Additional tube selection for removal for. analysis is in progress; the first additional tube is scheduled to be pulled on 1/15. The licensee has not detemined the. impact of the tube ion on their restart schedu R. E. Ginna 1/14 SRI Fax About 11 a.m., 1/13,'an' auxiliary operator noted Per MC 2515. (DN 50-244) a 3 drop / min pinhole leak from a weld in the 6 inch RHR suction piping from Containment Sump ')
"B " . Technical Specification relief was obtained to allow both the "A" RHR pump-and the "A" Reactor Coolant Drain Tank punp to be out of service simultaneously. The weld was repaired,and a hydro.was completed at 2 a.m., 1/14.
D
. - - - - . - - - - - - - - -- - - - - - - - . - ----.-----_.---_-a
.. e ,
g v y i . - DAILY-REPORT : REGION I 1/14/82 Facility Notification Item or Event ~ Regional Action DIVISION OF RESIDENT AND PROJECT INSPECTION (contd)' Salem 1/14 SRI Fax At 1:05 a.m.,' 1/14, while at 97 percent power, Per MC 2515. A PN Unit 1 turbine load reduction was initiated in response will be issued. (DN50-272)~ to a steam generator = feed pump low suction pres-sure alarm. After initial rod motion, " control
- rod urgent failure"'was annunciated and the #
-rods could not be moved. Boration was initiated '
to reduce.Tavg. High pressure in the steam generator caused safety valve 23MS15 to open. It failed to reseat-due to fouling of the manual operating arm. Primary pressure reached 2320 psig and was reduced by spray. Spray valve 2PS2 failed to reseat, reducing pressure to 2140 psig before manual pressure control was effected. At 2:10 a.m., the plant was stable at 46 percent _ power. The safety valve was reseated at 5:21 a.m. Rod con-trol was restored at 7:30 a.m. by replacing the firing circuit anda failed fuse. During'the transient, Tavg exceeded the 2-hour LCO limit.of 582_ degrees F for five. minutes, peaking at 592 degrees F. The plant is limited to less than 50 percent power for 24 hours due to accumulated axial flux difference (AFD) penalty minutes. Secondary parameters were )_ recorded and will be evaluated today to determine ' the cause of the initiating loss of feedwater _ suction pressure. No ENS call was made. Indian Point 1/14 RRI Phone PASNY President George Berry will retire next Information Item. Unit 3 month after more than 20 years of service. (DN 50-286) a
>y .y .
- i. * - M
, , ;1 L
( m-
- REGION I- '
1/14/82.
~
STATUS OF UNUSUAL INSPECTIONS OR INVESTIGATIONS No change from 1/7/82.- 2. 4 t
'3:
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- g . ,
NUREG/CR-2305. Vol.1
- ORNL/TM-7954 Distribution Category R5 e
Contract No. W-7405-eng-26 METALS AND CERAMICS DIVISION EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31, 1981 C. V. Dodd, W. E. Deeds, and R. W. McClung Manuscript Completed -- September 18, 1981 Date Published - October 1981 Notice: This document contains information of a preliminary nature. It is subject to revision or correction and therefore does not represent a final report. 6 Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, DC 20555 Under Interagency Agreement DOE 40-551-75 NRC FIN No. B0417 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 operated by UNION CARBIDE CORPORATION for the DEPARTMENT OF ENERGY
CONTENTS
SUMMARY
INTRODUCTION
............................... 1
....... ........ ............. 1 BACKGROUND OF THE ORNL PROGRAM FOR IMPROVED INSPECTION ....... 1 PROGRESS REPORT DURING QUARTER ENDING MARCH 31, 1981 ........ 2 0
4 iii
- ( 9 n ^
EDDY-CURRENT IN3PECTION FOR STEAM GENERATOR TU31NG PROGRAM yumyggty FEOsmE33 REPORT FOR PERIOD ENDING MARCH 31. 1981
'r
, W. E. Deeds, and * % y SUt9tARY Eddy-current methods provide the best in-service inspec-tion of steam generator tubing, but these techniques can pro-duce ambiguity because of the many independent variables that affect the signals. The current development program has used mathematical models and has developed or modified computer programs to design optinua probes, instrumentation, and tech-niques for multifrequency, multiproperty examinations.
Interactive calculations and experimental measurements have been ande with the use of me,dular eddy-current instrumentation and a minicomputer. These establish the coefficients for the complex equations that define the values of the desired proper-ties (and the attainable accuracy) despite changes in other significant variables. The computer programs for calculating the' accuracy with which various properties can be measured indicate that the tubing well thickness and the defect size can
,. be measured much more accurately than is currently required, even when other properties vary. Our experimental measurements have confirmed these results. We have made measurements at
.. scheduled field inspections of the StudiesIWt9detais4Seesis. steam generators and have obtained excellent data. We are continuing
.to improve the equipment and the data processing systems.
I INTRODUCTION This program was established to develop improved eddy-current tech-niques and equipment for the in-service inspection of steam generator i tubing. Our goal'is to separate the effects of variables such as denting, probe wobble, tubesheets, tube supports, and conductivity variations fros
- defect sise, depth, and wall thickness variations. Computer design of i
probes, instrumentation, and techniques is emphasised. 3ACKGROUND OF THE ORNL PROGRAM FOR IMPROVED INSPECTION 4 l l The ORNL program to develop improved eddy-current in-service inspec-tion for light-water-reactor steam generator tubing consists of design calculations based on theoretical models, construction of optimum r l ) 1 I .
4y. - 4 2 equipeast, laboratory tests of the best design, and field tests of the [' equipment. Using esdels established for eddy-currest coils in multiple cylindrical conductors, we calculated the electrical signals produced in ' the instrument for dif ferent frequencies, probe designs, and instrument ' l designs for enny test property variations. These variations span the range of these espected in the actual tests. A least squares fit of the test properties to the instrument readings and nonlinear functions of the instrument readings was then carried out. We repeated these calculations l s number of times with different coil and instrument parameters until an ! adequate system was obtained. . We assembled a prototype instrument from modular plug-in components. ! A probe was constructed, and the instrument was adjusted to confore to the design calculations described above. The instrument was connected to the f l parallel t rto,of the'MASBie ( M , ,
, j ryL We ende readings on_ tu
_ Ing test samples. . talthofesar.n ' r the range of anticipated test property variations for the Ginna and point Beach tests. i (Larger denting ranges and other property j variations will be run later.) We then did a least squares fit for all l the coefficients directly from the experimental data. , i Once the optinue coefficients had been determined the process was i reversed. Using the new coefficients, the etnicomputer continuously took ! readings, calculated the properties directly, and displayed the results on ! a CRT terminal in real time. -The calculated properties changed in the ! proper emaner as the probe was scanned past defects, tubesheets, tube j supports, and thin well regions. After the instrument successfully passed j these tests, its onboard eterocomputer was programmed to calculate the { properties in place of the ModCoop IV, and the instrument was retested. - Maskipe, g instgueseh,was7%ested te phe,. ! aeoten!.ppesstieg eggAA&ess. Changes have been ande in the progressing to improve the i accuracy of the tests, the ease of calibration, and the use of the
- instrument. The instrument contains an internal passive calibration cir- !
cuit that is tested against a set of reference standards. We will write operating instructions and testing procedures. [ FROGRESS REPORT DURING QUARTER ENDING MARCH 31, 1981. i We are continuing our task to improve the inspection of stese generator tubing, emphasising intergranular attack and contaminant buildup in the tubesheet region. We now record the raw eddy-current data rather than the calculated , tubing properties directly on digital essnetic tape. This enables us to process the data later in unanticipated ways if unusual conditions are l encountered and prevents our losing the raw data during the very limited time allowed for the on-site inspections. The calculated properties are still recorded on the strip chart as'the tube is scanned. The fit between calculated and actual properties is somewhat.better it we treat the tube differently in different regions of its length. The ' three essentially dif ferent types of region are shown in Fig.1. Spressee wedeutesteertong " Z 1hea.inseramenteroepense(wheat the atubeJa#ea$ sessemasse*bymembeseeppertheftwhengitdemiste&ysins,ide tgt,ug,spogt
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, FREE TUSE REGION lI i INTERFACE .
REGION N u s I
/
k s TUSESHEET
; on /
SUPPORT REGION Fig. 1. Types of inspection regions.
-=7 However, J.e41SR,.but.cmanotabenibe experimental measurements can psahfsJyt,ha,fqin,'thi be made as well .pesies.
interface region as in the others. Therefore, we have arranged the programs so that the com-puter can determine if the probe is in the free tube, the tubesheet, or an interface between the tube and tubesheet and then pick the proper set of coefficients to give the best fit. Although all property calculations are improved by this technique, tho' improvement is necessary only for the defect calculations. This concept is beleg tested on the ModComp and, if successful, will be programmed for the microcomputer. We completed the preparation of our equipment for field testing at a reactor site. A pickup truck was fitted as a mobile test laboratory with an enclosed compartment that is air-conditioned to protect the equipment. 8 Wiring cables permit the test laboratory to be located up to 120 m (400 ft) from the eddy-current instrument-in the reactor containment
, building (Fig. 2). Because the equipment inside the containment building i may become contaminated, a sealable box and a compartment to hold it have been constructed to permit its safe transportation back to ORNL.
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4 08tNI.-D*G 80-1247R INSIDE CONTAINMENT ( . STEAM GENERATOR H11 : 11il llII llIll PROSE 'p a POSITIONER CONTROL AND DATA sus $, ' PR E C DATA STATION k EDDY-CURRENT O @ "I INSTRUMENT Fig. 2. Mobile remotely controlled addy-current inspection system. Eddy-current field inspections have been made on the steam generators at. Point Beach and-Robert E. Ginna reactors for istergemaniar/4ttaokojint . ~ Some pre-tausemiigdesid, which was 'a' first field test for 'our equipment. lie arranged with pers liminary results are shown in Fig. 3. Although addi-plants to internesh'our inspection with their inspections.be ande in the calibration, da
.tional improvements can Our inspection at data reduction, the reactor schedule was paramount. The Ginna was on the critical path and RdiMI~a1EisT*6"h*TbM?desstium.'
Point Beach inspection was outside the critical path and required no addi-tional down time. Excellent data were obtained at both sites. Approximately 100 tubes 1passemesaned up3torchetfiest.asup istlhath; plants. The system was 307Antfl00715EIaSt371*MEs. An offset
~ operated at Ginna at frequencie .o l )~
.of all the data [(the wall thickness gave a value of 1.60'em This offset (63 mi wass rather than 1.27 mm (50 mils)) was observed at Ginna.
caused by temperature and cable dif ferences between our laboratory and Ginna. Otherwise, the data were' good, and's vector correction of the raw
. angnitudes and phasesThis made with thetechnique correction minicomputer restored the into was programmed readings the to their correct values.
-minicomputer for the Point Beach tests. The(frequenetesj.wereychangsdyo
. 26.ikHz to reduce the potential for offset errors.' 'These two I
c: * - , , -
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.).
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POINT BEACH SAMPLE PNL SAMPLE GINNA SAMPLE 38 pm (1.5 mils) 140 pm (5.5 mils) 25 pm (1.0 mils) Fig. 3. Preliminary results of intergranular attack measurements. )
- _ - - - _ - _ _ _ _ A
4 t . 6 changes eliminated the offset problem. he results from both Cinna and Point Beach show=d = 1 4 = 6* y t " - ' ; i- --- 44that-h ? m :75% which could bii attributed to intergranular attack, but the wall thinning
- was less than that observed in samples removed earlier from the steam generator.
The wall thickness of the Point Beach tubes appeared to be several mils thicker above the tubesheet. Af ter a study of the ef fects of simu-laced sludge of different nonconducting ferromagnetic materials, we specu-lated that the difference might result from a small layer of a good conductor such as copper that had plated out on the tubes. %== nyce og}y;spfMjes'lgf2thechigh-conductivityJeetehmould appear 411kes Walservdooonel. Af ter we had reported our findings, Westinghouse acknowledged that they had also seen these " copper signals" in their inspections and indicated that the signals were present in several dif-ferent generators, with Point Beach being one of the worst. Readings were excellent when made on standards immediately af ter the probe was removed f rom the steam generator at Point Beach. The tube scans at Ginna were
, much cleaner and showed little if any thickness change as the tube entered the tubesheet.
Figure 4 shows the raw readings of magnitudes and phases at the three test frequencies and the corresponding calculated properties measured near the tubesheet region of the steam generator at Point Beach, and Fig. 5 shows the corresponding quantities from the Ginna steam generator. 'pyp; i41TFtWickadesh'*vesd'ingswegreatenabove, .theatubesheet onnthe) Point: Beach 4uhing741ndi'estingWbui: 1;d up ?ofis'6medondticring5m'a liairia'l. The Ginna' wall' thickness is much more uniform. . Figures 6 and 7 show calibration checks on the wall thickness and radial clearance measurements, respectively, on machined standards. The
" measured" curves are those calculated by the eddy-current instrument, and the " actual" values are micrometer measurements.
To test the effect of copper buildup on the outside of the tube,
"""3^^ CT45'Jggadggggg1ppatod1en;5thgpu3sgeggfgtherwiset;nor f ps. Figure 8"sYows the raw readings for a sample with copper plated for a short distance on either side of a simulated tubesheet 75 en
-(3 in.) thick, and Fig. 9 shows the corresponding calculated properties.
The radial clearance channel remains essentially constant, as it should; the tubesheet channel clearly shows the presence of the tubesheet; and the thickness channel rejects the tubesheet (as it should) and indicates the presence of more material where the copper is plated. To test for possible buildup of magnetite (Fe30 4) on the outside of the tubes, tests were made with a powdered iron ring 19 en (0.75 in.) wide surrounding a tube. Figures 10 and 11 show the raw readings and calcu-lated properties, respectively, measured on the sample. The radial clearance channel correctly shows no ef fect; the wall thickness channel shows a slight decrease, probably caused by overcompensation for a sup-posed tubesheet; and a large indication occurs on the flaw channel. The least squares fit must be run with the magnetite around the tube. This will allow us to eliminate this signal from the thickness and defect chan- i nel and to measure the buildup on a magnetite channel if desired. Tests have also been made on dented tubing. Figure 12 shows the
- calculated properties for a flat-bottomed dent 0.19 mm (7.5 mils) deep.
The radial clearance channel clearly shows the correct amount of decrease,
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20 mms masa OM*0 3MAIO t i t 1 t t I t t t I t t t t t t I m# - . . i . . . . . . . . . . . ut so **a '"*'uss_ _ _ , _ im ao _ c aculaTE0 ntaones - am 30 - nacia cLEananct -4
,,,,,, i t i i , t i f f , , , , , i , , i I j ltEGION AgowE TugEspeET --
l TugESMEETRE8soes-= Fig. 4. Raw readings and calculated properties near tubesheet of Point Beach steam generator. samL. ems es.coseen i i I I I 5 I I I i i i 6 5 1 4 1 7 500 hMs Peta ,
'l
_ SCO eMr taas _ j
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~
- 20 kMr PMA -
.0 .M i
i I f f f I t I t
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- SW M: , , , , , , , , , , , , , , , ,
F_Las cMaassEL t.M 70 ' - - -- CALCULATED REAQs8eGS
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imeOf s am 30 - naal cttanamet - t t I t t t t i f f 1 t 1 f i f f I
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Fig. 5. Raw readings and calculated properties near tubesheet of Robert E. Ginna steam generator.
. . ,s 8
ORNL-DWG 84-40043R I I l - 4.3 50 4.2 45 - g - 4.4 _ 40 - N I 4.0 - 35 - 0.9 l l l - O.8 MEASURED - 1.232 1.029 0.838 1.029 0.864 ACTUAL 1.283 1.074 0.899 1.074 0.894 WALL THICKNESS (mm) Fig. 6. Calibration check on wall thickness measurements. OR NL- DWG 84 - 40014 R 50 I I I I I - i2 40 - g,o 2 V E 3 30 - t_ _
.8 E 0.6 20 l I I I O.4 MEASURED 0.691 0.732 0.711 0.869 1.074 ACTUAL 0.686 0.709 0.699 0.894 1.069 RADIAL- CLEA'RANCE (mm) o Fig. 7. Calibration checks on radial clearance measurements.
1 l
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9 a o m o e.-rore I I I 'l 1.
. 500 nHe PHA -
^
~
500 kHz MAG
=#emmm
_ 0.75 m. _ 400 kHf PHA _ 400 kHz MAG
~
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'with copper plated on outside of tube.
's ORM.-DwG 81-7%R NOMINAL 1.27mm PLATED TUBESHEET PLATED (0.050 in. TUBE) COPPER REGION COPPER I I I I I i i i i THICKNESS I*USI I**I
-CHANNEL -
50 1.27 40 1.02 RADIAL - 30 0.76
" ~
CLEARANCE O 20 0.51 v
.TUBESHEET -
- 'Di" 9 '
5' CHANNEL - f I l l l i I i l l
. . ... Fig. 9. ' Calculated properties for tube with copper plated on outside of tube on either. side of simulated tubesheet.
~/; .. . - . . , . .-. ...- .-.--- --. -- -. - . - - - - - . - = - - --
y -- na -: , J l ;~
.. ,m 10 CRNL-DwG 89-7076 I I I -l i I i i I
, e
' 500 kHz PHA -
~ ~
^
500 kHz MAG - 100 kHz PHA - 100 kHz MAJ _
~ ~
20 kHz PHA .
-EJkHtMAG '
RAirREADINGS WITH COPPER PRESENT Fig.-10. Raw readings at three frequencies with a powdered iron ring around outside of tube. OR.4L-DwG 09-7073R , I I l l l Ht9mm }+ - FLAW O.75 in. CHANNEL - t 0.25 mm (10 miis) _ _ l (mits) (mm) WALL - - 50 1.27 THICKNESS i-
.- 40 1.02 RADIAL 30 0.76 CLEARANCE I I- 1 l l 3 o, POWDERED IRON RING Fig. 11.. Calculated properties for tube surrounded by a powdered
- iron ring 19 umi (0.75 in.) wide.
v-
,... . .n :m 11 ORNL-DWG St-7077R e:
g g ; (mils) Imm) l MELL:
--_y _- qg 50 1.27 THICKNESS w - -
40 1.o;' RADIAL 30 n7f CLEARANCE 20 0.50 ROUND E y DENT V O.23 mm i - (9 mils) 0.43 mm (17 mils) Fig. 12. Calculated properties for dented tubing. N y and the well thickness channel shows nearly constant thickness, except at
, Li " the edges of the dont (where the radial thickness really is not constant).
Figure 13 shows similar esasurements on rounded dents. The shallower dent, 0.23 mm (9 mils) deep, produced satisfactory responses, but the deeper. dent. 0.43 mm (17 mils) deep, was outside the correctly compensated range and gave misleading signals. This shows the importance of including the full range of expected property deviations when calibrating the
~
instrument. Future studies will include the full range of dents, which should correct this problem. In summary, this type of multiple frequency inspection could show
' buildup of sludge, deposition of products on the tube walls, magnetite formation, and'other detrimental conditions so that corrective action can be taken before they start producing wastage, cracks, and other defects that are detrimental to the safety of the reactor.
. Personnel from Westinghouse and Zetec, as well as the reactor opera-tion personnel, were impressed with the system and expressed the desire to incorporate parts of it into their equipment. We will release these designs as they are completed.
s
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12 ORNL-Owe 31-707,g I l l (mik) (mm) ,
~
50 1.27
^
~
40 1.02
~
RADIAL - 30 0.7s CLEARANCE
, l l FLAT O.49 mm ,
DENT (7.5 mils) Fig. 13. Calculated properties for dented tubing. O e
- y. - .. . - .-- - - - - - - - - - - - . - - - - - - - - - - - - - - - - - - -
t ,e
- s. s f D 13 L
P NUREG/CR-2305, Vol. 1 OWL /TM-7954 Distribution ! Category R5 0 l INTERNAL DISTRIBUTION ' 1-2. Central Research Library 20. W. Fulkerson
- 3. - Document Reference section 21-23. M. R. Mill 4-5. laboratory Records Department 24. F. J. Homan ;
- 6. Laboratory Records, ORNL RC 25. R. C. Kryter
- 7. ORNL Patent Section 26. A. L. 14t t s
- 8. Nuclear Safety Information Center 27-31. R. W. McClung b13. W. E. Deeds 32. G. M. Slaughter 14-18. C. V. Dodd 33. D. E. Whitesides l
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'N 1983 NORTHEAST REGIONAL MEET!']!S IN PITTSBURG PITTSBURGH SECTION
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December 20, 1981 4 h_ n . n i r ;
%YMEETINGNOTICE
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DATE: dThursday, January 14, 1982 TPusTea PLACE: Skibo Hall - Carnegie-Mellon University B. P. rA #FARDI caiu n corporation TIME: 8:00 PM P. O. Box 1346 Pnt g g, rghf,ensyNanla 15230 SPEAXER: G.P. Airey Westinghouse Research and Development CHAIRMAN
- o. R. HALL PROGRAM: I Intergranular Corrosion in Inconel 600 sauere:..n cement company too osmma ortve N Steam Generator Tubing (4T[7s'fNr"" '" ' SOCIAL HOUR: 5:45 PM, Skibo Hall vics CHAIRMAN DINNER: 6:45 PM, Members and Non-Members - $10.00 w.o.OaRsTACxER Students' $5,00
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* "'ti. 'Re E x: ELECTION OF 0FFICERS WILL DE HELD AT THIS MEETING.
crucible neiearen center it s'b gh Pennsylvania 18230 (4ta 923 2965 ! I will attend the January NACE Mcoting. Please reserve MGMetRSHIP CHAIRMAN dinners. J.R.PoeTER Dravo Litre Company l 2500 Center City Towere l 660 $mithfleid Street Pittoburgh, Penneytvania 15222 l 14146#64419 Signature l l I
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[ SHEET.1,0F.L REACTOR VESSEL HEAD REMOVAL DESCRIPTION - RESP. PROCEDURE 12 24 , 36, ,48, NO ,72 ,84,, ,96, ,108, , REMOVE ulSSILE SillELDS MM (75F#4); REMOVE FTC CATWALK-2 MM (F4 R AISE FESTOON CABLE MM @
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INSTALL ALIGNMENT STUOS MM Os0 llNCOUPLE CRDM'S . MM 1504-12 CMidW4 DISC HEAD WATER LINES MM 1504-2 03 REMOVE IIEAD SERVICE LINE PLATFORMS MM
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SHEET LOF.ft. HEAD REMOVAL' SEQUENCE l HYDROGEN af0NITOR , ,,; (CAT ALYTIC) AUX /FH VENTILA-TION OPERABLE l PRELitIONARY (CHARC0AL FILTER RV HEAD REtA0mL - DIANGE OUT/ TESTING PREPARATIONS (48AINTENA80CE) VALVE INSP. EST M H FUEL l DH-V-8 WORK FUEL INSP-INSIDE HANDLING (PARTS CHEW) CONTAIN40ENT CRITMA TRAINif00 (VES/MO) IDENTIFY RX VESSEL HEAD FUEL SHUFFLE ISl TESTING R h WORK R 1
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(VES/NO) (t4AINTENANCE) 1505-2 (5 DAYSI REACTOR BLALDING CLEA80UP/ LASEL EQUSPteENT INCORE WORK SPENT FUEL HANDLING WORK f 3pfTERM
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SHEET.2.0F Jt. REACTOR' BUILDING CLIANING SCHEDULE DESCRIPTION RESP. FEBRUARY MARCH _ ss7 e 9 to it er is 14 is ta 17 se to 20 at 22 23 34 2a m 27 2e i e34 s s 7 e e s01t It 13 e4 is is 17 se se to T LABEL Etsb ?ENT/ MATERIAL g q RC, g COM*L ETE DUCT WORK CLEANING FUHRER gggggggpgg (X) OPERATING F100R CLEAN ONCLUDES HEAD STAPO, lHOE X FIXTURE, TRIP 00) ##MI U*U U HYOROGEN MONITOR / CLEAN POLAR CRANE / ggg CRANE RAIL / REMOVE SCAFFOLD
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FUEL TRANSFER CANAL CLEAN FUHRER gg jgqg r ()() TOP OF ELEVATOR FUHRER CCMfL1.Tl { CLEAN FH EQUtPMENT/ FUEL TRANSFER MECilANISM STATION OPERATIONS CCWLETE CLEAN TOP of D-RING / MISSILE SillELD UTluTY (MN) CCWU Tl REMOVE llEAD CASLE PROTECTION PLATF09M UT/MM (pys'wws'y4) dMIU TI CLE AN REACTOR VESSEL HEAD SUPER SC S_TRUCTURE (X) Ctf AN REACTOR VESSEL HEAD AREA a TRACK AREA UTIUTY Opwwwwww4) CLEAN INCORE AREA IC , (X)
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TIAL EQUIPMENT CATALYTIC ggggggggggggggggggyyg II i STUD ( AllONMENT) CLEAN MM (MGT) NOIEt SCstEDULE COVERS WORIC FROIA FTC AND ABOVE. REIAAINIMS WORK TO BE COVERED SEPAR ATELY. 1 4
SHEET.3.0F.f._ REACTOR VESSEL HEAD" REMOVAL PRE-PREP. DESCRIPTION RESP. FEBRUARY MARCH
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OPERA 8tE lN I'dCT HOSES ON MAIN / SPENT FH BRIDGES OPERATIONS l%~d' C(Mtl:TI: TRANSFER SYSTEM MOTOR CHECK OPERATIONS (p!lzp() C(Mtl:TI: REMOVE OLD SLINGS MECHANICAL 0 POLAR CRANE PHONE COMMUNICATIONS I&C MrsFseggad)(OnIPI El E FH EQUIPMENT ELECTRICAL PM'S , j ()WW4) s SPENT FUEL HANDLING BRIDGE (FH BLDO.) MM (REMOVE EQUIPMENT FROM BRIDGE) (h4)t'OdIPI'E ) REMOVE FTC SEAL PLATE FROM MM . O MISSILE SHIELD
SHEET.i 0F.A. REACTOR . VESSEL i-LEAD REMOVAL DESCRIPTION . RESP. PROCEDURE ,12,24:3.,4.,.0,T2,.4,,e,iOe%O)s2 i44,ise,ise,ieO,ise 204,2 s,22e, REMOVE MISSILE SillELDS MM WRITTEN (gg) REMOVE F TC CATWALI(-l (NORTH) MM N/A (O REMOVE RUST FROM R.V. FLANGE MM 1504-4 GO SUILD TEMPORARY SEAL PLATE UTILITY N/A OhW) DIS /RE E *Pl* CA8LES AND CRDM IC . 8504-2 OutO
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RACK AD ARE IC/EL/UT N/A REMOVE R.V. INSULATION MM 1504-3 OO CLE AN AREA AROUNO INSULATION UT N/A , OnO R.V. STUO-OIL OF WINTERGREEN MM 1504-5 On() REMOVE 69 CLOSURES / INSERTS MM 1504-12 (y UNCOUPLE ASPR1 MM 1504-13 bQ DISCONNECT *A* TO*8" CABLES (ASPR) ELECTRICAL 1504-2 h ) DISCONNECT TC CA8LES (ASPR) ISC 1504-2 h) TENSIONERS TO R.V. HEAD MM 1504-5 b) DETENSION R.V. HEAD MM 1504-5 (hwM) PARK INCORES ISC 1508-1 @MWEhO FH BRIDGE EQUIPMENT CHECKS I&C CkWIkW4kWk%1NT) REMOVE TENSIONERS FROM HEAD MM 8504-5 [j (D UNIT 2 LEADSCREW WORK MM WRITTEN (AWk%1kWkT PUT TRANSFER CARRIAGE ON TRACK OPS N/A (hC I
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' DESCRIPTlON RESP. PROCEDURE ys, i08:2o, o e, i44, Issy n, u 6, e e,so4,2 sy 8,seo,ast,a 47 s,a Myx> s z,see,s s se mo, PARK R.V llE AD STU0S MM 1504-5 (W)
REMOVE F T TUBE COVER PLATES (k80 lNSTALL FLOOO LINE COVER PLATE O) INSTALL AllGNMENT STUDS MM 1504-6 (O UNCOUPLE CRDerS MM , 15.0 4-12 O6C IIEMOVE CATWALK (SOUTH) C3 DISC HEAD WATER LINES MM 1504-2 T REMOVE HEAO SERVICE LINE PLATFORMS MM 1504-2 O) RAISE FESTOON CA8LE ,, ID ,- POLAR CRANE DILLION HOONUP ISC 1504-7 Q) REMOVE FTC HANDRAIL MM OC) SET UP EXCLUSION WATCH OPERATIONS AP1030 U) PREPARE TO REMOVE RV.HEA0 MM 1504-7 CQ REMOVE R.V. HEAO MM 1504-7 () INSTALL INDEX FIXTURE MM 1504-8 k% REMOVE TEMPORARY SEAL PLATE UT N/A y CLEAN / PAINT R.V. FLANGE MM 1504-4 ) INSIALL FTC SEALPL ATE MM 1504-4 bl60 INSPECT / CLEAN FTC-ORADE 8 Q.C. AP1020 (k?k ENGAGE / REMOVE PLENUM MM 1504-8 ) L 8' H IE MM 1504-6 (kWO
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A b et SHEET,ft,0F1-REACTOR VESSEL $EAD REMOVAL DESCRIPTION. RESP. PROCEDURE .w ion iao.im.i44.iu.n e no.ies aos.$ M. nee aseen.ase soose see ssa g gggg - - > -'- i.. , , .}}