ML20062C763

Orau 89/H-9, Confirmatory Radiological Survey of L-77 Reactor Facility Univ of Ca,Santa Barbara

ML20062C763
Person / Time
Site:	05000433
Issue date:	09/30/1989
From:	Landis M OAK RIDGE ASSOCIATED UNIVERSITIES
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
Shared Package
ML20062C759	List: ML20062C755 ML20062C763
References
ORAU-89-H-9, NUDOCS 9011020118
Download: ML20062C763 (42)

Text

{{#Wiki_filter:, Alh'C Z ory 2e k h h o*7 +,\\ CO+3rAPy-0 / ^ ORAU 89/H 9 u 0 ~ ( ~ L l j $afEd ** CONFIRMATORY RADIOLOGICAL SURVEY-ge Associated universities OF THE l j ["; fcl,', L-77 REACTOR FACILITY. Reg ' ? n's UNIVERSITY OF CALIFORNIA, SANTA BARBARA n ss "*0 " V i "ic SANTA BARBARA, CALIFORNIA Division of Radiation Safety, j and Safeguards, M. R. LANDIS i Emergency l Preparedness and l Radiological Protection Branch I i 1 1 I t Environmental Survey and Site Assessment Program Energy / Environment Systems Division FINAL REPORT ' SEPTEMBER 1989 - k ) 901102011e e91003 PDR ADOCK 05000433 P PNU

,.;..--.4-,;---- I lI -1 I g: I! v; I I! I I I I g I NOTIOES The opinione espressed herein so not noosseethy reflect the opinions of the sponeonng institutione of Oak Riope Associated - Univoretties. This report wee prepared as en account of work aponsored by the United Steles Governmora. Neither the United Statee Coverrment nor the U.S. Departmem of Energy, not any of their employees, meses any erettenly, express or implied, or eseumes any legalliebility or responalbility for the occuracy, comp 6eteness, or usefulness of any informators apparatus, product, or process discioned, or represents that its use would not intnnge privately owned rights, Reference herem to any specthe commercial product, process, or service by trade name, mark, manutecturer, or otherwise, does not necessertly constitute or lmply hs endorsement or recommende. tiert ortavoring bythe U.S. Government or eny egency tnereof.Tne views ond opiniorm ot euthors espressed hereln uo not nocessettly state or refwet those of the U.S. Government or any agency thereof.

DRAU 89/H 9 CONFIRMATORY RADIOLOGICAL SURVEY l 0F THE L 77 REACTOR FACILITY I UNIVERSITY OF CALIFORNIA, SANTA BARBARA SANTA BARBARA, CALIFORNIA i ^ Prepared by M. R. Landis -{ -i l Environmental Survey and Site Assessment Program. .i Energy / Environment Systems Division l Oak Ridge Associated Universitics 1 Oak Ridge, Tennessee-37831 0117 .+ I Project Staff S. F. Barnett E. A. Powell G. R. Foltz .C..F. Weaver M. J. Laudeman Prepared for i Division of Radiation Safety and Safeguards Emergency Preparedness and Radiological--Protection' Branch U. S. Nuclear Regulatory Commission Region V Office-i -r Final Report September 1989. i This report is: based -on work performed under -an Interagency Agreement; (NRC Fin. No. -A-9400) between the U. S'. Nuc' lear Regulatory Commission and the U.S. Department of' Energy.' I ~

I. -l TABLE OF CONTENTS I Page List of Figures 11 List of Tables . iii i Introduction and Site History 1. Site Description. 1 Survey Procedures.. 2 Results. 4 Comparison of Results With Guidelines 6'- Summary. 7 References 18 i Appendices: Appendix A: Major Sampling and Analytical Equipment Appendix B: Measurement and Analytical Procedures Appendix C: Regulatory Guide 1.86 - Termination of Operating. Licenses for Nuclear Reactors Appendix 0. Proposed Confirmatory; Survey Plan for the L-77 Reacter Facility University of California. Santa Barbara, California I I I

List of Figures Page FIGURE 1: North View of Lhe UCSB L 77 Reactor Facility. 8 FIGURE 2: Layout of UCSB L 77 Reactor Facility Indicating Measurement Locations on the Floor and Lower Walls, 9 FIGURE 3: Layout of UCSB L-77 Reactor Facility Indicating Measurement Locations.on the Upper Walls and Ceiling. 10 3 1 FIGURE 4: Layout of UCSB L-77 Reactor Facility Indicating Locations of Exposure Rate Measurements 11 l N l I i 11~

.] List of Tables-l Page Table 1: Surface Contamination Measurements - Floor and Lower Walls. 12 Table 2: Surface Contamination Measurements - Upper Walls and Ceiling . 15 Table 3: Exposure Rate Measurements i 17-1 4 3 i I L I 1 I I 111 l l

CONFIRMATORY RADIO 1hCICAL SURVEY - OF THE 1 L 77 REACTOR FACILITY UNIVERSITY OF CALIFORNIA, SANTA BARBARA SANTA BARBARA, CALIFORNIA I INTRODUCTION AND SITE HISTORY ~ The L-77 Reactor, located in Broida Hall (Building 572) of the University of California, Santa Barbara, California (UCCB), is a-Nuclear Regulatory I Commission (NRC) licensed facility (License'#R-124). The reactor originally was installed and operated at the University of Nevada, Reno, Nevada,.and was moved to UCSB in 1974. UCSB received a Facility Operating License on December 3, 1974, and was authorized for startup in January 1975. The maximum 1 operating power was.10 watts (thermal). The. reactor was operated:for' ') instructional purposes and activation analysis. The L 77 was a homogenous aqueous solution research reactor. The fuel solution was enriched uranyl sulfate ' dissolved in water, contained in the I Reactor Core Tank, which was contained in an inner shield ~ tank. 'The shielding-design permitted operation of the reactor while personnel-were present:in the Reactor Room. UCSB submitted a decommissioning plan to the NRC'and a dismantling order was issued on August 26, 1986. UCSB has completed decommissioning of the L ! facility and has submitted a close-out survey. report to the NRC. l l At the request of the NRC,. Region V, the Environmental Survey and Site l Assessment Program of Oak Ridge Associated Universities (ORAU) conducted a radiological survey to evaluate the radiological status.of.the L 77' reactor facility relative to the NRC guidelines for unrestricted use. i SITE DESCRIPTION-g License R 124 specifically designated the area of' operation as Room 1251, er in the eastern end of-Bui'1 ding 572. This room,is a high bay. area, constructed- ,3, Of Concrete-block walls, on concrete slab, poured on grade. A photograph of. 8 I 1 .)

the north view of this area is included as Figure 1. The area above the ceiling is an accessible roof structure. Room 1251 was previously divided into two areas by an internal wall: the northern area was the reactor zone and control console, and the southern area housed a Nuclear Chicago suberitical 5 assembly which was operated under a State of California radioactive materials license. An opening was cut into the northern wall of doom 1251 to facilitate 1 removal of the reactor core assembly. This opening provided access to Room 1356 which is currently being used for storage and as a work area. Both internal I walls have been removed and the eastern end of Building 572 presently consists of one large room identified as Room 1356. I 8 SURVEY PROCEDURES Document Review ORAU reviewed the licensee's documentation supportit.g the decommissioning 1 project. Documents reviewed included the dismantling order, the 2 decommissioning plan, the final survsy report 3, licensee /NRC correspondence, and other associated documents supporting the decommissioning activities. Facility Survey .5 on June 14-15, 1989, ORAU conducted a confirmatory survey of the L-77 reactor facility. The purpose of the survey was to verify the adequacy and f accuracy of the licensee's final survey, and to confirm the radiological condition of the facility relative to the decommissioning guidelines. Cridding The 3' x 3' grid established by the licensee on the floor was extended to include the lower walls (up to 6'). The upper walls were not gridded. 1 Measurements taken on the ungridded surfaces were referenced to the floor and i lower wall Brid, or to pertinent building features. 2

I Surfaco Measurements l

Thorough, systematic alpha, beta gamma, and gamma scans were performed on floors and lower vs11s (up to 6') using a gas proportional alpha / beta floor I

monitor, zinc sulfide alpha detectors,

" pancake" GM detectors, and NaI(T1) scintillation detectors coupled to scalers /ratemeters with audible indicators. Thirty five grid blocks on the floors and lower walls were randomly selected for surface contamination measurements (Figure 2). The grid blocks were numbered sterting with the southeast corner on the floor and going west to ~ east, east to west until all blocks had been assigned a number. The walls were numbered in the following order: east, west, north, south. Nu.nbers were-assigned starting with the upper left hand block of each wall and continuing 3 I across and down. Total measurements of alpha and beta gamma contamination levels were systematically performed at the center and four points, midway between the center and block corners. Smears for removable alpha and beta contamination were performed at the location in each grid block where the highest direct reading was obtained. Total and removable contamination levels were also measured at 20 locations on the upper walls, ceilings and miscellaneous overhead objects. l I Exposure Rate Measurements Camma exposure rates at 1 meter above the floor were measured at 6 locations within the L-77 area, using a pressurized ionization chamber. l Camma Spectroscopy Measurements i In situ gamma spectra were collected at several locations where gamma I exposure rate measurements were made. Spectra were collected using a NaI(TI) detector. The gamma spectra were used to identify the residual radionuclide contaminants. I 3

] h.. I Baseline Measurements Room 1207, located approximately 130 feet west of Room 1356, was used to establish the baseline for gamma exposure rate measurements. This area has the gg same construction history as Room 1356 and is located in a non restricted area which has no history of radioactive material use. Three exposure rate measurements were performed and two gamma spectra were collected using the l pressurized ionization chamber and gamma spectroscopy system. Sample Analysis and Interpretation

g 5

Smears were analyzed for gross alpha, gross beta and tritium activity. Additional information concerning major instrumentation, sampling equipment, 5 and analytical Procedures is provided in Appendices A and B. Findings of the independent measurements were compared to Regulatory Guide 1.86 (Appendix C) and the NRC requirement which states that ambient exposure rates nust not exceed 5 pR/h above background at 1 m from any surface. RESULTS 5 Document Review In general, the decontamination plan appears to have been adequately developed and implemented to ensure the NRC guidelines' were met. The 3 information contained in the final survey report adequately summarized the radiological status of the site. The licensee' final decommissioning report indicated that the lead reactor vessel and inner shield were activated and could not be shipped to an active burial site. This material is currently: o adequately packaged for transportation t.o an, as of yet, unspecified disposal site, o stored in a closed celt in the Engineering.Research Center, 4 l l

I o transferred to the State of California radioactive materials license as Amendment No. 41 to California License No. 1336 42. No other mixed hazardous wastes were identified by the licensee. I Tacility Survey Baseline Exposure Rate l Baseline exposure rate measurements from three background locations in Room 1207 ranged from 9 to 11 pR/h with an average of 10 pR/h. The gamma spectra collected only identified the presence of natural radionuclides normally present in building materials. Surface Scans Alpha, beta gamma, and gamma scans did not identify any areas of elevated contact radiation levels. I i Surface Activity Levels Results of total and removable contamination measurements performed on 35 i randomly selected floor and lower wall grid blocks are summarized in Table 1. The maximum alpha measurement was 100 o,m/100 cm2 and the maximum beta gamma 2 measurement was 1600 dpm/100 cm. The highest alpha grid block average was I 56 dpm/100 cm2 and the highest beta gamma grid block aversge was 2 1200 dpa/100 cm. The maximum removable alpha, beta, and tritium activity 2 2 2 levels were 5 dpm/100 cm, 8 dpm/100 cm, and 9 dpm/100 cm, respectively. Table 2 summarizes the resulta of 20 single point surface contamination l measurements performe:S on upper walls and ceilings. The total alpha activity i ranged from <23 dpm/100 cm2 a 2 to 68 dpm/100 cm, and the removable activity was 2 <3 dpm/100 cm. The total beta activity ranged from <450 dpm/100 cm2 to 2 1400 dpm/100 cm, the removable beta ' activity ranged from <6 dpm/100 cm2 i to 2 14 dpm/100 cm, and the removable tritium activity ranged from <4 dpm/100 cm2 2 to 8 dpm/100 cm. 1 8

I Exposure Rate Measurenents I Table 3 summarizes the exposure rate measurements, taken at six random locations. The exposure rate level in Room 1356 was 12 pR/h, compared to a lI baseline level of 10 R/h measured in Room 1207. The gamma spectra collected in Room 1356 only identified the presence of natural radionuclides normally present in building materials, no peaks associated with fission or activation products were noted. I COMPARISON OF RESULTS WITH CUIDELINES NRC surface contamination guidelines for release of facilities for unrestricted use are presented in Appendix C. The guidelines for residual g 5 alpha contaminacion, based on transuranics being the principal contaminant, are: I Total Contamination 1 300 dpm/100 cm2 (maximum in a 100 cm2 area) 100 dpm/100 cm2 (averaged over 1 m ) 2 i Removable Contamination 20 dpm/100 cm2 For residual beta. gamma contamination, the NRC guidelines for mixed fission products are: Total Contamination 15,000 dpm/100 cm2 (maximum in a 100 cm2 area) 5,000 dpm/100 cm2 (averaged over 1 m ) 2 Removable Contamination 2 1,000 dpm/100 cm l 6 m

m s These are the appropriate guidelines since it can be demonstrated that Sr 90 and 1 129, long lived fission products with more restrictive guidelines, are not present. The licensee has provf ded an acceptable justification for the exclusion of the Sr 90 guidelines which has been included in their final 3 decommissioning report as Appendix 7.5. Based on the yield of fission products, Cs 137 production levels (activity / fission) are a factor of greater than 105 those of I 129. For I 129 to be present at levels exceeding its 2 guidelines values (i.e. 100 dpm 100 cm, averaged over 1 square meter; 300 dpm/100 cm2 maximum over 100 cm2 and 20 dpm/100 cm, removable) the 2 corresponding level of Cs 137 would be a factor orders of magnitude above its .I guideline limit. Measurements made at this facility did not identify such levels of this more dominant fission product. It is therefore a logical conclusion that 1 129 is not present at significant levels and therefore the p less restrictive beta gamma guidelines values are applicable. All total and removable alpha and beta gamma levels, as well as removable contamination measurements, were within these guidelines. l All exposure rate measurements were within the guideline level of 5 pR/h above background. I

SUMMARY

On June 14 15, 1989, Oak Ridge Associated Universities performed a confirmatory radiological survey of the L 77 facility located in Broida Hall on the University of California, Santa Barbara Campus, in Santa Barbara, California, The survey included surface alpha, gamma and beta-gamma scans and, g p measurement of direct and removable contamination levels. The findings support the close out survey perb emed by the licensee, and confirm that the radiological conditions of the L 77 facility satisfy the NRC guidelines established for release for unrestricted use. l I 7

M M M _ - -.3 m . ;_ ;.;; 9 7 s. ~ i ' [ .f g'. c_ 3 I'd.,3.2.___.: _- s.

-y

f..

,,.~

s

_;.,e, n ya '.,_'.:= =E $. hMg g. % : ; ' ' ' }< m R gI--l,.f }. * ) L ~ h. ~;;' c r 4.- 1, _. -.. - g,.g : 4 -n. s ..e .]. 4: .,.{c . c..

.g z

'.y.

y -7 t. p-c

s., L :.;. '

., 4 : y ( jf = [i {. L. 1 ,.. i q,. ~ ~M-l :. 7 p _ ;c:, n ..,g a~ ~f .,y 4 ' 4 *' '- '-.,... n. y.,, r f - rl yp 5- , ~ %g;, ;,._ ;,.,g 1 (., ., j_ j : J-. e

y

.=

9 l-wr ... g...; ?. ]; _'- . r.. ..hll :Q ;)Q[ [. J _ [./ [ - l ,4

_' I.

..g_; [ !.

).

. p. -

y . p,, .w+

3. tie ' - -.

y l.

+g w;.;.-.

.: e y -; #Q ,1 ~ " - ' . +l - oo p.,., [ 4R ~ ? .Y. ~ l. ^ a h) y; a .: L-

f

'T - =w ;s. g: ' . _1 ;V " gnr;.

• ?y

.w,. 9. +w. k,. _h 'ety,.l 7.

g. 3 %%;.

g . l{: l$lJ:'._ l '* hhf~ ~ W 4 - (;?,' p 7. j [, ;. O c +jc , 7..f ~~'jsa y g {g.

f.

n f, xyp ;:2 4 ; 5 ' i* c. -~ e >yu y 'G: 79;$,. 3- .y %w= w 93 y 4 - 33 .Qp - .'tt -, y ':._ - g. u-n g; m:: ~. ., j - ~ ; ;.: '. t- .1 ... f: - %. o -. s. ~ c-py [ k': _..f ' ..f55E ;.

. i a

N::'i. % ~ .,4. -9?rwW A rMgw ysi' -. ~....... <. 2,w+G S :. - -JR.. _...x. . =: FIGURE 1: North View of the UCSB l-77 Reactor Facility

300 302 312 309 swas ll I 204 202 190 ^ I 1 172 163 167 157 154 5 146 148 151 246 = 125 121 110 216 265 106 93 94 ? 90 84 77 79

1. GRID BLOCK 8

SELECTED r0R 66 MEASUREMENT 61 224 295 43 0 256 = FEET I 3 LANDING 0 2 p STAJRS 8 321 323 8 I I . FIGURE 2: Layout of UCSB.L-77 Reactor Facility Indicating h Measurement Locations - on the Floor-ond Lower Wolls B

UCSlo J $TA!P% 31 92 93 l e i 7,g I De \\ i' f I' O MEASUREMENT LOCATIONS g12 of UPPER WALL j3 itg A CEIUNG N h g 15 h g 16, y gl7 N W 1E B O 6 d 2 LANDING 51 AIRS + l I FIGURE 3: Layout of UCSB L-77 Reactor Facility ' Indicating Measurement Locations on the-Upper Walls and Ceiling 10

] ucsid ) J e' [ e 2 I l 3 e'

1. MEASUREMENT O

LOCATIONS N h 4h ,5 h FEET 6 g o e uN ING METERS 1 FIGURE 4: Layout of UCSB L-77 Reactor Focility indicating Locations of Exposure Rote Measurements 11

E E TABLE 1 SURFACE CONTAMINATION MEASUREMENTS FLOOR AND LOWER WALLS L-77 FACILITY UNIVERSITY OF CALIFORNIA. SANTA BARBARA SANTA BARBARA, CALIFORMIA TOTAL ACTIVITY REMOVAOLE ACTIVITY 2 Alpha (dom /100 cm2) Beta-Gamma (dpm/100 cm ) Alpha Beta Trititre 2 2 Grid Block

• Range Average Range Average (dpm/100 cm2) (dom /100 cm )

(de-/100 cm ) 3 <23 - 34 23 890 - 1600 1100 5 8 <4 43 <23 <23 <450 - 960 520 <3 <6 <4 61 <23 <23 <450 - 560 <450 <3 <6 <4 66 <23 <23 <450 - 520 <450 5 8 <a 77 <23 <23 <450 - 630 560 <3 7 <4 79 <23 - 57 <23 <450 - 1100 740 <3 <6 <4 84 <23 <23 <450 - 600 <450 <3 <6 <4 2 90 <23 <23 <450 - 520 <450 <3 <6 <4 93 <23 <23 <450 <450 3 <6 <4 94 <23 <23 <450 - 520 <450 <3 <6 <4 106 <23 <23 <450 <450 <3 <6 <4 110 <23 <23 <450 <450 <3 46 <4 121 <23 <23 <450 - 1100 600 3 <6 <4 - --. ~. - -- - - - - - - g

Y M TABLE I (Cont inued) SURFACE Cor:TAMINAT10rs MEASUREMENTS FLOOR AND LOWER WALLS L-77 FACILITY UNIVERSITY OF CALIFORNIA. SANTA BARBARA SANTA BARBARA, CALIFORNIA TOTAL ACT!v!TY REMOVABLE ACTIVITY Alpha (dpm/100 cm2) Beta-Gama (opm/100 cm ) Alpha Beta Trittum 2 Grid 81ocka Range Average Range Average (dom /100 cm ) {dpm/100 cm2) { dom /100 cm2) 2 125 <23 <23 <450 - 780 <450 3 7 <4 146 <23 - 34 <23 <450 - 850 520 <3 <6 <4 148 -<23 <23 <450 - 1200 670 <3 <6 9 -to 151 <23 <23 <450 - 780 <450 3 <6 <4 154 <23 <23 <450 - 810 520 <3 <S <a 157 <23 <23 <450 - 1000 520 <3 <6 <4 167 <23 <23 <450 - 930 520 3 <6 <4 190 <23 - 34 23 <450 - 1400 1000 <3 <6 <4 204 <23 45 <23 <450 - 630 <450 <3 <6 <4 208 <23 - 100 56 <450 - 780 <450 <3 <6 <4 218 <23 - 79 34 <450 - 1100 700 5 <6 <4 224 <23 <23 <450 - 630 <450 <3 <6 <4 246 <23 - 56 23 <450 - 850 560 <3 ' <6 <4 - = _ _ -.- -..- - __.. _ _ =

w v w TABLE 1 (Continued) SURFACE CONTAMINATION MEASUREMENTS FLOOR AND LCWER WALLS L-77 FACILITV UNIVERSITY OF CALIFORMIA, SANTA BARBARA SANTA BARBARA, CALIFORNIA l TOTAL ACTIVITY REMOVASLE ACTIVITY 2 Alpha (dom /100 cm2) Bete-G m a (dpm/IOO em ) Alpha Beta Trit ium 2 2 Grid Block

• Range Average Range Average (dpm/100 cm ) (d m /100 cm2)

(dom /100 cm ) 256 <23 - 56 23 1100 - 1300 1200 <3 <6 5 265 <23 - 45 23 <450 - 1400 890 <3 <6 <4 295 <23 - 45 23 630 - 1400 930 <3 <6 <a n 300 <23 <23 <450 <250 5 <6 <4 302 <23 - 34 23 <450 <450 <3 <6 <4 309 <23 <23 <450 <450 <3 <6 <4 312 34 - 56 45 <450 - 1300 630 <3 <6 <4 321 <23 - 34 23 700 --1100 1000 <3 <6 <4 323 <23 - 34 <23 <450 - 1500 810 <3 <6 <4 ' Refer to F igure 1.

TABLE 2 SURF ACE CONT AP*INATION PE ASUSE*9EN75 UPPER WALLS AND CEILING L-77 FACILITY UNIVERSITY OF CALIFORNIA. SANTA BARBARA SANTA BARBARA, CALIFORNIA TOTAL ACTIVITY REMOVA9tE ACTIVITY Alpha Bet a-Gama Alpha Beta Tr it itri 2 2 2 Location" (dpm/100 cm2) fdom/100 cm ) (dpm/IDO cm2) (dom /100 cm ) (dce/100 cm ) -1 <23 <450 <3 <6 <4 2 56 590 <3 <6 <4 5 3 <23 700 <3 <6 <4 4 <23 <450 <3 <6 <4 5 <23 <450 <3 <6 <4 6 <23 1000 <3 <6 <4 7 34 670 <3 9 <4 8 34 630 <3 <6 <4 9 23 <450 <3 <6 <4 10 23 <450 <6 8 11 45 <450 <3 <6 <4 12 23 <450 <3 <6 <a 13 45 (450 3 <6 <a

9 r ) 2 m m c u i 0 t 0 4 4 4 4 4 4 4 i 1 r/ T m o ( d Y T I V I ) T 2 C m A c E a0 L t 0 S e1 6 9 4 6 6 6 6 A B/ 1 /h mo K d E ( R ) 2 mc a h 0 p0 l 1 3 3 3 3 3 3 3 A / A m R o A d S B ( T R N A E B M A E G A I P N T

• 4 U I N R

) d S L A O e A I SF u E E I n M C Y L i T A, A ) t N DI C a2 I n O N L N m C T C O A, mc o I A I R a ( A S A F R G 0 0 0 0 0 0 0 0 N L F I A - 0 1 5 9 3 0 0 0 2 I L L B a1 8 8 5 9 3 4 1 M A 7 A R t / 1 1 1 E A W7 C A e m L T B B o B N R L F d A O E O A ( T C P T Y P Y N T E U T A I C I S V A S I F R T R E C U V A S I N L U A T O T ) 2 m c a h 0 p 0 l 1 5 6 3 6 4 8 3 A / 4 5 2 5 3 6 2 m o d ( 2 erug a i n F io o t t a 4 5 6 7 8 9 0 c 1 1 1 1 1 1 2 r o e L fe R 5

I TABLE 3 EXPOSURE RATE MEASUREMENTS L-77 FACILITY I UNIVERSITY OF CALIFORNIA. SANTA BARBARA SANTA BARBARA, CALIFORNIA I Exposure Rate at 1 m Above the Surface i (pR/h) j Locationa 1 12 2 12 3 32 4 12 5 12 i 6 12 8 I 1 URefer to Figure 3, i i i 1 l 17

-- =-- E. 1 l REFERENCES 1. Docket No. 50 433, Order Authorizing Dismantling of Facility and Disposition of Component

Parts, United States Nuclear Regulatory Commission, Division of Pk'R Licensing B, August 26, 1986.

I 2. Decommissioning Plan for the L 77 Research Reactor, University of ~ California, Santa Barbare, Rocketdyne Division, Rockwell International Corporation, August 5, 1985. 3. Final Survey Report for the University of California, Santa Barbara, L 77 Reactor Dismantling, Rocketdyne

Division, Rockwell International I

Corporation, December 19, 1986. 5 I L I = 1 1 y I I 1 1 18

\\. = APPENDIX A MAJOR SAMPLING AND ANALYTICAL EQUIPMENT m

---

isnisn i i

---

amm-si

-sumsmi

a APPENDIX A MAJOR SAMPLINC AND ANALYTICAL EQUIPMENT ? ( The display or description of a specific product is not to be construed as an endorsement of that product or its manufacturer by the authors or their employer. A. Direct Radiation Measurements Therline " RASCAL" Portable Ratemeter Scaler Model PRS 1 l (Eberline, Santa Fe, NM) Eberline PRM 6 Portable Ratemeter (Eberline, Santa Fe, NM) Eberline Alpha Scintillation Detector Model AC 3 7 (Eberline, Santa Fe, NM) Eberline Beta Gamma " Pancake" Detector Model HP 260 (Eberline Santa Fe, NM) Harshaw Na1 Gamma Scintillation Probe Model 12S12/E (Harshaw Chemical Co., Solon, OH) Ludium Alpha / Beta Floor Monitor Model 239 1 (Ludium, Sweetwater, TX) Multichannel Analyzer Canberra Series 10 Plus (Canberra Instruments, Meridian, CT) Reuter Stokes Pressurized Ionization Chamber Model RSS 111 (Reuter Stokes, Cleveland, OH) Victoreen Beta Gamma " Pancake" Detector Model 489-110 4 (Victoreen, Cleveland, OH) Victoreen NaI Scintillation Detector Model 489 55 (Victoreen, Cleveland, OH) A-1

J B. Laboratory Analyses Low Background Alpha Beta Counter Model LB 5110 l (Tennelec, Oak Ridge, TN) Liquid Scintillation Counter Tricarb 300 (Packard Instrument Co., Inc., Downers Grove, IL) I I I 1 I l I l 1 A-2

g. 1!I I 1I I I l x,,-- I

• * * " " * * " ^ " ^ ~ ^ " " " * "

I I I I I I I 1 I 1

APPEND 1X B l i " Measurement and Analytical Procedures Camma Scintillation Measurement I Walkover surface scans were performed using Eberline Model PRM 6 portable b ratemeters with Victoreen Model 489 55 gamma scintillation probes. ,,.3 , Alpha and Beta Camma Scans and Measurements Floors were scanned for elevated alpha / beta levels by passing slowly over the surface with a Ludlum Model 239 1 Cas Alpha Proportional Floor Monitor with 2 a 600 cm sensitive area. Other surfaces were scanned for elevated levels by passing slowly over the surface with Eberline Model PRS 1 portable scaler /ratemeters coupled to Victoreen Model 489 110 beta gamma

• pancake" detectors and Eberline Model AC 3-7 alpha scintillation probes.

Measurements of total alpha radiation levels were performed using Eberline Model PRS 1 portable scaler /ratemeters with Model AC-3 7 alpha scintillation probes. Measurement of direct beta gamma radiation levels were performed using Eberline Model PRS 1 portable scaler /ratemeters with Model HP 260 thin window pancake GM probes. Count rates (cpm) were converted to disintegration rates 2 (dpm/100 cm ) by dividing the net rate by the 4x efficiency and correcting for active area of the detector. The effective window area was 59 cm2 for the alpha detectors and 15 cm2 for the GM detectors. The average background count rate was approximately 2 cpm for alpha probes and 40 cpm for the GM probes. Removable Contamination Measurements Cross Alpha and Cross Beta Smears for determination of removable contamination levels were collected on numbered filter paper disks 47 mm in diameter, thsn placed in individually I

l 1 1 l l-labeled envelopes with the location and other pertinent information recorded. I The smears were counted on a low background gas proportional alpha beta counter. Tritium For determination of removable tritium contamination, smear papers were i cut into small pieces approximately 1/4"

square, and combined with an appropriate scintillation fluid. Two unused smear papers were prepared in the same way. One was used to correct for background and the other was spiked with I

an NIST traceable tritius standard to provide an internal calibration factor and compensate for instrument drift. Camma Exposure Rate Measurements I 1 Measurements of gamma exposure rates were performed using a Reuter Stokes i pressurized ionization chamber. The average of a minimum of five readings was determined at a distance of 1 meter from the surface to the center of the chamber. Camma spectra were collected at each location where exposure rate measurements were made, using a portable Hal(T1) detector and multichannel analyrer system. Uncertainties and Detection Limits I The uncertainties associated with the analytical data presented in the tables of this report, represent the 9% (2a) confidence levels for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels. When the net sample count was less i than the (2a) statistical deviation of the background count, the sample concentration was reported as less than the minimum detectable concentration (<MDC). Because of variations in background levels and the effects of the Compton continuum caused by other radionuclides in the samples, the MDC's for specific radionuclides differ from sample to sample. I i

1 Calibration and Quality Assurance I Laboratory and field survey procedures are documented in manuals developed specifically for the Oak Ridge Associated Universities' Radiological Site Assessment Program. I With the exception of the measurements conducted with portable gamma I scintillation survey meters, instruments were calibrated with NBS traceable standards. The calibration procedures for th$ portable gamma instruments are performed by comparison with an NBS calibrated pressurized ionization chamber. I Quality control procedures on all instruments included daily background and check source measurements to confirm equipment operation within acceptable statistical fluctuations. The ORAU laboratory participates in the EPA and E.ML Quality Assurance Programs. -e 1 4 LI I i m M

.i.... .ni i.. ia-.- 9 3 s 5 l APPENDIX C REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS

7 ...s 1"$e/dk55U5.IYbRY GUIDE \\'e,run ca / DiaBCTORATE OF RESULATORY STANDARDS r REGULATORY GUIDE 1.88 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS ~ A, INTRODUCTION A bcensee having r. possasuomonly heanse must ] retam, with the Part 50 heense, authonzation for special g Section 50.51, "Durauon of license, renewn!," of 10 nudear matenal (10 CFR Pan 70, "Specul Nudear CFR Pan 50, "Licenang of Production and Utihr.ation Matenal"), byptoduct matenal (10 CFR Part 30," Rules Faciliues," toqmres that nach heense to operate a of General Appbcabihty to Licensms of Byproduct 1 production and utilizanon facihty be issued for a Matenal"), and source matenal (10 CFR Part 40, specified durauon. Upon expiration of the specified "Licensms of Source Matenal"), until the fuel, radio-penod, the license may be either renewed or termmated settve components, and sources are removed from the by the Comnussion. Section 50.82, "Applicauons for faciuty. Appropriate administrative controls and facility termint. tion cf beenses," recifies the requirements that reqmtements are imposed by the Part 50 license and the must be natufied to termmate an operating beense, technical specifications to assure that proper surveillance including the requirement that the dismantlement of the is perfonned and that the reactor facility is maintained facility and dispom! of the component parts not be in a safe condition and not operated. mimical to the common defense and secunty or to the health an! safety of the public. This guide describes A posesion only license permits various options and methods and procedures considered acceptable by the procedures for decommmientng, such as mothbaning, Regulatory staff for the terminaton of operstmg entombment. or dismanthns. The requirements imposed licenses for nudear reactors. The Advisory Committee tepend on the opuen selected. on Reactor Safeguards has been consulted concernmg this guide and has concuned in the regulatory postoon. Section 50.82 provides that the licensee may dis-mantle and dupose of the component parts of a nudear B. DISCUS $10N reactor in accordance with existing regulations. For research reactors and entical facihties, this has usually When a heensee decides to terminate hu nudear rneant the duassembly of a reactor and its shipment reacto operstmg license, he may, as a first step m tne offute, somenmes to another appropnately bcensed process, request that hn operstmg beense be amended to organt:.auen for further use. The ste from which a restnet him to possess but not operate the facihty The reactor has been temoved must be decontammated, as advantage to the beensoc of convening to such a necessary, and inspected by the Commtes on to deter. posesmon only bcense a reduced survelhance require, mine whether unrestncted access can be approved. In rnenu in that psnodic surveinance of equipment im. the caw of nudear power reacton, dismandeg has portant to the safety of reactor operation is no longer usually been accomplished by shipping fuel offsite, required. Once tha posesmon only license u issued, makmg the reactor inoperable, and d:sposms of some of reactor operation is mot permitted. Other acuvmes the radioactive componenu. related to cenation of operations such as unloadmg fuel from the reactor and placang tt in storage (either onnte Radioscuve cornponents may be either shipped off-of offsite) may be conunued. ste for bunal at an authonaed bunal, ground or secured uaASC REGULATORY cutDEs came e e.ee n, een =e .amma w go-w",=Y $*nsa a:I " Tee".anYeE I .I.dweepe mYses

a. see e.e e we

"" '*l"':ll" ::l"::"e; ll, lll"'em s,.: e, :'" :: 7.;:'d' ; I, '" "C".".? 60"l:'Ji"..'""'""~ """'"" "'- 7 ~ s

"" l:.::"'."'.t:".".47 """*.".'.7 ':l::",";,"* '." ".,,".

2 - e se er a gysum ssee t e.ur Am.s.sest.-

4. **eassa..

t. D 4e

1. -

e se$ ese aAusi e ~~ sus een eessmen sue

J on the ste. Thow radaoactive matenals remaarung on the Guids and waste should be removed from the ste. l ste mus be teolated from the pubbe by phyncal bamen Adequate radnauon rnonstoring, environmental surve0 l or othe means to prevent pubbc acces to hazardous lanes, and appropnete secunty procedures abould be ieveis of radiauon. Surveillance u necessary to amure the estabhahed under a posseemon only heense to ensure that long term mingnty of the bamen. The amount of the health and safety of the pubhc is not endangered. s l survealance required depends upon (1) the potenua! C hazard to the health and safety of the pubbc from

b. la. Place Entombasent. In. place entombment con-radioacuve matenal remanrung on the ate and (2) the sets of seahng all the remammg highly radioactin or integnty of the physical bamen. Before areas may be contammated compononu (e.g., the pressure vessel and l

rentased for unrestncted use, they must han been reactor mternals) withm a structure integral with the i / decontammated or the radioactmty must have decayed biolopcal shield after hovmg aD fuel asembbes, radio. [ to les than presenbed brruts (Table 1), acuve fluids and wastes, and cartam selected com. j ponenu shipped offsite. The structure should prende l The har.ard asociated with the retired factitty u mtepnty over the penod of time m which sipuncant [ evajusted by conadering the amount and type of quanuues (greater than Table I levels) of radioactmty = ( remauung contarranauen, the degree of confmernent of remam with the matena] m the entombment. An the remammg radioacuve metenais, the physical secuntv appropnate and contmumg surveulance program should provided by the confmement, the susceptibt,y to be estabushed under a possesmon only license. release of radiation as a result of natural,,.anorriena. = and the duration ofisquired surveillance

c. Removal of Radioactive Components and Dio.

l mentity All fuel anembbes, radioacuve Duids and C. REGULATORY POSITION waste, and other matenals havmg actmuss above ac. rn cepted unrestnetad scunty levels (Table 1) should be l

1. APPUCATION FOR A UCENSE TO POHN BUT removed from the site. The facihty owner may then have l

NOT OPERATE (POSSESSION ONLY UCENSE) unrestncted use of the site with no requirement for a f boense. If the facihty owner so desrea, the remamder of ) A request to amend an operstmg beense to a the reactor facihty may be dumantled and all vesuges ( ponesion only beense thould be made to the Dtreetor removed and disposed of. / of 1.acenang. U.S. Atonne Energy Commission, Washmg. ( ton, D.C. 20545. The request should mclude the

d. Conversion to a New Nucisar Symen or a Fosa0 l

foDowmg mformauon: Fuel System. Thu alternauve, which appbes only to i nuclear power plants, utihzes the existmg turbme system

a. A descnpuon of the current status of the facQ1ty.

with a new steam supply system. The onpnal nuclear steam supply rystem should be separated from the

b. A desenpuen of measures that will be taken to electric genersung system and disposed of m accordance i

I prevent enucahty or rescuvity changes and to rmrumtze with one of the prenous three rettnment alternauves. releases of radioacuvity from the facihty.

3. SURVE!LLANCE AND SECURITY FOR THE RE-1

c. Any propcsed changes to the technical spectSca-TIREMENT ALTERNATIVES WHOSE FINAL cons that reDect the possession only facihty status and STATUS REQUIRES A POMml0N ONLY the necessary duamambly/retuement acuvices to be UCENSE performed.

i A facibty winch has been licensed under a poses-

d. A safety analyas of both the actmues to be non only license may contam a significant amount of accomplahed and the proposed changes to the techrucal radioactmty m the form of acevated and contamtnated spectSest ans.

hardware and structural matenala. Suneinance and commensurate secunty abould be provided to asure that f

e. An inventory of acuvated matenals and their the pubhc health and safety are not endangered.

{ locaton in the facihty. }

a. Physical secunty to prevent inadvertent exposun

1. ALTERNATTVES FOR REACTOR RETIREMENT of personnel should be provided by muluple locked

{ bamers. The presence of these bamers should make it 1 Four alternauns for reurement of nuclear reactor extremely difncult for an.inasthonzed person to gatn facihties are consdered acceptable by the Regulatory acces to areas when radaauoc or contammation levels stafL These are: exceed those speciSed in Regulatory Pomuon C.4. To pavent inadvertent exposure, radtation areas abow $ i

a. Mothbalh4 Mothbaums of a nuclear reactor mR/hr, such as near the activated pnmary systern of a facuity conasts of puttmg the faciUty in a state of power plant, should be appropnately marked and abould protective storage. In general, the facihty may be left not be accessible except by cuttmg of welded closures or l

intact except that all fuel asembbes and the radsoacuve the disassembly and removal of substanual structures C-2 m

= D and/or shaeldmg matenal. Mears such as a remote. (1) Enytronmental surveys, l readout mtrumon alarm system should be provided to l in&cate to deugnated penormel when a phyncal barner (2) Facibry ratauon surtys. F is penetratec Secunty personnel that provide access control to the fachty may be used instead of the (3) Inspecuons of the phyucal bamers, and phyucal barnen and the mtruson alarm systerra. (4) Abnormal occurrences.

b. The physcal barners to unauthorned entrance mto the facihty, e.g., fences, buDd ngs, welded doon.

= 1 and access openmes, should be mapected at least

4. DECONTAMINATION FOR REl. EASE FOR UN-cuarterly to anure that these carnen have not detenor-RESTRICTED USE ated and that locks and lociang apparstus are mtact.

If it ts destred to termmate a heense and to thrntnate

c. A facthty radtauon survey should be performed at any further survealanca requirements, the fac:hty should least quarterly to venfy that no radioactive matenal is be sufLciently decontammated to prevent nsk to the

~ escaping or bemg traruported through the contamment pubbe h&alth and safety. After the decontarmnation u l bamen m the facdtty. Samphng should be done along seusfactortly accomplished and the site inspected by p the most probable path by which radioscuve matenal the Commasson, the Commuason may autherne the such as that stored m the mner containment repons beense to be termmated and the facility abandoned or Z could be transported to the outer regions of the facihty released for unrestncted use. The licensee should per. Z and ulumately to the environs. form the decontaminauon unng the followmg gmde-hnes: ~ d.An environmental radiation survey should be j performe1 at least semaannually to venfy that no

a. The bcensee should make a reasonable effort to

~l msnficant amounts of radaauon have been released to the ehmmate residual contammauon. environmtat from the facihty. Samples such as 500, g vegetat on, and water should be taken at locations for

b. No covermg should be appbed to radioactive j

wruch stausucal data has been estabbshed dunng reactor surfaces of equipment or structures by pamt, platmg, or y operations. other covermt matena) untilit is known that contanuna-uon levels (determmed by a survey and documented) are

c. A sste representauve should be des 2 nated to be below the hmru specfied m Table 1. In addition, a

= I responsible for controlhng authonted access mto and reasonable effort should be made (and documented) to movement withm the facArty. further mmunne contammat2on pnot to any such covenng. l

f. Admarustrative procedures should be estabhshed l

for the noufi:aton and reportmg of abnormal occur-

c. The radeanmty of the miener surfaces of pipes, rences such n (1) the entrance of an unauthonted dram hnes, or ductwork should be determined by penon or persons mio the facaity and (2) a ugnificant rnahang measuremenu at all traps and other apprepnate I

change m the radtauon or contammauon levels m the access potnu, provided contanunacon at these locatons facity or the offute envuonment. is likely to be representauve of contammation on the mtenor of the pipes, drun hnet, or ductwork. Surfaces s The following reporu should be made: of prerruses, ecmpment, or scrap which are likely to be l contammated bu: are of such ure, construeuon, or P (1) An armual reput to the Director of Lacentag, locauon as to maxe the surface inaccessiDie for purposes j U.S. Atorme Energy Commismon, Washmrton, D.C. of measurement snould be asumed to be contamtnated 20545, describing the resulu of the environmental and m excess of the permusable raiation hmits. i facihty radiauon surveys, the status of the facility, and an evaluauon of the priormance of secunty and

d. Upon request, the Comrmston may authorne a survenlance measurea, licensee to rehnquuh possesuon or control of prermses, l

equipment, or scrap having surfaces contanunated m l (2) An abnormal occurrence report to the Regula-excess of the hmits specified. Thu may include, but u tory Operanons Regional Office by telephone withm 24 not hmated to, spe=al ctreumstances such as the trsnsfer houn of dacowry of an abnormal occurrence. The of prermnes to another licensed organization Lat wiD l abnormal occurrence wGl also be reported m the armual contmue to work with radioacuve matenals. Requests y report described in the precedmg item. for such authonzauon should provide.

h. Records or loss relauve to the foDowmg items (1) DetaDed, specific informauen desen'bing the should be kept and retamed until the beanse is temu-premnes, eqmpment, scrap, and radioacuve contam)-

nated, after wtuch they may be stored with other piant nanu and the nature, extent, and degree of rendual records: surface contammauen. C-3

I (:) A detaled health and safety analysa indi. of a change m the techmcal specifications should be canng that the residual amounu of materiah on surface renewed and approved m accordance with the require. I areu, together with other consderations such as the ments of 10 CFR $50.59. prospecuve um of the premises, eoutpment,or scrap.are unlikely to result m an unreasonabit rak to the health lf rnejor structural changes to radioactive components and safety of the pubhc. of the facility are planned, sudi u removal of the I pressure vesel or mapor componenu of the pnmary

e. Pnot to release of the premtses for unrestncted rystem, a chamantlement plan ecludmg the irtformauon use, the licensee'snould make a compreheneve radtauon required by {50.8: should be subtrutted to the Comma.

survev estabhslung that contammation a withm the men. A diamantiement plan should be subtrutted for aD I hmiu'specified tri Table 1 A survey report should be the alterneuves of Regulatory Poution CO except filed witn the Director of Licenang. U.S. Atomic Energy mothbalhng. Howewr, mmor disasembly acuvines may Comrmuion, Washmrton D.C. 20545. with a copy to stiU be performed in the absence of such a plan, i the Director of the Regulatory Operations Reg onal provided they are permitted by eustmg operaung and Ofnce havmg juruchetton. Tie report should be filed at mantenance procedures. A dtemantlement plan should least 30 days pnor to the planned date of abandonment, include the fouowing' The survey report should: (1) Idenufy the premises;

b. A desenpuon of the dismantling activities and the I

(:) Show that ranonable effort has been made to precautions to be taken. reduce residual contammation to u low as pracucable levels;

c. A safety analyns of the dismantling activities including any etnuenu which may be released.

I (3) Describe the scope of the surwy and the general procedures followed; and

d. A safety analyas of the facility in su ultimate status.

(4) State the fmding of the surwy in uniu I specified in Table 1. Upon sausfactory fewew and approval of the dis. mantling plan, a dismantling order is issued by the After renew of the report, the Commision may Commission m accordance with (50.82. When dis. I mspect the faccities to confirm the survey prior to mantling is completed and the Commismen has been gnnung approval for abandonment. notified by letter, the apprepnate Regulatory Opera. uons Reponal Office inspecu the facility and verifies

5. REACTOR RETIREMENT PROCEDURES completon in accordance with the dismantlement plan.

I If residual radiauon levels do not exceed the values m As indicated in Regulatory Pontion C2, uveral Table 1, the Cornmmmon may terminate the hcense, If alternauves are acceptable for reactor facility reurement. these levels are exceeded, the licensee retuns the if mmer disassembly or "mothballing" is planned, this possession only bcense under which the dhmantling could be done by the ensung operstmg and mamte. activites have been conducted or, as an altern6tiw,may nance procedures under the license m effect. Any maxe appbcauon to the State (if an Agreement State) planned acuens involving an unreviewed safety question for a byproduct matenals beense, c.:.

TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS l REMOVABLEbe MAXIMUMbd NUCUDEa AVERAGEDC U.nat, U 235, U 238, and $ 000 dpm a/100 end !$ 000 dpm all00 cm2 1.000 dpm all00 cm2 assocated decay producu Transuranics, Ra 226, Ra. 28. 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th.030. Th 008, Pa 231. ~ Ac 027. I*125,I 129 Th.nat, Th 230. $r 90. 1000 dpm/100 em 3000 dpm/100 cm2 200 dpm/100 cm2 Ra2:3.Ra24. U 232, 1 126,1131.1 133 Beu pmme emitters (nuclides. 5000 dpm M/100 cm2 15,000 dpm M/100 cm2 1000 dpm h/100 cm2 with decay modes other than alpha emission or spontaneous fission) except St 90 and others noted above, aWhen surface monummeuen by both a)phe and beu pomma emitting nachdes exists. the bmits establahad ler alphe 6nd beta pmme emstting authees should apply independently. bas used in the table, dpm (dtentsprouens per nunute) means the rete of emismen by radioactive metenal as deurmaned by correcting the counu per meuw observed by an appropnote deuetor for backpound, emenney, and pometnc factors ansonsted with the innrumenuuom. 8 Measuremenu of evente conummant should not be aversped over more than 1 squan meter. For objecu of less surface area, the aversp should be dertved for each nach obpect 2 OThe snaximum conumanation level oppbes to an area of not more than 100 cm. 3 'The amount of removsale rodiesettre metenal per 100 cm of surfees ares should be determined by wiping that area with dry filter or soft absorbent papes, applymt moderete prouvre, and amesang the amount of radioactive materal on the wipe wuh an appropnate innrument of known emetency. When removable contamination on objecu of leu surfaa area is deternuned, the perunent leveh should be reduced proporuonally and the entare surface should be wiped. C-5 ~~

s .w 1 1 I I I I APPENDIX D i i PROPOSED CONFIRMATORY SURVEY PLAN g FOR THE L.77 REACTOR FACILITY i l UNIVERSITY OF CALIFORNIA 1 SANTA BARBARA, CALIFORNIA I a I L I i I f I i 1 4 i

1 l il y .I i PROPOSED CONFIRMATORY SURVEY PIAN FOR THE-8 L 77 REACTOR FACILITY UNIVERSITY OF CALIFORNIA: SANTA BARBARA, CALIFORNIA I. Site History and Description i The L-77 Reactor, located in Building 572 (Physics Unit 1) of the University of California, Santa Barbara, California (UCSB), is a Nuclear-1 Regulatory Commission (NRC) licensed facility (License #R 124). The-reactor originally was installed and operated at the University of 1

Nevada, Reno, Nevada, and was moved M UCSB in 1974 UCSB. received a i

Facility Operating License on Decembn e, 1974,_and was authorized for startup in January 1975. The maximum operating power was 10 watts I (thermal). j! License R 124 specifically designated the= area of operation as Room 1251, in the eastern end of Building 572. This room-is a hit,h bay area, constructed of concrete block walls, on concrete slab, poured on grade. The area above the ceiling is an accessible roof structure. Room 1251 was subdivided-into two areas: area. one was the reactor zone and { control console, and area two housed a Nuclear Chicago subcritical j assembly which was operated.under-a State'of California radioactive i materials license. An opening was cut into the northern wall of Room I 1251 to facilitate removal of the reactor core assembly. This opening provides access to Room 1356 which is currently being used for storage. The L 77 was a homogenous aqueous solution research reactor..The fuel solution was enriched uranyl sulfate dissolved in water, contained in _l the Reactor Core Tank, which was contained in an inner shield tank. The rhielding design permitted operation of the reactor while personnel =were present in the Reactor Room. Prepared by the Environmental Survey and-Site Assessment Program of Oak Ridge-Associated Universities. Oak Ridge',' TN, under interagency agreement (NRC. Fin. 8 No. A-9400) between tb U.S. Nuclear, Regulatory Commission and the U.S. Department of Energy. June 8, 1989 'D-1 m.

i o UCSB submitted a decommissioning plan to the NRC and a dismantling order k was issued on August 26, 1986. UCSB has completed decommissioning of l the L 77 facility and has submitted a close-out survey report to the-NRC. II, Purpose l I Oak Ridge Associated Universities has been requested by the Nuclear Regulatory Commission's Region V Office.to perform a survey to confirm l the radiological data presented by the University of California Santa Barbara, relative to release of the facility for unrestricted use.; I III. B.esponsibility Work described in this survey plan will be. performed under the supervision of Mr. J. D. Berger, Program Manager with the Environmental- ' i Survey and Site Assessment Program of Oak Ridge Associated Universities'- (ORAU). l' IV. Procedures 1. Oak Ridg,e Associated Universities. will review UCSB's close out survey report and supporting documentation concerning sitet decommissioning activities. 2. A one metsr grid will be established on the floor.and= lower wall (up to 2 m) areas of the reactor bay and associated control room. The upper walls will not be gridded. ' Measurements and samples from the ungridded surfaces will be.. referenced to.the floor and lower l wall grid, or to pertinent building features. 3. The floor-and lower walls.will be surface scanned using NaI(Tl). c,amma scintillation detectors (maximum distance from surface 5 cm)', ZnS alpha detectors (maximum distance from suhface 1 cm), and D-2

m . s. I P "p anc ake CM beta. gamma detectors (maximum distance from surface ~ 1 cm). Locations of elevated readings will be noted for further-l investigation. 7 l I 4. Exposure rate measurements will be made' with a pressurized l ionization chamber at a minimum of five locations at one meter from the floor, lower walls, and areas of elevated gamma radiation identified by surface scans. The background exposure rate will be .I established with the pressurized ionization chamber. The location for determination of the background exposure rate will be an area l 1 that is not radiologically contaminated, outside the restricted I area and of similar material and construction history. 5. Measurements of total and removable. alpha and. beta gamma cont; amination will be performed on a minimum of 20 of the floor and j 8 of the lower wall grid blocks, selected at random. -One set of-five direct measurements will be obtained for each surveyed grid

block, and one smear -will be taken ifor' each' s'et of five measurements.

Additional measurements will be performed -at t locations identified by the surface scan, I s 6. Direct measurements and smears will be obtained on the. upper walls and ceilings. Particular attention will be given to cracks, beams, j

piping, ledges, ducts, and other surfaces where material might settle or accumulate.

These surveys will include the inside surfaces of any drains and exhaust air duci.s. The number of survey locations will be determined by results as the survey progresses; however, a minimum of 20 measurements will'be. performed. l u 7. Direct measurements and smears-will be obtained at locations of i elevated contact radiation levels identified by the-surface scans. l 8. Samples of residues, including water, will be collected from floor cracks or joints, beams, inside and outside of piping, ledges, air ~ j ducts, and other surfaces as appropriate, l

]s 0 ^ 9. Other sample media and locations will be added to the survey, based: on findings as the survey progresses. l y l~ -l V. Data and Sample Analysis l Direct measurements, exposure ~ rate data,.and gamma spectra will be evaluated onsite to determine the need for additional decontamination. i Smears and other samples will be returned to ORAU laboratories in Oak' j

Ridge, Tennessee, for analysis.

Findings of the independent-measurements will be compared to Regulatory Guide 1.86 and the h1C I Docket No. 50-433 requirement which states that ambient exposure rates' j 4 must not exceed 5. pR/h above background at 1 m from any surface, VI. Tentative Schedule ? j Measurements and Samplin ; , June 12-16, 1989 e Sample Analysis July 1-15, 1989 l Draft Report August 15, 1989 1 I i i I I q I D-4 ll l llJ}}