ML20041E722

Forwards Responses to NRC 820224 Questions Re Final SAR, Including Supporting Data & 22 Replacement Pages for Sar. Revised Pages Correct Minor Typographical Errors or Update & Clarify Procedures

ML20041E722
Person / Time
Site:	05000134
Issue date:	03/03/1982
From:	Bolz R, Wilbur L WORCESTER POLYTECHNIC INSTITUTE, WORCESTER, MA
To:	John Miller Office of Nuclear Reactor Regulation
References
NUDOCS 8203110277
Download: ML20041E722 (58)

Text

{{#Wiki_filter:-_ WORCESTER Worcester POLYTECHNIC Massachusetts 01609 J INSTITUTE (617)793-5000 + 4 March 3, 1982 i Dr. James R. Miller, Chief Standardization and Special Project Branch Division of Licensing USNRC Washington, DC 20555 Docket No.' 50-134 i

Dear Dr. Miller:

In response to your letter of February 24, 1982, I enclose a list of answers to the 17 questions you asked together with supporting data. Also included are 19 copies'each of 22 replace-ment pages for our SAR. Most of the revised pages correct minor typographical errors i called to our attention by the NRC-LANL site visit team or update and clarify procedures in response to the questions asked. Sub-stantive changes appear on pages 43, 44 and 45 to satisfy question 6 of the questions you forwarded. If any further information is required, please contact us. i Sincerely, [ WA L. C. Wilbur, Professor of Mechanical Engineering i and i Director, Nuclear Reactor Facility L LCW/ns Approved: /' [ R. EbBol'z k ( Vice Preside.n [ nd Dean of Faculty bl 6 1g 8203110277 820303 PDR ADOCK 05000134 P PDR

WPI Nuclear Reactor Facility Docket # 50-134 March 3,1982 Response to NRC/LANL " Final Questions" 1. The WPI reactor power level instruments are calibrated at least twice a year, usually immediately following our annual maintenance program during the sumrer, and again during January or February. The results are recorded and subr.sitted to the Radiation, Health and Safeguards Committee (RHSC) at the October and April quarterly meetings. A copy of our current procedures for power calibration is enclosed (Exhibit 1). Basically, the procedure consists of irradiating a gold foil and counting it in a manner which enables us to determine the counting efficiency based on NBS irradiated foils. Knowing the cadmium ratio for gold at the chosen location, and the ratio of thermal flux at the irradiation location to the core center flux, and thereby to the core spatial average flux, we can calculate the true power level. If necessary we can then adjust the detector location to produce an accurate or slightly conservative instrument reading. 2. The t,eam port plug or external shield and the beam port shutter are visually checked as part of our start up procedure. A copy of our current start up check list is attached (Exhibit 2). This list must be completed prior to the first reactor operation of any day. 3. We have both a vacuum breaker system, as originally installed at the facility, and a relatively new back flow preventer system, specifications enclosod (Exhibit 3), mandated by the State. The bottom of the pool inlet water tine is located less than 4 inches below the maximum high water level in the pal so that if all systems could somehow fail, less than 4 inches of pool water could be syphoned. The radioactivity of the demineralized pool water is normally less than that of tap water. Four inches also defines per-missible variation in pool water level. The variation is caused by evapora-tion and manually controlled replenishment of the pool water. The city water line is normally valved off when not in supervised use to supply make-up water. 4. A one liter sample of pool water is forwarded to the Commonwealth of Massachusetts Lawrence Experiment Station Radiological Health Laboratory annually. No 8, y reading significantly above background has been found. In addition we measure the pool water gamma activity in house whenever the domineralizer is regenerated and at random intervals. The pool water purity is checked by measuring its resistivity during every checkout. 5. In case of power failure during operation, the WPI reactor automati-cally scrams and is then secured until power is restored. fio emergency back-up power is provided for the control systems. The area radiation monitors have a back-up system consisting of gel cells, trickle charged, which power the three fixed-position radiation monitors in case of power failure. The building evacuation alarm horns and strobe lights are battery powered by a system maintained by American District Telegraph.

-_ -. ~_ = - - -- _ _ _ _ _. - _. _ 2. 6. As we explained to the flRC/LAflL team, the results on pages 43 and 44 i --of-our-SAR-for 10 KW were obtained by scaling the original GE calculations 1 for a 1 kw reactor and making corrections for a different volume etc. The GE figures were sometimes grossly rounded off, as, for example, calling a cal-4 culated dose of 0.24 "less than 1" etc. More importantly they used releases in " curies" and failed to properly clarify the difference between case A, where 0.012% solids were assumed released, and case B where no solids were released. Since we do not possess GE's detailed calculations, we have elected to make the release and dose calculations ourselves, i In agreement with GE, we assumed 1.5% of the total fission product build-up is in the 1/4 melted element. 'Of this amount all the noble gases, i 1% of the halogens, and 1% of the. solids are released. We then used a uniform i mixing model and appropriate tables of 8s and ys emitted based on a Stone and Webster Radiation Protection Group Computer Program library, " Activity 2", and calculated a total dose of 35.5 rem /hr. In case 8 we used the GE assumption that only fission products which i vaporize at below 10000 F would be released from the single damaged plate, and that all the gas atoms within 5 microns of the surface of the plate are emitted. I Using the same techniques as in example A we found a dose rate of about 0.2 rem /h r. l We believe these numbers to be mort. accurate and realistic than the much larger values used by GE. A copy of our data base and calculations is attached (Exhibit 4). i 7. Our use of the WPI reactor and associated facilities is such that over the history of the facility we have never had a ubstantive radiation release or personnel exposure. Only trivial amounts of AR 41 are released at 10 kw operation, and this release is monitored for reactor operation at power levels over 1 kw. Both film badge records and dosimeter records indicate essentially background radiation doses to all personnel at all times. A copy l of our current Laboratory Safety Rules is attached (Exhibit 5). 8. WPI uses three five decade GM area monitoring systems, Victoreen Model 855, with ranges of 0.01 mR/hr to 1000 mR/hr. The maximum reading l. unocr all normal operating conditions is less than 40 mR/hr. on the highest l reading detector. A copy of the monitor specifications is attached (Exhibit 6). l Each area monitor is calibrated quarterly and the results are reported at the quarterly meeting of the RHSC. The calibration is accomplished by using a commercially manufactured and calibrated Cs 137 source positioned at several fixed distances from the detector, using a source and detector fixture con-structed in house and by following Victoreen electronic specifications. i 9 Whenever the WPI reactor is in operation or whenever license level radioisotopes are being used, all workers, students and visitors to the facility are required to wear personnel monitoring devices. Facili ty radia-tion workers use an external film badge monitoring service and all others wear pocket dosimeters. These requirements are instituted primarily for edu - cational purposes since the radiation exposures are virtually always below readable levels. The dosimeters are calibrated twice a year during student l experiments using Cs 137 or Co 60 sealed sources. The absence of readable levels of dose are consistently confirmed by the film Dadge records, t ~ ....r --. 7. ,y_..7, .-r,-, - -~~

~ 3. 10. The 50 mrem figure was based on the fact that our laboratory oriented courses are only a few hours a week and 7 weeks long and we have never had any sustained activity which would begin to test the limits of 10 CFR 20.202(a)(1). To clarify our compliance we have recently changed our RHSC rules to specify not more than 20 mrem in 7 consecutive days, bringing us into agreement with 10 CFR 20.202(a)(1). In light of item 9 above, the change will not have any effect on our operation. 11. No liquid or solid waste with radioactivity above normal background has been released or shipped from the WPI reactor during the past 10 years. Most of our irradiations involve solid foils, wires or powders which result in short lived radioactive elements. After suitable decay the samples are reused. Nearly all radioisotopes produced ir. liquid or solid form are made in micro-curie levels and have sufficiently short lifetimes that all measurable radio-activity is gone in less than a year from the irradiation date. The excep-tions are very small in quantity and activity and are stored on site. 12. The set point for the alarm on the radiation detector in the exhaust system is 2000 cpm. Based on flow measurements and manufacturer's data for the detector this translates to less than 0.15 Ci annually or 0.003 pu Ci/cc averaged over the total air flow for one year and based on 200 hours of reactor operation, all at 10 kw. In 22 years of operation the reactor has never been operated as much as 120 hours during a year. Also, we no.mally operate at well below the set point and approach it only during relatively long full power operation. 13. Our reactor records are based on chart use intervals rather than cal-endar years. Over 22 years we have used 26 charts for an average of about 10 months per chart. The total power produced from December 1959 to flay 1981 totals 6675 kw hr. The total number of hours during which the power level exceeded 0.1 watts from April 1968 to May 1981 is 1362 hours. Based on 28 actual weeks of academic activity per year, this would be an average of about 4 hours per academic work week. A table of chart dates, chart hours, kw hrs and total power produced is attached (Exhibit 7). 14. Three modifications have been made to the WPI Reactor since 1959. l a) During the summer of 1961 graphite was installed in the thermal column to replace the shielding bricks first used. The first critical opera-tion of tha reactor with the graphite in place was on August 24, 1961. b) On November 14, 1967, WPI was granted a license revision allcwing a maximum power level of 10 kwt, but contingent on the operation of an exhaust I system and Ar 41 monitor for all operations above 1 kwt. The exhaust system and monitor were installed over a period of several weeks, and the first operation of the WPI reactor at above 1 kwt took place January 30, 1968 with l the Ar 41 monitor on line. c) A neutron radiograph was installed in the beam tube of the WPI reactor and a concrete shield was designed and fabricated to replace the beam port shield plug. The shield block consists of about 3 tons of concrete and is about 171 cm long by 78 cm wide by 79 cm high. An internal cavity l contains the radiograph transport system. The reactor was first brought critical with the shield block in place for shielding evaluation on October 21, 1969.

9 .15. ,, Sprinkler heads are provided throughout the reactor bay and through-out Washburn Laboratory. Firemen and their supervisors have assenbled and toured the reactor facility on several occasions, the most recent being in October 1981. Evacuation fire alarm boxes are located in Washburn but no fire alarm system in Washburn goes directly to Security or the Worcester Fire Department. The nearest fire station to Washburn is on Grove Street, approximately 0.6 car miles away. There are two dry chemical fire extin-guishers, maintained by WPI Security, located inside the Reactor Facility. 16. In addition to the ADT monitored fixed area monitors within the facility, periodic surveys made with portable detectors, and semi-annual wipe tests made in the facility, air dust samples are obtained from the facility exhaust duct and counted semi-annually. Also semi-annually wipe tests are made at six campus locations external to the facility as listed in our SAR Pg. E18. During the history of our license no wipe test or dust sample has shown any evidence af being in excess of normal background. 1 I

9 17. 5 9 President of WPI Dr. E. Cranch V. President & Dean of Faculty 4 Dr. R. E. Bolz 1 I RHSC ME Department j Dr. R. Goloskie, Assoc. Prof. Prof. D. N. Zwiep Head of Physics, Chairman I l i!- l Reactor Facility Director i Prof. L. C. Wilbur, Prof. of ME i f Nuclear Specialist RSO I SR0 i Mr. Thomas White

E k LA l-s . - (. i9 RHSC / l-23-76 I Power Level Calibration Procedure The reactor power level is calculated from the equation Ot If V p, 10 3.1 X 10 A gamma scan of the fuel indicated that the ratio of 0 center to U235 is 577, the non-h factor is 0 avg is 1.7. We assume that of 0.97 and the core loading is 3269 gm of U235. We further assume that h for the fuel is approximately 1. On this basis the core 6 averaged flux based on maxwellian average velocity is 7.46 x 10 neutrons per square cm per second at I watt. These are conservative 6 figures compared to GE who used 7.7 x 10 and a higher ratio. Substituting, and converting to most probable velocity, we obtain p(center) ny 7 1.123 x 10 A special holder has been constructed of plexiglas that enables us to insert gold foils in the center of the core. The cadmium ratio has been measured there many times and is taken to be 4.08. Gold foils have been irradiated at the NBS pile and counted here at WPI \\in several counting arrays so that the counting efficiency for 1 this geometry foil is well established. A gold foil is inserted in the core center for at least one ~ minute, removed, cooled for an appropriate length of time and counted. The power level can then be calculated and compared with the log record of instrument readings. Gold foils may also be counted in the F3 location in a machined polyethylene block fabricated for the purpose. The thermal flux 'l ratio between 05 and F3 has been found to be about 1.80. The cadmium ratio in F3 is between 16 and 18. These numbers may be refined as further data is collected. Page 18

' Approved RHSC' July 1980 WORCESTER POLYTECHNIC INSTITUTE OPEN POOL REACTOR Startup Check List Date: l. Visual Check of core and facilities Beam port shielded Shutter down Fuel LoadingI I elements Pool level normal Source in Thermal column closed Demineralizerreadingl l B10 Counter in lowest position Pump Running 2. Area monitors checked Set at (50-20-20) mr/hr 3. Chamber power supply voltage set at (+)I Ivolts; (-)l l volts Chamber connected C-1 (+) C-2 (+) (-) Log N (+) Hv Monitor B10 Scaler connection Scaler high voltage set at 4. Recorder check (on and inking) Linear Power Log N Recorder current OK Ln Linear Power Charts dated 5. Scaler 60 cycle check I I Set Gain at 10 Discriminator set at 8.0 6. Master switch in test position 7. Log N Calibrate lo Cal Scram Hi Cal ( 8. Rod Drive Test

1. 1 Withdraw Insert Digicon Manual Rundown
2. 2 Withdraw Insert Digicon Manual Rundown
3. 3 Withdraw Insert Digicon Manual Rundown Reg. Rod Withd:aw Insert Digicon Jog In dog Out Manual Rundown 9.

Master switch in operate Magnet currents 10. Source checks. Trips from 4" out Blade 1 C-1 Trip Pointl IScale 10-8 Scram Reset First Rod permissive check (Log count rate less than 3000 cpm) #1

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11. Period Alarm Set at 15 sec. 12. Source in core position D2 13. Second Rod Permissive check Control blade / reg rod off in limits

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Bladel IManual Scram 15. Log N scram flag set atl IScram OK Recorder on again 16. Source readings Now Before last critical run ( Log CR Metert .l scalerl lmeterl lscalerl I C-1 Down Scale 10-8 C-2 Down Scale 10-8 17 Scaler in continuous mode PERFORMED BY:

quorments of U.it.. coundation f or Cross Connection Control Standard p. for reduced pressure pnnciple backflow preventers. No. 900 Series %",1",1%",1%",2" g

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For Higher Hazard Applications CONSTRUCTION OF BASIC VALVE j . Watts No. 900 Series provides the complete concept in cross S W STRmER W SHOWN) connection control for the protection of potable water, mee ' DIRECTION OF NORMAL FLOW the performance requirements of A.S.S.E. Standard 1013, AWW Standard C 506 and U.S.C. Foundation for Cross Connection UNION O . Control Standard for reduced ressure principle backflow pre. INLET THIRD (Back up) venters. Also, accepted by U.S. i ublic Health Service. CONNECTION CHECK VALVE Beyond its size, weight, and cost advantages, the No. 900 Series offer many other advantages of value to anyone concerned with RELIEF FIRST efficient cross connection control such as design simplicity, low VALVE CHECK flow resistance, quiet operation, simplified installation and OlSC VALVE service; and features a third (back up) check valve for added ASSEMBLY safety. The No. 900 assembly is the only backflow preventer furnished RELIEF complete with strainer, test cocks, and gate valves. Its compact, VALVE well balanced, practical design now brings complete prote tion to SEAT thousands of installations which were not economically possible before such realistic considerations were proven and available. No. 900 is suitat,le for supply pressures up to 175 psi and for REllEF TEST l water temperature up to 140* F. VALVE COCK 2 1 VENT t o Union Connections e Compact size for ease of oStainless steel internal parts installation M ER % TEST 8P " COCK 3 j OStandardly furnis!.ed wnh e Maximum total of 10 parts vE T bronze body strainer required for complete service 3 VALVE) J o Maximum flow at low

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CHECK (SECON A VALVE CHECK i ASSEMBLY VALVE) l h C [ O [ O LET TEST t: -dC GNN ECTION COCK 4 0 ./ "1" ) e y g;,,, Dimensions (Inches) vvei t (In.) A B C D libs.) Patent No. 3,636.968 Patent No. 3,747,621 18 % 7Vs 4 3Vs 14 V2 other U.S. and Foreign Patents Pending 1 21 % 813/is 5% 39/is 15 1% 20 % 9% 6 3% 16 IV2 28 % 10 % 6% 4 39 g 2 29 % 12Vs 8% 43/a 42 Maximum Supply Pressure 175 lbs. No.900 Furnished complete with gate valves, strainer and test cocks, as shown. l l Also available with 0.S. & Y. gate valves. All Bronze Construction 4 For Additional Detailed Information. send for Folder F900 Ey kb

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mm_.a_mu.. -. _. Ikx role irl thc U)VE Cirn (bol 77aisiq (<cGcicr Sahbj. ( Volurn e c f S a liltj ' 30,000 ff') The dos' mit 5 ddcfmined {ir e fun cases descobcci behu). Tiie cases are atadcled for Hie osa rea. dor cperala9 of tokw fir I yea.r. kever, il should be. ncied he nOxthtum recorded tw1 uxG af ID kw (or 8 hoors, otte. Over 0. 20 tjelr pertbd. / 6ssA ... usith as7. reactiieRy thserhart poshdalek losed an GE logic : 19 0f a.ful dentenfmay(met The Emint paiarl kiild-u si ihat niach Gal af 41e core cat cr corresgnds -{o alcat t.y ohne lo(at beuld up at he cn.. 98 psviWe., thcagh unhkety, Mat Some. of he rm% alarmham impid 'nacF wH4 . Jhe alease. af rodthacHre. nble gass Abrn oc tallot attert of lie care proaala liv oilly means & caitanithahh9 ts air rurroatiding &c tudor. ZFis li pssibk. Statimces of stat hudn findads and halo 7ats will be carned alog taith bubbles d noble tpe and dispeisedin itic air abwe. he pool. (linoilages rdensd Asm mdted fad: 1007. nebtegas, t.o1. Infogens, t.oJ Tdidd. Unh9 ute. Hssibn prdact th&:rtfat9 of the _ Adituu z hanun,, lhe dose tale lo a ptsarl th the facibly, assamihq a. anilbrai'mirinj niedel, s t.1 rent /hr gomma. dosc and 3't.t rYn/w bela dass, gerity o folal of 35 5 r m Vnr. _ DSE 8 5'upporc lhof a kal( derefers and he pl draths al a indetale at(c. and Hie daddhy on one or more futt elemeit{s s ofready dann)el. 7a~ simphYy Ihs pratent H & assumed Hnf a saigle plate ofone elentent is Ormpiddy Sinkped of dadding and thatUte kn prodads rdeced 1 are Hnse a>hich Valmije bdaa) toco'f (Ke, h'r, $r, r). d(so, assante the gaseous Ambn prodack (a be untforatty deiabafcd throughoolMe cniefud and-thaf all fks qas afams wim acoil range- (5~ micrents) of fhe surfact art emn/d. Htis results th

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66. CO: VERSION FACTORS _ THYROID DOSES _ d Test 1. TID-14844 (Calculation of Distance Factors for Power an f conversion factors Ret.ctor Sites, Table III, Page 25) gives the table o ld id dose to amount of radioactive iodine inha e iod !)*/Ar = the dose delivered during an infinitely long per relating thyro per curie of radioactive iodine inhaled). (where: tion If the breathing rate is R (m3/hr) in an iodine concentra dose, per hour of A (C1/m ), the total long-term contribution to the 3 breathing, is D = AR (D=/Ar) rem. tive portion Assuming a breathgng rate characteristic of the acm3/sec), t of the work day (3 47 x 10-tion to dose can,be calculated: D=KA IODINE THYROID DOSE CONVERSION FACTORS _ TABLE _C1: E D=/Ar (ren/hr per Ci/m )_ 3 Biological (Rad /Ci_ Iodine Ha* f-Life (Seel 6 Isotope 6 1.85 x 10 5 1.46 x 10 6.57 x 10 k I131 k 6.67 x 10 3 5 35 x lo 8.39 x 10 5 I132 4 0 x 105 5.'00 x 10 r 7.52 x 10 4 I133 4 3.13 x 10 3 2.5 x 10 3.11 x 10 5 Il34 5 1.55 x 10 k 1.24 x 10 2.42 x lo Il35 yMOLE BODY DOSE FROM SEMI-INFINITE CLOUD _ tanding 2. The vbole body beta and ga==a dose rates to a persoa staining x Cur at the center of a hemispherical cloud conradioactive isotope

1. 3. Section C.2.e:

= 0.23 E x Dg g and- = 0.25 E p X f i 1 I v

E i 67. 1 where: Eg and Ej are the average betas and ga-n energies per disintekration (Mev/ dis) and D is the dose rate (rad /sec) .E These average values can be developed fmm the data in the " Table of Isotopes" by Iederer, Hollander an'd Perinnn. This was originally ~ done by R. E. Miles (SW Interoffice Memo, Average Gam =a and Beta Energies, 10 Aug.1972) and is sum arized in the following Table which lists the appropriate conversion factors: D' = K X g g D' = K X Y 5 I where: D' are the dose rates in rem /hr I i ~ I l 1 l k

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1 J 69 s WOLE BODY DOSE IN A ROOM OF FINITE SIZE 3 A. BETA DOSE Since the range of beta particles in air is considerably less ] than the dimensions of typical reactor building compartments and control j rooms, the beta dose can be evaluated by using the semi-infinite cloud model and the Kg conversion factors listed in Table C2. B. CAMMA DOSE The dose rate in a room may be appmximated by assuscing that ] the dose point is at the center of a hemisphere of the same volume as J the room. in air is large compared Since the mean fue path of most gammaa to the dimensions of most rooms, it is appmpriate (and slightJy conservative) In fact, for to call this product of air attenuation and buildup unity. ]. UR < 1 ve have the approximations: e " ";61 '- pR ~ B(r) f 1 + pR and the product of attenuation and buildup e" B(r) # 1 - (pR) W l ~ Then, the dose at the dose point may be calculated by considering the dose and summing over all such due to a typical differential shell element at Ry elements: s 4xr hur For a monoenergetic gamma source R D = f dD = C S R o [ d

f 70. where: = Dose rate at the dose point (rem /hr) D

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I hx$\\. Is. ^h ,T f0y _ SAFETY RULES FOR OPERATI0t1 Iti THE REACTOR LABORATORY ( These regulations, based on those presented by Overman and Clark, " Radioisotope Techniques", apply to all persons who work with or in an area containing radioactive materials. Specifically this includes those persons who are performing any experimental work in (or visiting) the nuclear reactor facility. General 1. Before starting work with radioactive materials each person shall make known any previous work with radioactive materials or radiation sources and any exposure over the maximum permissible dose. 2. Topcoats, hats, and other personal belongings including books (other than those required for work) should not be brought into the laboratory. 3. Eating, drinking, smoking, and application of cosmetics in the laboratory work areas are forbidden. 4. High standards of cleanliness and good housekeeping should be maintained throughout the laboratory: 5. Pipetting liquids of any type by mouth or the performance of any similar operation by mouth suction is forbidden. 6. If, in the course of work, personal contamination is suspected, a survey with a suitable instrument shall be made immediately, to be ( followed by the required cleansing. Routine precautionary surveys should be made at frequent intervals when using liquid or powdered radioactive materials. l Safe Handling 7. As a rule of thumb the total gamma irradiation of any part [ of the body should not exceed 100 mrem / week. The RHSC rules must be l complied with. 8. Pocket ioni::ation chambers or film badges sbli be worn by all persons entering the Nuclear Reactor Facility when the reactor is at power or when radioisotopes requiring a byproduct material license such as WPI Ifcense 20-03680-2 are to be used. 9. Survey meters must be used to check the dosage level at appropriate stages in experiments with radioactive materials. 10. Approved warning signs must be properly displayed in all areas where there is a radiation hazard. This is essential for the pro-tection of everyone who might have cause to enter the laboratory and would, therefore, include not only other students and staff, but also the janitor, watchman, fireman, etc. l ( Page 20 l

Ril5C /' 3-78 ff'(

11. All containers containing radioactive materials, including

( sealed sources and standard sources, should be labeled with radiation warning tape. The isotope, amount, and date should be indicated.

12. Radioactive material, including portable sources, should be properly stored when not in use, and there should be sufficient shielding to reduce the radiation level below 2 mr/hr at the surface of the shield container.

13. As a gUtde to the use of protective gloves it is helpful to remember that the safest policy is to adopt the use of aseptic techniques in handling radioactive materials.

This does not imply that protective gloves must be' worn routinely, but when there is a strong possibility of hand contamination they should be worn as, for example, when handling radioactive liquids. Suitable disposable plastic gloves are available on request.

14. A person with breaks in the skin of the hands must wear protective gloves and should consult his instructor or adviser before starting work with liquid or powdered radioisotopes.

'_Contamina tion l

15. Any suspected contamination of the body or clothing or apparatus must be reported to the staff at once.

( Disposal of Radioactive Waste 16. Disposal of radioactive waste shall confom with 10 CFR 20. l Such disposal shall be carried out only under the supervision of the facility staff. l j. i l l l ( Page 21

b xi. L,'f C SECTION I GENERAL DESCRIPTION 1-1. PURPOSE. G-M Arca Monitoring Systems.dctect and measure gross gamma activity. The Victorcen 855 G-M Area Monitoring System achieves this in and around facilitics where radiation emitting material are handled or processed. The Code of Federal Regulations determines limits on radiatiim exposure. Part of Title 10 of the CFR specifically states that persons having in their possession fissionabic materials must have an operating criticality alarm system. The Victorcen 855 G-M Ar,ca Monitor provides these necessary criticality alarm capabilities. Typically, an arca monitoring system will utilize a detector in any location where personnel might possibly be exposed to an adverse amount of radiation. For reactor facilitics, these areas consist of the following: fuel storage and handling areas, reactor beam ports, hold-up tanks, coolant loops, normal working areas such as labs, hallways and control rooms. Monitoring of pulsed radiation is required around those facilitics, which means that the detector must respond properly to this type of radiation The readout console itself including the alarm set controls should be located in areas under supervisory control. In addition to issuing a warning of high radiation levels, the alarm trips can be used to actuate interlock devices or any of a number of other safety features. i 1-2. SPECIFIbATIONS. 1-2.1 System Specifications. Ringes: Five decade;.01 mR/hr to 1,000 mR/hr or 0.1 mR/hr to 10 R/l.r. Accuracy: Decade, within *20%.oF Ra h). 19u-o a Circuitry: Solid state. Type of Radiation: Gamma. j Energy Dependence: *15% from 100 kcV to 2.5 mcV. Direction Dependence: Less than 30% from any direction. f Type of Detector: G-M tube. 0 0 { Temperature Limits: Detector and Remote Meter / Alarm; -20 F to 140 F. 0 l i Readout Module; 320F to 120 F. ) Ilumidity: Detector; O to 100%. Readout Module; 0 to 95%. l Illgh Alarm: Adjustable trip level, Indicated by meter deflection when function l l switch is turned to Alarm Set position. External Alarm Contacts: Illgh Alarm; 5 amps,120V. i Alarm Level Adjustment: The Ifigh Alarm level trip points are adjusted by means of a 15-turn potentiometer, located on the printed circuit board. Ifigh Alarm Reset: Standard mode or latching type alarm action with manual reset. Iligh Alarm reset is accomplished by depressing the " Red Light" push-button switch. Fail Indicator: Indicates a failure in the system power supply, II. V. supply, low l I l 3 )

T ~~T~ 7 i .f voltage power supply and loss of power. The indication is by means of the green ' Cont light turning off. f( Recorder Output: 0 to 10 mV, *3%. Mou Computer Output: 0 to 50 mV, *3%. Dim Input Power: 115/230 volts, *15%, 50/60 Hz. [, Wei 4 Power Supply: +600 volts regulated; +22 volta unregulated; +10 volta regulated; -6.8 volts regulated. '(5) Bl: Auxiliary Power: 15 to 18. volts, 300 ma, maximum. Detector Connector: AN3102-18-1P. App Readout Module Connector: Rear terminal strips on back of printed circuit boards 3 Com Mot (accessible from the front of the case). Detector Dimensions: 3 inches diameter, 6-2/2 inches high. (7.S3 cm,16.'S cm) $. Din Detector Weight: 1 pound (0.45 Kg). ].1 We Mounting of Detector: Wall bracket. ]g) 3 1-2,2 Accessories. .a g .(1) Remote Alarm (858-1). D Ou Visual Alarm: Red light,1 inch square,1/2 inch thick (2.54 cm,.1.27 cm). .e Audible Alarm: Loud buzzer activates with alarm light. Logic: A red pushbutton light is provided on the front panel, his light will go ON when the radiation exceeds the preset level. Depressing this epring loaded button l will reset the alarm w}ien the radfation has subsided below the preset level. ne alarm will not reset'while the radiation level is above the alarm set point. The alarm level trip point is adjusted by means of a 15-turn potentiometer, located u on the printed circuit board. Temperature Limits: -200F to 1400F. p Ilumidity Limits: 0 to 95% (weather-proof). r Mounting: Iieavy duty industrin1 junction box with flanges for wall mounting. Dimensions: 7' inches high, 7 inches wide,- 4 inch'es deep. (17. 8cm,17. 8cm,10.2cm). Weight: 4-3/4 pounds. (2.15 Kg). (2) Remote Alarm / Meter (858-2, Lo Channel & 858-3, Hi Channel). The specifications are the same as for Model 858-1 except: e w.: . Weight: 5 pounds. (2.27 Kg). Meter: 3-1/2 inches wide with 5 decade display. (3) Single Channel Readout Enclosure (848-3). Application: Single channel area monitoring readout module. l Construction: Wrap-around, welded steel case, with channel guide for module Insertion. 5d Mounting: Rubber pads for bench or shelf mounting. Dimensions: 4-1/2 inches high, 6-1/2 inches wide,12-1/2 inches deep. (11.4 cm,16.S cr y ? 31.6 cm). Weight: 5-3/4 pounds (2.60 Kg). (4) Rack Chassis (848-1), Application: Multi-channel area monitoring readout modules.. j 4

[ yhi bit ] v POWER PRODUCTION Data taken from In N recorder and processed ay ICM 1620. Program by Wm. J. Museler used through June 1964 and thel program by R. Richardson used the rea f ter. Chart hours at Chart Dates Above 0.1 watts Power E Power Dec 1959 to Jan 1960 5.343 kw hrs 5.343 kw hrs Jan 1960 to Nov 1960 4.947 10.290 Nov 1960 to May 1961 4.170 14.460 May 1961 to Feb 1962 4.600 19.060 Feb 1962 to Jun 1963 5.97 25.030 Jun 1963 to Jun 1964 30.0 55.030 4 Jun 1964 to Apr 1965 22.25 77.280 Apr 1965 to Apr 1966 29.11 106.39 Apr 1966 to Mar 1967 25.58 131.97 Mar 1967 to Apr 1968 278.5 410.5 Apr 1968 to Jan 1969 102.39 416 826.5 Jan 1969 to Apr 1969 77.80 458.6 1285.1 Apr 1969 to Feb 1970 72.33 398.2 1683.3 Feb 1970 to May 1970 99.50 574.7 2258.00 May 1970 to Feb 1971 93.5 384.1 2642.1 Feb 1971 to Nov 1971 55 307.4 2949.5 Nov 1971 to Jun 1972 101.6 681 3630.5 Jun 1972 to Feb 1973 48.6 325 3955.5 Feb 1973 to Nov 1973 98.5 559.3 4514.8 Nov 1973 to Oct 1974 106 381.4 4896.3 i Oct 1974 to Apr 1975 93.3 297.7 5194 I Apr 1975 to Jan 1976 34.1 185.5 5380 Jan 1976 to Feb 1977 60.2 257.6 5637.6 i Feb 1977 to Dec 1977 67.2 335.7 5973.3 Dec 1977 to Mar 1979 90.2 279.7 6253.0 t Mar 1979 to Feb 1980 83.8 247.8 6501 Feb 1980 to May 1981 77.8 174.2 6675 Fuel Burnup U235 consumed from Dec 1959 to May 1981 - 0.34 grams

TABLE OF CONTENTS (continued) Page (\\~# SECTION 5 - REACTOR OPERATING CHARACTERISTICS 28 5.1 Introduction 28 5.2 Nuclear Characteristics 28 5.2.1 Critical Mass and Loading 5.2.2 Neutron and Gamma Flux 5.2 3 control System 5.2.4 Eurn-up 5.2.5 Temperature and void coe ficient r 5.2.6 Neutron Lifetime 5.2.7 Alteration of Core Geometry SECTION 6 - REACTOR OPERATING PROCEDURES 31 6.1 Reactor !bnagement 31 6.2 Health Physics and Safety 31 6.2.1 Access Requirements 6.2.2 Health Physics and Safety 6.3 Operating Standards 32 6.3.1 Normal Operation 6.3.2 Core Alterations 6.3.3 Reactor Refuelinc 6.3.4 Approval of Experiments 6.3.5 Operation of Experiments 6.3.6 Limitation of Reactivity Insertion 6.4 Waste Disposal 36 (]) 6.5 Emergency Plan 36 SECTIQi 7 - SAFEGUARD EVALUATION 39 7.1 General 39 7.2 Accidents of Mechanical Type 39 7.2.1 Power Failure 7.2.2 Fuel Element Failure 7.2.3 Binding of Control Blades 7.2.4 Loss of Coolant 7.3 Accidents of Operating Type 41 7.3 1 Startup Accident 7.3.2 Refueling Accident 7.3.3 Mishandling of '.aineralizer Resin 7.4 Accidents of Experimental Type 42 7.4.1 Flooding Beam Port 7.4.2 Maximum Credible Accidents 7.4.3 Dropping Fuel Element on Full Core 7.4.4 Collapse of In-Core Experiment APPENDIX A - TECHNICAL SPECIFICATIONS APPENDIX B - REACTOR DATA - Figures i through 22 APPENDIX C - LIST OF TYPICAL EXPERIMENTS () APPENDIX D - FISCAL STATEMENT APPENDIX E - RADIATION, HEALTH AND SAFEGUARDS COMMITTEE RADIATION REGULATIONS Memorandum 1 - 60 APPENDIX F - REQUALIFICATION PROGRAM 1-27-82

f - 23 Summary of Calcula*ed Reactor Data (Furnished by the General Electric Company Tabulated below are the significant design parameters which are used in this reactor. Reactor Materials: Fuel Uranium aluminum alloy, fully enriched Moderator High purity light water Reflector High p Tity light water and graphite Coolant High purity light water Control Boral and stainless steel Structural material Alumin um Shield Water and aluminum lined concrete Structural Dimensions: Pool 8 x 8 by 15 f t, deep Core (active portion) 15 x 15 by 24 inches high Grid box 9 x 6 array of 3 inch modules Beam port One, 6 inch diameter i Thermal column One, 40 x 40 inches in O cro -eectio# Strategic Materials: Fissionable material 3.4 Kg U-235 Bur n-up Approximately 15 U-235 Fuel life Limited by factors other than burn-up Thermal Characteristics. (Calculated) Operating power 10 kw (maximtm) Tenperature, water 130 deg. F (maximum) Hot-spot factor 2.8 Haximum heat flux 400 Btu /hr.-sq. ft. Specific power (clean, cold) 3.0 watt /gm U-235 Maximum gamma heat in core 10 watt / liter Nuclear Characteristics: ( Calculated) l Average thermal flux 9 x 10 0"" Average fast flux 23 x 10 nv Haximum operating excess reactivity 0.5% a keff. Critical mass 3.3 K8 -4 A keff. per Temperature coefficient -0.5 x 10 () degregC Void coefficient -2.0 x 10-A keff. per l 8.2x10-gid 15 v Prompt neutron lifetime seconds 1-27-82

geometry which has been previously loaded and for which an excess reactivity measurement has previously been made, criti-cality checks shall be made in loading the last 3 elements in lieu of the loading step requirements of Paragraph 6(a). When a change of core configuration involving a single grid position is being made, two safety blades shall be cocked at the half withdrawn position and the third shall be fully inserted during the fuel transfer. For a pre-viously untried configuration the removable plate element shall be used first in the new position with only two plates present. Thereafter not more than I two plates shall be loaded in any set. Core excess reactivity measurements shall be made after each step to insure that the total excess reactivity after each fuel insertion will be below the maximum permissible value of 0.5% A k/k. l 6.3.3 Reactor Refueling The reactor is refueled manually, using either a hook grapple or a fuel element grapple. Each fuel element may be carried to one of the storage racks located in the pool. 3 The water surface is at least 10 feet above the top of any fuel element positioned in either the core grid box or a fuel rack. During transfer the water coverage varies as the element is manipulated, but af ter the element clears the grid box it is usually under about 8 feet of water. Radiation levels at the pool surface during the entire transfer are negligible. The geometry of the storage racks is such that poisoning is not required. The I fuel elements can be loaded with the locating plate in place, and the con:rol drives connected. This is desirable from the safety standpoint, since it pre-vent s the control blades being raised by hand. 6.3.4 Approval of Experiments Experimental procedures are reviewed periodically by the Radiation, i l Health and Safeguards Committee. When high radiation levels or potentially dangerous experimenes are envisioned, the details of the experiment are sub-mitted to the Radiation, Health and Safeguards Committee for approval. If, in the Committee's opinion, the experiment is not safe, final approval of the experiment will be withheld by the Facility Director until adequate safeguard provision has been made by redesign of the experiment of inclusion of required i safety circuits. 6.3.5 Operation of Experiments Experimental facilities will be kept clean to minimize the amount of foreign matter likely to become activated and escape into the building. The interior of the thermal column will be checked periodically for corrosion. i 6.3.6 Limitation of Reactivity Insertion The reactor is designed to be inherently safe by holding the excess reactivity generated in any accident within a safe limit extrapolated " rom the Borax tests. (1, 2) l l l 1-27-82,

7.3 Accidents of operating Type 7.3 1 Startup Accident v Analysis indicates that the fuel elements will not melt if a startup accident should occur. A pessimistic estimate of the energy released ir a startup accident has been made under the assumption the log N-period channel fails to operate, and REVISION NEEDED that scram does not take place until the power reached 1-1/2 times nomal operating level. The maximum withdrawal speed of the safety blades is 7.5 in/ min. The time delay from generation of a scram signal to the instant when the safety blades are free to drop is less than 100 milliseconds. During sc raat, the blades are assumed to accelerate ( for purposes of the accident evaluation) at rates corresponding to the 1-second drop curve in Figure 11. The energy release is about 40 kw-sec, giving a peak temperature rise of less than 2 degrees F. Fuel heat capacity is 35 kw-sec/ degrees F. The calculation assumes the neutron source to be in placegnd uses a conservative ratio of trip power to source power, equal to 10 g the source is absent from the core, this ratio may be of the order of 10 for a clean core, g ("0.3 fission /sec/kgthen ( grresponds to the spontaneous source po we r" fission rate of U and a multiplication factor of 6, due to 17% shutdown reactivity. Startup under this condition is S normally prevented by an interlock which requires the B10 startup counter to read at least 50 counts /sec (Section 4.3.10). If, however, the interlock were inoperative, and the gdes were withdrawn, the final period would be only about 10% shorter than in the case considered above, with unimportant effect upon the fuel temperature rise. 7.3.2 Refueling Accident Erroneous loading of a fuel element should be easy to avoid because of the small core size, good visibility of the core, and easily recognizable reference points. All fuel loading will be under the supervision of a senior licensed operator and will be in accord with the provisions of Section 6. It is not normally possible to go critical with the safety blades in the core. In the event a loading error had taken place, the reactor could conceivably go critical with one safety blade partially in the core. Since the position of the blades is indicated on the centrol console, the operator would know that an error had been made and could shut down the reactor before going to power. 7.3.3 Mishandling the Demineralizer Resin Following an accident such as cladding failure of a fuel element, the resin would become radioactive and the dose rate at the demineralizer may tem po ra rily rise as high as 10 rem /hr. To avoid such peaks of g) radioactivity, replacement of the resin, would be scheduled when the dose 41-1-27-82

from a horizontal 14-inch pipe into water at temperatures in the range of 100 to 150 degrees F. to determine the depth of water required to condense all the steam. This depth was found to be never more than 6 feet. It is con-cluded that 10 feet of water above the core is more than sufficient to insure that no fission products are carried to the water surface by way of steam bubbles generated in the nuclear excursion. It is possible, though very unlikely, that some of the molten alumunum might react with water during the excursion. The heat evolved in the reaction of 30 pounds of aluminum would raise the pool temperature by about 1 degree F. The fuel contains about 170 pounds of aluminum. Actually, no more than a small fraction of this amount could get into the molden condition necessary for rapid reaction, and it is clear that the effect upon the course of the accident is negligible. The postulated accident creates no mechanism, other than a violent reaction, by which solid particles might be propelled through the water and into the atmoaphere. The aluminum-water reaction, if it takes place at all, would be too weak to accomplish such expulsion. The release of radioactive noble gases from the molten portion of the core provides the only means for contaminating the air above and around the reactor. It is possible that traces of solid fissioa products and halogens will be carried along with the bubbles of noble gas.snd dispersed in the air above the pool. The following calculation gives an order-of-magnitude estimate of this release, assuming that the " traces" constitute 1% of the solids and 1% of the halogens in the molten fuel. 1. Total energy release rate by fission products in core after one year of operation at 10 kw: 1.05 E 15 MeV/see from ys 1.03 E 15 MeV/see from 8s 2. Distribution of fission products released to atmosphere: Noble Cases Halogens Solids Total Portion of total 1?% 10% 80% 100% l F.P. in core ~ l Portion of total F.P. in core contained in .015(10)=.15%.015(10)=.15%.015(80)=1.2% 1.5% melted 1/4 fuel element Portion of total F.P. in core released from 1(.15)=.15%. 01 (.15 ) =. 0015%, (. 01) (1. 2 ) =. 012 % 0.16% l melted 1/4 fuel element l MeV/sec from ys released 1.79 E 12 2.86 E 10 1.11 E 11 1.93 E 12 I MeV/sec from 8s released 9.91 E 11 9.30 E 9 1.36 E 11 1.14 E 12 With a comoartment volume of over 30,000 cubic feet, the concentration of fission products inside would create an exposure rate of the order of 40 rem /hr, giving personnel adequate time to take measures for the protection of the 43 1.7_n?

~ _ _ _ facility and for their own safety. The dose rate outside the building does i not constitute a significant hazard to the public. These figures are very l conservative in that in actuality the reactor is at power less than 10% of the total hours in a year. I Loss of Coolant with Damaged Fuel i Suppose that a leak develops and the pool drains at a moderate rate, sim" to the accident reviewed in Section 7.2.4, but made more serious by the fact that the cladding on one or more fuel elements is already damaged. As before, assume that the reactor had been running at full power up.to the beginning of the accident, and that no emergency measui s are taken to prevent pool drainage or to cool the core. 4 To simplify a probably much more complex situation, suppose that a single plate of one element is completely stripped of cladding, and that the fission products released are those which vaporize below 1000 degrees F., i.e., xenon, l krypton, bromine, and iodine. Assume that the gaseous fission products are uniformly distributed throughout the core fuel and that all the gas atoms within recoil rangs (5 microns) of the surface are emitted. (1) The release of fission. products is calculated as follows: Volume fraction released to building: 1 1 element x 1 plate x l 25 elements 10 plates 2 x 5 microns x 1 mil ~ -0.4 x 10 = 39 mils 25.4 microns Total energy release rate by fission products released to building: 1.30 E 10 MeV/sec from ys a 5.39 E 9 MeV/see from 8s 4 With a building volume of over 30,000 cubic feet, the concentration of fission products inside would create an exposure rate of the order of 0.2 rem / hour, with no significant hazard to the public. Spill of a Radioisotope If a vial ot an isotope solution being irradiated should disperse the i isotope in the pool water, it would be removed by the pool cleanup demineral-izer. The worat condition of this type is considered to be the spilling of a l vial containing AuC13 in solution after it has been irradiated for five days in a flux of 1011 nv. It is calculated that a 10 gram sample dispersed in the i pool water would produce 140 mrem /hr at the pool surface and area monitoring equipment will sound'an alarm. The room would be evacuated until such time as the activity had decayed to a safe level and/or'the contaminant had been removed by the pool cleanup domineralizer. l In the event of the spill of a volatile highly radioactive isotope on the ~ floor, area monitoring equipment will sound an alarm. The room will be evacuated until such time as the activity has decayed to a safe level and/or j the contaminant gases have been removed by the ventilating system. The worst condition of this type is considered to be the spilling of ten grams of mathyl i 44 i 3-3-82

_-= -. l 4 \\ iodide, CH 1, after it had been irradiated for 4 hours with the reactor at 3 full power.. Assume that the I-128 formed is similar to I-131 except that the Y-128 has a half-life of only 25 min. As a first approximation, to determine j the limiting case, assume that the air is still, that the entire solution evaporates in approximately 30 seconds into an 8 cubic meter volume and that decay is a negligible. factor. These are all conservative assumptions. Cal-culations indicate that the maximum permissible ingestion dose could be inhaled in the first 0.2 seconds. In a short time the operation of.the exhaust system would increase this time figure appreciably. If the vapor were uniformly distributed throughout the reactor compartment, for example, the dilution would be increased by a factor of 100. With an exhaust duct velocity of over 800 feet per minute and 3,000 cu. ft, per min. exhaust air, there will be four room changes per hour and the exhaust-gas dilution is such that the radiation hazard at the duct exit is negligible. 7.4.3 Dropping Fuel Element on Full Core 4 This accident would increase reactivity by less than 0.5%. ~ 7.4.4 Collapse of In-Core Experiment i Each in-core experiment must be evaluated to assure that no more than 0.6% reactivity is added by the worst mode of failure. It is possible, though unlikely, that the reactivity due to one of the i above accidents could be inserted in less time than it takes the reactor to scram. Assuming that 0.6% reactivity were added instantaneously when the i reactor is operating at normal power, the heat released in the first second [ (no scram) would be about 5 kw-sec. At the same time, the fuel temperature would rise less than 1 degree F at the core center. Actually, scram of even a single one of the three safety blades will shut the reactor down in a traction of a second. The flux scram level of 1.5 times normal corresponds to the prompt rise after the insertion of only O.25% excess reactivity. The accident is thus judged to create no hazard to the facility, much less to personnel. 1.

Barnes, R.S.,

et al, " Swelling and Inert Gas Diffusion in Irradiated Uranium," A/ CONF.15/P/81,1958 (Pg. 54) 4 l i t f l 3-3-82 45 j l

I 2.13.5 All samples or experiments shall be doubly encapsulated () and ensured leak tight if release of the contained materials could cause corrosive attack to the facility or excessive contamination of the pool water. 2.13.6 No experiment shall be installed in such a manner that (1) it could significantly shadow the nuclear instrumenta-tion system monitors, (2) failure of the experiment could interfere with the insertion of a safety rod, (3) failure of the experiment could damage the reactor, or (h) failure of the experiment could release excessive airborne con-tamination. 2.13.7 No explosive or other materials which could combine violently shall be irradiated in the reactor or in ex-ternal experiment facilities, in quantities greater than the equivalent of 25 milligrams of TNT. In addition, the stress that would be produced in the experiment container upon detonation of the explosive shall be calculated and/or experimentally determined to be less than the yield stress of the container. 2.13.8 If a container fails and releases a terial which could damage the reactor fuel or structur by corrosion or other means, physical inspection shall be performed to determine the consequences and need for corrective action. D 5-3.0 surveillance Requirements 3.1 Daily Prior to each day's critical operation (with exception of those experiments which require the reactor to be operated con-tinuously for more than one full day), the two safety channels, the log-N channel and the console annunciator system shall be checked and ensured to be operational. 3.2 Quarterly The area radiation monitoring systems and the pool vater level switch shall be checked and ensured to be operational quarterly. i 3.3 Semi-Annually At least semi-annually, a reactor inspection shall be performed consisting of the following: a. The excess reactivity of the core above cold clean critical shall be measured, b. The console instrumentation shall be calibrated by (} a foil activation measurement of reactor power where A6 1-27-62

cppliccbla er otharwisa celibrctsd cnd chacksd for proper conditions. ( Pool water pH shall be measured and conductivity and c. pH devices shall be calibrated. 3.4 Annually At least once every 15 months, all fuel elements shall be removed from the core and placed in the storage racks. While the fuel elements are thus stored the safety blades will be brought to the surface and visually inspected and the rod drives lubricated. Rod drop times and magnet release times shall be measured for each safety blade and a plot of rod drop times versus distance shall be obtained for each safety blade and compared with data of previous years. Abnormal deviation from previous data will be investigated and reviewed by the Radiation, Health and Safeguards Committee. 3.5 Action to be Taken If maintenance or recalibration is required for any of the items, it shall be performed and the instrument shall be rechecked before reactor startup proceeds. 3.6 Radiation Detection 3.6.1 Area Monitors n k_) Area radiation sensors capable of detecting gamma radia-tion in the range of 0.1 to 100 mR per hour shall be installed near the beam port, demineralizer, thermal column door, fuel storage area and less than one metre above the core pool surface. Upon indication of radia-tion levels in excess of 50 mR/hr (20 mR/hr for fuel storage) these monitors shall actuate audible alarms in the reactor room and in the second and third floor areas above the reactor pool. Portable area monitors capable of detecting gamma radia-tion in the range of 0.10 to 50 mR per hour may temporarily replace fixed area monitors described above provided that the required alarms are operational. Area monitors and the building alarm system may be disabled for maintenance and testing if the reactor is in the shut down condition and a senior operator is continuously present. l 3.6.2 Portable Monitors During reactor operation operable portable survey instru-ments shall be readily available to the reactor operator for measuring beta-gamma exposure rates in the range 0.01 l (^) mr per hour to 50 r per hour, and fast plus thermal neutron (_/ dose rates from 0.04 to 1000 millirem per hour. One or more portable survey instruments for measuring beta-gamma exposure rates in the range 2 mr/hr to 50 r/hr will be kept A7 ! l-27-82

5.5 Operating Records (,,) In addition to records required elsewhere in the license application, the following records shall be kept: (1) Reactor operating records, including power levels and periods of operation at each power level; (2) Records showing maximum radioactivity released or dis-charged into the air or water beyond the effective control of the licensee as measured at or prior to the point of such release or discharge; (3) Records of emergency shutdowns and inadvertent scrams, including reasons for emergency shutdowns; (4) Records of maintenance operations involving substitu-tion or replacement of reactor equipment or components; (5) Records of experimer ~.s installed including description, reactivity worths, locations, exposure time, total irradiation and any unusual events involved in their performance and in their handling; (6) Records of tests and measurements performed pursuant to the Technical Specifications, and h' (7) Records of in-pile irradiations. 5.6 Reports In addition to reports otherwise required under this license and applicable regulations: (1) The licensee shall inform the Commission of any inci-dent or condition relating to the operation of the reactor which prevented or could have prevented a nuclear system from performing its safety function as described in the Technical Specifications. For each such occur-rence, Worcester Polytechnic Institute shall promptly notify by telephone or telegraph, the Director of the appropriate Nuclear Regulatory Commission Regional Com-pliance Office listed in Appendix D of 10 CFR 20 and shall submit within ten (10) days a report in writing to the Director. Division of Reactor Licensing (hereinaf ter, Director, DRL) with a copy to the Regional Compliance Office. (2) The licensee shall report to the Director, DRL, in writing within thirty (30) days of its observed occurrence any substantial variance disclosed by operation of the Reactor from performance specifications contained in the ()s ( Safety Analysis Report or the Technical Specifications. i i A12 [1-27-82

(3) The licensee shall report to the Director, DRL, in writing within thirty (30) days of its occurrence any significant gS changes in transient or accident analysis as described in (_j the Safety Analysis Report. 5.T Annual Operating Reports A report covering the previous year shall be submitted to the Director of the appropriate Regional Inspection and Enforcement Office by March 31 of each year. It shall include the following: (1) Operations Su= mary A su==ary of operating experience having safety signi-ficance occurring during the reporting period including: (a) Changes in facility design. (b) Performance characteristics (e.g., equipment and fuel performance). (c) Changes in operating procedures which relate to the safety of facility operations. (d) Any abnormal results of surveillance tests and inspec-tions required by these technical specifications. (e) f-)s A brief su==ary of those changes, tests, and experi-(_ ments which required authorization from the Commission pursuant to 10 CFR ' 50.59(a). (f) Changes in the plant operating staff serving in the following positions: 1. Reactor Facility Director 2. Health Physicist. 3. Radiation, Health and Safety Comm1ttee members. (2) Power Generation The most current su= mary of thennal output of the facility available together with a summary of the total thermal t power generated over the life of the reactor. (3) Shutdowns A listing of unscheduled shutdowns which have occurred during the reporting period, tabulated according to cause, and a brief discussion of the actions taken to prevent recurrence. A13 1-27-82

(4) Maintenance A km) A discussion of corrective mairtSnance (excluding pre-ventative maintenance) performed during the reporting period on safety related systems and components. (5) Changes, Tests and Experiments A brief description and a summary of the safety evalua-tion for those changes, tests, and experiments which were carried out without prior Commission approval, pursuant to the requirements of 10 CFR 50.59(a). (6) Radioactive Effluent Releases A statement of the maximum quantities of radioactive effluents released from the plant. 5.8 Fuel Storage Two fuel storage racks are located on opposite sides of the reactor pool. Each rack shall be designed to contain not more than 18 fuel elements. When the reactor contain0 a c.itical mass, all additional fuel elements not in the core shall be locked in place except as authorized by the licensed senior operator in charge. O (_,/ A fuel element shall not be stored outside of the reactor pool unless it produces radiation dose levels of less than 100 mr/hr at the storage container surface. Storage containers of fuel elements shall be locked closed when unattended. All fuel element transfers to or from the reactor core shall be conducted by a staff of not less than three persons which shall include a licensed senior operator in charge and a 11 censed operator. Staff members will continuously monitor the opera-tions using appropriate radiation monitoring and core nuclear instrumentation. 5.9 Initial Startup of Altered Core Configuration 5.9.1 During a critical experiment of a new (not previously used) core configuration, subcritical multiplication plots shall be obtained from at least two 1 astrumentation channels. 5.9.2 When a change of core configuration involving a single grid position is being made, two safety blades shall be cocked at the half withdrawn position and the third I shall be fully inserted during the fuel transfer. For a previously untried configuration the removable plate element shall be used first in the new position with only two plates present. Thereafter not more than two plates A14 ll-27-82

i The migration area in graphite can be computed from the p diffusion length and age of neutrons. d Experiment 13. NEUTRON AGE IN GRAPHITE OBJECT: To measure Fermi Age in the Thermal Coluin Graphite. The neutron flux distribution in the half of the graphite thez=al column close to the reactor core face in measured by the use'of cadmium covered indium foils. By the application of age theory, considering the core as a finite plane cource of neutrons, and correctire for self absorption in indium, the neutron distribution in the z direction (perpendicular to reactor core face) yields the neutron age to the indium resonance energy in graphite. Experiment lb. THERMAL NEUTRON CROSS-SECTIONS AND BEAM TEMPERATURE OBJECT: To study neutron bea=s and ascertain the cross-section of materials inserted in the beam. The beam port is used to provide a neutron beam. A BF3 counter is exposed to the beam and the beam is attenuated by materials placed in it. Using appropriate measurement techniques and cadmium filters the beam tecperature or the material cross-section may be determined. Experiment 15. NEUTRON RADIOGRAPHY OBJECT: To introduce the student to neutron radiography techniques. A neutron radiograph consistin6 of a cone shaped collimator and target p holder.is placed in the beam port and small samples placed on dysprosium, indium or gold foils are radiographed using the transfer technique. v f l t I O C3 1-27-82

k/ i i 12 10 O 7 -- 6.- S 4 hh',- 3 ??f - Fast Flux 'E. t 3.', h , fl. - Therml Flux gI ' 10 h., N kh 6, - N / \\ 3 7 s N \\ \\ } ___\\. \\ s m zy ~ E 3 - \\\\ 8 b 8 r.: 10 10 9 \\ 1( 0 7 \\ s - 6 4 3 2 y'.. \\\\ l 7 = Core c Water 9 0 i O E 3 W I Distance from core center (es) FIGURE 6 6 l Average Neutron Flux at 10 kw ~ l l O l 1-27-82

- :.- d e i O O O e TEMPERATURE COEFFICIENT ( ak X 10 / *C -6 k i e m N I I i m m u u u O u o u o o O _u o u o u O l s i l I j O a ut i ln -4> O -4 '2 m g i j* i 1 m= y 2 O ) -1 d' r Z ' i r,e 8 l Cc

D g m

l N l l O i M - m O oM m j 3 - .F m O T l p 7i -m n 2 O l lO m d 2 O i m j 2 n I s Or h 4 m a I i j i ta l l l i. o 4

==-e.. e, -o .c a

Fig ure /0 VOID COEFFICIENT g s ~O 45 l0 -- - - = S og Al AL l y f CORE H

• -CORE---+

32 o 2 o -N ~ n ' O_. x -5 g x <3 -80 9 m OE -15 a u. W O O - 20 a O - 25 ~ - - - " l -35 -40 0 4 8 12 16 20 24 28 DISTANCE FROM CENTER OF CORE (CM) 1-27-82

WPI OPEN P00L REACTOR O s0uTs FuEt RACx 1 2 3 4 5 6 7 8 9 A B C D E F X X X X X 2 X X X S X X 3 X F7 F1 F2 F3 n g w .a Thennal 2 Column X F9 s F4 F5 F6 5 F10 Side O BEAM PORT 5 X Fil Z F12 F13 F14 F15 X F19 F16 F17 F18 F20 6 0 7 X F8 F21 F22 F23 F24 l e X X X X X E! X l l X X X X 2 X X X (denotes removable plug) TYPICAL CORE ARRANGEMENT NORTH FUEL RACKS O 1 l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 FIGURE 13 l 1/27/82

V ffv RHSC April, 1980 REPORT FORMS DUE JULY 1. Area Monitors & Pool Level Monitor 5. Survey Instrument Calibration 6. Reactor Facility Inspection 7. Campus Wipe Test 8. Sealed Source Wipe Test 9. Activities and Releases 12. Evacuation Drills 13. Fuel Inventory 14. RS0 Report 16. Security Measures N '- OCTOBER 1. Area Monitors & Pool Level Monitor 2. Excess Reactivity Measurements 3. Power Level Calibration 4. Pool pH Readings and Conductivity Instrument Calibration 5. Survey Instrument Calibration 6. Reactor Facility Inspection 10. Annual Inspection Report 14. RSO Report (. 15. Materials Status Report 17. Reactor Facility 8, y, n Survey t JANUARY 1. Area Monitors & Pool Level Monitor 5. Survey Instrument Calibration 6. Reactor Facility. Inspection 7. Campus Wipe Test 8. Sealed Source Wipe test 9. Activities and Releases 11. NRC Annual Report 12. Evacuation Drills 14. RS0 Report 16. Security Measures APRIL 1. Area Monitors & Pool Level Monitor 2. Excess Reactivity Measurements 3. Power Level Calibration 4 Pool pH Readings and Conductivity Instrument Calibration 5. Survey Instrument Calibration 6. Reactor Facility Inspection 14. RS0 Report 15. Material Status Report 17. Reactor Facility 8, y, n Survey -l E9 3-2-82

r j,o [h'RHSC -April 1980 [ Reports and Meetings Relative to the Nuclear Reactor Facility I. REVIEWS AND INSPECTIONS LISTED BY REPORT FORM NUMBER 1. Area monitors shall be periodically checked with Cs or Co sources and the pool level monitor shall be tripped. The method used, date, and results shall be reported in writing to the RHSC at each quarterly meeting (July, Oct., Jan., April). 2. Excess reactivity measurement shall be made semi-annually and the results shall be reported in writing to the RHSC at the October and April meetings. 3. Power Level Calibration data shall be obtained semi-annually and the results reported to the RHSC in writing at the October and April meeting. 4. Pool pH readings and calibration values for the pH meter and resistivity probe shall be obtained semi-annually and reported to the RHSC in writing at the October and April meetings. 5. Beta Gansna survey instrument calibration dat~ shall be re-viewed by the RHSC at each meeting to insure ;ampliance with 10 CFR 34.24. 6. The RHSC shall tour and inspect the reactor facility at each quarterly meeting. The log book, film badge records, and facil-ity maintenance records shall be reviewed at these times. 7. Wipe tests on selected areas of the reactor facility and at selected locations throughout the campus shall be made and the data submitted to the RHSC at the July and January meetings. I 8. Wipe tests on all licensed sealed sources shall be made and the data submitted to the RHSC at the July and January meetings. 9. A written report to the RHSC sgl1 be made at the July and January meetings concerning Ar release data, pool water activity, air dust sample activity, effluent releases, and solid waste disposal or storage. 10. Rod drop and release time data and annual inspection results shall be submitted to the RHSC at the October meeting. 11. The Reactor Facility annual report to the NRC shall be reviewed by the RHSC at the January meeting. 12. The RHSC shall conduct and review at least two evacuation drills per calendar year and report at the January and July meetings. E10 3-2-82

h" .RHSC April 1980 [ 13. The RHSC shall annually review in July the fuel inventory report of the reactor facility. 14. The RHSC shall review a report to be suumitted by the RS0 at each quarterly meeting. 15. The RHSC shall review at the October and April meetings the semi-annual materials status reports submitted by the facility director to the NRC. 16. The RHSC shall discuss emergency and security measures at least annually. 17. Reactor Facility 8, y, n Survey shall be made semi-annually at a power level greater than 5 kw. II. RESPONSIBILITY 1. It shall be the resonsibility of the Reactor Facility Director to assure that all inspection reports to the RHSC listed in Part I are properly forwarded to the Secretary of the Committee and to the. Radiological Safety Officer. 2. It shall be the responsibility of the Radiological Safety I Officer to continuously maintain an adequate file of the field data supporting the reports submitted to the RHSC. 3. It shall be the responsibility of the RHSC Secretary to mail to each member of the comittee at least one week prior to the quar-terly meeting a meeting notice listing the agenda topics and item-izing the reports to be submitted and reviewed at the meeting. 4. It shall be the responsibility of the RHSC Secretary to mail to each member of the comittee a copy of the comittee meeting minutes within two weeks of the meeting date. l 5. It shall be the joint responsibility of the RHSC Secretary, the l Radiological Safety Officer, and the Reactor Facility Director to each maintain a file of the RHSC minutes. I ( f Ell 3-2-82

N# RHSC April 1980 DATE SUBMITTED ( OCTOBER APRIL-REPORT FORM 17 REACTOR FACILI_TY 8, y, n SURVEY t DATE PERFORMED: POWER LEVEL (GREATER THAN 5 kw) COMMENTS \\ d SIGNED: FACILITY DIRECTOR I' I t i E29 I t 3-2-82 _ _ _.}}