CLI-84-11, Forwards Listed Handwritten Notes & Documents Used by NRC in Preparing SER Re Certification of Equipment Qualification, Per CLI-84-11.Documents Will Be Available in PDR & Lpdr
| ML20133G017 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 07/24/1985 |
| From: | Thompson O Office of Nuclear Reactor Regulation |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| CLI-84-11, NUDOCS 8508080530 | |
| Download: ML20133G017 (43) | |
Text
{{#Wiki_filter:_ _ _ _ . -_ _ . . _ . _ __ _ . _ _ . . _ _ _ _ _
-' . . July 24,1985 i
f .
' Docket No. 50-289 ,
4 I
- MEMORANDUM FOR
- John F. Stolz, Chief t j Operating Reactors Branch #4, DL FROM: Owen Thompson, Project Manager Operating Reacters Branch #4, DL i
SUBJECT:
DOCUMENTS REQUESTED BY UCS REGARDING EQUIPMENT , QUALIFICATION AT TMI-1 PER CLI-84-11 l By letter dated May 16, 1985, the Union of Concerned Scientists (UCS) i requested that the Comission direct the staff to provide "the underlying
; data and documentation concerning the SER conclusions" that relate to the staff's certification of equipment qualification for TMI-1 per CLI-84-11. ; Subsequently, on June 19, 1985, Comissioner Asselstine requested additional
- information about the available documentation.
The following documerits, which are currently in the NRC Document Control l System (DCS) with an accession number, have been sent to the Record Services j 4 Branch (RECSB) with instructions to n.ake the documents available in the Public Document Room (PDR) and Local PDR. l o Memorandum from Darrell Eisenhut, Director, Division of Licensing to
- Richard Vollmer, Director, Division of Engineering, dated August 3, 1984, l subject: TMI-1 Restart Proceeding Environmental Qualification Certification 1
1 o Memorandum from Brian W. Sheron, Chief, Reactor Systems Branch, DSI to , j Vincent Noonan, Chief, Equipment Qualification Branch, DE, dated December 12, ; 1984, subject: TMI-1 Equipment Subject to a Harsh Radiological Environment - The enclosed informal notes and documents that were used by the staff in i i preparing the Safety Evaluation are to become available in the PDR and Local i l PDR by distribution of this memorandum. 4
. o Notes made by NRC staff during its review of GPUN's submittals, telecons ,
( with GPUN and audits of the TMI-1 EQ files o Copy of "SB LOCA Radiation Qualification File Index" provided to NRC staff by GPUN during September 6 and 7,1984 TMI-1 EQ file audit ) I o Telecopy, dated February 21, 1985 from GPUN to NRC, providing i i information on incore thermocouple extension cable l I i ) l 8508080530 850724
; DR ADOCK 0
- -- -. . . . .- . - - - . . =. . . . -
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- r
.
- Memo to Stolz '
,L i Documents, including test results, analyses, calculations, evaluations, etc.
f which were relied upon by GPUN to demonstrate equipment qualification are in the licensee's possession and therefore the staff cannot make those documents available. i All documents in the staff's possession that were used in preparing the ;
- Safety Evaluaton in response to CLI-84-11 will be available in the PDR and
. Local PDR as soon as this memorandum is processed by RECSB. ,
; 8TQ DINAL $ U M N l Owen Thompson, Project Manager
- Operating Reactors Branch #4, DL
Enclosures:
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$ LOCA RADIATION OUALIFICATION FILE INDEX f_E7 _ _ _GE.8 EO-T1-101 ,/ BIW Cable /6Es aE6 EQ-T1-102 i States Terminal Block JE7 a.o/EB EQ-TM-103A F J - Limitorque VMO (Rx Bldg) gE(,
i 467 EQ-TM-105A(1)' / Limitorque VMO (Aux Bldg) r m V JE7- OE 7
- Q-TM-105A(2) s' Dings Brakes (Rx Bldg) 7 P S 13E6C EQ-TM-105A(3) y Dings Brakes (Aux Bldg)
AE7 DES .EQ-T1-108 / Anaconda Cable JE7 J.o6E8 - EQ-T1-109 # continental cable AE7 @ES . EQ-T1-lll ~ Kerite Cable OC6 S E6 _ EQ-TM-ll2A ' GE Terminal Block
' 55ES .
G E 7.. EQ-TM-ll3A 's ASCO Solenoid Valve (NUREG-0588) 5'N_ If0 EQ-TM-ll4A/~ ~ T ASCO Solenoid Valve (DOR Guidelines) p' f6 35TES p 'EQ-TM-116A O s Static-O-Ring Press. Sw. (DOR Guidelines)
'2E5 3.5 E7 EO-TM-ll7A #
Static-O-Ring Press. Sw. (NUREG-0588) ZE7 SE7 EQ-TM-ll8A.4 s __Rosemount Transmitter (ll53D) 656 SE6 EQ-TM-119A / Westinghouse Motors y 6SE F #E7 EQ-TM-122A(1) / ' Foxboro Transmitter (Aux Bldg) 385 6 /E7 EQ-TM-122A(2) v' Foxboro Transmitter (Rx Bldg)
#i*/. . IE8_.. EQ-TM-123A -
GE Fan Motor ME6 AEG EQ-TM-127A ./ NAMCO Limit Switch OE7 4E7 EQ-TM-128A s- Bailey Transmitter
.%.._6_56 EQ-TM-129A / Rosemount Transmitter (1152)
E *7 DE7 e EO-TM-130A O/ Rosemount RTD y -- 2E7 f.iBE8 EO-TM-131A(1) s Conax Electrical Seal (PL Series) 57 a.2ES . EQ-TM-131A(2) - Conax Electrical Seal (75900 Series) 7dE6 /EG s EQ-TM-132A Microswitch Limit Switch WG SE7 EO-TM-133A # ' --- Transzorb (Diode) ' If7 Gd8 EQ-T1-134 s Raychem Heat Shrink Tubing SSES /E8 s EQ-TM-135 g
- Target Rock Solenoid Valve C17 3.03E 8 EQ-TM-136A s' Weed RTD
!E7
- _.--fE8 y. EQ-TM-137A s GE Penetrations i \ ",, 9 ,'- 3 76ES /, 7 DES". EQ-TM-138A/' O/ Ross Solenoid Valves @ EEf mE8_ EQ-TM-14 0 A .- s Samuel Monte Cable 1"# g.g \ ". s e SEE NOTE 7 cp RJ l
TES/GA/4-TL'T O u </D A l-908 f.cGtt fpp hcVrr No. So-2Tri e
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OOPIES WILL BE DESTROYED UNLESS SPECIFIED DESTROY 4 l m . C' DATE .---
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SCtW-TI-628-803 Sheet I or 2 1114tM (IBIPINEST EVAtlin1XON but SRttT Proper Facility: fest unit: 1 BUEEE 0588 I1Ol II.F.2 Chected [N d sectet: Se-2ee Appr g
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Incere t=-ettsuttaa gualltteattert outstandine Systas: slensterlee Operettne cuattrennian se=' h-d I t si= Plant Is an. 19 hours Ylee 1 Test & Analysts AA-lega-thras 1994 1
- Campeaant: Cable Temperatesre J See Accident
(*F) - profile el t Test a Analysts Fisure 6.B-13 Itenefacturer: . Cent $aental wire Pressure See Accident 2 IP514) Profile #2 Test a Analyste 5, iI Itedel Insuber: __ Fleere 6.B 14 glGA3B/DeEK-15L l i Se14 tite 999 senseldsty 2 Fasictlen: (tl Test 4 Analysis Eise. Connector Chemical 0.5-11 Accurecy: Spec: N/A Spray (ple) I Besest: N/A p ,, 4 M t & Aaelysis % hI mediatten 2.14ste' Service: (samt 4 5 3ncere Tmennocenotes - Test & Analysis Asl 134/40 tocation: (*r/ Tears) 3 note A Containment Pase 2a Test a Analysis r1. ses' e 16e* lF1eedLevel * ' _ . ._ ce R/A } Eles: 246.66 ft 3 E/A s/A ED Aheve Fleed Level: Tes Pase 2a i Ul an Men geferences:
- 9. 29 To *139*/f SELES FILE M_ ta TJ 139
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- 3. Crum 1es 242 sev. 4 Aree 32 i in C. EPos C-Iles-625-5350-ee7
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SCEW-T1-61. 003 Sheet 2 cf 2 - COMPONENT MATERIALS EVALUATION SNEET INCORE MONOTORING
, Plant I.D. No.1 AR-149A thru 199A Component: Cable -
Manufacturers CONTINENTAL WIRE AND CABLE CO. Model No.: CAI B/M EK 1$L D o THERMAL ACINC RADIATION PARTS LIST MATERIALS LIST QUALIFIED LIFE REFERENCE QUALIFICATION REFERENCE Wire Insulation Tefloo"(FEP) 130'F/2.47x10' yr CPUN 1.1x10'R CPUN h . 1101-5340-75 Rev 1 C-1101-625-5350-004 N air F Wrapping Aluminum / Mylar 130*F/3.95x10* yr CPUN 1x10*R EPRI RP 1707-3 C-1101-600-5350-002 Rev 1 Jacket Teflon (FEP) 130*F/2.47x10' yr CPUN
~
1.1x10*R CPUN 1101-5350-75 Rev 1 C-1101-625-$350-004 E u 2 D . 5 - E A 0680N/pg 18 o Page 16 of 16
. m x rumrhagn . .
Nucleer cate C-1101-425-5 350 -eay m.......y........ ..3.... : Radiation Normal Service plus Accident LOCA ' "IO/25N4"* "";" WEcT... Conditione for*the lacove 'Detectictr* Cable"AsseRbl7 " "" uj
. .......... .. ................................................... .......... ............ em. e=n Mh , [
1.0 purpose Determine service the Incore detector co penetration cable assembly normal 40 year life plus accident radiation exposure requireaanta. 2.0 Conclusion The incore detector to penetration cable assembly will be exposed to a total intergrated radiation dose of 4.79 x 106 RADS. 3.0 taferences 3.1 GPW Calculation C-1101-625-5350-004 3.2 IUIEG-0020 vol. 5 No. 7 July 1981 e 4o 3.3 TMI-1 FSAR page 1.1-1 dated 7/82 3.4 TMI-1 FSAR Table 6.6-3 3.5 CPW Consral Arrangement Dvs. IE-153-02-004/007 3.6 Radiological Health Handbook 1970 3.7 D0R Cuidelines Appendix 3 . I y( 3.8 CAI Radiation maps I-001-052 dated 8/23/71
/
4.0 calculation /Ar.alysis The incore detector cable assembly le required to function fort
\
*4.1 Normal Service *
\ p '
N 40 year 6 30 MR/hr Normal Operation (Refer 3.8) 350400 Houra x 50 MA/ hours = 1.752 x 10I MR or 1.751 x 10,4 Rads in 40 Fear service 4.2 Bets Accident 1.0CA Dose (Refer 3.1) O ih CPW calculation concluded that the jacket and shield reduced the dose h'j p from leta radiation that reaches the wire insulation by a factor of 10. 4.3 Camms Accident !.OCA Uo_se (Refer 3.7) 1 l l DOR Figures 1 through 4 provide factors to be opplied to the I conservative dose to correct the following plant specific parameters (1) reactor power levell (2) containment volusal (3) shielding; and (4) compartment volume. Acco 0018 stee 810 LE0 *0N tM3T)rH ikfD 61 GI 50/IE/20
p , ... (o Nu^1rt eu.,.ot-1101:-625.-5350.3.*e 7 O i. DnatTneo.... 2. . OF sasseT.,. Radiation Wormal Service plus Accident LOCA ***' """"I E/2 5 /8 5. .. . 3. '. condttrons"for"the"Incors Detecti68 card"Arsatty" COM9.9YIDATt.?" ""5 W$
.........................................................................................CMsCD.SY/ Daft j1s The radiation service condition for the incore cable assembly appitention specific parameters are Reactor power level - 2.535 MWeh (Refer 3.3)
Containment volume - 2.126 x 106 gc3 (Refer 3.4)
# + Compartment volume - 102,203 ft3 (Refer 3.5) thickness of D-Eing shield wall (concrete) - 48" (Rafer 3.5)
+ Thickness of Steel Door - 1 f t (Refer 3.5)
Time equipment is required to remain functional - 30 Days
*The compartment volume was assumed to be the line of sight area {
above and below the incore detector seal plate. Free GPUN General ' Arrangement Drawings this volume was calculated to be 40 f t a 21 f t. x 105 f t = 88.200 f c3 above and 19 ft x 11 ft x 67 ft = 14,003 ft3 below the incore seal plate. ' Total = 102.203 ft3 (Refer 3.5)
+The Density of Concrete 2.25 ga/cm2 and steel 7.86 ga/ca2. are assumed to provide equivalent shields (Refer 3.6)
T'n e problem is to aske a reasonable estimate of the dose that the equipment could be expected to receive la order to evaluate the - odequacy of the radiation service condition specification. Step 1 1 Inter the nosogram in Figure 1 at 2,333 ) Nth reactor power level and l 2.126 a 106 ft3 gentainment volume and read a 30-day integrated I does of 1.4 x 10' RADS. Step 2 Znter figure 2 at a dose of l'.4 a 107 MAD 5 and 48" of concrete shielding for the soapartment the equipment is located in and read
> 1 x 103 RADS. This is the dose the equipment receives from sources i outside the compartment. To this must be addes the dose from sources j
, inside the compartment (Step 3).
se.,1 1 p Enter F18ure 3 at 102,203 ft3 and read a correction factor of 0.31. Q The dose due to sources insige the compartment (line of sight volume) would then be 0.31 (1.4 x 10') = 4.34 m 104 RADS. Tne suma of the doses from steps 2 and 3 equals: 1 a 103 RADS + 0.31 (1.4 x 107) RADS a 4.34 x 104 meco este s*4e GTO 420*0N W3Tni ndD 82 Et GB/TE/ZO
~ ~
Nuclear cate. woc-2.lat :415 .525a-F87 SHssT NC' " 3 Os ""3 Raciation Normal SeWvice plus Accident LOCA ' "ICI " "C:: nd it ie nd "f at "t W Itic titii ' De res tT6W ' Cib1'e" Xi s sim bTf"~~~~~ . dan o o o o YD7I5 / 8 coup. ev/04Ts NT/4M %
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .' .. .1+)y
. . .r . .
4.4 Reta + Caams Accident Dosa 4.34 x 106 RADS Gamme 30 LOCA Intergrated Exposura 0.10 (4.34 x 10 )6 NADS neta 30 LOCA Intergrated Exposure 4.34 x 106 , 4,34 ~ ~ ~x
~ 1c5 = 4.774 x 106 BADS Total Accident Intergrated Exposure 4.5 Suasary The incore detector to pentration cable assembly will be exposed tot 1.752 x 10' RADS during Normal Service-4.774 x 106 RADS LOCA Accident-Total 4.79 x 106 3Agg eum I
Accocow es4e EB /E0*D4 W3 Tnt ridD 02 GT G8/TC/E0
\ - Nuclear @ cuc.wo...cr.110.1-625-5350.-007 hr i
- i. susucer...... Radiation Normal Service plus Accident LOCA BHEEr NO. . .. OF........3 f5 CW t i m ' ro r" tw' Insc re"De t Ec yy"ppf ogg j t i bn" COST 6"AY teh 1.0 Pu_rpose Determine service the incore detector to penetration cable assembly normal 40 year life plus accident radiation exposure requirements.
2.0 Conclusion The incere detector to penetration cable assegbly will be exposed to a total intergrated radiation dose of g x 10 RADS. 3.0 References 3.1 GPUN Calculation C-1101-625-5350-004 3.2 NUREG-0020 Vol. 5 No. 7 July 1981 3.3 TMI-1 FSAR page 1.1-1 dated 7/82 3.4 TMI-1 FSAR Table 6.6-5 3.5 GPUN General Arrangement Dwg. IE-153-02-004/007 3.6 Radiological Health Handbook 1970 3.7 D0R Guidelines Appendix B 3.8 GAI Raolation maps E-001-052 dated 8/23/71 4.0 Calculation / Analysis The incere d'atector cable assembly is required to function'fors 4.1 Normal Service 40 year 9 50 Mt/nr Hermal Operation (Refer 3 l 350400 Hours x 50 MR/ hours = 1.752 x 10 I MR
.8) or 1.752 x 104 Rads in 40 year service 4.2 Beta Accident LOCA Ocse (Refer 3.1) ;
GPUN calculation concluded that the jacket and shield reduced.the dose from Seta radiation that reaches the wire insulation by a factor of 10. 4.3 Gama Accident LOCA Dose (Refer 3.7) 00R Figures 1 through 4 provide factors to be applied to the conservative dose to correct the following plant specific parameters: (1) reactor power level; (2) containment volumes (3) shielding; and (4) compartment volume. .
. 0211p pg 74 aoooonw . .
120 1.20 W N M TEIET E8/TE/20 .
SMucht h calc.McC .1101-625-5350..007 #evJ
**"' '" Radiation Normal Service plus Accident LOCA "Y0/lN84"" ""
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....................................~......................n...... CHK'D. ev/DATE ....
The radiation service condition for the incore cable assembly application specific parameters are: Reactor power level - 2.535 MWt (Refer 3.3) ! Containment volume - 2.126 x 10 ftJ (Refer 3.4)
- Compartment volume - 102,203 ft (Refer 3.5)
+ Thickness of D-Ring shield wall (concrete) - 48" (Refer 3.5)
+ Thickness of Steel Door - 1 f t (Refer 3.5)
Time equipment is required to remain functional - 30 Days
*The compartment volume was assumed to be the line of sight area above and bt'ow the incore detector seal plate. From GPUN General Arrangement Drawings this volume was calculated to be 40 f t x 21 f t x 105 f t = 88.200 t3 above and 19.ft x 11 ft x 67 ft = 14,003 ft below the incore seal plate. Total = 102,203 ftJ (Refer 3.5)
+The Density of Concrete 2.25 ge/cm2 and steel /.86 gt't/cm2 are assumed to provide equivalent shields (Refer 3.6) the problem is to make a reasonable estimate of the dose that the equipment could be expected condition to receive in order to evaluate the adequacy of the radiation service specification.
Sten 1 l Enter ft the nomogram in Figure I at 2.53h MWth rea'ctor power level ano 2.120 a 10 6 3 containment volume and read a 30-day integrated dose of 1.4 x 10 7 RADS. Steo 2 Enter Figure 2 at a dose of 1.4 x 107 RAUS and 48" of concrete snielding for the compartment the equipment is located in and resa I x 103 RA03. This is the dose the equipment receives from sources outstee the compartment. (Step 3). To this must be adaed the dose from sources insioe the compartment Step 3 anter Figure 3 at 10'2,203 ft3 and read a. correction factor of 0.31. The dose due to sources inside e compartment (line or sight volume) would then be 0.31 (1.4 x 101) = 4.34 x 1 RADS. The sums of the doses from steps 2 and 3 equals: 1 x 103 NADS + 0.31 (1.4 x 107 ) AAOS = 4.34 x 105 ,,
- 021lp pg 15 i
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ascar n o. .. 3,. ... on . . . .3. . . . . .
SuaJECT .R ad i at io n .16a rnal. .S e rv i ce pl us . Acc iden t. .LOCA. . . . . . . . . . . . . . . . oATc . . //4f/ E .. Conditions for the.Incore Detection Cable Assen.bly coup. sy/oArm ...
........................................................................................CHK'D.SY/DA ...
Steo 4 Enter Figure 4 at 10 hour and read a correction factor of Q4E . Apply this factor to the sum of the doses determined from steps 2 and 3 to correct the 30 day total dose to the equipment inside the compartrnent to 10 hour. !' O.45 (4.34 x 106) .1.95 x 106 4.4 Beta + Gama Accident Dose 1.95 x 106 RADS Gama 10 Hour Integrated Exposure 6 0.10(1.9)x10)RADSBeta10HourgntegratedExposure 1.95 x 10' + 1.95 x 105 = 2.145 x 10 RADS Total Accident Integrated Exposure 4.5 Sumary - The incore detector to penetration cable assembly will be exposed to: 1.752 x 10 RADS during Normal Service 2.145 x 10 RAD 5 LOCA Accident Total 2.16 x 100 RADS b@
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6 DEC 3 1:34 NUCESSF Memorandum
)
Subject TtI-2 Radiation Exposure of lacore %ermacouple system External Extension cable Date: November 27, 1984 From: Radiological Engineering Support operations Manager al-2 J.A. Flanigan Location: 21-2 g Manager EQ 9240-84-2592
.0 s.J. Miliati In response to your request of i
the following information. 11-13-84 (RefsEPEI/84/1913) we are providing We most reliable data presently available support a total dose of 1.0E7 Rads to the teflon cover of the extension cable. Thir total is comprised of four contributors'11the dose due to incore table contamination, 2)the dosetoreceived due by MPR-213, 3)the done due to Krypton-85, and 4)the dose Xenon-133. Rads,1.75E6 Rads and 6.21E6 Rads respectively.The individual dose estimates are 1 Details regarding each of these estimates are also included. We 6ase estimate based on incore table contamination was obtained through the use of documented surveys by Health Physics Technicians. The surveys were from 10-16-80 using thirteen total surveys. Reactor Building Entry 83 through 6-26-84 Entry 9394 , Several decontamination attec; pts have been made results. on the table, therefore reducing the emphasis placed on smear survey usually performed with an RO-2A.All surveys used for dose estimation were S/Y dose rate surve elevated garina exposure rates.h e exposure estimate based on MPR-213 was used to rep ination Results of the Three Mile Island Radiation Detector HPR 213This estimate was
. HPR-213 is a general area radiation monitor located adjacent to the incore~ table t dose received by the detector was based on transistor testing. This dose.does The l
detector was not* contaminated.not include any exposure from gassa or contamination as the ; nuilding on May 28, 1981 Entry 011.We detector was removed from the Reactor I The expos Report No. 44~ure basedactivity to convert on the prehence levels of skin to Rec /hr Krypton-85 gas was estimated using RCR dose Krypton-85 Building Purge Kr-E5 Venting, activity levels in the Reactor Building were taken ' from GE throughout the building. with gas concentrations assumed to be equal analysis reports for 1980 to validate accuracy. Total Kr-85 activity was cross checked with e , assumed to be ended at the start of purging June Exposure 28, 1980. due to ur-a5 was l v20 LE0 *0N Aooco648 s s3 W3'OnN ndD EZ ST GB/TE/E0
J. . :' :- (LJ t pag. 2 November 27, 1984 9240-84-2592 - s.J. Milioti hemajorcontribubontothedoseestimateisthatfromexposureto Xenon-133 gas. error, his n e r is also the most likely source of significant he estimate of the X-133 contribution was determined by using ICRP-30 to convert activity levels to Rem /hr skin doso. X-133 activity determination was done by plotting a half life curve to aero time from a Beactor Building air sample taken on May 4, 1979. His is in accordance with the logic uesd in GEND-INF-032 vol. I, Radionuclide Mass Balance for the 'rMI-2 Accident: Data Base system and Prelir.inary Mass Balance Vol. I. here is presently an effort underway to provide further information on f I-133 levels which may allow for revision of the number presented here. Se exposure estimates represented are rough estimates based on currently available information. Any questions concerning their accuracy or the need for refinement may be addressed to S. Layendecker at Extension 8364.
*** s J.A. F anigan Radiological Engineering support Operations Manager TMI-2 nsl.K '
calculations sy: s. Layendecker
. Radiological Engineering TMI-2 Checked By:h.E.* a inian
- Radiological Engineering support Effluent Assessment Manager TMI-2
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Approved By: W
... O . Slobodien, Man'ger a
diological Engineering TNI-2 JAF/8L/ JET /K3S/wsm attachments ec: Carirs
- J.E. Hildebrand-Radiological Controls Directcr TMI-2 * *
- C.A. Kuehn-Manager, Radiological Controls TMI-1
- R.P. shaw-Radiological Engineering Manager TMI-1 l
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