ML19248D376

Appl for Amend 1 to License R-127 Modifying Procedures & Tech Specs to Permit Operating Levels Up to 20 Watts & Authorizing Addl 700 G U-235

ML19248D376
Person / Time
Site:	05000538
Issue date:	07/31/1979
From:	Bradley Jones MEMPHIS STATE UNIV., MEMPHIS, TN
To:
Shared Package
ML19248D375	List: ML19248D374 ML19248D376
References
NUDOCS 7908150693
Download: ML19248D376 (55)

Text

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APPLICATION FOR P1ENDMENT NO. 1

.T,0 MEMPHIS STATE UNIVERSITY FACILITY OPERATING LICEi!SE NO. R-127, DOCKET 50-538 Submitted by

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Billy M,'Jonis President Memphis State University STATE OF TENNESSEE COUNTY OF SHELBY Billy M. Jones being duly sworn, states that he is President of Memphis State University; that he executed this document for the purpose set forth; that the statements made herein are true to the best of his F.nowledge, information and belief; and he is authorized to execute this document on behalf of said University.

Sworn and attested this day before me July 7_//, 1979.

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,s CONTENTS Pa".LR 1

1, GENERAL INFOR.",ATION..

A.

Introduction 1

1 B.

The Applicant II.

PROPOSED FACILITY MODIFICATIONS.

2 2

A.

Introduction B.

AGN-201, Serial 103 Reactor Modifications.

4 1.

Core Tank Assen bly 4

2.

Shield Water Tank Assembly 4

6 3.

Instrumentation and Safety Systems 9

4.

Gas Handling System.

5.

Safety and Control Rod Drive Assenblies.

11 C.

Reactor Building and Shielding 11 D.

Reactor Characteristics 16 17 E.

Adminictrative Controls.

17 F.

Training and Requalification III.

PROPOSED AMEND"ENT TO OPERATING LICENSE R-127 19 A.

Reactor Fuel.

19 B.

Maximum Power Level 19 19 C.

Technical Specifications 1.

Safety Linits and Limiting Safety Systems 19 Settings.

2.

Liriting Conditions for Operation:

Control and Safety Systems 20 3.

Limiting Conditicns for Operatien:

Shielding 24 25 4.

Design Features.

IV.

SAFETY EVALUATION 27 A.

AGN-201, Serial 103 Reactor.

27 27 1.

Core Tank isssembly 2.

Shield Water lank Assembly 27 3.

Instrumentatio7 and Safety Systems 27 29 4

Gas Handling System 5.

Safety and Control Rod Drive Assemblies 30 tOf i n1 i ui ii

F CONTENTS C01TIhijEn

_P a c_e_

30 B.

Peactor Building and shielding.

C.

Technical Specifications.........

31 D.

Transportation and Storage of fuel.

37 E.

Environmental Considerations........

3B F.

Emergency and Security Planning 39 V.

REFERENCES.

40 VI.

APPENDICES...

Al Al A.

Dose Rate Calculations B.

Transinnt f.nalysis......

B1 111...

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V c_

ILLUSTRATIONS

?n392 II-1.

AGti-201 Reactor Facility (existing)..

3 11.2.

AGN-201 Core Tank Assembly Podifications 5

7 II-3.

Instrumentation Ranges 10 II-4.

Gas Handling Systen.

II-5.

Safety and Control Rod Drive Assembly feodifications 12 II-6.

Proposed AGN-201 Shield.

13 15 II-7.

Proposed AGN-201 Reactor facility fi,

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iv

i 1.

GENERAL It.F0P"ATIOT1 A.

Introductia Pursuant to the Code of Federal Requlations, Title 10, Chapter 1, rart 50.90, Memphis State University (MSU) hereby applies for amendment to its Class 104 (c)

Facility Operating License No. P,-127, Docket 50-533, to:

1.

Operate the Model AGN-201, serial 108 Nuclear Peactor at continuous power levels up to and including 20 Watts (thern;al) and intermitten'.ly at power levels up to and including 1000 Watts (thermal) and, 2.

transport up to 700 grams of contained U-235 between Oak Ridge National Laboratories, Oak Ridge, TN and the Memphis State University, South Campus, Memphis, TN as supplemental fuel loading far the reactor and, 3.

receive and possess up to 1400 grams of contained U-235 in connection with operation of the facility.

Ap;)roval of reactor operation at the requested power levels is essential to the perfornance of scheduled research which requires radiation dases in excess of those obtainable un-der the current operating license and, in general, will significantly increase the overall educational and research capabilities of the A3N-201 Reactor.

Approval of transport, receipt, and possession of an additional 700 grams of con-tained U-235 is necessary to ensure the continued availa-bility of replacement fuel which was removed from an AGN-201 reactor that has been scrapped, and which is in tem-parary storage at Oak Ridge National Laboratories.

If approved, the existing facility will be modified as de-scribed in part II of this application.

B.

The Applicant.

General information required by 10CFR50.33 concerning the applicant, except as shown below, is contained in the Application for Construction Permit and License to Operate the Mndel AGN-201, Serial 108 Nuclear Reactor at Mcmphis State University dated April 11, 1975, as amended (see Docket 50-533).

1.

Earliest date for completion of alteration:

9/1/79 2.

Latest date for conpletion of alteration:

11/30/79 1

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II.

PROPOSED FACILITY MODIFICATIONS A.

I n t ro d uc ti on.

The core of the AGN-201, serial 103 reactor fs composed of a homogeneous mixture of approximately 20L en-riched UO2 in polyethylene surrounded by a graphi te reflector and lead and water shielding.

The assenbly is contained in a reactor tank, 6b feet diareter by 95 feet high, and is lo-cated in the Reactor Room of Building 113, MSU South Canpus.

A control console is connected to the reactor assembly by in-stru; entation and control cabling and is located approximately 14 feet fron the reactor in a separate Reactor Control Room The existing facility arrangement is shown in Figure 11-1.

A core conplete description of the Reactor Assed'y and control console is given in references a and b.

The MSU Reactor has been operated for approxima tely eir;hteen years ( Argonne National Laboratory, 1957 - 1972 ; MSU, 1976 to present) with no apparent indication of fuel deterioration, dnd has not been operated at power levels greater ti an 0.1 Watt.

By the addition cf concrete shielding and Dj extending the instrunentation, other AGN-201 reactors have been operated at continuous power levels up to 5 Watts with no signi ficant increases in operational hazards over the past twenty years.

Moreover, by modi fications similar to those proposed in this application, the AGN-201, serial 100 Reactor was successfully operated at continuous power levels up to 20 Watts, and in-termittently at power levels up to 1000 Watts by the U.S.

Naval Post gradua te Scncol (USTGS) at Manterey, Cali fornia for a period of approxir ately eight years (references c, d, e).

This reactor was subsequently transferred to the Cali fornia Polytechnic Insti tute where it has been satisfactorily operated a t po',,er l evel s up tu 0.1 '.'a tt s ince 197 3.

Experience with the serial 100, AGN-201 Reactor a t USNPLS has shown that a maxirum apparent core temperature of 34.20C was achieved for a single operation concencing at roon temperature and reraining at 1000 Watts for nine minutes.

Core terperatures versus tire at various power levels are suc rarized in reference e.

Operations at power levels greater than 20 Watts also re-sulted in gas di f fusion from the AGN-201 fuel for periods as long as two-to-three days following such opera tions.

The evolved gas was predominantly hydrogen, containing detectable concentrations of fission product gases, and was estimated to resul t in a reactivity loss of (.0135 *.001) per KW-hr of operation.

The increased neutron flux in void areas such as the AGN-201 Glory Hole (when empty) and the reactor skirt area produced rneasureable concentrations of Argon-41 (References d,e).

Proposed modifications described in subsequent sections of this application consider the increased operational concerns de-scribed above, as well as necessary additional shielding and radiological controls, and an extension of existing ins trumenta-tion necessary to sa fely operate the AGN-201, serial 103, reactor at the power levels requested.

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.'c FIGURE II-1. /G! 201 REACTOR FACILITY (existing) -.3 e...M Y,e'.,

4 B. AGI'-201, Serial 103 Roactor M difications 1. Core Tank Assenbly (see Figure 11-2). The lower core-half support asserhiy (aluminun rod) will be replaced by a longer support assembly which will extend through the louer core tank cover and which ulil accon todate installation of a ter pe ra ture sensor. The penetration through the core tank cover will be sealed to maintain gas tight integrity of the core tank. The new support assembly will be suf fi-ciently long to extend through the bottom of the AGN- '01 -tor Tank and into the rod drive cavity. An adapter will be fabricated which penetrates the battom core tank cover, and v.hich will interface with the Gas Handlirg System discussed in part II.D.4 of this application. ihe penetration throu';h the core tank cover will be sealed to r.aintain gas tight in-tegrity of the core tank, The adapter will be suffi-ciently long to extend through the botton of the AGN-201 Peactor Tank and into the rod drive cavity. The core tank assembly will be pressure tested to at least 6 plig to verify c:as-tight integrity f ol l m. i ng installation of the preposed alterations. 2. Shield Water Tank Assembly. a. The J.eactor Tank and bottom lead shield only have penetrations to accorrodate the four Safety and Control Rod thi:Qles. Two additional holes will be bored thtough the lead shield and Reactor Tank to accept the extended lower core-half support assembly and the Gas Handling System Adapter discussed in Part II.C.l. These additional penetrations will not violate tho secondary gas seal established by original AGM-201 de

sign, b.

The existing Shield Water Tank has only one instru-rent tube (4V I.D. X 17"). The Channel 1 and 2 neutron detectort are located in Access Ports 4 and I c^espec-tively. To accoc: >odate cor pensated ion chacbers for Channels 2 and 3 (2stinated c erall length 24 inches), ne. instrument tubes (aluminum) will be fabricated. The Channel 1 detector (CF3 proportional counter) will ren ain in Access Port No. 4. c. The Red Warning L ght n. unted on top of the Shield Water Tank will be removed to eliminate interference with the top shield proposed in pc t II.C. This light serves no useful purpose under the current facility configuration. 4 '!4 ,O/

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h s 3. Instrumentation and Safety Systems (Figure II-3), a. Nuclear Safety No. 1. The existing system utilizes ~~~~ ~ ~ a Tracer 1ab SC-34X T0 4 50K CFM) Countrate Instrunont with BF proportional counter and preamplifier. The system hrovides indication from below source level up to approximately 55 milliwatts of reactor power and pro-vides a Low Level Scram for countrates below 120 CFM. A redundant High Level Scram function is also provided at 90-95. of full scale meter deflection. The redundant High Level Scram will be eliminated. The existing high voltage switch for the Channel 1 neutron detector will be modified such that it will simultaneously disable the Low Level Scram function by insertion of a dummy electrical signal and remove high voltage from the detector This operator action will be controlled by operating procedures which will permit high voltage removal above 90J full scale meter deflection and require reenergization bekw 40 milli-v:atts. An automatic time delay will be incorporated into the Low Level Scram disabling circuit to ensure on-scale indication prior to removing the dun signal 9 and re-arming the Low Level Scram function. The ti' e delay viill be necessary to prevent a scram fror occur-ring sinultaneously with reenergi:'ing the proportional counter high voltage. b. Nuclear Safety No. 2. The existing system utilizes a Keithley - 412 Log Picoan neter (10 13 7 AMPS) with an uncompensated ion chamber. The system provides Log-N indication from belov. source level up to approxi-mately 1.6 Watts of reactor povier and provides an analog signal for remote Reactor Period measurcrent. Nuclear Safety No. 2 nrovides a reactor Low Level Scram at source levels less than 0.5 X 10-13 A.*PS, a High Level Scram at power greater than 0.2 Watts, and a Peried Scram at periods less than 5 seconds. Tne picoammter will be replaced with a nev, Keithley-25012 Log-N Period Amplifier (10 10-4 AMPS), or equivalent, and the uncompensated ion charber will be replaced with an equivalent co' pensated ion charber This system should provide Log-N and period indication from approximately 45 micro Watts (slightly above source level) to 4.5 Kilowatts of reactor pocier The existing period reasurement circuit will be removed and the period scram circuitry modified to interface wi th the new Log-N period ampli fier The existing Channel 2 Lov. Level Scram will be elininated since on-scale indication may not be achieved until after reac-tor startup has commenced. 6 is. < us.> L. I d/

.3 - - 10 10 ~ - 10' 4 !(T 4 -4 1000 - F- \\ 5 \\ 3 \\ - go \\ - 10 -10~ k ,0.c \\ - 'oo ix 3 s _,y .6 ) - 20 WATTS -10 / ~ j / ._ g o / / - t0-f l / / I Q ~7 -10 -10 - f.O H 1.0 / 8 _3 g 0 . l(7 -IO / -- O.1 O.1 9 250 K - 10 9 10 If / / .b / /i ',250 K / / / 5 ~ -.O -10 / .O[ O / / / m / / / ~:O .10 / o. uj .01 / / / / 10 / ' !O a , / 10 / -/ j [ b / .iO / / / / ! S__ / / / - DOI ~' u 00: / . lO-I I IO !i [g " ~ _11 .000l - - 10 l2 A 10 12 130 CPM SOURCE-SOURCE LEVEL y ~30j;W . O O 001 .0000: .13 -I3 _gO-13A -!O A (A) EXISTING SYSTEM (B) PROPOSED SYSTEM Channel 1: Linear countrate (0-250K CPM) "/BF P.C. Channel 1: Linear countrate (0-250K CPM) "/BF 3 3 ~ ~ -13 -7 -1 Channel 2: Log-N Picoammeter (10 -10 AMPS) "/II.I.C. Lmannel 2: Log-tl Picoammeter (10 AMPS) "/C.I.f -13 -1 -3 Channel 3: Linear Picoammeter (10 AMPS) "/U.I.C. Channel 3: Linear Picoammeter (10 -10 gggg) w/C.I. ] F / / / Present range of critical operation I/ / / / / ';ormal range of critical operation '\\ \\\\\\ Intermittent operation FIGURE II-3. INSTRUMENTATION RANGES e.-

c. Nuclear Safety No. 3. The existing systen utilizes a Keithley-410 Linear Picoacre ter (10-13 _10-3 Arps ) with an uncompensated ion chamber. The system provides linear indication from below source level to approximate-ly 16 kilowatts of reactor power. fluclear Safety flo. 3 provides a Low Level Scram at source levels less than 5: of full scale and a High Level Scram at power great-er than 0.2 watts. The uncompensated ion chamber will be replaced with an equivalent compensated ion chamber A mechanical in-terlock will be incorporated with the existing range selector suitch to prevent exceeding the 10 x 10-7 At@S scale (approxicately 30 watts) except for approved high power operation and instrument calibration. This op-erator action will be controlled by operating procedures which will only permit removal of the interlock for approved operations above 20 watts, or for instrument calibration when the reactor is shutdown. fio changes in the scram circuitry are anticipated for Channel 3. d. I n_t.e r_lnci_lifte.. A High Core Tank Pressure trip function will be added to the existing Interlock Line continui ty ci rcui t. This function will be provided by a pressure-sensitive transducer which will be lo-cated in the Gas Handling Systen discussed in part II.B.4 of this application. The sensor will be in-stalled such that the Interlock Line will be inter-rupted and an automatic reactor scram initiated at core tank pressure greater than 5 psig. e. Te~ne ra t_u r_elo n.i_t o r_i n g. The present system only provides local indication of Shield Water Tank temper-ature and utilizes a well-type bimetallic thermometer An additional temperature monitoring system will be added to monitor in-core terperatures and shield wat: - temperature of the reactor assembly. The system will utilize thermistor or thermocouple sensor elements with a conmon remote readout instrurent located at the con-trol console. The sensor eierent(s) used for the in-core monitor will be selected fr om materials that would result in a negligible reactivity effect on the reactor Core. The core temperature sensor will be installed into the lower core-half support assembly previously discussed in part II.B.1 of this application. Since the sensor will be in contact with the aluminun, tube which sup-ports the core fuse rather than in direct centact with the fuse, it is anticipated that the indicated temper-ature will be slightly less than actual fuse tempera-ture while temperature is rising (Reference e). 8 / 1ii

Interconnecting wiring from the core temperature sensor to the console will be routed through the existing conduit between the Rod Drive Assembly cavity and the reactor skirt at oa. Thus, the original AGN-201 design secondary gas sea t, will not be violated. 4. Gas Handling System (Figure II-4). Due to radiation damage in polyethylene, hydrogen evo-lution with detectable concentrations of fission prod-uct gases are expected to result from reactor operations greater than 20 watts. Thus, a system capable of mon-itoring, sampling, and handling these gasses will be necessary to ensure that compliance with 10CFR20 can be accomplished as well as to prevent formation of haz-ardous concentrations of hydrogen. A means must also be provided to permit air sanples to be drawn f rom within the reactor shielding for the purpose of deter-mining Argon 41 concentrations which may accumulate in the AGN-201 Reactor Skirt area and other void spaces during high power operations. A Gas Handling System will be fabricated to perform the following functions: a. Maintain a dry nitrogen blanket in void spaces of the AGN-201 Core Tank Assembly. b. Monitor core tank pressure. c. Interrupt the Interlock Line Safety System and thereby initiate a reactor scram for core tank pressure greater than 5 psig, d. Prevent core tank overpre Jrization. e. Provide sample capability for reactor skirt area and other void spaces within the reactor shield. f. Provide radiation detection capability for gases within the handling system. g. Pro'.ide gas discharge capability. It is anticipated that NSU will fabricate and use a system similar to that shown in Figure 11-4 and which is discussed in detail in reference f. The system will interface with the core tank bottom cover adapter dis-cussed in part II.B.1 of this application. Gases that may be released from this system will be routed via a permanent extension of the systen shown in Figure II-4 to an existing exhaust fan in the east wall of the Reactor Room. TL s fan is approximately 17 feet above ground level and the discharge flow path is di-rectly from the east wall of the room, approximately 3 feet below the roof and 3 feet from the northeast Corner. ,u, }_ 9

V <\\ e Core Tank ,e W ead Shiele L -1 / /,5 c u -' Soldered Joint @X n^ j - --Pressure Switch Insi 11 ctor Cas S =pling Tube I l Gas Manifold Outside Reactor Shield _._ _ __ __ ___ __ __ __ _ j 9 1 i Mercury Mano eter i I l 1 I l I l v Auxiliary Gas l K$ e@ Sampling Tube / (A X$s '7 Pressure Release i Pressure l l Valve i I Gauge j _ _ _ _ _ _ _ __ _ __ ___ __ __ _ _ J O t_ _ Pressure }' Reducing ~ Valve Le G-M Chamber (L s se i i i - G-M Tube i s Nitrogen i i Tank % p Pump ' - To Building Exhaust Yan on EA>T1.ia.i aaux Ku.u. FIGURE II-4: GAS HA:DLIr;G SYSTETt 10 i /h

The existing facility utilizes a fixed Air Particulate Sampling system which continuously monitors and returns air in the reactor room. This system provides audible and visual alarms. MSU intends to supplement the ca-pabilities of this system by purchasing a portable high volume sampler, and by developing a portable gaseous activity sampling capability. 5. Safety and Control Rod Drive Assemblies (Fiqure II-51. In order to provide additional free void volume for accumulation of gases from the Safety and Control Rod Fuel Capsules, and to provide a method for disposing of these gases,,Todifications to the fuel capsules and rod follower tubes for each of the rods are necessary. M5U proposes to modify the AGN-201, serial 100 reactor Rod Drive Assemblies sinilar to the manner shown in Figure II-5 (a), and to fabricate a Gas-Release Tool similar to that shown in Figure II-5 (b). The modifi-cations should not interfere with nor in any way alter the existing Control and Safety Rod motion and response times. These alterations are discussed in more detail in references e and g. C. Reactor Building and Shielding (Figures II-6 and II-7). The existing concrete block partition shown at the southeast end of the reactor room in Figure II-1 will be removed in order to provide greater freedon of ac-cess to the AGN-201 experimental facilities. In addition to the lead and water shielding provided by the existing AGN-201 design, a cylindrical shield corposed of both borated and ordinary concrete block will be constructed around the reactor assembly (Figure II-6). The shield will consist of two concentric cyl-inders of blocks, one cylinder compressed by borated concrete, and arranged so that seams will overlap to minimize radiation streaming and so that a total wal, thickness of 44 inches is provided. Penetrations through the shield will be made in order to extend the Glory Hole and Access Port liner tubes out to the shirld face. Theso extensions will nornally contain shield filler plugs to minimize streataing. A renovable shield plug at the base of the 5ield will provide access to the reactor. The cylindrical shield will have a top support and shield assembly consisting of aluminum angle and flatbar frame, approximately 18 inches of borated paraffin, and a wooden walking deck This top shield will a lso contain a removable shield plug approximately 5 feet in dianeter to provide access to the reactor. Seams will be arranged to minimize radiation s trearm.ng. -. 1 i't 11

... g- - - New metal cylinder Qug -- - - jL' l a-

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QQ b C L__ --__. _ __w 1_ _r te s _ _. .----~_.}_. Fuel can Steel plug Rod follower Valve assembly Proposed Rod Assembly (a) A \\ Pressure gauge 2> To gas ha-d-ling system l l j )0-ring ,, l p 1 / Fv 4}a. ' ^ -qd i i -~ e' t i.'i tr l l L.. ~.<w _m a jf,', s J~Q .} g I W LaJ oh<dMS-1 c oa rod Plun e,e r Sh;t-off Valve " (b) Rod-Gas-Release Device (Portable) FIGtJRE 11-5. SAFETY A*:D CONTROL RCD ' RIVE ASSEM3LY MODIFICATIONS 12 j \\3

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t --- J SECTION A-A

^ a-FIGURE II-6. PROPOSED AGN-201 SHIELD

Instrumentation and control cables, and gas handling sy; tem piping / tubing, will be routed through an exist-ir.g cable channel which is recessed into the reactor room floor. This channel is normally covered by steel plate and will pass beneath the proposed shield below floor level. A chain link fence topped with strands of barbed wire will be constructed around the existing Reactor Room and Control Room (Figure 11-7). The fence will estab-lish an area of controlled access extending approximate-ly 20 feet from the east and north walls of the facility. The fence will contain a gate leading to the Control Room east entrance, and a gate to acconnodate access to the north loading dock For reactor operations above 20 watts, the fenced area will be posted as a Radiation Area and the gates will be locked. For op-erations at and below 90 watts, it is expected that radiation dose rates external to the north and east walls of the facility will not exceed limits for un-restricted access to the fenced area. Posting of signs and access requirements will be controlled by operating procedures for the facility. The Reactor Room and tae roof of the reactor bui'. ding will be posted Radiation Areas for operations helow 20 watts, and High Radiation Areas for operations above 20 watts. Access shall be controlled by facility pro-cedures, and except for approved radiation surveys, access to the Reactor Room and the roof shall be pro-hibited for reactor operations above 20 watts. Rooms 142, 144, and the Reactor Contral Roem !ill be posted as Radiation Areas for operations above 20 nat's. Access will be controlled in accordance with existing facility procedures. Pooms 248 and 252, which are located on the second floor directly above roons 142 and 144 respectively, will be posted as Radiation Areas for operations above 20 watts unless operational surveys show that these rooms may be classified for unrestricted access. ,< ? 34 >. ') : ,i!

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D. P ea c to r C h a ra c_t e r i c. t_i_c s_. t Maximum continuous power 20 watts (thermal) Peak thermal neutron flux 9 x 108 n/cm -sec. 2 Gar:ma field at vertical shield face 0.8 r;R/hr. fleutron field at vertical shield face <.01 rirem/hr f41ximum i ntermi tten t power 4.5 x 10g (thercal)n/cm2-sec. 1000 wat Peak thermal neutron flux Gamr:a field at vertical shield fa:e 38 mR/hr. fleutt an field at vertical shield face.06 mrem /hr. Total Stored Fission Products 200 Curies (Irrediately after shutdown) Long-lived Fission Products 20 Curies (1 day after shutdown) Total Thickness of Concrete Shield 44 inches Total Shield Thickness above Reactor 18 inches borated paraffin 'laxirumestimatedoperatingtjreat

10. 3 c:inutes 1000 watts (reactor core 20 C at startup) 16 m '/ h

$ l )

E. Administrative Controls. Necessary changes to existing facility operating and emer;ency procedures will be made to ensure that reac-tor operation is in compliance with approved license and technical specification changes. In addition, new procedures will be developed which include the follow-ing: 1. Gas Handling System operation under norml and abnormal conditions (see reference f). 2. High Power Operation (operations greater than 20 watts) to include a Pre-operation Checklist, precau-tions, limitations, and exclusion area controls. 3. Maintenance Operations including defueling and re-fueling procedures to satisfy increased radiation con-trol concerns resulting from operation in excess of 20 watts (see reference f). 4. Calibration and operation procedures for new in-strumentation proposed in part II.B of this application. 5. Test procedures to include post-modification fuel loading, startup, physics tests, and radiation surveys. 6. Fuel Storage and Handling precedures to preclude simultaneous storage of more than 700 gran.s of con-tained U-235 in the er.isting fuel stor age facility or the AG'i-201 Reactor Facili ty. New procedures and necessary changes to existing pro-cedures will be processed in accordance with Section 6.0, Administrative Coltrols, of the existing facility technical specifications prior to implementation. F. Trainina and Requalification. The facility staff currently includes two licensed Senior Operators. One additional person is under-going training and it is anticipated that he will com-plete Senior Operator license exminations during July, 1979. A specific training and requalification program will be scheduled by the Reactor Supervisor to ensure that all staf f Operators and Senior Operators are properly informed and knowledgeable of aoproved facility alter-ations and procedure changes. This program will con-sist of f ormal lectures and oral and written examina-tions. It is anticipated that operational training 17 L ') n ,la

will not be required since facility 3difications should be completed within thc .. constraints nor-mally specified to maintain operational proficiency and, except for intermittent high power operations, reactor operation up to the proposed steady-state limit will not be significantly different from exist-ing procedures. The training program shall be approved by the Reactor Administrator and reviewed by the MSU Reactor Safety Committee. In addition to scheduled training, the facility licensed Operators and Senior Operators will participate in prep-aration of procedures, procedure changes, and in im-plementation of approved alterations. Examinations will be administered and evaluated, and training records will be maintained in accordance with the MSU Operator Requalification Program approved by USNRC Operator Licensing Branch letter dated January 11, 1978 (Docket 50-538). 18 ,,\\ L ~ ') l -

III. PROPOSED AMEi;DMEllT TO OPERATII;b LICEf;SE R-127 A. Reactor cuel (Part 2.B). Arend the existing authorization for reactor fuel to: 1. Transport up to 700 grams of contained U-235 as re-serve and interchangeable fuel replacement loading for the reactor. Such shipment authorized between Oak Ridge flational Laboratory, Oak Ridge, Tennessee and the Memphis State University South Campus, Memphis, Tennessee. 2. Receive and possess up to 1400 grams of contained U-235 in connection with operation of the facility. B. MaximumPowerLevelJPart2.Cl. Amend the existing au-thorization for operation to: 1. Operate the reactor at power levels not to exceed 1000 watts (thermal), and 2. Operate the reactor at power levels of 1000 watts (thermal) or less in such a manner that the inte-grated power level for any seven consecutive days shall not exceed 3.36 kilowatt-hours. C. Technical Snecifications,_ App _endix A of Operatino License R-127. 1. Sa_fet1 L i m i t s a n d L i ni t i_nL a f,e_tr_Sy s t om S e t t i n a s S Replace existing section 2.0 with the fello>ing: 2.0. SAFETY lit:ITS AND LIMITI!;G SAFETY SYSTEM SE,TTI ES_ 2.1. Safety Linits. Applicability. These specifications apoly to the maximum core temperature and minimum snield water temperature and level during steady state or tran-sient operation. Obiective. To assure that the inteority of the fuel raterial is maintained and that essentially all fission fragments are retained in the core ~ ratrix. 2.1.1. Specification. The maximun core temperature shall C not exceed 200 C during either steady state or transient operation. Bases. Thepofyethylenecorenaterialdoesr.ot rolt belon 200 C and therefore assures inteqrity ofthecoreandretentionofessentiallyalf fission fragments at tenperatures below 200 C. 'l (O< ') ^> /_ /7

2.1.2 Sne c i fi ca t. i on. The reactor shield tank water temper-ature shall be maintained above 10 C, and the water level in the tank shall not be more than 12 inche Se-low the top of the reactor shield tank. Bases. Low reactor shield tank temperature may result in freezing of the water. The resu't of expansion due to freezing of the water may damage the shield tank and other reactor components. This condition would degrade core contairmtent and shielding capability. U A safety limit of 10 C prosides a margin for confidence that the reactor will not be operated with frozen shield water. The shield water level of 12 inches below the top of the tank provides an adequate medium for continuous neutron flux monitoring during reactor operation and ensures adequacy of the facility secondary radiation shield for operations greater than 100 milliwatts. 2.2 Limiting Safety Svstem Settinos. /mpl i ca bi l i ty. These specifications apply to the narts of the reactor safety system which '..ill limit maxir.m core temperature. O b i ec t i v e_. To assure that auton tic protective action is initiated to prevent a safety limit from being ex-ceeded during steady state or transient operation. 2.2.1 Specification. The core therrgl fuse shall melt when heated to a temperature of 120'C or less resulting in core separation and a reactivity loss greater than 5 ok/k. Cases. In the event of failure of the reactor to scram, the self-liniting characteristics due to the high neg-ative temperature coefficient, and the rfelting of the thermal fuse at a temperature below 120 C will assure safg shutdown without exceeding a core temperature of 200 C. 2. Limitino Conditions for Operation: Control and Safety Systers a. Revise the existino specification 3.1.a. to read: "The available excess reactivity with all control and safety rods fully inserted and including the potantial reactivity worth of all experiments shall not exceed 0.65 'i/ k. b. Revise the existing specification 3.1.b. to read: "lhe shutdown margin with the most reactive safety or control rod fully inserted shall be at least 2 '1/k, referenced to 200C. 20 t' Q ', 197 d) /' r

c. Replace existing section 3.2 with the following: 3.2 Control and Safety Systems App l i ca bi l i t v._ These specifications apply l.o the rcactor control and safety systems. Objective. To specify the lowest acceptable level of performance, instrument set points, and the minimum numbers of operable components for the reactor control and safety systems. 3.2.1 Specification. The reactor shall not be made critical unless the following specifications are met: a. The total scram withdrawal time of the safety rods and coarse control rod shall be less than 200 milliseconds, b. The maximu;;. reactivity addition rate for each rod shall not excecd 0.04 1/k/sec. c. The safety rods and coarse control rod shall be interlocled such that: 1. Reactor startup cannot convence unless both safety rods and coarse control rod are fully withdrawn from the core. 2. Only one safety rod can be inserted at a time. 3. The coarse control rod cannot be inserted unless both safety rods are fully inserted, d. A loss of electric power shall cause the reac-tor to scram. e. The reactor source level reasured by the ';uclear Safety channel l:o.1 countrate instrument is more than 120 CP.". f. The reactor core tanF is pressurized with dry nitrogen to at least 1 psig. g. All reactor safety syst en instrumentation shall be operable in accordance i.ith Table 3.1 with the following c.llowable exception: 1. !;uclear Safety Channel i;o. I may be bypassed for re ctor aperations conducted above 40 ailli-watts provided I;uclear Safety Channels 2 and 3 're verified to be operat>le. n '/ li l['t 21

TELE 3.1 SAFETY CHANNEL SETPOINT FUNCTION Nuclear Safety =1 12/ full scale Scram at source levels Low Countrate < 12< of full scale I;uclear Safety *2 (Log) $ 2000 watts Scram at power >2000 watts High Power Reactor Period >_ 5 seconds Scram at periods <5 sec. I!uclear Safety e3 (linear) High Power 5 95' full scale Scram at power >95L Low Power ? 5' full scale Scram at source levels < 5" of full scale Shield Water Terrperature

15 C Scram at temperature <l5 C U

Shield Water Level

10.5 inches Scram at levels >10.5 inches below top of shield water tank Seisnic Displacerent

- 1/16 inch 5 cram at displacements >l/16" Core Tank Pressure - 5 psig Scram at core tank Manual Scram pressure >5 psig Scram at operator option Radiation Monitor Alarm at or below level set to meet requirements of 10CFR Part 20 Air Particulate Monitor Alarm at or below level set to rreet require:r.ents of 10CFR Part 20 Bases. The specifications on scram reactivity rate in conjunction with the safety system instru"'enta-tion and set points assure safe reactor shutdown during the most severe foreseeable transients. The limitations on reactivity addition rates allow only relatively slow increases of reactivity so that amole time will be available for manual or auto-catic scram during any operating conditions. Inter-locks on control and safety rods assure an orderly approach to criticality and an adequate shutdown ca pa b i 1 i ty. The minir un, reactor source level assures that an adequate neutron countrate from which to conduct an orderly and controlled startup is registered on the startup channel before a reactor startup begins. n ') q l[b 22

Pressurizing the core tank with dry nitrogen to at least 1 psig assures that evolved hydrogen from high power operation cannot agglomerate into hazardous concentrations. The neutron detector channels (nuclear safety chan-nels 1 through 3) assure that reactor power levels are adequately monitored during reactor startup and operation. Requirements on minimum neutron levels will prevent reactor startup unless the startup channels (nuclear safety 1 and 2) are operable and responding, and will cause a scram in the event of instrumentation failure. In order to provide as-surance that at least two nuclear safety channels are operative for all ranges of reactor operation and to prevent overranging che channel 1 startup instrument, Nuclear W ety thannel No. 1 is allowed to be bypassed for o;m ations above 40 milliwatts only if the remaining two channels are verified to be operable. Since the AGN-201 core negative temperature coeffi-cient of reactivity in conjunction with the raximum notential excess reactivi ty specified in 3.1 prevents reactor operation at high power levels for time in-tervals necessary to approach established safety limits or limiting safety system settings, the high power level scrams are established to provide re-dundant automatic protective action at levels low enough to assure safe shutdown during rapid reactivity transients and to prevent exceeding requirements for design of the facility radiation shield. The period scram conservatively limits tne rate of rise of re-actor power to periods which are manually controllable and will autonatically scram the reactor in the event of large reactivity additions. The AGN-201's negative temperature coefficient of reactivity causes a reactivity increase with decreas-ing core temperature. The shield water temperature safety channel will pgevent reactor operation at temperatures below 15 C thereby limiting potential reactivity additions associated with temperature de-Creases. Water in the shield tank is an important component of the reactor shield aad operation without the wa-ter may produce excessive radiation levels and in-adequate neutron flux nonitoring capabilities. The shield tank water level safety channel will prevent reactor operation without adequate water levels in the shield tank, '10 ..' b 23

The reactor is designed to withstand 0.69 accelera-tions and 6 cm displacements. A seismic instrument causes a reactor scram whenever the instrument re-ceives a horizontal acceleration that causes a hori-zontal displacement of 1/16 inch or greater. The seismic displacement safety channel assures that the reactor will be scrammed and brought to a subcritical configuration during any seisnic disturbance that may cause damage to the reactor or its components. The core tank high pressure scram prevents reactor operation with internal tank pressures above that for which core tank integrity is assured. The manual scram allows the operator to manually shut down the reactor if an unsafe or otherwise abnormal condition occurs that does not otherwise scran the reactor. A loss of electrical power de-energizes the safety and coarse control rod holding nagnets causing a reactor scram thus assuring safe and im-mediate shutdown in case of a power outage. A radiation monitor nust always be available to op-erating personnel to provide an indication of any abnormally high radiation levels and an air particu-late monitor must be available to warn operating personnel of a degradation in core tank or gas r.on-itoring system integrity so that appropriate action can t'e taken to shut the reactor down and assess the hazards to personnel. 3. Linitino Conditions for Operations: Shielding Replace existing section 3.4 with the following: 3.4 Shieldinq. Apolicability. This specification applies to reac-tor shielding required during reactor operation. Obiective. The objective is to protect facility personnel and the public from radiation exposure, 3.4.1 Specification. The followinq shielding requirc-ments shall be fulfilled er to raattor startup and during reactor operati a. The reactor shield tank shall be filled with water to a height within 10 inches of the nighest point on tne manhole opening. b. The thenral column shall be filled wi th water or graphite. Access to the reactor building roof area above the reactor shall be n stricted during reactor operation. 24 .o, j ' I

c. The facility secondary shield shall be in place with removable shield plugs installed. 3.4.2 Speci fi ca tion. Access to the reactor room shall be prohibited, except for radiation surveys, during op-erations above 20 watts. Cases. The facility shielding in conjunction with ~~-'in~ated restricted radiation areas is designed desi to limit radiation d>ses to facility personnei and to the public to a level below 10 CFR 20 limits un-der operating conditions, and to a level below cri-terion 19, Appendix A, 10 CFR 50 recommendations un-der accident conditions. 4. Design Features. Peplace existing section 5.1 with the following: 5.1 Re ac to r, a. The reactor core, including control and safety rods, contains approxi;rately 660 grams of U-235 in the for: of 20 enrichea UO2 dispersed in approximate-ly 11 kilogra".s of polyethylene. The lower section of the core is supported by an aluminum rod and a thermal fuse. Tne fuse nelts at temperatures below 1200C causing the lower core section to fall away from the upper section reducing reactivity by at least 5. ik/k. Sufficient clearance between core and reflectors is provided to ensure free fall of the bottom half of the core during the rost severe transient. b. The core is surrounded by a 20 cm thick high density (1175 gm/ cms) graphite reflector followed by a 10 cn thick lead gan ra shield. The core and part of the graphite reflector are sealed in a fluid-tight aluminum core tank designed to contain any fission gases that right leak fror the core. A val ved gas handling system is permanently connected to tha core tank assembly to permit monitoring and disposal of gases which may accumulate in the tank from high power operations. c. The core, reflector, and lead shieldino are en-closed in ar d supported by a fluid-tight steel reac-tor tank. An upper or " thermal column tank" may serve as a shield tank when filled with water or a thermal column when filled with graphite. 25 -96 I 2Id

d. The 6': foot diameter, fluid-tight shield tank is filled with water constituting a 55 cm thick fast neutron shield. The fast neutron shield is formed by filling the tank with 1000 gallons of wa-ter. A 44 inch thick concrete block shield support-ing a top cover that contains approximately 18 inch thick borated paraffin encloses the 6'; foot diameter shield tank to provide a secondary neutron and gan. 'a shield for operations above 100 milliwatts. The com-plete reactor shield shall limit doses to operating personnel in restricted and unrestricted areas to levels less than permitted in 10 CFR 20 under oper-ating conditions. Two safety rods and one control rod (identical e. in size) contain up to 20 grams of U-235 each in the same form as the core material. These rods are lift-ed into the core by electromagnets, driven by revers-ible DC motors through lead screw assemblies. De-energizing the magnets causes a spring-driven, grav-i ty-assisted scram. The fourth rod or fine control rod (approximately one-half the diameter of the other rods) is driven directly by a lead screw. This rod may contain fueled or unfueled polyethylene. 1 ') 0 6 J g' 26

IV. SAFETY EVALUATION AGN-201, serial 108 Reactor design features, characteristics, and operating conditions have been previously evaluated in the Safety Evaluation by the Of fice of Nuclear Reactor Requ-lation in support of Docket 50-538 (reference a). Modifica-tions similar to those proposed in this application were evaluated for the AGN-201, serial 100 Reactor in support of Docket 50-43, and the safety and reliability of the modified serial 100 reactor was demonstrated for several years at the U.S. Naval Postgraduate Scnool (USNPGS) in Monterey, Calitor-nia (reference c). A. AGN-201, Serial 103 Reactor The following evaluation applies to alterations proposed in Part II.B. of this applica-tion. 1. Modifications to the core tank assembly wculd extend the tank containment boundary into a permanently installed gas handling piping systen and provide access for core temoerature instrumentation. Following installation, the core tank and piping system will be pressure-tested to 6 psig internal pressure which is approximately 200 of the expected pressure increase resulting from high power oper-ation (reference d). Prior to each reactor startup, fa-cility operating procedures will require verification that the core tank is pressurized with dry Nitrogen to at least 1 psig, but not more than 2 psig. A pressure sensor will initiate an automatic reactor scram for core tank pressures above 5 psig (maximum). MSU considers that the proposed test pressure, prestartup procedure require-ment, and automatic protective action will assure that the reactor will not be operated without the core tank integrity provided by the previously evaluated AGN-201 design (reference a). 2. Modifications to the Shield h'ater Tank Assembly would not compromise the secondary fluid-tight seals nor alter the mechanical strength and shielding characteristics of the assembly previously evaluated in reference (a). 3. Modifications to Instrumentation and Safety Systems would assure a minimum of at least two neutron detection and safety channels or monitoring reactor source level, startup, steady state and transient operations. The use of compensated ion.hambers in place of the existing un-compensated chambers,;0uld ensure accurate low level neu-tron detection for mixed radiation fields with thermal neutron to garma t atios much less than those possible in the AGN-201 Reactar. The scope of existing instrumenta-tion and safety system capabilities will be extended to provide an in-ccre temperature monitor and an interlock to assure gas-tight integrity of the core tank assembly. 4 Q r+ 1 7 r; ./ ist 27

Nuclear Safety No. 1 would provide linear countrate indi-cation from the minimum reactor source level to more than one decade beyond the power level at which criticality is normally achieved in the AGN-201 Reactor, This system will initiate an autematic reactor scram in the event of instrument or detector failure, and for countrates less than the minimum specified source level which assures that adequate neutron countrates are registered on this start-up channel before startup can begin. This low level scram would be deactivated at power levels above which channel saturation would occur provided the Channel 2 log-N in-strument has on-scale indication and has been verified to be operating properly. MSU considers that this mode of operation will not compromise reactor safety since the objective of the low-level scram feature will have been fulfilled prior to deactivation and the instrument will be restored to operation at power levels below which chan-nel saturation will not occur, and since the Channel 3 low-level scram will be functional for all ranges of op-eration. Due to the self-limiting action of the large neoative temperature coefficient in conjunction with op-eration of the core fuse, safe reactor shutdown without core damage is assured for instantaneous reactivity in-sertions as high as 2.0 ik/k. This protective action has been evaluated independently frcn any instrumentation channel carabilities (reference a). MSU considers, there-fore, that this assurance in addition to the redundancy provided by high power scram functions of Nuclear Safety Channels 2 and 3, provides an adequate margin for safety such that the Channel 1 high level scram function can be eliminated without degradation of existing safety system objectives. Nuclear Safety No. 2 would provide log-N and reactor pe-riod indication from the subtritical range, slightly above the minimur source level, to levels greater than the max-imum licensed power requested in this application. This safety system will initiate an automatic reactor scram at specified high power levels and at specified reactor periods which will ensure a controlled rate of power rise. MSU considers that raising the threshold of on-scale in-dication ' eliminating the low-level scram function for Nuclear Saiety No. 2 will not degrade safe operating per-formance since only subcritical operation is practical below the proposed threshold level and it is expected that Channel 2 will have on-scale indication following inser-tion of the two safety rods. In addition, a measure of redurdancy for protection against low neutron levels is provided by Channels 1 and 3. In the event of Channel 2 instrument or detector failure, operating procedures re-qui re imredia te reactor s hutdown <] 28 Y" I)

Nuclear Safety No. 3 will provide linear indication of neutron levels from the minimum reactor source level to more than one decade beyond the naximum licensed power requested in this application. This safety system will initiate an automatic scram in the event of instrument or detector failure and will thus provide redundancy to assure that levels above the minimum specified for at least two channels are registered on the instruments be-fore startup can begin. Nuclear Safety No. 3 will also initiate an automatic reactor scram at specified high power levels ano will therefore asauie high level scram redundancy for a minimum of two safety channels. Speci fications for the minimtm! reactor source level, max-imum rate of reactivity insertion, maximum potential ex-cess reactivity, and safety and control rod insertion se-quence interlocks in conjunction with existing facility startup procedures will result in an additional degree of instrumentation redundancy since all three safety chan-nels will have on-scale indication in the range for which criticality is normally achieved. Inus, MSU concludes that the proposed alterations to existing instrumentation and safety systems will provide adequate neutron "oniter-ing capabilities and suf ficient safety system redundancy to support controlled and safe reactor operation. The modification to the Interlock Line continuity circuit would ensure that reactor operations could not be conduct-ed with internal core tank pressures above those for which gas-tight integrity will have been tested. The addition of an in-core temperature monitor will provide a margin of confidence beyond that previously evaluated to ensure that Limiting Safety System Settings and Safety Limits will not be exceeded, and will not significantly al+1r core reactivity. Thus, MSU considers that the Interlock Line alteration and addition of in-core temperature mon-itor will provide assurances of reactor safety beyond those provided by existing AGN-201 design. Similar mod-ifications have been evaluated in reference c. 4. The Gas Handling System proposed in Part II.B.4 of this application, in conjunction with facility operating procedures, will perform functions and have capabilities equal to or better than the system used to operate the ACN-201, serial 100 reactor (reference f) which was pre-viously evaluated in reference c. The design, construc-tion, fuel loadin6, void spaces, neutron flux levels, and general operating characteristics of the MSU serial 108, AGN-201 Reactor are similar to those documented in refer-ence c. Due to the AGN-201 similarities, MSU considers that radiation damage to the serial 108 Reactor fuel discs, hydrogen and fission product gas diffusion rates from the polyethylene fuel discs, and Argon-41 production rates in reactor voids will not exceed those documented in ref erences d through g for the serial 100 reactor. 29 Lu, s a) i

Therefore, MSU concludes that the proposed Gas Handling Systen and associated operating procedures, in coniunction with the existing continuous air particulate activity monitor and alarm system, will ensure that the personnel protection requirements of 10CFR20 for airborne radio-activity can be adequately met. 5. The Safety and Control Rod Assembly modifications proposed in Part II.B.5 of this application are sinilar to those described in references e and g, which were previously evaluated in reference c. Due to similarities between MSU's serial 108, AGN-201 Reactor and the serial 100 reactor evaluated in reference c. MSU considers that the proposed modifications and corresponding gas handling procedures should pose no threat to safe operation beyond that previously evaluated in reference c. B. Reactor Buildina and Shieldinq. The following evaluation applies to alterations proposed in Part II.C of this applica-tion. 1. Continuous Operation at 20 Watts. Calculations for 44" concrete shielding around the AGN-201 Reactor (Appendix A) indicate that dose rates from gamma radiati,on at the outer sur%e af the concrete shield will be 1 0.8 r7/hr l'eutron radiation will not be transmitted througn the shield. Calculations for the surface of the consisting top shield consisting of 13" of borated paraffin show that the maximum dose rates would be 168 mR/hr gana and _0.2 mrem /hr fast neutrons for a water-filled therroal ccl-umn, and - 612 mR/hr gama and - 6 nrem/hr fast neutrons for a graphi te-filled thermal column. At 10 feet above the shield (height of roof), these dose rates decrease to 7 mR/hr gada and 5 0.1 mrem /hr fast neutrons (water filled), and 63 mR/hr ganma and < 0.6 mrem /hr fast neu-trons (graphite filled). Thus, the radiation dose rates outside the reactor roon will cnly exceed linits specified for unrestricted access (10CFR20) on the roof directly above the reactor. Existing procedures prohibit access to this area during reactor operation. 2. Intermittent Operation at Pc'.:er Levels Greater than 20 t.'atts. Calculations for the proposed shielding at 1000 watt op-eration (Appendix A) indicate that dose rates at the sur-face of the concrete shield would be 138 mR/hr garma and 1.06 mren/ hour neutrons. The borated paraf fin top shield will limit dose rates to f 3.4 R/hr g wma and 9 rjrem/hr neutron for a water-filled therral column, 2 and 2 31 R/hr gamma and 2 270 orem/hr neutrons for a graphite filled tnermal column. These values would de-crease to 1340 cP,/hr ganea and _1 mrem / hour neutrons (water filled thermal column) or 13.1 R/hr gan.na and i 27 nrer/ hour neutrons (graphite filled thercal column) at a height of 10 feet above the reactor (height of roof). 30 /G, 1 ~ j' , _s

Assuming the reactor could be operated at 1000 watts for 15 minutes (approximately 50' longer than expected), the highest dose available would be 2 8 rem on top of the poly shield, but access to this area will be prohibited by physical barriers during high power operation and the area is within the viewing range of the console operator via a window in the reactor room to control room wall. The highest dose in an area not in visible range of the operator would be available on the roof while operating with a granhite-filled thermal column. This dose would be 2 780 m'R gamma and I 7 mrem neutrons. The dose at the e shield would be 5 10 mR gamma surface of the corcre and '.02 mrem neutrons. Since not more than one high power operation could be conducted within a one hour time interval, and since access to the roof is prohibited dur-ing all reactor operations, MSU considers the administra-tive controls governing access to posted restricted areas to be sufficient to assure personnel protection from ra-diation exposure. 3. Accident Conditions. The original design analysis (Docket F-l$) Calculated Values for radiation doses available at the AGM-201 shield tank exterior wall for a hyrothetical 2 step change of reactivity. Assuming the potential dose for the duration of the accident to be 6 rem at the shield tank wall, the addition of the proposed concrete shielding would decrease this dose by a factor of approximately 2600 at the surface of the concrete, and by a factor of approximately 5700 at 10 feet from the concrete surface (control room area). Thus, the dose available to personnel in the control room for the dura-tion of such an accident wr id be approximately 1 nillirm and would not exceed the limitations of 10CFR50, Appendix A, Criterion 19 for accident conditions. Based upon the preceding information, MSU considers the additional shielding and extension of restricted and post-ed areas proposed by this application, in conjunc'i with m written rad ological controls procedures, to provide rea-sonable assrrance that maximum personnel radiation expo-sures will t e in compliance with federal regulations dur-ing normal nd potential accident conditions. C. Technical Srecifications. The followinq evaluation applies to technical spccification changes propused in Part III.C of this applicatior. 1. The revised Safety Limits would delete the existing specification which limits the maximum steady-state power level to less than 100 watts. The negative temperature coefficient inherent in AGN-201 reectors in conjunction with the maximum potential excess reactivity, experinents included, provides assurance that the core cannot sustain significant power levels for pe-riods of tim' necessary (cteady-state) to approach melting temperatures of the polyetaylene fuel matrix. 31 '9' '39

Data supporting this conclusion has been documented in reference e from actual operations up to and including 1000 watts (thermal). Maximum operating times experienced in the AGN-201, serial 100 Reactor at steady power levels ranging from 100-1000 watts with 0.5? excess reactivity were recorded when control rods reached full insertion indicating that temperature effects had reduced the re-activity to zero. Based upon this data, it was estimated that temperature effects would not impose a time limit on operation at powers below about 175 watts provided +he reactor contained 0.7; excess reactivity, which is greater than the maximum potential reactivity presently permittod for AGN-201 reactors including any installed experiments. The calculated steady-state temperature rise of the AGN-201 core is 0.50C/ watt (Docket F-16) and has tradition-ally b un used for AGN-201 core evaluations over the past twenty years. This value would yield a steady-state temperature rise of 37.5"C above ambient for continuous operation at 175 watts. The temperature at the acre cen-ter would be greater by a factor of approximate ly 1.S4 (reference h), or 1350C above ambient. Assuming the core fuse would not function to shutdown the reactor, the rt-sulting te:rperatures remain less than the melting temper-ature specified for the polyethylene fuel matrix. These steady-state values at 175 viatts appear to be very con-servative since the negative temperature coefficient of reactivity, both calculated and measured, is significantly greater than is implied by the postulated steady-state temperature rise vs. available excess reactivity, and, though it is recognized only as qualitative data, the apparent core temperature fo, SS minutes operation at 200 watts documented in reference e is significantly lower than would be expected. For these reasons, MSU considers that deleting a maximum power value from Section 2.1 of the technical specifica-tions will not increase the probability that fission prod-uct retentien and fuel matrix integrity would not be r.:aintained during the maximun steady-state operation achievable, nor would establishing this parareter o a Safety Linit provide additional protection beyond that offered by the core fuse. Rather, it is considered that a liniting value for reactor power appears to he irrelevant to the stated objcctives for Safety Limits and would be more apropriate as a Limiting Condition for Operation. 32 L' Q, 1 ?C /4 1JJ

2. The basis for the shield tank water level Safety Linit would be reworded to address the proposed modifi-cations and represents no change to the existing speci-fication for reactor operation. 3. The revised Limiting Safety System Settings would delete the existing specificgtion which requires a re-actor scram at power levels - 0.2 watts. In view of the objectives for LSSS, evaluation of this action is the same as that for the proposed change to the Safety Limits. Reactor high power scram requirements would be specified as Limiting Conditions for Operation. 4. Proposed revisions to the wording of reactivity limits specified as Liriiting Conditions for Operation would de-lete the reference to 20" C for available excess reactivity, and insert the reference to 20 C for minimum shutdown margin. 0 The available excess reactivity referenced to 20 C would permit an actual excess reactivity of more than 0.65; to exist for reactor operations conducted at terperatures U below 20 C and is therefnre not consistent with the basis for this specification. MSU considers that tae available excess reactivity limit should be referenced to the actual condi tions existing during reactor operation. Thus, de-0 leting the reference to 20 C will provide additional clarification that the excess reactivity shall not exceed 0.65' for any condition of reactor operation. The specification for ninimum shutdown margin with the most reactive safety or control rod fully inserted has no coron reference for measuremant. Althouch the max-imum reactivity worth of control and safety reds for AGh'- 201 reactors in conjunction with the maximun available excess reactivity, including potential reactivity worth of expericients, is such that a 2 k/k specification for shutdown margin at any temperature typical of reactor op-eration will ensure the reactor can be brought and main-tained subcritical, MSU considers that referencinq this specification to 20 C will provide clarity and a rare consistent basis for comparison during the conduct of annual surveillance requirements. 5. Proposed changes to the Limiting Conditions fo: Oper-ation of Control and Safety Systens would require a min-inum of two nuclear safety channels to be in operation, ensure that gas-tight integrity of the core tank is main-tained, and ensure that hazardous concentrations of free hydrogen would not be formed during all conditions af re-actor operation. A requirement for an air particulate monitor to be in service during reactor operation would also be added. Specifications for interlocks, minimum indicated neutron level for startup, shield water temper-ature and level, seismic displacement, manual scram oper-ability, reactor period, and radiation monitors would main as required by the existing license. /# U 33

a. Since the core tank and gas-handling system will be tested to at least 6 psig, specifications for at least 1 psig nitrogen pressure, and initiation of re-actor scram at pressures greater than 5 psig (inct eas-ing) in the core tank would ensure that the reactor is not operated without core tank and cas handling system pressure-tight integrity. In aduition, the minimum pressure specification will ensure that suf-ficient nitrogen is maintained in free-void spaces of the core tank to minimize the probability of an ignition or explosion of hydrogen which may be re-leased from the UO - polyethylene discs. This mode of operation has pfeviously been evaluated as provid-ing an adequate margin for safety (reference c), and operation of the AGt4-201, serial 100 reactor has dem-onstrated the validity of this evaluation. Due to the similarities of design, construction, operating characteristics, and modes of oneration, MSU considers the safety evaluation of reference c to be applicable to the AGM-201, serial 108 reactor a' d that the pro-posed Limiting Conditions for Operation for core tank pressures do not represent an unreviewed question of reactor scfety. b. The existing technical specifications require three nuclear safety channels to be operable with the exception that either Channel 1 or Channel 3 nay be bypassed for up to 12 consecutive hours provided the reraining two channels are verified to be oper-able. The proposed change to this specification would permit only Channel 1 to be bypassed for oper-ations above approximately 40 milliwatts, provided both Channels 2 and 3 are verified to be operable, to prevent overranging this instru: rent. Operation of the reactor would not be permitted with Ct.annel 2 or Channel 3 out of service, and all three nuclear safety channels would be operable for reactor startup and the rarge in which criticality is normally achieved. MSU considers that the proposed change will provide adequate safety channel and neutron c onitoring re-dundancy for all ranges of reactor operation and will not lessen the degree of redundancy required by the existing license. c. The ninimum indicated source level for reactor startup is presently specified as an instrument set-point for Nuclear Safety Channel No. 1 which is es-tablished at 120 CPM. Since the Channel 1 instrument is a linear countrate instrument emaloying a range selector swi tch, a fixed neutron level at which the low level scram setpoint can be established is only appropriate for one corresponding position of this switch. As the operatar changes ranges to tronitor increasing or decreasing neutron levels, the percent of full scale reter deflection corresponding to a fixed countrate (neutron level) also changes. 34

• V" b[

Therefore, the proposed alteration would clearly spec-ify 120 CP'l as the minimum indicated source level for commencing a reactor startup, and would establish the instrument low level scram setpoint at 5 12 of full-scale deflection which is sufficiently greater than the noise level for all positions of the range selec-tion switch. MSU considers that the proposed change will increase the deq'ee of confidence that the reac-tor will not be star'ed up without an adequate source level and that the s tpoir.t in terms of percent full scale meter deflection will ensure that the instrument will be responding sufficiently above electronic noise levels for all positions of the range selector switch. d. Scram protection from high power levels would be provided by Nuclear Safety Channels 2 (log N) and 3 (linear). The Channel 3 setpoint would be established at 2 95" full scale meter deflection and Channel 2 would provide redundancy at 5 2000 watts. Channel 2 would also provide a scram for reactor periods less than 5 seconds. Since Channel 3 contains a linear piccarreter which erploys a range selector switch, a scran setpoint based upon a fixed neutron level would only be appro-priate for one corresponding position of this switch. As the operator changes ranges to monitor increasing or decreasing neutron levels, the percent full scale deflection corresponding to a fixed neutron level also changes. Thus, the, Channel 3 scram setpoint would be established at 2 95 full scale meter de-flection. This instrument will also initiate a scram signal for indications < 5' full-scale deflection to ensure the range selected is providing indication greater than instrumentation noise levels. Since it would not be possible for an operator to operate the range switch rapidly enough to maintain on-scale in-dication for Channel 3 during a rapid reactivity in-sertion that resulted in a pcriod less than that normally permitted, a scram signal would be initiated on the same scale for which the transient began and within a maximum possible interval of 0.; M ade me-ter indication above the initial power level, In the event Channel 3 failed to scram, the Channel 2 log-N instru: rent would provide redundant high power scram protection at a fixed power level 5 2000 watts. The negative temperature coefficient of reactivity that nos been demonstrated in AGN-201 Reactor cores, in conjunction with the maximum potential excess reactivity, provides assurance that safety limits would not be exceeded at the maximum steady-state power levels achievable (Part IV.C.1). 35 4 o " ') 4 i 30

Moreover, the hazards summary contained in references h and i show that, without automatic action from neu-tron monitoring instruments, degradation in fission product containment should not occur from a posiLive step change in reactivity as large as 2. More re-cently, a transient analysis was conducted for the AGN-201 Reactor which assumed a step change in reac-tivity corresponding to one-dollar and considered termination of the transient by scram rod withdrawal one second later, which is a time equivalent to a factor of more than 3 times the estimated instrumenta-tion response characteristics (reference j). MSU considers this latter anaylsis to also be applicable to the AGN-201, serial 101 Reactor. Utilizing the graphs shown in Figures 2 and 3 of ref-erence j which are based upon a one-dollar step change in reactivity initiated from a power level of 0.1 watt. a scram actuated by Channel 3 at 0.9 decades above P(t=0) would correspond to rod withdrawal being in-itiated at t=0.34 seconds (Appendix B). This assump-tion considers a trip at 0.9 watts indicated plus 0.3 seconds response time, and shows a reak transient of < 30 watts. Considering a scran from Cnannel 2 at 2000 watts, rod withdrawal would occur at a point on the graph (extrapolated) located at t-1.3 seconds (Appendix B). The peak transient would be < 20 kilo-watts. The total energy release would be < 3000 joules and resultant radiation dose to a person stand-ing next to the reactor (without the additional shield-ing proposed in this application) would be < 33 mrem MSU considers that this analysis pr'vides a conserva-tive point of view since the transient begins at a power level below that for which temperature coef fi-cient feedback effects are significant and thus rep-resents a more rapid power rise during the interval between transient initiation and rod withdrawal than if calculated for Po > 0.1 watt. In addition, the Channel 2 period scram is not considered in the anal-ysis, nor is the additional shielding that is proposed in this application. Therefore, MSU considers that the proposed high power scran setpoints would provide reasonable assurance that personnel exposure limits or reactor safety limits would not be exceeded during the maximum credible transient. e. The proposed Limiting Conditions for Operation would require the existing Air Particulate Mor.itor to be in service during reactor operation. This mon-itor is located in tne Contrcl Room and continuously samoles and returns air to the Reactor Roum atnosphere. The monitor has a continuous readout capability and provides an audible and visual alarm if airborne par-ticulate levels in the Reactor Room exceed an adjust-able alarm setting. 36 74 bI

MSU considers that the continuous Air Particulate Monitor, in conjunction with operating procedures, would provide redundant protection to ensure the re-actor is not operated without pressure-tight integrity of the core tank ard gas monitoring system. In ad-di+. ion, the monitor would provide sufficient warning so that an assessment of personnel hazards from air-borne particulate activity can be made prior to enter-ing the Reactor Room. f. The proposed Limiting Conditions f or Operation applicable to reactor shielding would delete the re-quirer:ent to prnhibit entry to all areas in which dose rate is > 1 mr/hr (measured at licensed reactor power). The additional shielding proposed in this application, in conjunction with reactor operating and facility radiological controls procedures, provide assurance that potential radiation doses will be in compliance with the requirenents of 10CFR Part 20 and Criterion 19 of Appendix A to 10CFR Part 50 for accident condi-tions. For this reason,I",SU considers that a basis no longer exists for the prohibition anainst entry into areas in which the dose rate > 1 mr/hr. However, since Operation at the maximum power level requested by this application would result in high rddiation areas direCtly on top of the reactor and pro-posed additional top shield, a requirement has been added to prohibit access to the reactor room, except for radiation surveys, during operations greater than 20 watts. Since the area of concern is visible to the console operator via a viewing window, and since access to the reactor room is only available through a norolly locked door from inside the control room, MSU considers that this requirement should increase the degree of confidence that 10CFR20 requirements will be met during high power operations. D. Transoortation and Storage of Fuel, Personnel from Memphis State University will supervise the transport of reactor fuel from Oak Ridge I ational Laboratory to Neghis, TN. Preparatien and transfer will be in accordance with 10CFR Parts 70 and 71. The fuel will be packaged for transport as Fissile Class III shipments under the general license provisions for shipment of licensed material. The fuel will be divided and stored in 5 sealed DOT specifica-tion 6C containers. Each container will be packed with vermiculite and placed atop an empty 6C container inside a DOT specification 6J container, which will then be filled wi th vermiculi te and sealed. The containers will be trans-ported in two separate shipments in an enclosed and locked van with two drivers. Because calculations indicate that the critical mass for AGN-201 fuel is more than 650 grans of U-235, packaging and transporting the fuel in this man-ner precludes accidental criticality. 37 1 t n. I..,, i / .v

MSU considers that the fuel can be transported without haz-ard and will be adequately protected against theft by using the preceding method. Mc +4c State University will store the fuel on site as re-p i a m.. - -' 4-the installed AGN-201 Reactor core. The fuel wil: ren,u.. m the 6J containers which will be locked in an area which re!uires access through at least two lock-ed doors and which is protected by an electronic burglar alarm and is patrolled by a security force (reference b). MSU considers that the storage precautions will provide adequate security since the licensee will not possess an amount of special nuclear material that is equal to or greater than the formula quantities specified in 10CFR part 73. Since facility procedures will limit the quantity of con-tained U-235 which will be handled, used, or stored in the proposed fuel storage area or the reactor room to 5 700 grams at any one time, MSU does not consider that the crit-icality monitors specified in 10CFR Part 70.24 will be necessary. E. Enviro: mtal Considerations. Environmental considera-tions f or licensing the Mc:phis State University Pe:earch Reactor to operate at 0.1 watt are contained in the Negative Declaration and Environrental Impact appraisal dated June 14, 1976, which was issued with Construction Permit No. CpRR-122 on June 15, 1976 (reference k). MSU considers that operation of the AGN-201 Reactor at the power levels requested by this application will not significantly alter the factors evaluated in that appraisal. The fence to be constructed as described in Part II.C of this application would not enclose or alter areas occupied by wildlife, vegetation, nearby waters, cr aquatic life. Other physical alterations to be performed are within the boundaries previously evaluated in reference k for site preparation and facility construction. Additional thermal effluents from operation of the reactor at powers up to and including 1000 watts would be rejected to the surrounding water tank and eventually to the atnos-phere by means of conduction and radiation. There would be no release of liquid effluents. However, the potential for gaseous effluents is increased due to diffusion of ra-diogases from the polyethylene fuel discs of the reactor core following operations at power levels significantly greater than 20 watts. Operation of the AGN-201, serial 100 reactor demonstrated that a total activity of 40 p Ci could be present in ap-proximately 2700 cc of evolved gas corresponding to a 3 psi pressure increase in the core tank assembly

• ~) h kkk 3,0

The accumulation of the evolved gases, most of which was hydrogen, took place over g period of 2 to 3 days follow-ing a high power run of 10 watt-minutes. The total activ-ity rtyresented that activity present in two-day old gas. There was no evidence of gross fission products other than the inert radiogases with a long-lived component that was pi esumed to be Krypton-85. The amount of evolved gas that was released at any one tire amounted to 5 11 of the total gas in the core tank, and was released in a controlled man-ner to ensure that effluents were below the maximum permis-sible concentratiens for release to unrestricted areas. The procedure, assu:rptions, and calculations for such re-leases are contained in reference f. Although the potential for formation of releasable gaseous activity to the environa:ent is increased from high power operation of the AGN-201 Reactor, MSU considers that the additional orecautions including the higher than design core tank and gas-handling system test pressure, high core-tank pressure ccram interlock, and proposed technical speci-fications correspondingly increase the degree of confidence that cas-tight integrity of the core tar,k will be maintained. Therefore, the potential for an unplanned release of gaseous effluents should be no greater than previously evaluated in reference L Planned releases of evolved gases will be necessary. Since the proposed gas-handling system, method of operation, and method of discharge and dilution to be used with the MSU re-actor are closely similar to that previously evaluated in reference f, MSU considers that evaluation te be applicable to the alterations proposed for the serial 108 AGN-201 Re-actor. Because the planned releases arc controlled to en-sure that maxir:um permissible concentrations for release to unrestricted areas are not exceeded, MSU considers the environnental effects from gaseous effluents to be insignif-icant. MSU considers that operation of the reactor at the power level requested by this application, in conjunction with the proposed alterations to radiation shielding, posted and controlled restricted areas, and technical specifica-tions will not alter other environmental considerations appraised in reference k F. trercency and Security Plannino. MSU does not propose any changes to the Security Plan or the Emergency Plan previously evaluated and approved in reference a. .n

• ,ju

-, {

REFERrnc[s a. Memphis State University facility Operating License R-127, effective December 10, 1976 (Docket 50-538). b. Application for Construction Permit and License to Operate the Model AGN-201, serial 108 Nuclear Reactor at Memphis State University, dated April 11, 1975, as amended (Docket 50-538). c. U.S. Naval Postgraduate School Facility Operating License No. R-11, as amendcd (Docket 50-43). d. Report of Operation of Reactor Facility License No. R-11 Amendment No. 1, transnitted by ltr. U.S. Naval Postgraduate School to U.S. Atomic Energy Commission, dated March 30, 1962 (Docket 50-43). e. Report of Operation No. II of the U.S. Naval Postgraduate School's ASN-201 Reactor Facility, dated January 6,1964 (Docket 50-43). f. Supplement I to Application for Amendment to Facility License R-11 to Permit Disassembly and Reassembly of Reactor Core, dated June 10, 1963 (Docket 50-43). g. Request for Amendment to Facility License R-11 to establish new procedures for nonitoring for Argon-41 and to modify the control and safety rods..., dated January 8, 1964 (Docket 50-43). h. Aerojet-General Nucleonics, Elementarv Reactor Experimentation, A. T. Biehl, et. al., Ed. (San Ramon, California:

October, 1957).

i. " Hazards Sunmary Report for the AGN-201 Reactor," Aerojet-General Nucleonics, August, 1956 (Docket F-15). j. "A Safety Analysis for the Georgia Tech AGN-201", J. Narl Davidson, July, 1976 (Docket 50-276). k. " Envi ron: ental Considerations Regarding the Licensing of the Memphis State Univers, j Research Reactor," Construction Permit CPRR-122, USNRC, June 15, 1976 (Docket 50-533). ^ '/ 4 h3 40

APPET; DIX A DOSE R,UE Cid.CULATIONS 1 [} i tj z: s

APPEjijlIX A DOSE RATE CALCULATIONS FOR 44" CONCP,ETE SHIELD WITH 18" POLYETHYLEilE (BORATED) TOP A. Introduction. Radiation dose rates measured during low power operatian of the AGil-201, serial 103 reactor are used as a basis for assuming the conservativs values which are used in the dose-rate estimates for 20 watt and 1000 watt operation. The measurements are made with portable survey instruments and have been corsistent over a period of more than two years. The gamma instrum^nts are low range (0.50 n:R/hr. ) and are caljbrateousingaCobalt-60sourcewhosestrengthisknown to - 1.6? The neutron instrument utilizes a ten-inch poly-ethylene snhere with sc ntillation detector and has a spectral i response closely approximating the dose curve for neutron energies from thermal to / MeV, anc retains capabilities through 12 MeV. The neutron survey instrument is regularly calibrateg using a California -252 source whose strength is known to - 3' P O',lE R LOCATION G A'"4A TEUTRO1 50 m Watt Shield Tank exterior 3.5 -R/hr. < 0.1 mrw/hr (Glory Hole plane) 50 m Watt Thernal Column Top 2.3 mR/hr. 0.2 mrem / hr. Cover (Water filled) The measured dose-rates are higher than those estimated in the AGN-201 Preliminary Design Analysis (docket F-15), pre-sumably due to radiation scatter ani due to operation with-out boron additions to the shield water, - qu I 7 6 Al

B. _As s upp t i o ns. 1. Dose-rates at 100 milliwatt operation are conservatively assumed as: LOCATIO:1 GAT"4A i;EUTRO 1 Shield Tank exterior 10 mR/hr. 0.3 mrem /hr. Thermal column top (water filled) 6.6 rR/hr. 0.6 mrem /hr. Thermal column top (graphite filled)* 60 mR/hr. 18 mrem /hr.

• Dose-rates at top of graphite filled thernal column are es-timated to be greater than water filled by factors of 9 (gacra) and 30 (neutron).(1) 2.

The assumed dose rates for 100 n!lliwatt operation are considered to be those caused by an equivalent point-source located 10 cm. within the core tank assembly. The AG:4-201 shielding thus consists of 10 cm. core, 20 cm. graphite, 10 cm. lead, and 55 cm. water corprising a total distance of 95 cm between shield tank exterior and source of radiation (see docket F-15). 3. An v erage attenuation coef ficient for gamma photons is assu"^ed for 44 inches of ordinary concrete, and is Nsed Jp-on the following factors: y Ei;ERGY n/a(2)r(2) ] 1 'i i B( ) B.elili_ -3 7.7 MeV .0?4 2.35 .056 112 6.26 2.9 5.5 X 10 -3 6.0 MeV .0c/ 2.35 .064 112 7.15 3.5 2.7 X 10 -4 3.0 MeV .036 2.35 .035 112 9.50 6.7 5.0 X 10 -4 2.2 MeV .043 2.35 .101 112 11.30 11.4 1.4 X 10 -0 1.0 MeV .064 2.35 .150 112 16.30 38 1.9 Y 10 AVERAGE ATTEi;UATIC:1 COEFFICIE IT = '- by i11 = 1.8 X 10-3 The average coefficient is considered conservative since it represents less attenuation (higher transmitted dose-rate) than for values corre ponding to energies below approximately 6 MeV. Gamma photons with energy greater than 6 l'eV consti-tute only a small fraction of the spectrum outside the design AG:1-201 shield (docket F-15). 1Biehl, et. al., Elementary Reactor Experimentation, Aerojet-General l;ucleonics, October, 1 ~7 2Aill-5800, Reactor Physics Constants, Second Edition, Argonne i;ational Labora tory"(USAEC: July, 1963T ~~ 4 3 [ Ih ' - ' I: A2

4 A representative relaxation length, Ai, for ganma photons in polyethylene is assumed to be 21 cm. (docket F-15). 5. Representative relaxation lengths,.ti, for fast neutrons are assumed to be }2.1 cm. in ordinary concrete and 8 cm. in polyethylene (Docket F-15, (2)). 6. it is assumed that the radiation dose-rates will be pro-portional to the rate of energy release in the reactor, and that the dose rates at 100 milliwatts can be extrapolated using the following relationship: b=K(Y)( ) A, where d b=extrapolateddose-rate K = dose / Joule, and is determined from the assumed dose-rate at 0.1 watt, i.e., K= 0.1 (.1N) (3600 J/w-br) -f = rate of energy release in Joules / hour. Ro = 95 cm R1 = distance in cm frco equivalent point source to the point of interes t. A = Attenuation coei.icient for the proposed additional shield-ing. 7. No credit is taken foi the free air distance between the reactor shield tank and the proposed 44" thick concrete shield, which will be approximately 16 inches. In addition, no credit is taken for the 8" concrete block walls that enclose the Narth, West, and South boundaries of the Reactor Room. /'

• i A3

C. Calculations. 1. Ga:nma Dose Rates at 20 Watts a. Concrete shield: b = (10 n2 ) (7.2 X 10 S-- 5 hr.) (207) (1.8 X 10-3) = 0.76 "hr.O-- y hr. 360 J/hr, b. Polyethylene top co /er (water filled thernr1 column): (1) b = ( p0--- -) ( 7. 2 X 10 )( )2 6 ex. {- - } = 68 nR/hr. where R1 = 95 ! 45.7 = 141 cm. Xi = 45.7 cm. ), i = 21 cm. (2) Ten feet above top cover (roof level) b = (68 ) (f {f{} =,7 nL/jui R c. Polyethylene top cover (graphite filled water column) (I) (CL h-) (9) = 61_2_gRfhr. = y (2) Ten feet above top cover (roof level) rD (7 hr.) (9) = 63 tR/hr, D 1 = y 2. Ganra Dose Rates at !nUO Watts a. Concrete shield: b=(0 i - ---l) ( 3. 6 X 10 'f) (2 7) (1.8 X 10-3) = 38 rP/hr. 6 9 b. Polyethylene top cover (uuur filled thermal colu:rn) 6 ( c 4) (3.6 X 10 J/hr.) ( ) exp. f pf (1) D = = 3.4 R/hr. (2) Ten feet above top cover (roof level) (1.4 R/hr.) ( -) = y,0 3 /hr_ D, = c. Polyethylene top cover (graphite filled thermal colunn) (1) D (3.4 R/hr.) (9) = 30.6 R/hr = 7 I () L ! d (2) Ten feet above top cover (roof level) b = (340 _) (9) = 3.1 R/hr y A4

3. fleutron Dose Rates at 20 Watts a. Concrete shield: i) = (0,l p iyg/nn. ) ( 7. 2 X 10 J/h r. ) / 5 -)2 { Xi 4 9 exp. n 3bd J,n r. '207 li =.0012 mren/hr where Xi = 112 cm li = 12.1 cm-b. Polyethylene top ccver (water filled thermal column) b " ( 10 -)(/.2X10'J/hr.)(h)exp. f -l n = 0.18 rnren/hr. where R1 = 95 + 45.7 = 141 cm. Xi = 45.7 cm. Ai = 8 cm. c. Polyethylene top cover (graphit. filled thermal column) (0.13 rrem/hr.) (30) = 5 4 nrer/hr (1) D = n (2) Jen feet above top cover (roof level) (5.4 mrem /hr..)(f{c) = 0.54 mrep/hr _c D = n f;0TE : i;o credit taken for 5 "/o Boron in polyethylene or con-crete. 4. Tieutron Dose Rates at 1000 Watts Concrete shield: a. 6 112cm n 360 J/hr. ,/hr.) (95 )2 0.3 miem/hr.) ( ~~ exp. 12.1 cm f D \\ 207 =.06 mrem /hr b. Polyethylene top cover (water filled thermal column) n J/hr. ) (f-f1)2 'h_r ) (3.6 X 106 f-l'- } / (1) b = ( ~ _z exp. = 9 mrem /hr (2) Ten feet above top cover (roof level) D = (9 mr /hr.. ) (f-1) = 0.9 mrem /hr. n A5 gg, ,/, Q j ./

c. Polyethylene top cover (graphite filled thermal column) (1) D = (9 mrem /hr.) (30) 270 mrem /br = n (2) Ten feet above top cover (roof level) D = (0.9 mrem /hr.) (30) 27 mrem /hr. = n NOTE: fio credit taken from 5 "'o boron in polyethylene or con-crete. 5. Thermal Neutrons a. Assuming a typical oly-Coron mixture with density 0.94 g/cc containing 5 [/o Coron additive, the ef fective Boron density would be: e = (.94) (.05)=.047 g/cc The renoval cross-section would be: t No , wher e

.t herm a l AW

= densi ty, g/cc No = Avogadro's number, atoms /rmle

a = Microscopic absorption cross-section, cm AW = Atomic weight

'037_) (._. __JJ_5_9_[ = 1. 95 cm-1 ' thermal b. The thermal neutron attenuation by the 18 inch top shield can be estimated by: exp. {-: (i), where lo I = thermal neutrons from the shield Io = thermal neutrons in to the shield = renoval cross-section Xi = shield thickness I -30 Thus, yo = exp. { - 1. 9 5 (4 5. 7 cm. ) : = 2 X 10 A6 i 'cn i" .) U

^ O c. Since it is intended that one cylinder of the blocks comprising the 44" thick concrete cylindrical shield also be borated to = 5', no thermal neutrons are expected to be transmitted through the proposed additional shielding. 4 A7

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