Text

Radiation & Shielding Measurements During Initial Operation of Dresden Nuclear Power Station
	ML19340A517
Person / Time
Site:	Dresden
Issue date:	02/15/1961
From:	Craig W; GENERAL ELECTRIC CO.
To:	;
References
	GEAP-3404, NUDOCS 8008070702
	Download: ML19340A517 (72)
	v • d • e

{{#Wiki_filter:_ _ _ _. GEAP-3404 O - / W t. c w/t.' 8~1pMg /d >L. d7 ,.,e ? -- & ul4'M. to,/2.fQ WOJks 4 egg} RADIATION AND SHIELDING MEASUREMENTS DURING INITIAL OPERATION OF DRESDEN NUCLEAR POWER STATION By W. D. Craig ~ February 15, 1961 GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT ~ r ;N E ~~~ .\\ TTY EEM MIS 'i J F ' SAN JOSE, CALIFORNIA k. sooao2opog LSUl.ATORY 00CXET fil.E COPY .,. p1 p

RADIATION AND SHIELDING MEASUREMENTS DURING INITIAL OPERATION OF IEESDEN NUCLEAR POWER STATION By W. D. Craig February 15, 1961

1 PROPRIETARY INFORMATION NOTICE This material is prepared for the use of Commonwealth Edison Company, Nuclear Power Group Members, and General Electric Company in conjunction with Dresden Nuclear Power Station. It is released for general information and is not to be published or quoted in whole or in part without specific approval of the General Electric Company. l 1 .--..r.r. g + e

e e FC USE CF C=E EMPLOYEEJ CMLY GEN ER AL $ ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT TECHNICAL INFORMATION SERIES TsTLE PAaE NO.R-61 APE-25 AUTNOR susaECT CLAmelr: CATION GEAP-3404 W. D. Craig DRESDEN TESTING NUCLEAR oATE RADIATION MEASUREMENTS 2/15/61 TITLE RADIATION AND SHIELDING MEASUREMENTS DURING INITIAL npmATron nv mvRwn nitri.RAR PntJA RTATTOM ASETRACT This report summarizes the radiation and shielding measurements, and experience gained, during the initial operation of the 180 MW(e) Dresden Nuclear Power Station. G.E. CLASE. REPRODUCIBLE COPY FILED AT NO.PAGES TT ^ APED Library 69 oav. class. NONE CONCLUSIONS As the first large boiling vater reactor power plant, Dresden afforded the first comprehensive check of many radiation and shielding questiens. Rs.diation scatter-above and below the reacter vessel f0rced the addition of some concrete shield biccks. With this exception, the shielding as designed ve.a adequate, and conservative, for operation and maintenance of the plant. No shielding was required around the turbine. Measurements of neutron flux and gP-a heating in the ree.ctor vessel showed both to be vell vithin design criterie.. Residual radiation at reactor shutdown is nct a major facter in the main-tenance of the plant at this time. Good agreement with theoretical predictions was noted for Nitrogen-16 radia-tien, controlling access to steam equipment and primary water circuit. S1 nificant savings in shielding of the 6 reactor and plant eeui nent a e possible in the design of future piarts. bisseh upr da.tc. es.thered a_t Dremden. By cutting out this rectangle and folding on the center line, the above information can be fitted into a standard card file. For list of contents-drawings, photos. etc. and for distribution see next page (FN-610 2). IOMIC PCW R QUIPMENT DEPARIMENT INFORM ATION PREPARE FC. b Tears MAoE eV ! SECTION ENGINEERING C OU NTE RSIG N E 1 V. A. Elliott eUILDING AND ROOM NO. A 70 LOCATION S"" aIO 9 P - IP l i f0MI R ares-les Ci-st2

A G E N E R A L 't) E L E C T R I C Atomic Power Equipment Department TECHNICAL INFORMATION MEMORANDUM CONTENTS OF REPORT GEAP-34Eh No. of Pages 69 No. of Charts Drawing Nos. 142F822 142F823 142F824 142F825 142F826 142F827 142F828 142F829 142F830 142F700 142F701 142F702 142F703 142F704 142F(05 142F706 932C449 Distribution R. J. Ascher1 I. R. Kobsa Co=monwealth Edison J. A. Bailey R. G. Lorraine R. K. Baird A. J. McCrocklin H. K. Hoyt A. S. Bartu D. McDaniel J. H. Hughes - 2 R. O. Brugge R. A. Mickle N. A. Kershaw C. D. Carroll C. E. Morris G. L. Redman K. P. Cohen J. L. Murray C. B. Zitek F. E. Cooke V. D. Nixon Nuclear Power Group - 16 W. D. Craig J. F. O'Mara APED Library - 7 V. A. Elliott E. R. Gwen W. H. Ellis G. R. Parkos C. F. Falk K. T. Perkins M. A. Freemon F. C. Rally J. Forster R. C. Reid L. E. Foster R. B. Richards

0. J. Foster B. F. Rider R. B. Gile T. F. Robinson J. A. Haaga G. M. Roy W. A. Hartman G.

Sege G. L. Helgeson F. C. Skopec F. A. Hollenbach W. R. Smith D. H. Imhoff T. Trocki J. Jacobson G. Urabe W. R. Kanne M. L. Weiss ATPE 6A

4 ACKNOWLEDGEME1CS The author wishes to recognize the efforts of R. A. Mickle, General Electric, in assisting in the preparation and the performance of the Dresden radiation and shielding tests. Acknowledgement is also given to R. K. Baird and M. L. Brewer, General Electric, and to J. H. Hughes, J. McIntyre, J. McAsey, W. Kaiser, and R. Pavlik of Commonwealth Edison for their assistance in the performance of the tests. r I 1

JAQkgQfQQffffgg Page No. I. Introduction 1 II. Summary and Conclusions 3 III. Radiation Measurements 5 A. Reactor & Enclosure Surveys During Operation 6 B. Turbine Building Surveys During Operation 20 C. Shutdown Radiation Surveys 27 D. Environs Surveys 29 IV. Discussion of Measurements 30 A. Reactor Surveys 30 B. Turbine Building Surveys 33 C. Radiation Measurements at Reactor Vessel 38 D. Shutdown Radiation Measurements 45 V. Appendix 47 A. Plant and Elevation Drawings of Reactor Sphere k8 and Turbine Building B. Neutron Flux Capsule Carrier Assembly Drawing 64

1 i l I. INTRODUCTION This report summarizes the radiation and shielding experience gained during the startup and initial operation of the Dresden Nuclear Power Station. In this report are the first possible comparisons of measured dose rates versus predictions for a large boiling water reactor power plant. As in many other areas of Dresden's technology, the program of radiation measurements during startup and initial operation was necessary for three principal reasons: 1. A proof test of Dresden shielding adequacy, and/or determination of required shield revisions. 2. A gauge of the accuracy to which BWR, shielding calculations may be made, as a guide in the design of succeeding plants. 3 Extension of basic shielding theory by measurements of those parameters upcn which little information exists. In order to satisfy these three requirements, a schedule of dose measurements was made for the several plant areas in which radiation vould exist. This schedule called for general surveys, and also for particular measurements of points of interest. A copy of this' schedule appears in the Appendix (Section VI) of this report. Points of specific interest in the measurements included: A. Reactor Enclosure Radiation 1. Measurement of direct radiation through reactor shields. 2. Determination of the radiation scattering problems above the top head, and below the bottom of the reactor. 4 --

3 Measurements of radiation fran primary water system components. B. Turbine Building Radiation 1. Determination of required. turbine shielding, and/or maintenance access problems. 2. Measurement of Nitrogen-16 concentration in off-gas system and in condenser hot-vell. C. Radiation at Pressure Vessel Wall 1. Determination of gamma heating at vessel vall by calorimetry. 2. Measurement of neutron flux and spectrum at vessel vall. D. Shutdown Radiation Measurements l l.

II. SLDOUCY AND CONCLUSIONS A. Shielding and radiation protection at Dresden Station is adequate for safe operation, maintenance, and inspection of the plant. B. All shields designed to protect against direct radiation in the Dresden plant are conservative, and such sbields are being reduced in future plauts. Shielding technology for large boiling vater reactor power plants is now well established. Concrete thicknesses actually needed around representative components of the Dresden system, in order to reduce the dose rate to less than 0 5 mrem /hr,are: Turbine - None (with controlled-access area approximately 30-150 feet around turbine pedestal) Condenser 1/2 - 3 feet Air Ejector and Off-Gas System - 4 feet Feedvater Heaters - 2 4 feet, varying from first to last stage Feedvater Pumps - None PrimarySteamPipes-3-1/2 feet Secondary System - None Feedvater Piping - None Condensate Demineralizers 2-1/2 feet Primary Recirculating Water System and Secondary Steam Generators 4 ft. PrimarySteamDrum-4-1/2 feet Cleanup Demineralizer 3-1/2 feet C. The influence of scattered radiation, and radiation streaming, is shown to be a factor in the design of large reactor power plant design. D. The buildup of corrosion product activity with operating time in the primary system may be a major factor in influencing maintenance pro-cedure of recirculation loop components. More operating time must be logged before meaningful predictions may be made. E. Nitrogen-16 radiation, controlling plant shielding and access (during operation) to steam ead recirculating water equipment, conformed reasonably to theoretical predictions. A significant example is the turbine operating floor, which was shown to require no shielding for routine inspection and maintenance. F. Measurements were made of neutron flux and gamma heating just outside the reactor pressure vessel vall. The 40 year neutron irradiation exposure at the inside radius of the vessel vall, on core midplane 19 elevation, vill be about 1.6 x 10 neutrons /cm (nyt) from neutrons of greater than 0.1 Mev energy. This compares favorably with APED 19 design criteria permitting no more than 5 x 10 nyt ( ) 0.1 Mev) in the design life f a pressure vessel. Gaw heating in the vessel steel is a maxi =um of 23,500 BTU /hr-ft.3 (0.03vatts/gn). This results in a calculated maximum tangential thermal stress of about 2,200 psi, tensile on the inside radius of the vessel. Since 10,000 psi is allowable under ASME Boiler and Pressure l Vessel Code Case 1234, this is quite cons-m tive. .k-I I

III. RADIATION MEASUREMENTS Gamma dose rates are entered in units of milliroentgens per hour, neutron dose rates in millirem per hour. Since the conversion from Sa-a milliroentgens to millirem is unity, the neutron and gama dose rates are additive to units of millrem per hour. Neutron dose rates are followed by the designator "n". Gamma dose rates are undesignated. Gamma measurements above 1 mr/hr were made with " Cutie Pie" portable ionization chamber ratemeters. Gamma measurements less than 1 mr/hr were made with portable Geiger counter-ratemeters and with precision dosimeters. fieutron measurements at 10, 22 5, 35, and 42 megawatts were made with portable BF3 counter-ratemeters and with Rudolph fast neutron ratemeters. Above 42 megawatts, neutron surveys were made with a double-moderator neutron dosin.eter. Unless otherwise noted, measurements vt. made at a distance of 1 meter from the source surface.

For Description See

" Double Moderator Neutron Dosimeter," J. De Pangher, HW-57293, July 15, 1958. i

A. Reactor and Enclosure Surveys During Operation (See corresponding figures showing plan views of survey area and plant drawings in Appendix) 1. Special Radiation Survey Sheet No. 1 (Refer to Drawing 142F822 in Appendix) Reactor Power (MWT) 309 315 626 Dual Single Dual Point 10 22 5 42 Cycle Cycle Cycle 1. Primary steam drum, 80 5000-8000 (Est. ) east-center 22 n 2. Primary e,eam drum, 20 2000 ( st.) south-end 25 n 1200 n 3 Riser at southeast of 120 steam drum 5n ~ k. Outside steam drum 0 05 enclosure, E1.634 0.02 n 0.1 n south 5 Above steam drum 0 25 enclosure, EL.649'6" 0.002 n 2n,

. = i 21 4 5 6 I i 20 i 3 N 2 16 11 22 7 18 ~ l 8 19 19 T 14 12 11 10 13 RADIATION MEASUREMENTS Af ELEVATIONS 565 ft. & 584 ft. (For greater detail, see Drawing 142F823 in Appendix) _7

. = _ _ i 1 i ) I j l 16 13 l'.3. 7 l 1 l 12 i l '~ N 11 4 I __ a 4, 10 r::. =:.- E k r, L i 9 8 7 6 \\ / 2 i RADIATION MEASUREMENTS AT ELEVATION $48 ft. 1 l (For greater detail, see Drawing 142F824 in Appendix) 1 1, I f

2. Special Radiaticn Survey Sheet No. 2 (Ccatinued) Reactor Power (MWT) 309 315 626 Dual Dual Dual 10 22 5 35 42 cycle cycle cycle 21. E1 565, 15' east of Rm. H 0.8 0 5 1. n 0.1 n 22. El. 584, above reactor top 5 75 125 head 46 n 18 n 35 n Surveys at 1, 22 5, and 35 MWT made before installation of reactor top heat shielding beams. After the 35 MWT survey, a 3 foot thickness of concrete shielding beams was installed above, and to the south of, the reactor top head to completely enclose reactor. l

2. Special Radiation Survey Sheet No. 2 Reactor Power (MWT) 309 315 626 Dual Dual Dual lo 22 5 35 42 cycle cycle cycle 2. El. 565, north or reactor o.3 o.6 1 o o o o 35a 6.5 n 9 5 n o.1 n o.25 n 0 58 n o.8 n 4. El. 565, east sec. shield o o o vall o.1 n o.1 n o.1 a 6. El. 565, east sec. shield 0.2 05 o o o vall 1 n 1.8 n 0.1 n o.1 n o.1 n 7 El. 565, south canal vall o.2 05 o o 0 o 1n 31n o.1 n o.1 n o.1 n 0.1 n 9 El. 565, elevator entrance o.1 o.4 o o o 05n 4n o.1 n 0.2 n o.2. n 12. El. 565, vest sec. shield 0.2 05 0.2 0.2 03 vall 1n 4n 0.1 n o.1 n 'o.1 n 13 El. 565, over sec.stn. gen. o.2 0.4 o 0.2 o o D 1n o.1 n 0.1 n o.2 n o.1 n 15 El. 565, vest sec. shield 0.2 0.2 0.2 0.2 vall 1n 0.1 n o.1 n o.1 n 16. El. 560 '-6", bottom of stairs o.2 o.2 0.2 0.2 below vire guns 2n o.2 n o.4 n o.4 n 17 El.565, vire scanning o.4 0.2 03 03 platform 4n o.2 n 0.2 n o.3 n 18. El. 560 '-6", center of leak o.4 o.5 o o o-detector room 2n 56n 4n 0.1 n o.1 n o.2 n 19 El 565, in instrument Rm. H o o 0.1 n 0.1 n 3 Special Radiation Survey Sheet No. 3 Reactor Power (MWT) 309 Dual Point 10 22 5 35 42 cycle 1. El. 549, SW corner of O O.4 0 0 valkvay 0.2 n 1.4 n 0.1 n 0.1 n 3 El. 549, vest walkway 0 0.4 0 0.2 n 05n 0.1 n 5 El. 549, NW corner of O O valkvay On 0.1 n 8. El. 536'-6", Platform 0.4 0 vent 13n 0.1 n

11. El. 548, Platform O

O.4 O. south 03n 2n 01n 13 El. 549, Inside SSG Room (These measurements reported on Special Radiation B on elevated platform Survey Sheet No. k) 15 El. 548, Inside Nuclear O O Steam Inst. Room C On 0.1 n 17 El. 549, North valkway 0.2 0.4 0 1 n 1.8 n-O. 15 n See note at bottom of data, Survey Sheet No. 2. + .h -u- .,-1. 6 -

I 7 8 i 6 3 O>> q_ 4 1 9 .Lo 11 \\ / RADIATION MEASUREMENTS AT ELEVATION 529 - 517 Ft. l (For greater detail, see Drawings lh2F825 and 142F826 in Appendix),_ _

k. Special Radiation Survey Sheet No. 4 Reactor Power (MWT) 309 315 626 Dual Dual Dual Point lo 22 5 35 42 cycle cycle cycle '2. El. 529'6', North, below 0 0.4 0 0 o O crane hatch o.25 n 1,3 n o.1 n o.22 a 0.2 n o.3 n 7 El. 533, Downcomer SSG 22 5 900 1000 1800 Room A 15 n 15 n 27 n 8. El. 533, SSG A 7 llo 120 250 11 n 20 n 10. El. 529'-6" outside south o O O sec. shield vall o.1 n 0.2 n El. 535, Ext. Ion Ch. O O O ~O ~ Guide Tube #1 30 n 200 n 1.6 a 25a El. 517, Between Primary 300 300 and Secondary Shields 1600 n 240 n El. 517, NE, in corridor 1 1 outside door to pipeway 18 n 1 n vestibule See note at. bottom of data, Survey Sheet No. 2 l 1 1 i J 4 10 ~ 12. 7.~ l l la 13 ~ 4_5 6 7 3 l ~ .T N 2 r l i 1 . - -1 1 ik 1 / \\ 1 RADIATION MEAS'REMENTS AT ELEVATION 502 FT. (For greater detail, see Drawing 142F827 in Appendix) -14 4

5 Special Radiation Survey Sheet No. 5 Reactor Power (MWT) 309 315 626 Dual Single Dual Point 10 22 5 35 42 cycle cycle cycle 1. El. 502, corridor N of 0 0 05 15 05 scram valve room 1n 14 n 5n 1n 2. El. 502, corridor N of 0 0 0 05 15 1 scram valve room 2n 75n 21 n 5n 3n 3 El. 502, corridor E of 0 8.5 03 c.4 sub-pile room 25n 9n 0.2 n 0.2 n _ 5 El. 502, corridor E of 15 15 13 5 03 0.4 sub-pile roan 5n 17 n 37n 14 n 0.2 n 0.2 n 7 El. 502, corridor E of 0 05 0.2 sub-pile room 0.1 n 0.2 n 8. El. 502, Instrument 0 0 0.4 Room A, doorway 0.1 n 0.1 n 11. El. 502, Instrument 0 0.1 n Room A 12. El. 502, Doorway to 0.25 0 Cleanup Demin. A. 0.1 n 13 El. 502, Between cleanup 0.25 70 Demin. A & Reg. HX A 14. El. 502, Between Cleanup 0 2 200 Demin. B & Reg. HX B 0n 4n Between the 35 and the 42 MWT surveys, a partial thickness of shield blocks was placed in the control drive installation ports in the primary shield at El. 502. This was completed between the 309 and 315 MWT surveys. A laminated steel, cadmium and wood sliding door and a concrete labyrinth were placed in the passageway between the sub-pile room and the drive accumulator room, E. 488,' between the 35 and 626 MWT surveys. Between the 315 and 626 MWT surveys, a wooden neutron shield cribbing was placed around the control drive hydraulic line penetrations of the primary shield at El. 502. o.-._

,......_s, {.. ~ l / 3 11 -7 7' 5 0 N 10 12 8 13 6 2 N-/ 4 9 1 \\ Y. _/ N.,._..._......"' RADIATION MEASIREMENTS AT ELEVATION 488 FT. (For greater detail, see Drawing 142F827 in Appendix) ]

6. Special Radiation Survey Sheet No. 6 Reactor Power (MWT) 309 315 626 Dual Single Dual Point 10 22 5 35 42 cycle cycle Cycle i 1. El. 488, Cor:idor S of 0 0 0 0 Reactor 0n 0.1 n 2. El. 488, Corridor S of 0 0 0 0 Reactor 0n 0n 0.1 n 6. El. 488, Control Drive 0 0 0 0 Instrument Room On 0n 0n d 7 El. 488, Control Drive O O O Instrument Room On 0n 0n 8. El. 487, Doorway Between 15 7 100 90 h ~ Accum. Room & Sub-Pile 10 n 12 n 40 n 30 n 6n Room 9 El. 487, Corridor N of 0 05 2 03 Accumulat r Room On 15n 1n 10. El. 487, Corridor N of 0 2.0 25 05 Accumulator Room 0n 2.4 n 2.2 n 25n 12. El. 487, In Accumulator O 25 35 05 Room 07n 6n 4n 3n 13 El. 487, Below Reactor 10 70 1200 (Est) in Sub-Pile Room ko n 5000 n (Est) See note at bottom of data, Survey Sheet No. 5 - i i 39 3a 37 35 36 l i i L j 3h 33 _ ~~'" T - j t j 32 29 ~ ~ - I f 1 l li3 26 E 1-;. 4 31 28 f 25 a 1 L f i N 24 I 3o 27 4 N W 23 E Y j 15 1 1 L7 E 4 E } >\\ 12 0 10 7 J t f i 13 11 0 5 1 l t 9 4 4 3 ,2 1 RADIATION MEASlEDENTS AT ELEVATIolt 551 FT., 'RIRBINE BUILDING l (For greater detail, see Drawing 142F700 in Appendix) .. ~,.,.. - - -. _,. _, - -.

B. Turbine Building Surveys During Operation 1. Special Radiation Survey Sheet No. 8 (nanna Only) Reactor Power (MWT) i 309 315 626 Dual Dual Dual Point Cycle Cycle Cycle 1. El. 551, Stairs W. or Turbine o o 0.6 2. El. 542'-6" Stairs W. of Turbine 0 0 0.8 l 3 El. 534, Stairs W. of Turbine o o 0.8 4. El. 551, Turbine Floor West 0 0 1.2 5 o o 2. 6. O O.5 2.8 7 0 05 2.6 8. o 0 33 n n n n 9, o o 1 10. 0 0 03 11. O O 1 12. 0 0 2.2 13 0 0 1.4 14. El. 551,. Stair to Cond. Pump O 1 1 15 El. 531'-6" O 1.6 4.

16. El. 521 '-6" o

1.4 45 17 El. 551, Hatch 0 0.8 4 18. El. 551, Hatch above Con. Demin. 05 1.8 lo 19 El. 551'-6", Curb S of Turbine o.8 3 11-20. 4 13 29 5 21. 5 16 31-22. 65 19 39.

O Special Radiation Survey Sheet No. 8 Continued Reactor Power (MWT) 309 315 626 Dual Dual Dual Cycle Cycle Cycle 23 El. 551, Floor Joint 1 4 10 S of Turbine 24. 5 17 39 25 7 19 5 39 26. 17 5 65 88 27 E1 551, Turbine Floor 15 4 22 Over FWH Comp. 28. 45 11 36 29 6 15 40

30. E1 551, Turbine Floor at 35 7

26 Deck-Plate above WH Comp. 31. 55 15 ko 32. 11 5 13 5 45 33 E1 551, Turbine Front Std. k 11 19 t.ver Floor Grating 3k. 4 11 19 35 El. 551'-3" Stairs E. of 0 0 0 Turbine kO. Contact with H.P. Steam Line 35 190 300 entering turbine steam chest _21 2

If0ffffTiTI 6 E 5 7 N E k -,3 C C b E3 C 1 RADIATION MEASIREMENTS AT ELEVATION 531 FT., WRBINE BUILDING (For greater detail, see Drawing 142F701 in Appendix) 22

2. Special Radiation Survey Sheet No. 9 (Gamma Only) Reactor Power (MWT) 309 315 626 Dual Dual Dual Point Cycle Cycle Cycle 1. El. 531'-6" cond. Demin.' c. 17 50 75, 2. El. 531'-6" Window Penetration 2 10 14 to Demin. 3 El. 531'-6" Window Penetration 2 6 8 to Demin. k. El. 331'-6" Valve Gallery 0 0 0 5 0 0 0 6. 0 0 0 7 0 0 0 t 4 -

.- ~ 26 21 J Li,3 _.--, 39 Of N 37 L i 32 2c l 1.. lC 3 31 15 23 21 N 22 7J r-e O ~" e 19 M 10 6_ L._ l 14 10 ~' 16l _ [9 "L. 17 15 - b 3 L...d._, 3 2 1 I RADIATION MEAStRDENTS AT ELEVATION 495 FT. - 517 FT. IN 11)RBINE BUILDING (For greater detail, see Drawing 142F702 in Appendix) 24.

3 Special Radiation Survey Sheet No.10 (m-na Only) Reactor Power (MWT) i 309 315 626 Dual Single Dual Point Cycle Cycle Cycle 1. El. 517'-6" w. Corridor 0 0 0 2. 0 0 0 h. 0 0 0 5 E. 517'-6", N. Corridor 0 0 0 6. O O O 7 Off-cas 05 3-7- Sampler 1 8. El. 517'-6", S. Corridor 0 0 0 9 0 0 0 10. 0 0 0 11. " by NH 0 0 0 Comp. #1 12. 0 0 0 13 0 0 0

14. El. 517'-6" Pri. Feed Pump A 0

0 0 15 El. 514, Cond. Pump Rm. 0 0 0 16. 0 0 0 17 O O O 18. O O O 19 El. k95, Bel 5w Hot-vell 2 11. 12-50 range 20. El. 510'-6" Pri. Steam 235 390 800 Line

21. El. 517'-6" Air Ejector Valve 0

0 0 Sta.

22. El. 517'-6" N. Side Air Eject 5r A 245 0 Shut Down 800 23 El, 517'-6" E. Side 60 0

1000 - 2 5-

Special Radiation Survey Sheet No. 10 Continued Reactor Power (MWT) 309 315 626 Dual Single Dual j Point Cycle Cycle Cycle

24. El. 517'-6" Air Ejector Comp. A,N.

5 0 Shut Down 500 1 25 El. 517 '-6" ,E. 05 0 50 ) 26. El. 517'-6", Lube 011 Room 0 0 0 27 0 0 0 28. 0 0 0 29 El. 512, M Comp. 52 12 31: 120 31. El. 512, m, Comp. #3 28 130 30 32. 17 180 150 33 22 70 120 34. 12. 48 30 35 80 280 300-36. 70 180 200 37 El. 517 Trackway 0 0 0 38. 0 0 0 39 0 0 0 40. El. 517'-6", Off-Oas Monitor 0 0 0 41. El. 517'-6", Air Ejector Rm.B 1 380 1 42A. El. 512, M Comp. #2 15 25 80 42B. 11 14 80 05 30 11A. El. 512, M Comp. #1

C. Shutdown Radiation Surveys The data listed below are representative of dose rates prevailing in areas within shielded compartments which will require periodic access for main-tenance. Dose Point, and Time After Shutdown Dose Rate (mr/hr) 1. Reactor vessel top heat (contact) a. 2 5 br. after shutdown from 315 MWT, core exposure 5,000 MWDT* (5/1/60) 17 b. 24 hr. after shutdown from 315 MWT, core exposure 5,000 MWDT (5/2/60) 15 36 hr. after shutdown from 425 MWT, core c. exposure 37,600 MWDT (11/16/60) 5 (Core history before this shutdown: For 4 hrs. previous to shutdown, 425 MWT; previous 12 hrs. at 626 MWT; previous 12 hrs. at 425 MWT; etc.) 2. Below control drives, reactor vessel botta head (15-30 ft from vessel) a. 0 3 hr. after shutdown from 315 MWT, core exposure 5,000 MWDT (5/1/60) 4 b. 1 5 hr. after shutdown from 315 MWT, core exposure 5,000 MWDT (5/1/60) 3 28 hr. after shutdown from 315 MWT, c. core exposure 5,000 MWDT (5/2/60) 05 c. 52 hr. after shutdown from 425 MWT, 5 core exposure 37,600 MWDT (11/16/60) (Operating hinory as described in (le) above.

Megawatt-day (thermal) 3 Reactor bottom head,/60).

contact with control 10 driveflanges(11/16 (Core history as described in (lc) above) 4. Cleanup demineralizer room, between demineralizer tank and regenerative heat exchanger a. 1 5 hrs. after shutdown from 315 MWT, core exposure 5,000 MWDT (5/1/60) 130 b. 3 7 hrs. after shutdown from 315 MWT, core exposure 5,000 MWDT (5/1/60) 120 c. 27 7 hrs. after shutdown from 315 MWT, core exposure 5,000 MWDT (5/1/60) 35 d. 18 days after shutdown from 315 MWT, core. exposure 5,000 MWDT (5/19/60) ( Per F. Brutschy & R..Osborne, VE) 15 14 days after shutdown from 626 MWT, e. core exposure 12,500 MWDT (9/9/60) (Per R. S. Gilbert, VE). 60 5 Secondary steam generator, near bottom head a. 3 hrs. after shutd m from 315 MWT, 15 side core exposure 5,000 MWDT (5/1/60) 18 bottom b. 27 bra, after shutdown from 315 MWT, 1.2 side core exposure 5,000 MWDT (5/2/60) 2 bottom c. 18 days after shutdown from 315 MWT, 0.8 side core exposure 5,000 MW (5/19/60) 7 bottom (Per Brutschy & Osborne, VE) d. 14 days after shutdown from 626 MWT, 2 side core exposure 12,500 MWDT (8/8/60) 70 avg. bottom (Per R. S. Gilbert, VE) 36 brs, after shutdown from 425 MWT, core 230 bottom e. exposure 37,600 MM (11/16/60) (Core history as described in (le) above). D. Environs Surveys After addition of the top head shielding blocks, measurements at environs stations and survey points around sphere and turbine building during operation show dose rates to be quite low. At full power, the source of greatest influence on environs dose rates is the turbine, including crossover and moisture separator. Measurements taken with precision dosimeters located at various points on the roof of the aininistration building and access control building give dose rates of 0 3 - 0 5 mr/hr at 626 MWT, full steam flow. These are the highest dose rates at any uncontrolled environs tsrea, and are within the 05 mrem /hrlimitationofuncontrolledaccessareasatDresden. Dose rates in office areas below the roof are even lower. IV. DISCUSSION OF MEASUREMENTS A. Reactor Surveys Radiation surveys were begun at a reactor power of 10 Wt (April 1,1960). This and subsequent surveys during the next two days at reactor powers up to 40 W t showed that neutrons streaming above and below the reactor vessel would create untendable dose rates at increased reactor powers. These neutrons were escaping from the reactor vessel radially at the core elevation and were then streaming up and down the annulus between vessel and primary shield. Above the vessel top head, the radiation scattered from the vertical canal valls, and the valls and top of the craneway, finally escaping from the shield through the large crane openings at the north and the south of the shield (El. 588'). Once outside the shield, these neutrons were able to shover all elevations of the sphere above El. 529', by scattering from the - sphere shell itself. In addition, neutron dose rates were readily measurable in the areas around the sphere. Dose rates at a reactor power of about 22 5 Wt were: El. 565' North (Point 2, Survey Sheet 2) 6 mrem /hrneutrons (Highest floor designed to be accessible during reactor operation) 0.6mr/hrgammas El. 58h' Above Reactor Top Head (before 90 mrem /hrneutrons installation of shield blocks) 10mr/hrgammas Below the bottom head of the reactor, scattered neutron radiation into elevations 488' and 502' gave dose rates in the order of 50 mrem /hr at 25 W t in the control drive scram valve and accumulator rooms..

These conditions were corrected by: 1) The addition of a 3 foot thick a concrete shield over the top head of the reactor, 2) Plugging holes in the concrete biological shield.below the core, through which the control drive hydraulic piping passes into the drive valve room, 3) Adding a one foot thick concrete labyrinth and a steel sliding door between the sub-pile area (below the bottom head) and the control drive accumulator room, 4) Stacking concrete blocks to full primary shield thickness in the pipe passa6es through the primary biological shield, through which the risers and return recirculation piping passes, and 5) Stacking concrete blocks (2 feet thick) in large holes in primary shield,just below the bottczn head, through which the control drives were handled during their installation. A tabulation of neutron and gan:ma dose rates at the various power levels ~ appears in Section III of this report. Classically, radiation scattering from a vall follcws this empirical fomula: A / J / 6tAascos@ t = r2 c T 2 \\ 9**Ge p

where, sesttered flux measured at detector D

= incident flux on scattering area A = A scattering area = g( albedo (reflecticn or scattering coefficient) of vall area = A for scatter through included angle 4 tQ r distance from A to receiver point 2 l

4 For strict application, A <A r. This condition may be met by breaking the 2 scattering surface into parts. For the case of neutron flux escaping out the north crane opening in the shield, after scatter off the roof and (east and west) side shields, this condition may be met. The total of the 2 effective scattering areas is estimated to be 7,500 ft ; the average distance from scattering areas to the 565 elevation dose point was 50 feet. At 22 MWt, dose rate above reactor top head in the canal was 80 mrem /hr. At the same power, the dose rate at the 565 elevation north point was 6 5 mrem /hr. The average value of % is 30*. Then the albedo may be roughly 1 calculated as, 9ebY S'I = <;b; 4 cos R s G,. c x 2. 5x io = 8 Xio x 7. 5 X/o's d.16 6 J./%/o~

== for scattering off the concrete valls through an average included angle of 60*. As seen in the tables of dose rate measuremants, the dose rate above the reactor top head shield blocks (3 foot thickness of concrete) at full power (626 MWt) is 35 mrem /hr from neutrons, and 125 mr/hr from gansnas. At the dose point on elevation 565, north, the scattered dose rate at full power is0.8 mrem /hr. Average energy of the scattered neutrons is 0.6.Mev. During full power operation, dose rates in all areas within the sphere are lessthan1 mrem /hr(neutronsandgammas)exceptforthepointslisted below: t I 1. At the external ion chamber drive area (El, 535, norta), the ion chamt3r guide tubes are a source of neutron streaming. Dose rates at some tubes during half power operation were as high as 75 mrem /hr. This situation was remedied by insertion of cylindrical tube plugs with a length of 12 inches of polyethylene and one inc> of steel and lead. Dose rates from the '2ottest tubes are now reduced to 1-2 mrem /hr. 2. Spare instrument tubes leading down the annulus between reactor vessel and primary shield penetrate the shield at an elevation above the El. 548 floor. Some neutron streaming ( m15 mrem /hr) is detectee at the tubes. Since the streaming beam is well above the floor, over the head of any passerby, no plugging is necessary. ~ 3 Neutron dose rates in the order of 5 mrem /hr exist in the drive valve room, locally around the hydraulic lines penetrating the primary biolo6 cal shield. 1 4. In the pipeway vestibule area, El. 517, between primary and secondary shields, neutron leakage out the recirculation line penetrations in the primary shield gives rise to neutron dose rates of 60-100 mrem /hr. Since this area contains the recircult. tion piping and the primary steam lines with their associated high gamma fluxes from Nitrogen-16, no accessibility restriction accrues. However, neutron scattering in this area results in local dose rates of a 5 mrem /hr in the El. 517 valkway near the doors into the vestibule area. B. Turbine Building Surveys Although reactor core and pressure vessel shielding was comparatively well known, the radiation levels and shielding requirements in the -

primary cooling circuit were yet to be proven. The accessibility problems presented by radioactive isotopes in the steam and condensate systems of the plant were found to be of a r smaller order of magnitude than contemplated during design. Carryover of Nitrogen-16 in the steam is the only sizeable contributor to dose rates in the steam and condensate systems, with negligible corrosion product carryover. Of 400 curies of Nitrogen-16 produced per second in the reactor core et full power, only 5 curies per second reach the turbine. The turbine is unshielded, and dose rates of 10-60 mr/hr t a experienced by an operator during an inspection of bearings, front standard controls, etc. Dose rate at the generator collector rings is less than 0 5 mr/hr. Radiation levels around the condenser vary from 3 mr/hr at the water boxes to 20 mr/hr beneath the hotvell, with a local hot spot (200 mr/hr) at the off-gas line take off. As a non-condensible, the majority of the Nitrogen-16 goes with the off-gas to the shielded air ejector room, in which radiation levels are as high as 1 r/hr. The delay time (approximately 2 minutes) provided by the hotvell allows the residual Nitrogen-16 activity to die off before the consensate reaches feedvater pumps and heaters. The shielded feedvater heater i compartments have radiation levels between 1 and 250 mr/hr within the compartments, depending upon the extraction stage of the regeneration steam to individual heaters and upon the condensate water level in the heaters and drain tanks. -34 s I

Dose rates from primary recirculating lines in the secondary steam generator compartments average 2 r/hr at full power, with steam 4 generator dose rates lightly less. Neutron dose rates of 20 mrem /hr exist in the secondary steam generator compartmerits. The majority of this is due to Nitrogen-17 decay, with some possible contribution due to scattered neutrons from the reactor vessel. Correlation with Calculations Fair correlation was found between the currently used method of calcu-lation and measurements of Nitrogen-16 dose rates. The calculation method may be summarized by the following equations: E= m l, /), _ MNs?) Arte,S New MN k = Nitrogen-16 nuclei produced per fission neutron incore /7 fissi eutr a pr duced in energy group E + AE = E per fissier neutron in core, based upon the Watt-spectrum. energy-dependent activation cross-section for the formation = AcT(E) 16 16 of Nitrogen-16 in the 0 p) N reaction.

(Based upon BNL-325, Harvey and Hughes, Neutron Cross Sections,

7/1/55,p.85). [ g = energy-dependent cross-section of core for removal of neutron from its initial energy group E + AE to a lower energy group. This includes hydrogen mode ntion, uranium capture, inelastic scattering by oxygen, structure, and fuel, and elastic scattering by oxygen. The calculation of the contributions of the various energy groups is thus based upon a modified age calculation of the fission neutrons born above the reactor threshold of about 10 5 Mev. The result of this summation of the contributions is: -6 N = 3,5x/o Nitrogen-16 nuclei produced per fission neutron Total production of Nitrogen-16 when the core is operating at the full thermal rating of 626 megawatts is thus, = g,ggff"' N-16 nuclei fiss.neut. fiss. I N 7g 3jgjo g g g watts p fiss.neut. fission vatt - sec /4 N = /.7X/d Nitrogen-16 nuclei produced in core per second. p This may be converted to curies by /4 C =/,p/O N-16 nuclei x d,0945 disint. x / curie-see p sec. see-nucleus g g io disint. 433 curies per second of Nitrogen-16 released from core. c = p As may be seen from above, this method of calculating Nitrogen-16 production does not depend upon a calculation of the neutron fluxes of the various energies in the core. Rather, it is assumed that all neutrons born above 10 5 Mev are degraded below the threshold before leaving the core. -.

The splitting of this activity between water and steam may be predicted by two methods. One is based upon observations at EBWR, and the other is based upon a chemical analysis of the reversible reactions involving ammonical and nitrate radical forms. Both agree that, at the average Dresden core exit steam quality of 5 5%, the percentage of Nitrogen-16 escaping the core in a steam-cmpatible form is about 5 5%. On this basis, 23 5 curies per second start from the core in the steam. With a calculated 17-second delay time from core to turbine throttle, the entering rate of Nitrogen-16 flow through the turbine is 4 7 curies per second. On this basis, the calculated dose rate one meter from one 18-inch diametersteampipejustupstreamoftheturbineis410mr/hr. The measured dose rate at this pointvas 600 mr/br. Recent measurements of turbine dose rates (11/60) have shown that'a general" reduction of about 35% has occurred after air in-leakage to the primary feedvater system has been reduced. This may be explained, not in terms of reduced production of N-16, but as a reduction of carryover of N-16 in the steam from the core and drum. With substantial air in-leakage, the primary water is nearly saturated with atmospheric nitrogen (N-14). This inhibits N-16 solution in water upon its formation and increases its carryover with steam. Minimizing air saturation of the primary water has resulted in reducing this carryover, and dose rates are closer to predictions. It is interesting to note, as substantiating evidence, that Nitrogen-13 release from the stack dropped by a factor of four upon correction of 3 the air in-leakage. Nitrogen-13 is produced by the (n, 2n) N

reaction, with the target Nitrogen-14 nuclei supplied by air leakage into the system.

The fourfold reduction in Nitrogen-13 is much greater than the Nitrogen-16 reduction, since the target source of Nitrogen-13 vaa reduced. No less Nitrogen-16 is now being produced; the equilibrium has just been forced more in favor of the water phase. C. Radiation Measurements at Reactor Vessel Ganana heating and neutron flux were measured outside the reactor vessel vall near the core midplane, during full power operation measured values show each to be lower than the predicted values. This is principally due to the 452-channel core loading which existed during the measurements; the predictions were made on thesbasis of a full 488 channel core. 1. Neutron Fluxes in Vessel Wall The neutron flux and spectrum measurements were made with the use of a ~ series of threshold-reaction activation samples. The sample assembly (Capsule Carrier Assembly) is shcr.rn on Drawing 932C449 in the Appendix. ' The assembly was mounted in place of the " Dark Channel" external ion chamber in ion chamber guide tube #12, and positioned as close to the reactor vessel as possible. In position, the carrier was separated fromthereactorvessel(atcoreelevation)by1/2inchofsteel (inner shield caisson), the vessel insulation, and approximately 6 inches of air. At that pocition, two assemblies were irradiated for one hour. The results of specific activation deteminations by B. F. Ridar at Vallecitos Atomic Laboratory were: 0 0 5 Mev neutron flux 2.2 x 10 n/cm2,3,c, (by Cl35 (n,OC) P32) 2 9 Mev neutron flux 7 2 (by S32 (n, p) P32) 36x10 n/cm -see.

5 0 Mev neutron flux 2 (by Ni5o (n, p)CoS8) 2.2 x 107 n/cm,,,, 8.1 Mev neutron flux (by Al 27 (n,d) Na24) 4.2 x 10 n/cm-sec 0.1 Mev neutron flux ( extrapolated fromabove) 8 x 10 n/cm-sec Further extrapolation back to the outside radius of the pressure vessel gives j g (.17X/,7f)ng ) 8X/0 2 X(.aco/ Q>O./Mty = /,' /X / O n/cm-sec = It was calculated that the neutron flux on the outside radius of the 9 9 2 pressure vessel would be betveen 3 x 10 / and 7 x 10 n/cm_,,, ( > 0.1 Mev). Thir was based on full core loading of 488 fuel bundles, or channels. The lower value determined experimentally may be partly reconciled with the theoretical value by examining th decrease in core radius in the 452 channel core. The absence of channels in the outer periphery increased the water reflector thickness in the direction of the samples by 3 5 inches. Tais increase in water path vould theoretically decrease the fast (7 0.1 Mev) flux by a factor of ,/2 X 3.SX 2. Sf z.9 e = This factne, applied to the range of theoretical prediction above, would 9 9 2 decrease ' theoretical prediction of 3 x 10 - 7 x 10 n/cm-seetoa 9 2 range of 1 x 10 - 2.4 x ' ) n/cm -sec. The effect of water annulus thickness upon vessel neutron ?xposure is evident.,

i i The neutron flux at the inside radius of the vessel can be extrapolated from the measured value at the shield caisson. As noted above, extra-polation to the outside radius of the vessel gave a neutron flux of 9 2 1.1 x 10 n/m -sec (7 0.1 Mev). Similarly, further extrapolation to the l inside radius of the vessel gives a fast (> 0.1 Mev) neutron flux of 9 U E/SS goo /0 g /,/ X/0 J G p yy * /,6X/O n/cm -sec with the 452 channel 6 ve a 40-year neutron exposure at the vessel inside core. This would 1 radius of 1.6 x 1019 2 n/cm with a plant load factor of 0.85 If the core were loaded to its periphery, the 40-year exposure at the inside radius 2 of the vessel vould be 4.6 x 1019 n/cm. Present APED design criteria 19 permits a maximum neutron exposure of 5 x 10 n/cm2(7g,1g,y)1,y,,,,1 steel, based upon consideration of embrittlement and rise of transition temperature in the steel. Gamma Heating in Vessel Wall With the use of a special calorimeter, gamma heating was measured at the same point (inside channel #12 guide tube) as the neutron fluxes described above. The calorimeter used was composed of a slug of stainless steel of known mass, suspended and insulated within an evacuated thin-valled case. Both slug and case were fitted with thermocouples. Kncving the heat capacity of the slug, the heat transfer rate between alug and case, and the rate of rise of slug temperature in a Eamma energy flux, the gamma heating coul(. be evaluated. Thus, -40

~ t l I d7 where is the time rate of temperature rise in the slug. h CM %en =- p g mr. Sws V 3 ga-a heating in slug B'IU/hr-ft. Q = C specific heat of slug, BTU /16 *F. p mass of slug, lb m = total heat transfer coefficient between slug and case. H = H=H (T,,T,) T, slug temperature, 'F = case temperature, 'F T = e 3 volume of slug, ft, V = H (T,, T,) was determined experimentally before the gamma heating measure-ment. At full reactor power, this data of slug and case temperature vs. time was recorded: July 16, 1960 Ts Ts Tc Tc Ts - Te Time (Hrs) hv] (*F) (mv) (*F) 1222 1.86 95 5 1.86 95 5 0 1325 1 99 99 1 90 97 2 1343 2.03 100. 1 92 97 5 3 1430 2.07 101 5 1 94 98 35 At case temperatures of 95 - 98*F, ~* //(tf5)=6,fsrio BTU /g. *p -3 Cm= 8,5 rio btu /o, p V ~ A 42 rio # 0.' ~ -h1-

3 Evaluating, Qvariesbetween775and825 BTU /hr-ft. The average gamma heating, calculated by averaging the data, is 3 800 BTU /hr-ft Q = insteel,atapoint,justoutsidethe1/2-inchthicksteelinnershield Caisson. To extrapolate the gamma heating to the outside and to the inside radii of the reactor pressure vessel, use may be made of the theoretically predicted slope of the gamma heating in the pressure vessel steel. The theoretically predicted gamma heating in the vessel steel was: 4 - 4./ d Q = 2.6 y/o G BTU /hr-tt3 thickness of steel, in feet, from the inside radius t = of pressure vessel. Thus, at the inside radius t = 0 and gamma heating was predicted to be 3 26,000 BTU /hr-ft Q = at the outside radius of the pressure vansel, following d e prediction. The gamma heating was calculated to be b - 6, / (0, S0) Q

2. 6 X to 2

3 1235 BTU /hr-ft, = Using the same exponential slope (6.1 ft'1) to extrapolate the average measured value beyond the caisson, to the pressure vessel outside radius, " Gamma Eeating in Dresden Pressure Vessel", by R. A. Mickle & W. D. Craig, March 9,1960 :

fo,/(eaissou 7%Iekness) ,,g foO Q Y = Zoo caisson thickness = 0 5 inch = 0.0416 ft. Then the extrapolated heating value at the vessel outside radius becomes 4,/ X 0. 04/ fa 2tf 800 Q X Q = gee 3 1110 BTU /hr-ft, = 3 This compares to the predicted value previously noted of 1235 BTU /hr-ft, 3 The gamma heating at the inside radius, corresponding to 1110 BTU /hr-ft 3 at the outside radius, is 23,400 BTU /hr-ft for the 452 channel core. The corresponding maximum tangential thermal stress is about 2000 psi (tensile) at the inside radius of the vessel, assuming perfect vessel insulation. For less than perfect insulation, maximum thermal stress should be lower. Again, the theoretical value was based upon a 488-channel core loading and the measurements were made with a h52-channel core. Using the increase in water reflector thickness calculated and used in the neutron flux analysis above 3 6 inches, the gamma heating at any point outside the water reflector r may be theoretically compared for the full h88-channel core and the 452-channel core. The effective one-exponent fit of water attenutation of the gamma heating spectrum in Dresden reflector is -1 y,ff = 0.03 cm 43-I

For a water path increase of 3 5 inches, the ensuins attenuation is o.03 x 3 5 x 2 54 g , o,p, Then, for a full k88-channel core, the gamma heating in the vessel hy, or 13 times the gamma heating in the 452-channel steel vill be core. on the basis of the measured value of heating outside the caisson with the 452-channel core, the value of gama heating at the inside radius of 3 the pressure vessel with a k88-channel loading vill be 30,400 BTU /hr-ft, This corresponds approximately to a maximum tangential thermal stress of 2600 psi (tensile) at the inside radius of the vessel. The measured values of gama heating cannot be taken as particularly precise data. The heat generation rate and the temperature differentials are lower than those for which the instrument was designed. The gamma heating measured represents an upper limit. It should be noted that a series of thermocouples were located on the outside diameter of the pressure vecel during installation. It was originally planned to check gama heating and thennal stress by noting the temperature differential between core inlet temperature (approximately corresponding to water temperature oa the inside radius of the vessel) and the temperature indicated by a thermocouple on the outside of the vessel at core mid-plane height. This method did not give the desired results, due to: 44-

1) The less-than-theoretical efficiency of the vessel insulation, which caused a lowering of the temperature differential across the vessel vall.

2) The small temperature differential (~10*F), which would exist across vessel vall at the measured rate of gamma heating, even with perfect vessel insulation.

3) The inability of the thermocouples to measure within an accuracy of 1-2% (or 5-10*F at 550*F).

D. Shutdown Radiation Measurements The program of shutdown radiation measuremente, nas just begun. Several ~ years of operation with attendant shutdowns and surveys must be accumulated before a real pattern of activity buildup vill be evident. At this time, after about two months of effective full power operation, onl+ Mnor quantitites of cobalt-60 (5 3 year half-life) are present. Iron-J3 (45 day halfelife), cobalt-58 (72 day half-life), and zirconium-95 (65 day half life) have yet to exhibit the full magnitude which they will reach. The majority of plant shutdown radiation, outside the core, at the present time is caused by manganese-56 (2 5 hour half-life) and copper-64 (12.8 hour half-life). Tables of shutdown dose rates around steam drum, steam generators, and piping are given in Section III-B of this report. Direct radiation from the core is negligible after shutdown, in such areas as reactor top head. Dose rate above reactor top head during such operations as stud-heating for head removal, is controlled by scattered radiation from the core and structure around the core streaming up the annulus between vessel and primary shield. The dose rate from this source isabout50mr/hr,24hoursaftershutdown. This is attenuated by filling the refueling canal with about two feet of water Another source is top head activation. At the present time, top head dose rates, at contact, are 5 mr/hr 36 hours after shutdown from full' power operation. Within 12 hours after shutdown, the dose rate below the control drives in the sub-pileroomislessthan5mr/hr. At shutdown, dose rates in secondary steam generator compartments are of the order of 100-200 mr/hr.nearrecirculationpipi"gandsecondarysteam generators. This decays to about 25 mr/hr in 24 hours except near local hot spots. Substantial buildup with reactor operating time may be expected, particularly around valves, flanges, pumps., etc. The two cleanup demineralizers and their associated regenerative heat exchan grs represent the most concentrated radiation sources (outside reactor vessel)'after reactor shutdown. Dose rates of 100-200mr/hr exist in the cleanup demineralizer rooms one week after shutdown. In the first 48 hours after shutdown, dose rates around cleanup demineralizers and heat exchangers are several hundred mr/hr. To date, however, the demineralizer recirculating piping and pump exhibit small contamination; and the radiation levels in the room containing the desineralizer recir-culating pump are 10 mr/h1 m less, one week after shutdown. In the tu?bine building, the condensate demineralizers contain the only signifi-cant source at shutdown. All other equipment is capable of direct access within a few seconds of reactor shutdown..

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