ML20010H229

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Forwards Info Re Equipment Environ Qualification/Radiation Profiles in Response to 810915 Request
ML20010H229
Person / Time
Site: Fermi 
Issue date: 09/18/1981
From: Colbert W
DETROIT EDISON CO.
To: Kintner L
Office of Nuclear Reactor Regulation
References
EF2-54-825, IEB-79-01B, IEB-79-1B, NUDOCS 8109240241
Download: ML20010H229 (47)
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{{#Wiki_filter:. De'Toit Edison RE!EEN" September 18, 1981 EF2 - 54,825 j y 4 IQ k Mr. L. L. Kintner Q dp T_g Division of Project Management %f, cp U.S. Nuclear Regulatory Commission 9' EI g y [1 cp C7 s Washington DC 20555

Reference:

Enrico Fermi Atomic Power Plant-Unit //. ( W

Subject:

Equipment Environmental Qualification - Radiation Profiles. l'er.c Mr. Kintner: er your request, the following documents are enclosed: r 1. General Electric letter TDEC-4034 dated 8/25/81 and document number 22A3019 Rev.1; Radiation entries in FSAR table 3.11."; for normal plant operation. 2. Sargent & Lundy letters SLM(NI)-223 dated 8/12/81 and SLM(NI)-207 dated 7/8/81 and attachment, Equip-ment Qualification Dose Analysis during Post-LOC,", accident. The above information was requested during a telephone con-versation on September 15, 1981 between the NRC (Frank Akstulewicz) and Detroit Edison regresentatives (Dick Beaudry, Lou Bregni and Len Fron). Should you require any further information, please let me know. Very truly yours, Wthk 9 W. F. Colbert Technical Director Enrico Fermi Unit 2 Project WFC/QHD/mb O O I 5 /// 8109240241 01091T PDR ADOCK 05000341 A PDR

y GENER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECT C g.tP 75gng AVE.. SAN JOSE, CALIFORNIA 95125 tugust 25, 1981 1]EC-4034 Mr. W. F. Colbert, Project Engineer Enrico Fermi 2 Project The Detroit Edison Company Documentation Control - Room 361 2000 Second Avenue Detroit, MI 48226 Attention: Mr. L. Sherman Gentlemen:

SUBJECT:

FERMI 2 RADIATION ENVIRONMENTAL CONDITIONS The radiation environmental condition entries in FSAR Table 3.11-5 were extracted from GE Specification 22A3019, BWR Equip: vent Environmental Requirements. The values in this specification were derived by a combination of analysis and extrapolation of actual operating plant measurements. Very truly yours, t I)fg((o~je. son P ct Mana:ar I Enhco Fermi 2 Project CMJ:sem/2I ~ cc: T. J. Evans, GE Southfield F. Gregor, Edison S. K. e...er, Edi son R. Pratt, GE Site T. Hintun, GE Site O n. 9

G4NERAL h ELECTRIC NUCLEAR ENERGY DIVISION Document No. 22A3019 Rev.3 (c. -3 TRANSMITTAL General Electric Class PROJECT (S) ENRICO FERMI 2 & 3 TITLE OF DOCUMENT nWR EQUIPMENT ENVIRONMENTAL REQUIRDfENTS TYPE OF [] PURCHASE SPECIFICATION REPLACES DOCUMENT: N SYSTEM DESIGN SPECIFICATION DOCUMENT NO. (..,j [] INSTALLATION SPECIFICATION [] PIPING OR COOLING SYSTEM INVOLVED , RESPONSIBLE ENGINEER VM DOCHEZ ISSUED BY EA HARTMAN DATE REFERENCES MASTER PARTS LIST (MPL) NOS. A61-4270 SPECIFICATIONS DRAWINGS OTHER REVISION RECORD (' REVISED PER (ECA, ECN, ETC.) NE 35496 SHEETS AFFECTED 5,6,7,8,9,10,11,12 & 14 REVISION IDENTIFIED WITH & COMMENTS: DISTRTEITTION l NAME MAIL CODE COPIES NAME MAIL CODE COPIES l VM DOCHEZ 761 1 PH GREGORY 717 1 PR04 EN1.ICO FERMI KH1 l 723 3-DCC TOTAL 13+:}f l 742 1-Chuc L Proj.Cnt. F: le 366 12+1H l 595 LIB 1 Johnson 369 1 (-) l \\_ 711 1 PCF ECN - 12 753 1 PR04 EN1LICO FERMI :l LCl DOCUMENT CONTROL 250F 1 TOTAL 4+1H 250U 'l Proj.Cnt. F. le 366 3+1M . S A -0 19 A30-0 1..0 -000 772n 1 Johnson 369 1 660CF 1 PCF ECN - 12 FEB121973 g-- 361MHF 1 ...uj HRM' S$-- Co.N LTR. NO Ii (i APPROVI!D FOR ENRICO FERMI UNIT #2 l in /r /AG 4

GEN ER AL h ELECTRIC DOCUMENT NO _22A3019 REV. I HUCLE AR ENERGY DlVl510N ' (- ATOMIC POWER EQUIPMENT DEPARTMENT APPLICATION (. Son Jose, Californie MPL A61-4270 QSPECIFICATION DRAWING TYPE LESIGN DOCUMENT TITLE BWR EQUIPMENT ENVIRONMENTAL REQUIREMEhTS G ' REVISIONS q 1 Per ECN NE35496. Sheets 5,6,7,8,9, 10,11,12,14. Revisions were identified with a spade. ($) 3 73 ( g u 3 1 1973 I G ngggog u.n-ac y,ocy73ag~inde$'2-s-7, ~~. DEC 161970 -u' 2 -~ 1 REVISION STAFUS SHEET ,ye j NED GOS 63-4 9 .mm... CL - ~. ~ ,...->+a.. w, n. m.- " -'1 w J - ~ - s es u nn

GENER AL h ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT q DESIGN SPECIFICATION t-jho 2 co o .-uf 3 TIT L E BWR EQUIPMENT ENVIRONMENTAL REQUIREMENTS 1. SCOPE 1.1. This document specifies indoor environmental data to be used for design of equipment supplied by the Atomic Power Equipment Department (APED). 1.2. Scismic requirements, and vibration levels are not included in the scope of this document. 2. APPLICABLE DOCUMENIS, CODES, AND STANDARDS 2.1. General Electric Company Doc-cs 2.1.1. This specification, or applicable portions, thereof, represent the controlling environmental data for use in design of the specific equipment supplied by APED. If there is a conflict with respect to environmental data between this specification and other design documents, the requirements of this specification shall govern. ( 2.2. Codes and Standards N. 2.2.1. The following documents are to be used in conjunction with e.l.la specification to the extent specified herein. a. Atomic Energy Commission (AEC) - Criterion 1 of the AEC General Design Criteria, 10CFRSO, Appendix A. 3. DESCRIPTION 3.1. Incorporation of appropriate enukonmental design data is necessary to ensure proper functional performance of the system or equipment during all design modes of operation. 4. REQUIREMENTS 4.1. Normal Conditions 4.1.1. Normal conditions are defined as those conditions existing during routine plant operations. Environmental requirements stated i-Tables as Normal or Operating are those which shall be used for design, and represen ormal, maximum and minimum espected conditicas that may exist during routine plant aration. DEC 161970

GENER AL h ELECTRIC s e ATCMIC POTER ECUIPMENT DEPARTMENT ( spec.~o. 22A3019 a c v. ~o.1 s (. DESIGN SPECIFICATION j ~ ~o. 3 c o~, o~.~ a s, i 4.1.2. Tables summarize the enviromental conditions which shall be used for component or system design within the plant locations stated. All components shall be designed to operate under the normal conditions. 4.2. Abnormal Conditions (? 4.2.1. Abnormal enviromental conditions are defined as those which deviate from the conditions described in Paragraph 4.1, preceding. The most significant abnoraal condition is the environment during. and following postulated design basis accidents. Other ambient conditions, including cmall continuous steam leaks which generate high temperature and/or high humidity, and test or operator-controlled conditioas, shall also be considered. 4.2.2. Essential components and safety systems shall be designed to operate or be in a fail-saf e condition, as given in Section A,11-111 of the following tables. Essential components are those which are essential to the prevention of accidents which could affect the public health and safety or mitigate their consequences, according to the definition and interpretation rt v. iterion 1 of the AEC General Design Criteria,10CFR50, Appendix A. 4.3. Tables (- 4.3.1. The following tables are divided into Sections A and B. Section A defines the pressure, temperature, and humidity environmental conditions. Section B defines the radiation environmental conditions. 4.3.1.1. Section A is further subdivided into three L.srts: Section A-I includes all equipment operating under normal conditons. Section A-II and A-III delineate the abnormal conditions for essential s oponents. The tables in Section A-II and A-III for essential components represent an envelope of abncreal conditions in which the systems or components are required to be functional or in a fail-safe condition, as noted. The specified envelope is not based on one specific design basis accident, but on all postulated accidents relevant to this envelope. 4.4. Drywell Zones 4 A diagram follows (Figure 1) showing typical drywell zone locations within ( j pr.4.1. imary containment. l t 4 9 tuc D-DEC 1 S 1970 7 .. _,.s mm m,

.a o GENERAL-' ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT s re uo.22A3019 = c v. ~ o. ' - DESIGN SPECIFICATION . ~,. o 4 c o~, ou.mc c, 5 u l ZouE 1 Vicinity Recirculation Same as above 128' Average Same as above Pump Motors - Minimum Qm Zone 4 135' Maximum m r3 m -< Area Beneath RPV - Same as above 135' Average Same as above $2 Zone 3 100* Minimum (5) 28 165* Maximum (4) 2 H II. Reactor Building Range from (-)0.10" 70* Normal 40% Normal (Not otherwise noted) to (-) 1.0" Water 104* Maximum 90% Maximum gage, static pressure 40* Minimum 20% Minimum 1 Reactor Building Standby Same as above $104* Maximum Same as above Liquid Control Area 70* Minimum II HPCI, RCIC Equipment Same as above 70* Normal Same as above 5 Area 104

  • Maximum (6) 5 (6) 60* nl..tmum ll v, h g

M{ Core Spray and RHR Same as above 70* Normal Same as above ,h c')9 Equipment Area 104

  • Maximum (6)

M (6) C) 40* Minimum IA N (Notes are given at end of table) f,g, o @ H

0 Section A-I (cont.) PRESSURE, TEMPERATURE, RELATIVE HUMIDITY ENVIRONMENTAL CONDITIONS g J vs Normal Conditions - Plant Operating h I Area Pressure as Noted Temperature Relative Humidity

  • F O

n Reac or Building (con.) r II. Steam Tunnel Range from (-)0.10" 10' Normal 40-50% Normal 5 5 2 E cs to (-)1.0" Water 130* Maximum 90-98% Maximum gage, static pressure 40* Minimum 20% Minimum III. Turbine Building (3) Range 0.0" to (-) h 120' Mafchu %$dEggjer40% Normal N 90% Maximum g$ G.25" water gage 20'Nrmdl'(kClectr[ cal)20% Minimum E. u. ) Max 2 mum static pressure 104 IV. Radwaste Building (3) Range 0.0" to (-) 70* Normal 40% Normal 2m 0.25" water gage $ 104* Maximum 90% Maximum y m" static pressure 40* Minimum 20% Minimum qQ $= Radwaste Building 0.0" to (-) 0.5" 70* Normal 40% Normal yg Equipment Cells water gage static 120' Maximum 90% Maximum g pressure 40* Minimum 20% Minimum V. Control Room Range 0.10" to 1.0" 60*-90* Normal 40-50% Normal water gage static 120* Maximum 460% Maximum (7) pressure 40' Minimum 10% Minimum Notes: (1) Primary containment atmosphere during normal operation may be inerted with 96 percent nitrogen, 4 percent oxygen. (2) Whenever the residual heat removal and core spray motor and the emergency core cooling system are running, during test periods area space coolers may be required to maintain I1 the ambient temperature listed. g" (3) Components located in turbine building or radwaste building required to operate under ab- ~ g normal conditions, if any, should be designed for equivalent conditions as shown for mn reactor building. 5 (4) During loss of offsite power, and other emergencies, except during Design Basis Accident { hla temperature of and area underneath the reactor pressu e vessel will be maintained at o (7 165'F or lower for up to 30 minutes. e M (5) The same minimum temperature (100*F) shall apply inside base of the shield wall. Air velocity over vessel insulation and exposed vessel parts shall be approximately 6 f t/sec. ;= c) i-* (6) The maximum temperature and humidity will occur simultaneously in these spaces less than

s a r il 1 percent of the time.

, o failure conditions Relative Humidity may approach 90, percent for (7) During HVAC equipment 100 hours, but l'0*F would not occur simultaneously.

N .R O 'm.' tio e-og g RSSDrTIAL EQUIPMElef o mm FRESSURE TEMPERATURE, REIATIVE E21Dh., N IRONMENTAL CCNDIT1055 Section A-II ~ 8 Inside Prinary Contaisunent - Ahnormal Conditiona mM) Componente must be operable under the following conditiona - 0 Condition C4 :t 1 Core sprey injection check valve Temperature 340'F 340*F 320'F 250*F 200*F OE 4F3 n LFCI-RRR injection check valve Pressure (1) $-2 to 56peig(2) -2 to 35 pets -2 to 35peig ) to 25pois o to 20 pois m 100 % 100 % 100 % 100 % y2 Reactor shutdown cooling suction Re1. htsnidity g00 % valve including operator and cable Duration (3) 45 sec (3) 3 hours (3) 6 hours 1 day (3) 100 daye s m m llI3> talief velve including r-" operator and cable oC Vessel level indicator Structural ccerponente (e.g. loop m E restrainte, vessel skirt, etc.) $m l i I 2 Feedwater Check Valve Temperature 340'F 340'F 320'T n ev, m c3 $ EPCI staan line isolattan Fressure -2 to 56peig -2 to 35 pets -2 to 35peig i -.4 valve including operator and Rel, humidity 100 % 100 % 100 % i Duration 45 sec 3 hours 6 hours gg cable RCIC eteam line isolation valve t C7 includinF operator and cable Q Reactor Kater cleanup suction valve includios. operator and cable Reactor water sample line valve including operator and cable Lices 2 inches and smaller (Isolation) Valves, Operatore, Cabling) Cables to intermediate range monitors and process radiation esonitor Reactor vease) head spray isolation valve including operator and cahle - 3 Main steam isolation velve Temperature 340*F 340'F includies operator and cable Pressure -2 to 56paig -2 to 35peig y; Main steam drain toolation valve Ra1. htsnidity 100 % 100 % ,a" o includirs operator and cable Duration 45 see 1 hour Standby liquid control injection check valve U b 4 Recirculation valves (; main valves, Temperature 310*F 285'F g bypase valves, equaliser valve) Pressure -2 to 56psig -2 to 35 pets n including operators and cables Rel. htseidity 100 % 100 % CD a Duration 45 sec 30 min I O 3 4 n c) e 4 .r m o <=> 4 CD M es

A m ./ 9 tn M } v, sss m 1AL acc1Fw m 7 l rn n Section A-II (cont.) FRESSUR?. TD(FERATURZ, REIATIVE HIMIDITT, '37IR019fmAL CONDITIONS -ri e-e n $ Inside Primary Containment Ahnormal Conditions 3= d ,d_, Yalves not required to be operable but must not f ail open under the o Condition Component following conditions g g m m 5 Feedwater check valve Temperature 250*F 200*F g EFCI steam line isolation Pressure 0 to 25pois 0 to 20 pois Qm m y l valve iricluding operator and Rel. hunidity 100 Z 100 % 2 l cable Luration 1 day 100 days p m g RCIC ateam line isolation valve l c including operator and cable g j Recirculation valves (main valves m bypass valves, equalizar valve) m l m includirs operators and cables y Reactor vessel head spray isolation o m Valve it.cluding operator and cable m C2 Reactor water cleanup suction valve j M includits operator and cable m W Reactor water sample line valve y3 includirs eparator and cable m Lines 2 inches and smaller Z (isolation valves, operatore, cabling) 6 Main steam isolation vetta Temperature 340'F 320'F 250*F 200*F includies operator and cable Fressure -2 to 35 pets -2 to 35psig 0 to 25psig 0 to 20psis Main stes:s drain isolation valve Rel. humidity 100 % 100 % 100 Z 100 I including operator and cable Dura tion 3 hours 6 hours 1 day 1CO days Standby liquid control injection check valve Notees (1) The equipment inside the primary containment will be subjected to 62 pets and 135'T for a maximum of 3 dcys during periodic leak testing.

  • E (2) 56 pois la 90 percent of maximmi containment internal pressure of 62 peig, as allowed by ASME Boiler and Fressure Yesaal Code,

$a 2 Nuclear Vessels, Section III, Article 13, Paragraph N-1312, Sub-Faragraph (2) o $(3) Durations shown are terminatien times measured f rom the initiation of the postulated accident, i.e. Condition 1, the 3 hour duration, is the period from 45 seconds through 3 hours, the 1 day duration is the period from 6 hours g 1 7 through 1 uy (24 hours). g s a m M o 3 M 2 m e '4 I N 2 m o l t

5 GENER AL O sticinic ATCMIC POWER EQUIPMENT DEPARTMENT (' ..cc no. 22A3019 ,,,,,,, 1 ( -DESIGN SPECIFICATION 9 c o~, o .-s c,10 .-~o. LEGEND SECTION A - 11 This legend is a ccapilation of basic abnormal environmental pressures and temperatures together with the time durations expected. The full spectrum of simultaneous environ-mental possibilities is not *p csented in a series of curves, but rather as a descrip-ion of the boundarica within which designated equipment must operate at discrete times during the cycles / modes of tha reactor's operation. 1. Temperatures: 340*F Upper bound on maximum superheat temperature for a steam leak. This maximum can occur only when the reactor is at a pressure of 400 to 500 psi and a containment pressure of 50 psia. For higher or lower reactor pressures or lower containment pressure the temperature is less. 320*F Haximum superheat temperature during shutdown cooling line flush af ter reactor has been depressurized to 150 psia. 310*F Upper bound on saturation temperature at containment design pressure. This temperature applies only to the recirculation valves which must be functional only in the event of a recirculation line break. In the event of a steam leak (' that causes high superheat temperatures closure of the recire valves is not required to flood the core. 285'F Saturation temperature at 35 psig (plus 4*F nargin). This temperature applies only to the recirculation valves which must be functional only in the event of a recirculation line break. 250*F This represents the maximum long term temperature in the containment during the first day following a DBA. 200*F This represents the extended long term temperature in the containment following a postulated DBA. 2, Pressures: -2 psig Negative design pressure of the primary containment 56 psig Positive design pressure of this primary containment. 35 psig The containment pressure corresponding to all the non-condensibics initially in the drywell being transferred to the wetwell. 25 psig Upper bound on long term pressure response up to one day following a postulated DBA. 20 psig Upper bound on extended long term pressure at one day and longer following a postulated DBA. I.f uED:

4 O GENERAL ELECTRIC ATOMIC power EQUIPMENT DEPARTMENT (,~DESIGNSPECIFICATION ..cc. 22A3019 ,, c... I ... 10 cou,o~....,il 3. Durations: 45 seconds Conservative time duration to cover peak containment pressure. 30 minutea In the event of a recirculation line break the recirc valve. must be operable to insure core flooding. This represents a conse.vative duration. (,'t I hour Applies to valves that isolate automatically on low RPV pressure or high drywell pressure. 'Ihis time represents a conservative duration during which the valves must be operable. 3 hourn a) Conservative duration of time to depressurize the RPV, at a rate not exceeding 100*P/lir, down to 150 psia, b) Conservative duration of time to flush shutdown cooling lines and de-pressurize reactor below 50 psia. ( O I l ...u a o-

l ~ e m' M. s , b Section A-III ESSENTIAL EQU1FMENT M Outside Primary Containment - Abnormal Conditions mn Condition Component Componente must be cperable under the followir.g conditions y w 1 HFCI system isolation valves Temperature (41 148'r (1) Q> h including operator and cable Pressure 7" w.g. Q) -4 d ~ RFCI pump, turbtne, control. Rel. Humidity 1001 Q1 o E instrumeatation and electrical Duration 1 hour .Z 5 m equipment Q) E RCIC system isolation valves, oa m including operator and cable RCIC pump turbin., controle. Q33 instrumentation ele <.trical m equipment Q) o P s C I Y 5 - Main steam isolation 3alves Temperature (4} 143 7 g} g g2 Pressure 7" w.g. Q) z m l Rel. Esidity 100% (1) d r"" Duration I hour O " ] C3"4 m m Feedwater isolation valves. TemperatureC4) 148'? Q) $ :33 +3 including operator and cable fressure 7" w.g. Q) w Reactor water cleanup isolation Rel. Htssidity 100% Q) gO valves, including operator and Duration I hour z d cable 4 RER system isolation velves, Tempera ture(4) 148'? for 61eent$s Q) g including operatore and cable Pressure 7" w.g.for 1 hour RER pumpe, heat exchanger, controle, Rel. Utssidity earo inches v.g.for 6 months ) instrtementation and electrical Duration 100% R.E.for 1 hour ] equipment O) 901 R.E.for 6 months i Core spray system isolation valves, including operator and cable Core spray pumpe, controle, i inattumentation and electrical equipment U) y; l r 0 I Valves not required to be operable but must not fail open under the following conditions w o g l g Reactor water cleanup isolation Temperature (4) 148'? for 1 hour (1) 3 4 velves, inclisding operator and Fressure 7" w.g. for i hour Q) g j cable Rel. hissidity 100% R.R. for 1 hour (1) g Duration CD E n m 6 HFCI system teolation valves. Temperature (4) 148'Ffor 1 hour Q) 0 n, g including operator and cable Pressure 7" w.g. for 1 hour Q) ,i e RCIC system isolation valves, Rel. Htssidity 100% for t hour (1) o a including operator and cable Duration Z " 7 1 Main stems teolation valves x - -1 in eteen tunnel including 5 operators. i e-I

GENER AL h ELECTRIC J.TCMIC POWER EQUIPMENT DEPARTMENT

s. s c. ~* 2?A3019

"""o1 g-., (.. DESIGN SPECIFICATION = ~ w o. 12

      • '*"'"'"13 Section A-III (continued)

Notes (1) 148'F,100 percent R.H., ar.d 7 inches static pressure may occur concurrently for the 1 hour as given, but R.H. and static pressure will decay af ter this (! period. (2) Motors rated for continuous operation in an ambient temperature of 104F will operate in a higher ambient temperature with decreased life expectancy. Space cooling may be required to limit the ambient to an acceptable level. (3) Temperature based on RIIR equipment operating. (4) Temperatures given do not take into account any temperature rise caused by direct steam impingement. g (5) Equipment unable to withstand 100% R.H. and elevated temperatures (Motor Control Centers and Electrical Switchgear) for safety systems shall be located outside of secondary containment, or in separate rooms ventilated independently from the remainder of the reactor building. b essuto: ,4. ggw-. s h e 'n.r T* v-5 " W 7 'l N ' ~ ~ # ^

p p (S ~~. e RADIATION ENVIRONMENTAL t'UNDITIONS Section B. I. Inside Primary Containment C .n e Equipment or Area Radiation 1) Operatir., Dose Rate Design Basis Accident 2) Integrated DosI 9 ] g, Type ' Plant Oper System Oper Type Dose Rate Norr.a1 Accident z E c3 n m a Drywell, No Gansna 6.5 x 104 LOCA 1.3 x 106 2.3x1010 2.6x107 E m: 7essel Shield Neutron 6.3 x 107 7.9x10.6 9" 1 ge With Vessel Shield SQ Zone 1 Gamma 25.0 m. LOCA 1.3 x 106 8.8x106 2.6x107 E $# Above Core Neutron 5 x 10' 6.3x1013 2m O m Zone 2 Gamma 50.0

  • LOCA 1.3 x 100 1.8x107 2.6x107 Q"g Co-e Region Neutron 1.4 x 105 1.8x1014 A as 2

~n Zo e 3 Gamma 7.2 4.

  • 0CA 1.3 x 106 2.5x10g 2.6x107 E

Unvir Vessel Neutron <1 cl.3x10 Zone 4 Gamma 25.0 k 'LCCA 1.3 x 106 8.8x106 2.6x107 t. Near Recire. Neutron 2 x 103 2.5x1012 Zone 5 Camma 4.0 h LOCA 1.3 y 100 1.4x10b 2.6x107 >l5 ft. Recirc Neutron 2 x 103 2.5x1012 Zone 6 Gamma 0.1 b~'LOCA 1.3 x 100 3.5x10 2.6x10/ Torus Neutron 2 x 102 g{ .-. 5 N

  • LOCA Loss of Coolant Accident

{ = 7

1) Gansna Dose Rate Rads (Carbog)/ hour

- 100 percent load factor at rated power O C:: Neutron flux Neutrons /cm -sec j 2) Camma Dose Rads (Carbon) e rr {-LOCAAnt.lysiswasbasedontheassumption h Neutron fluence Neutrons /cm2 (NVT) that 100 percent of the noble gases, 50 e. Normal Conditions Integrated Oyer 40 years Percent of time halogens, and 1 percent

{

2 Accident Conditions Integrated over 6 months) f the solid fission products were re-gu leased from the core i; c>

9

O.,,.. D. g RADIATION ENVIR0hWDrrAL CONDITIONS 5! Section B. (Con.) II. Inside Secondary Containment wvN Equipment or Area Radiation

1) Operating Dose Rate Design Basis Accident
2) Integrated Dose 2

Type Plant Oper System Oper Type l Dose Rate Normal Accident 9 8 5 z E o 1 General Y1oor Area Gamma 0.001

  • LOCA 6.5 x 102 3.5 x 10 1.7 x 105 0*

2 9m ~ w 5 >'~ l HPCI Gamma 0.015 0.200

  • LOCA 1.6 x 102 3,3 x 103 4.5 x 104 0

and RCIC Area 'l m RER and HPCS Gama 0.015 0.030

  • LOCA 1.6 x 102 5.3 x 103 4.5 x 104 m"

n m, Area y 5m 24" Pipe Cama 0.0

  • LOCA 1.4 x 104 0.0 7.9 x 105 gn Containing Torus 5

Water (Typf al Pipe) k Cleanup Systems a) Heat Exchanger Gama 15.0

  • LOCA 6.5 x 102 8.76 x 10(

1.7 x 105 b) Pump Room Gama >0.05

  • LOCA 6.5 x 102 1.8 x 104 1.7 x 105 c) Filters & Tanks Gama 10.0
  • LOCA 6.5 x 102 3.6 x 106 1,7 x 105 Steam Tunnel Gama 5
  • LOCA 1.6 x 102 1.8 x 106 4.5 x 104 I%

Rod 2.5 x 102 >2.5 x 102 5 Drop r?

  • w i

Standby Gas Gamma 0.001

  • LOCA 5.7 x 105 3.8 x 104 CE Treatment System a

rro E U =*

1) Gamma Dose Rate Rads (Carbon)/ hour

- 10 Percent load factor at rated power. e. pg 2 Neutron flux Neutrons /cm -sec e LOCA Ana1 sta was hased on the assumption i ~ 7 1 Rads (Carbog)(NVT) that 100 percent of the noble gases, 50 iS

2) Ca=ma Dose 09 Normal Conditions Integrated over 40 years l Percent of time halogens, and 1 percent

@g Neutrons /cm Neutron fluence Accident Conditions Inteersted over 6 monthes of the solid fission products were re-leased from the core.

R, p p\\ \\ l E, v RADIATION ENVIRONMENTAL CONDITIONS } Section B. (Con.) III. Turbine Building j Equipment or Area Radiation 1)0perating Dose Rate Design Basis Accident

2) Integrated Dose

} Type Plant Oper System Oper Type ' Dose Rate Normal Accident 3 i 5 o = E m i General Areas Ga=ma 0.001 4 x 103 Qm ) Protected by Sm " m Shields o r-Operating Floor Gama 0.005 - 77.0x104 5 3 General 0.020 E z m -4 r= o m Contact HPT Gamma 0.5 1.8x105 m n 5 -4 5m Contact LPT C== = 0.1 3.5x104 gn 5 { Equipment Bay Camrum 0.05 - 1.8x106 i (Heaters, con-5.0 densers, etc.) Steam Jet Air Gamma 15 R/hr 5.3x106 Ejector Condensate Treat-Cansna 10 R/hr 3.5x106 ment o 'U m

1) Camma Done Rate Rads (Carbon)/ hour P - 100 percent load factor at rated power y

i Neutron flux Neutrons /cm -sec - LOCA Analysis was based on the assumption O 2 ' JG

2) Cam =a Dose Rads (Carbon) that 100 percent of the noble gases, 50 o

Neutron fluence Neutrons /cm2 (NVT) percent of the halogens, and 1 percent I w Normal Conditions Integrated over 40 year of the solid fission products were re-IA c3 Accident Conditions Integrated over 6 months-Icased from the core.

  • ?

-o e, .w [* c-

\\ p, p A. p, / % v R 3 RADIATION ENVIRONMENTAL CONDITIONS Section B. (Con.) IV. Rad-Waste Building rn i n n i ~ Equipment or Area Radiation

1) Operating Dose Rate Accident Doa-
2) Integrated Dose Q

n Type Plant Oper System Oper Type Dose Rate Normal Accident 5 5 g z E e n m Control Room Gamma 0.001 3.5 x 102 h m:z: n m o = E ';,~ Valve & Fump Rooms Gama 0.020 7.0 x 103 f iO Storage Tanks Gama 20.0 7.0 x 106 gm (Unprocassed) ]m l Centrifuge Gamma 100.0 1 x 10 7 m O -4 l m = -e 5O i z l V. REACTOR C0hTROL ROOM f Control Room Gamma 0.0005 1.75 x 10 3 x 10 2 1* a r 100 percent load factor at rated power ?, l 90

1) Gamma Dose Rate Rads (Carbon)/ hour i

LOCA Analysis was based on the assumption ci r-Neutron flux Neutrons /cm -sec that 100 percent or the noble gases, to y 2

2) Gamma Dose Rads (Carbon) percent of the halogens, and 1 percent C

t of the solid fission products were re-S c{ Neutron fluence Neutrons /cm2 (NVT), m Normal Conditions Integrated over 40 yearsg i icased from the core. E'

  1. a Accident Conditions Integrsted over 6 months /

F' ~ o a CD z a r - 3 .a () m

  • r3

.N 4 I hnu SARGENT a1. UNDY ^* 4 2 E fJ G I N E C H S \\ s ,a n uo uns u nser L.C. PRON c u e c a r.o. e t t e ra o t s, o n o 3 t t LrPHont 3:2 2 6 o.2 000 4 SDI (til) -?23 Augunt 12, 1981 Project l'o. G139-30 The Detroit Edison Company 1:nrico Permi Atonic Power Plt.nt - Unit 2 [ Post-Accident Radiation Dones to Equipment; Phase I Update o The Detroit Edison Company Enrico Perni - Unit 2 Project Docte.cnt Control Of fice - 110 S.D. 2000 Second Avenue Detroit, Michigan 48226 Attention: Mr. R. J. Beaudry Roon No. 310 ECT Referencec: 1. John S. Drtis to R.J. Ecaudry,"Pont-Accident Radiction 11oses to Ecuipnent, Phase I Recults Updato", Letter !!o. SL'i(NI)-193, Dated June 10, 1981.

2. John S.

Drtis to R.J. Deaudry," Post-Accident Radiation Dosen to Equipment: Phase II Results", l* Letter No. SL'i(NI)-207, Dated July S, 1901. Attached for your unc in en update of the tr.ble yhich van trans-mitted in reference 1. As we have discussed, a change in cur method of nodeling the doses due to pipes carrying pcst-LOCA radioactivity resulted in our reporting slightly hic:her (.oses l in the final Phase II results (reference 2) then were reported in the Phase I results (reference 1). (Specifically: the l Phase II ovaluation reflect the increase of dose along the 45 bisect of a 900 bend in the line over that which vould bc l calculated for a totally straight line. The 2ffect of taking this into account is an increase of epproxinately 50';.) l 1 d o Ex .. A

~ SAftGENT (k LUNDY // E N C. I N L C H S cmcaco j N p. Hr. R. J. Beaudry August 12, 1981 The Detroit 1;dison Company P a.,g e 2 4 Please place the attached table in your filen and mark the earlier version as obcolote to avoid any confusion over this adjustment. If there are any questions, please let me P.now. Yours very truly, J.S.DRTIC John S. Brtis Supervisor Shielding & Radiological Safety Section JSD:cim In Duplicate Attachant Copics: M. G. Sigetich (1/1) ~ F..Gregor (1/1) C. Seibert (1/1) Q. Duong (1/1) F. Featham (1/1) J. S. Loomis (1/1) G. P. Lahti (1/1) F. P. Tsai (1/1) NSLD File: 4C-17-Al 9 O e 9 4 9 9 t 9 ,9 @mti.;fN Qg.i u- _7 m

fC'sl.. ~ lgf, SIX MONTII POST-LOCA EOUIPMENT RADIATION DOSES ,8 NUTECH Calculated Area Designation Elevation Columns Rows Dose (RAD) SGTS Cubicles 3.1.1 677'-6" F to G 12 to 17 1.6 x 10 8 HVAC Area 3.1.2 677'-6" F to H 9 to 12 5 x 10 3 HVAC Area 3.1.3 677'-6" G to H 12 to 17 1.5 x 10 I Refueling Floor 3.2 684'-6" A to F 9 to 17 3.7 x 10 5 i. HVAC Area 3.3 659'-6" F te II 9 to 13 3.1 x 10 4 1 - Switch Gear Roon. 3.4 643'-6" F to H 9 to 11 3.~1 x 10 4 Rx. Bldg. 3rd Fir. 3.5 643'-6" A to F 9 to 11 2.4 x 10, 6 i 4 .Sw tch Gear Room 3.6 613'-6" F to H 9 to 11 3.1 x 10 Rx. Bldg. 1st Fir. 3.7 583'-6" C to G 9 to 11 5.4 x 10 b 6 Rx. Bldg. 1st F1r. 3.8 583'-6" C to G 13 to 17 5.4 x 10 b 6 West Corner kms. 3.9.1 562'-0" 6 5.4 x 10 b East Corner Rms. 3.9.2 562'-0" 6 5.4 x 10 b CRD Punp Room 3.10 562'0" G to H 9 to 11 3.1 x 10 b 6 Compressor Room 3.11 562'-0" G to H 9 to 17 3.1 x 10 4 East Corner Room 3.12.1 540'-0" 6 5.4 x 10 b West Corner Room 3.12.2 540'-0" 6 5.4 x 10 b i 6 HPCI Room 3.13 540'-0" G to H 9 to 11 5.4 x 10 b 6 Rx. Bldg. 2nd Fir. 4.1 613'-6" A to C '9 to 17 5.4 x 10 b i ~

'D- ' : 24. SIX MONTl! POST-I.OCA EOUTPMENT RADI ATION DOSES (Cont'd) ' -l$1, 4..\\ '? h Calculated" NUTCCH Area Designation Elevation Columns Rows Dose (RAD) 6 .9. Rx. Bldg. 2nd Fir. 4.2 613'-6" C to F 9 to 12 5.4 x 10 b 5 !0. Rx. Bldg. 4th Flr. 4.3.1 659'-6" A to B 11 to 13 3.7 x 10 b 5 !1. Rx. Bldg. 4th Fir. 4.3.2 659'-6" E to F 9 to 10 3.7 x 10 b 8 Y: 1.49 x 10 !?.. Drywell None Y +B i 1. 89 x 10' Y: 4.81 x 10 !3. Torus None 8 Y+S: 5.91 x 10 6 !4. Undesignated Rx. None All A to G 9 to 17 5.4 x 10 a,b Bldg. Areas 6 F to G 11'to 13 5.4 x 10 a,b 15. Steam Tunnel None The calculated dose to the internals of the hydrogen recombiners and other equipment 8 9 s. carrying drywell atmosphere is 9.4 x 10. The recommended dose is 1 x 10. T'ac calculated dose to the internals'of primary water carrying equipment is 3. 1.6 x 107 rad. The recommended test dose is 2 x 107 rad. The maximum dose calculated in this area. h a

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)* ^ E N C51 N C C l4 5 se t Asv mon aOr 9 tat ti CHIC A GO. IL LINot S 6060 3 L C.FRON T E L i e-H O N t 392 269 7000 AUG 241981 SLM(NI) 207 July 8, 1981 '~ Project No. 6139-30 The Detroit Edison Company Enrico Fermi Atomic Power Plant - Unit 2 Post-Accident Radiation Doses to Equipments Phase II Resulta The Detroit Edison Company Enribo l'ermi - Unit 2 Project Docur.cnt Control Office - 110 S.B. 2000 Second Avenue Detroit, Michigan 48226 Attentions !!r. R. J. Beaudry V Room No. 310 ECT =

Reference:

John S. Drtis to R. J. Beaudry, "Pcst-Accident Radiation Dcscs to Equipment: Phase II Results," Letter Number SLf!(NI)-177, dated May 28, 1981.

Dear Mr. Beaudry:

The post-accident radiation zone naps and the asoociated report (reference) have been revised to inco:porate your corm 3ents and changes found to be necessary during our review and approval process. A copy of each 40 attached for your use. The revised zone maps can be identified by tho "G-81" designation in their lower right hand corner. The earlier sets should no longer be used. e e .e 9 @O 9

k CARGENT& LUNDY E tJ G I N E E R S CHICAGO I Mr. R. J. Beaudry July 8, 1981 The Detroit Edison Company ,Page 2 If there,are any questions, please let me know. Yours very truly, DOHN S. bRTl3 John S. Drtis Supervisor Shielding & Radiological Safety Section JSBacjr In Duplicate Enclosure Copics: M. GL Sigetich (1/1) F. Gregor (1/1) C. Scibert (1/1) Q. Duong (1/1) M. Feathan (1/1) J. S. Loomis (1/1) G. P. Lahti (1/1) F. P. Tsai (1/1) HSLD Files 4C-17-Al (1/1) 4 5 e \\ e 9 l t @O ..~ ..az- .c -_n. ...u____,n_ c-

L C.FRON o t ( EQUIPMENT QUALIFICATION, RADIATION DOSE ANALYSIS FOR THE ENRICO -FERMI, UNIT 2 STATION ~ 1.0 Introduction The purpose of this report is to outline the sources and assumptions used in the NUREG-0588 post accident equipment qualification dose calculation, for the Enrico Fermi - Unit II station. Integrated beta and gamma doses are calculated for equipment exposed to airbor.de and aqueous mixtures of released fission products, in accordance with the guidance of NUREG-0588. 2.0 Shutdown Core Fission Product Inventories The core fission product inventories are calculated using the RIBD (Radio Isotope Buildup and De 'ay) subroutine of the RUNT-II computer program.1 A tabulation of the input parameters and assumptions are given in the. attached table, " Parameters and Assumptions for NUREG-0588 Analyse,s." [These parameters were the same as those used by General Electric to generate the shutdown activities supplied in Reference 2. The only difference is that 3430 MWth is used for this study instead of the 1 MWth used by GE. 3.0 Airborne Source Dispersion A schematic of the release model used to calculate the activity of airborne noble gas and halogen radionuclides in various parts of the Reactor Building are shown in Figure 1. Note that initial release assumptions for a BWR are not explicitly given in NUREG-0588, so reasonable extensions of the PWR model must be made. For this study it is assumed that 100% of the core noble gas inventory is initially airborne in the combined drywell and torus free volumes. In addition 25% of the core halogen inventory is assumed to be airborne in the drywell free volume and, 25% of the core halogen inventory is assumed to be plated-out o,n_drywell su_rfaces at_t=,0. The 25_%, plate,out_____ assumption, is _ conservative, though somewhat arbitrary since_ a realistic halogen deposition model for the drywell of a i BWR has not been formulated. The drywell and torus are assumed to leak at a constant rate of 0.5 volume percent per day throughout the course of the accident. Main Steam Isolation Value (MSIV) leakage does not contribute to the secondary containment source because of the pressurized leakage control system. 6

o When determining airborne concentrations in the reactor building outside the primary containment, it is assumed that radionuclides relcased from the. primary containment are completely mixed in the reactor, building free volume before being exhausted through the Standby Gas Treatment System (SGTS). Conversely, when calculating SGTS filter inventories and exhaust plume concentrations, it is assumed that there is no mixing in the reactor building free volume and that the sources go directly to the SGTS. Airborne radionuclide source activities are calculated using the Baffle portion of the RUNT-III program. In Baffle the core inventories calculated by RIBD are distributed throughout a multicompartment model of the sour;c regions described above. Time dependent buildup and decay of radionuclides are taken into account in Baffle. 4.0 Liquid Soitrees Baffle is also used to model the time dependent activity of halogens and fission solids mixed in the primary coolant and suppression pool water. A schematic of the release model used is shown in Figure 2. As mentioned, above, explicit release assumptions for a BWR are not provided in NUREG-0588 and thus a reasonable adaptation of the PWR methodology must be made. For the PWR case, the action of spray removal systems produc_s a net liquid source term composed of almost 50% of the core halogens and 1% of the core fission solids. Since credit for spray removal is not allowed on BWR's, no comparable mechanism exists for transporting airborne halogens into the suppression pool water. However, for this analysis the conservative assumptions of Regulatory Guide 1.73 (i. e., 50% of the core halogens and 1% of core fission solids mixed in the suppression pool water) are used. The RUNT-II computer code has a special option that allows noble gas daughter products evolved from halogen parents to be removed from the source region. This option is used in the analysis of liquid source terms. 50 Integrated Doses 1 Integrated doses are also calculated using RUNT-II A source geomgtry model is selected from those available in the ISOSHLD portion of the program and dose rates are determined at times at which source activities were calculated in Baffle. Total integrated doses are then obtained by a simple trapezoidal integration of the dose rates.

.___._.____.__._~..____.___.sa- -_m.-._ 0 Descriptions of the modeling used in several of the integrated dose determinations are given below*., 5.1 Drywell Centerline Doses from Airborne Nuclides Integrated gamma doses in the drywell from airborne noble gas and halogen radionuclides are conservatively calculated by treating the spherical portion of the drywell bulb as a ~ right circular cylinder with an equivalent interior volume. No credit is taken for shielding afforded by the sacrificial shield or from piping and equipment contained in the drywell. Note that gamma shine from halogens plated-out on the interior surfaces of the drywell also contributes to the centerline dose. The calculation of the plate-out shine dose is dis-cussed in the following section. Integrated beta doses in the drywell are calculated using a semi-infinite cloud beta dose model with the nuclide concen-tration in the cloud determined by mixing the assumed ncble gas release in the combined drywell and wetwell free volumes and the assumed halogen release in the drywell free volume only. NUREG-0588 indicates that an infinite cloud beta dose model should be used to determine beta doses at the drywell centerline. However, it also states that half of the in-finite cloud dose is appropriate for cables in cable trays to account for beta self-shielding. Since even relatively small thicknesses of dense materials provide complete self-shielding against beta particles, the semi-infinite cloud model is used. In addition to the semi-infinite cloud beta surface doses, beta depth doses are also calculated for a 10 mil thickness of 2 gm/cm3 material. NUREG-0588 allows credit for beta 3 attenuation in 10 mils of 2 gm/cm material when calculating exposures to surfaces in contact with the cloud in the drywell. 5.2 Drywell Centerline Gamma Dose from Plated-Out Halogens As mentioned in the previous section, halogens plated-out on the interior surfaces of the drywell also contribute to the drywell centerline dose. For this calculation it is assumed that 25% of the core halogen inventory is instantaneously plated-out in the drywell at time zero. This source is assumed to be uniformly distributed over the steel and concrete surfaces within the drywell. The total plate-out area is given in the EF-2 FSAR in Table 6.2-8. Because the exact orientation and position of all the plate-out surfaces in the drywell are unknown, it will be assumed that the shine dose at the center of the drywell can be approximated by the calculated dose from a spherical shell source e

O 4_ whose radius is equal to the interior radius of the drywell bulb. Again, shielding afforded by structural materials and piping inside the drywell is ignored. In ISOSHLD the dose from the spherical shell source geometry is de,termined by using an equivalent point source located at the sphere's radial distance from the dose point. 5.3 Drywell Wall Surface Doses from Plated-Out Halogens and Airborne Sources In addition to contributing to the drywell centerline dose, the halogens plated-out in the drywell also contribute to surface exposure doses. For this study the surface dose contribution from plated-out halogens are conservatively calculated by using an infinite plane source model with the source strength per unit area determined by spreading 25% of the core halogen inventory unformly over the total drywell plate-out area. In ISOSHLD both beta and gamma depth doses are calculated at points 10, 30, and 50 mils below the surface of a 2gm/cm3 material. Since the beta and gamma point kernel dose equations for an infinite plane source are singular for zero depth, it is not possible to directly determine surface doses using ISOSHLD. Surface doses are extrapolated from depth dose results. ~ The contribution to the gamma dose from airborne noble gases and halogens is assumed to be half of tL drywell centerline value, while the semi-infinite cloud beta dose is assumed to apply at both locations. 5.4 Torus Immersion Doses from Airborne Noble Gases Immersion doses from noble gas radionuclides airborne in the torus air space are calculated by modeling half of the free volume in the torus as a circular cylinder. The length and radius of the cylinder are determined by the amount of the airborne source actually seen by a dose point located anywhere on the torus centerline. No credit is taken for shielding afforded by piping or other equipment contained in the torus air space. As noted in the source term development discussion, the initial release of 100% of the core noble gas inventory is assumed to be uniformly mixed in the combined drywell and torus free volumes. None of the postulated halogen release to the drywell is assumed to enter the toros air. O

o ~5-5.5 Imm?rsion Doses from Halogens and Fission Solids Released into the Suppression Pool Immersion doses from halegens and fission solids released into the suppression pool are calculated using the same cylinder model as was used in determining the air immersion doses in the torus. The only differences are that water is the attenuation medium instead of air and that source dilution factors are based on liquid rather than air volumes. The initial release of 50% of the core halogen inventory and 1% of the core fission. solid inventory is assumed to be mixed in the combined primary coolant and suppression pool liquid volumes. Note that beta doses are semi-infinite cloud values. 5.6 Refueling Floor Centerline Immersion Doses Immersion doses at the refuleing floor centerline from airborne noble gas and halogen radionuclides that have leaked from the drywell into the reactor building are estimated by modeling the volume above the refueling floor as an air filled cylinder with a height equal to the distance from the operating floor to the roof of the reactor building. Leakage from the drywell is assumed to be completely mixed in the entire volume ex-hausted by the Standby Gas Treatment System (SGTS). No credit is taken for shielding afforded by equipment resident in the refueling volume. ~ 5.7 Shine Through the Drywell Wall into the Reactor Building Gamma shine through the drywell wall into the reactor building from airborne noble gases and halogens is estimated by modeling the drywell bulb as a spherical source region surrounded by a concrete shell. The sphere radius is taken to be the interior radius of the drywell bulb, 33', and the shell thickness is taken to be the minimum drywell wall thickness, 6'. No credit is taken for source attenuation due to structures within the drywell. Doses are calculated at a point l' outside the concrete shield wall. 5.8 Dose from Halogens and Particulates Trapped on the SGTS Filter To calculate the gamma shine dose from halogens and particulates trapped on the charcoal filter in the Standby Gas Treatment System (SGTS), it is aasumed that all radionuclides captured in the SGTS are deposiv.ed uniformly throughout the charcoal mass. In ISOSHLD the charcoal is modeled as a rectangular prism filled with a matrix of carbon and source nuclides. N'T'N $-,-c .[, M' r e,

No credit is taken for shielding afforded by casing materials surrounding the filter region. It is also conservatively assumed that nuclides leaking from the'drywell are immediately exhausted through the SGTS without mixing in the reactor building free volume. 5.9 Shine from SGTS Filter Through 5' Thick Shield Wall The shine dose through the concrete wall that surrounds the SGTS filter trains is calculated using the same model used to determine the dose from ur. shielded SGTS filters. The only differences are that a 5' thick ordinary concrete wall is placed seven feet from the center of the filter and that dose points are located outside the wall. 5.10 Doses from Unshielded Pipes Containing Halogens and Fission Solids Gamma shine doses from unshielded pipes containing halogen and fission solid radionuclides that have been mixed with the pri-mary coolant and suppression pool waters are calculated using a simple, water filled cylinder geometry. Pipe wall thicknesses are based on Schedule 40 pipe specifications, while the pipe length is arbitrarily taken to be 15'. Nominal pipe diameters ranging from 4" to 24" are examined with dose points positioned from 1 inch to 50' away from the pipe. 5.11 Shine from 16" Pipe Through 1.5' Thick Wall Gamma shine doses from a 16" Schedule 40 pipe that is shielded by a 1.5' thick ordinary concrete wall are calculated using the same ISOSHLD models used for the unshielded pipe cases described previously. The only differences are that a 1.5' thick shield is placed 3.5' from the center of the pipe and that dose points are placed outside the shield wall. 5.12 Plume Immersion Dose from SGTS Releases To calculate an immersion dose from the radionuclides released from the SGTS vent, the source nuclides are assumed to be i uniformly distributed in a large, finite cylindrical volume adjacent to the wall of the reactor building. The volumetric source strength is determined by the SGTS release rate and a simple x/O based on building wake mixing. The radius of the cylinder is related to the reactor building cross section area, while its length is taken to be six times the distance O

o ~7" from the recctor building wall to the dose coint, i.e., 6x50'=300'. No credit is taken for mixing in the reactor building prior to release through the SGTS. 5.13 " Shine Doses from Refuleing Floor Volume to fy.terior Points If complete mixing in the reactor building free volume is assumed, the airborne noble gases and halogens contained in c the volume above the refueling floor will contribute to the expesure doses received by equipment located outside the readtor building. To account for this source, the volume above the refueling floor is modeled as a rectangular prism. An ~ exterior dose point is conservatively assumed to be located on the same elevation as the center of the refueling floor volume. No credit is taken for shielding afforded by the wall of the reactor building above the refueling floor ele'ation. Dose points are positioned at 10 to 500 feet from tae wall of the reactor building to assess the impact of this source at various distances from the building. 5.14 Doses to the Internal Components of the Hydrogen, Recombiner The internal components of the recombiner are exposed to the undiluted drywell airborne source of noble gas and halogen nuclides. To estimate the dose to these components, the free volume within the recombiner heater box is conservatively modeled as an equivalent cylindrical source. This geometry ignores the actual piping layout in the heater box and neglects the dose contribution from the air filled piping outside the heater box. The radionuclide source concentra-tion in the cylinder is assured to ha the same as the airborne concentration in the drywell. 5.15 Shine Doses from the Hydrogen Recombiners To estimate the dose at points away from the hydrogen re-combiners from airborne noble gases and halogens passing l through the device, the free volume in the recombiner heater l box ic modeled as a sphere source surrounded by an iron a shell. The shell accounts for the metal casing and the pipe walls that define the actual air flow path through the recombiner. The contribution to the shine dose from piping located outside the heater box is accounted for by scaling the source strength such that the total activity in the sphere is equal to the total activity in all the source volume associated with the recombiner piping. Integrated doses are calculated at points 15" to 50' away from the sphere center. a, 7 -T' ~"'^;~'~.~' y+'"'",QT" FTi 3,Q 9

,8 t 5.16 External @line Dose to the Control Room Ventilation Equipment' Area ~, e The external shine dose to the control room ventilation equip-ment orca comes from airborne noble gas and halogen radio-nuclides in the reactor building. Shine from nuclides in the reactor building below the refueling floor level must pass through the l'8" thick reactor building wall before entering the equipment area, while nuclides above the refueling floor are only attenuated by the 4" thick roof slab over the equipment area. For the purposes of this calculation the shine through the reactor building wall is ignored. The shine through the root is modeled as a slab abutted to the reactor building wall above the refueling floor level with dose points located from 3' to 50' beyond the shield wall. 5.17 Dose at a Point Outside the Torus Liner from Airborne and Liquid Sources in the Torus Exposure doses to points outside the torus liner come from airborne noble gases in the torus and from halogens and fission solids mixed in the torus water. Since the gaseous and liquid phases in the torus occupy approximately equal volumes, doses outside the torus are calculated by treating each phase separately, while ennservatively~taking the net ~ ~ ~ ~ dose to be the sum of the cont ibhtions from each phase. The dose from the airborne nur iides is determined'by modeli'Eg ~ half the torus air volume as a long, rectangular prism source. An infinite slab model is used for the liquid dose calculation occause of the short mean free path for gammas in water. In both cases a thin iron shield is used to account for attenuation afforded by the torus liner. The dose point is one inch outside the torus liner. 5.18 Doses from 4" Diameter Hydrogen Recombiner Pipe Containing Drywell Air The gamma shine dose from an unshielded 4" diameter pipe containing drywell air is calculated using a simple, air filled cylinder geometry. The pipe wall thickness is based on schedule 40 pipe specifications, while the length in arbitrarily taken to be 15'. Dose points are positioned from 1 inch to 50 feet away from the pipe. 9 , 5lF ,,,,;wr-G

7

6.0 Radiation Zone Maps It should be pointed out that the enclosed zone maps are based on a conservative interpretation of NUREG-0588 for the Enrico Fermi Unit - 2 nuclear power plant. Doses to the equipment, integrated over a six month period postaccident, were calculated using the models given in section 5 of this report. These doses were then used to create a set of radiation zone maps for the integrated dose to equipment located in the Reactor Building and specified areas of tne Auxiliary Building. 6.1 Airbo_r,ne Sour ces in Reactor Buildir.g In determining the zones, consideration was first given to the airborne noble, halogen and particulate sources. The dose due to beta particles was calculated using an infinite cloud for-mulation which gave a value of 3.1 x 105 rad everywhere in the Reactor Building HVAC Boundary. The dose dud to gamma rays was based on a finite cloud immersion dose model for the refueling floor. Dosos to smaller rooms were scaled as the volume to the 1/3 power based on an unattenuated spherical source model. Values for various volumes are given in Table 1. 6.2 Shine Doses from Refueling Floor Airborne Sources The Shine Dose to the Auxiliary and Reactor Buildings were based upon the model given in section 5.16. The 4 inches of concrete shielding would keep these doses to a value of less than 104 rad which would render them insignificant. + 6.3 Doses from Unshielded Pipes Containing Radioactive Liouids Doses from pipe sources were based on nodels described in section 5.10. Values obtained from these models were used to generate a set of " Dose vs. Distance" curves for varying diame'.er pipes. These curves are shown in Figure 3. Doses for pipes of diameters not modeled were obtained from calcu-lations based upon the next larger pipe. Zones were defined by obtaining the highest zone that could be realistically shown (usually only to within 5 feet _of the source) on the plot. 6.4 Unsbielded Pipes Containing Radioactive Airborne Sources These doses were plotted in the same way as radioactive liquid sources except a different graph was used to find the Doses. This graph is shown in Figure 4.

~ 6.5 Shine Doses from the Hydrogen Recombiners The Doses from the Hydrogen Recombinch use the model described in section 5.15. These doses were plotted in the same way as the Radioactive Liquid sources except that the Doses were read from a diff

nt graph.

This graph is enclosed as Fi*gure 5. ~ 6.6 Shine from SGTS Filters Sources Doses from the SGTS filter bank were based on the model de-scribed in section 5.8. These doses were plotted in the same manner as those for Radioactive Liquids and the graph is enclosed as Figure 6. 6.7 other Sources Doses that were shielded by thick shields (i.e., greater than 5 feet of concrete) all prove to be insignificant. These included shine from the Primary Containment, SGTS filtere, the Torus, and Refueling Room shielded by Reactor Building walls and floors. The Doses from the SGTS Plume also was insignificant. The Primary Containment area, due to its large sources, was designation as Zone I. The Torus area, due to contact dose to the torus of 5.1 x 106 Rad, was designated as Zone _I. 7.0 Conclusions Postaccident doses were calculated for all equipment that could contain radioactive sources. These doses were combined to give the total integrated dose to any point in the Reactor Building and designated areas of the Auxiliary Building, and were then plotted on a set of General Arrangement Figures. To use these drawings, one should locate the area of interest an' 'ind the zone that you are in. If there are no highly ra active sources, (this can be determined by inspection of ,the 'EQ" set of figures), use the highest value of that zone as the dose value. If there are Highly Radioactive Sources in your area, you will have to read your Dose from the tables given for the Highly Radioactive Sources on each drawing. If, however, you are in contact with a radioactive source, you should use the Doses given under Notes 4, 5 and 6 of the first page of the figures. In this manner the dose to any equipment can be found.

i R3forences 1) Pichurski, D. J. " RUNT-II A Computer Program for Determining Time Integrated Doses," S&L Program Number RAC 09.8.034-2.20, December 1979 2) beneral Electric (GE) Letter to BWR Owner's Group, " Radiation Source Term Information for NUREG-0578 Implementation," November 1979 3) Regulatory Guide 1.7, " Control of Combustible Gas Conner.trations.in Containment Following a Loss-of-Coolant Accident," Revision 2, November 1978 4) Pichurski, D. J. "ISOSHLD A Computer Program for General Purpose Isotope Shielding Analysis; S&L Program Number ISOO9.8.029-2.20", January 1981 5) Enrico Fermi-2 Final Safety Analysis Report (EF-2 FSAR) t m.

0 L2 ss i SGTS Refueling Floor Drywell Torus & (Primary Suppression Containment) Pool W D Reactor Building Barrier 1 Filter 1 S urce Secondary SGTS L Containment L Filter p Con med y 2 1 r E 100%/ 0.5%/ m day day e l 100% Noble 25% Halogen 0% Jther Figure 1 Airborne Source Dispersion Model =- = t-"

o 0% Noble 25% IIalogen 0% Other Equilibrium Plate Out Core to Sources A I Drywell Surface 0% Noble 50% !!alogen 1% Other V Liquid Sources Primary Coolant plus Suppression Pool Water l l l Figure 2 Plate Out and Liquid Sources O An s=h : - %%... a~ d. e, u w,.- w % ' - - - w ^--^^^-N- --^-----'T'~ ^d -^- - - -7" ~ ^ ' ~

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May, 1981 Paga 1 of 4 DETROIT EDISON COMPANY, FERMI-2 FROJECT NO. 6139-30 PARAMETERS AND ASSUMPTIONS FOR NUREG-0588[AEALiS55[

  • 's, ITEM VALUE REFERENCES COMMENTS NSSS Power Level 3430 MWt FSAR p. ISB 6-37 For Source Developmeut Burnup 1095 days Reference 2 Full Power Operation Conversion Ratio

.250 Reference 2 ~ PU239/U235 Fissions 4 x 10 Reference 2 U-235 ABS. XSECTION 325 b Reference 2 Source Code RIBD Reference 2 Water Density 0.74 g/cc FSAR Fig. 4.3-1 Normal Operation Steam Flow 1.416 x 10 lb/hr FSAR p. 4.4-28 Normal Operation Core Coolant Flow 1.0 x 10 lb/hr FSAR p. 4.4-28 Normal Opera,tibn Steam Pressure 1035 psia 13 Core Ave. Thermal Flux 2.9 x 10 NEDO-20948 SOURCE REI. EASE Dryvell Airborne Nobles 25% Ref. 1* drywell plus Airborne Halogens vetwell free air 0% Ref. 1 volume Airborne Others 25% Ref. 1* Distributed over Placed out !!alogens the surface area of the Drywell u A non-mechanistic & conservative interpretation of NUREG-0588 guidance e

l. ~ i May, 1?M1 Page 2 of 4 PARAMETERS AND ASSUMPTIONS FOR THE NUREG-6'~88 ANALYSIS (Cont'd) 5 Item Value Reference Comments Wetwell Airborne Nobles 100% Ref. 1 ~ \\' Well plus wetwell Dilut,cd in the dry-1 Airborne Halogens 0% Ref. 1 free air volume Airborne Others 0% Ref. 1 & 6 Waterborne Nobles 0% Ref. 1 Diluted in the RPV Waterborne Halogens 50% Ref. 1 plus the suppression pool water volume for s Waterborne Others 1% Ref. 1 the systems drawing from the suppression pool Airborne Source Mitigation Suppression Pool Air Volume 130,900 ft FSAR p.6.2-4 Dryvell Free Air Volume 163,730 ft FSAR Tbl. 6.2-1 Total Primary Air Dilution Volume 294,630 ft From above Primary Containment Leak Rate 0.5%/ day FSAR p.15B.6-37 & Ref. 6 A Secondary Containment Free Volume 2,800,000 ft FSAR p. 6.2-106 Refueling Floor Volume 995,600 ft Ref. 3 Drywell Plate-Out Area 41,280 ft FSAR Tbl. 6.2-8 h STANDBY CAS TREATMENT SYSTEM SGTS Flow Rate 100%/ day FSAR p.15B.6-37 ...of secondary con-

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SGTS Filter Efficiencies 99% Iodines FSAR p. 15B.6-37 SGTS Filter Efficiencies 99% Particulates FSAR p.15B.6-37 SCTS Charcoal Weight 3600 lb FSAR Tbl. 6.2-11 3 b SGTS Charcoal Volume 140 ft FSAR Tbl. 6.2-11 SGTS Charcoal Bed Depth 6" FSAR Tbl. 6.2-11 7 i

May, 1981 PARAMETERS AND ASSUMPTIONS FOR Tile NUREG-0588 ANALYSIS (Cont'd) Item Value Reference Comments Waterborne Source Mitigation Reactor Coolant Liquid Volume 11,390 ft FSAR Tbl. 6.2-1 Suppression Pool Water Volume 117,450 ft FSAR p. 6.2-4 Total Liquid Source Dilution Volume 128,840 ft From above Plume Immersion X/Q 4.37x10 sec/m Ref. 3 Hydrogen Recombiner R'f. 3 Free Air Volume 11,800 in e Systems Assumed to be Contaminated Reactor Pressure Boundary to the Second Isolation Valves (Primary L;'. quid) Dryvell (Primary gases) i Wetwell (Torus Water & gases) ilPCI (Torus Water and Containment Atmosphere on Turbine Side) CSS (Torus Water) RilR/LPCI (Reactor Liquid) l RCIC (Torus Water and Containment Atmosphere on Turbine Side) CRD System (Torus Water) - Scram discharge header & holdup pipe flydrogen Recombiner (Primary Containment Atmosphere) Secondary Containment Atmosphere (Primary Containment Leakage & Evolution from ESF Equipment Leakage) Secondary Containment Drains & Sumps (ESF Equipment Leakage) SGTS (Secondary Containment Atn.osphere) Primary Sampic Systems (Reactor liquid,, gases) Torus Water Management System (Torus Water) 8

9 May, 1981 Pcg2 4 of 4 PARAMETERS AND ASSUMPTIONS FOR THE NUREG-05T48 ANALYSIS (Cont'd) ISOLATION All non ESF paths are assuned to be isolated with respect to the primary centainment at the second isolation valve. Other than SGTS & sampling, all paths out of the reactor building are assumed to be isolated. Isolation valve leakage will be ignored. RWCU is isolated post-LOCA. References 1. NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment". December 1.979 2. F. E. Gregor to F. Tsai, "Information Transmitted Concerning the Technical Support Center Design", Letter No. EF2-48409, dated March 12, 1980 3. Pichurski, D. J., "NUREG-0588 Post Accident Equipment Qualification Doses", EF2-TMI-EQ-04, March, 1981. 4. Notes of Meeting March 20, 1980, " Discussion of the Progress of the Post-Accident TMI Related k'ork, Being Performed by Sargent & Lundy." (attached) m 5. Enrico Fermi-2 Final Safety Analysis Report (EF-2 FSAR) 6. Notes of Meeting, December 18, 1980, " Equipment Qualification," Dated December 31,1981. (Attached) O [ .}}

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