ML19350C154

Ctr for Nuclear Studies,Memphis State Univ,1980 Annual Rept,Nuclear Reactor Operations.

ML19350C154
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Site:	05000538
Issue date:	03/31/1981
From:	Dietz R, David Jones MEMPHIS STATE UNIV., MEMPHIS, TN
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_ Center for Nuclear Studies Memphis State University 1980 Annual Report Nuclear Reactor Operations License R-127, Docket 50-538 i AGN-201 Nuclear Reactor, Serial 108 i

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Center for riuclear Studies fiemphis State University 1980 ANNUAL REPORT of NUCLEAR REACTOR OPERATIONS AGN-201 Nuclear Reactor, Serial 108 Facility Operating License R-127, Docket 50-538 i

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Mr. R. L. 'Dietz Uf." 7. W. Jones Supervisor / Nuclear Operations Dirgetor a

ABSTRACT The 1980 Annual. Report of Nuclear Reactor Operations is prepared in accordance with Specification 6.9 of Appendix A to the Memphis State University Facility Operating License R-127, Docket 50-538. The report includes the period from January 1 to December 31, 1980. The

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results of certain reactivity measurements made in 1981 are also in-cluded in order to maintain continuity of reported data.

Reactor encs tions during 1980 were primarily for the purpose of

operator train?ng and no new or previously untried experiments were performed. The maximum steady-state power level at which the reactor was operated was 60 milliwatts. The reactor and associated systems functioned as designed and no safety-related maintenance was required.

There were no unscheduled shutdowns. Various measurements of core parameters and raciation survey measurements of the facility do not significantly differ from those contained in previous reports of AGN-201 operations.

Amendment No. I to the facility operating license was issued by the USNRC in 1930.. This amendment authorizes steady-state operation up to 20 watts and intermittent operation up to 1000 watts (thermal) provided that modifications to the- facility are made in accordance with the application for amendment, and that written notification is made to the USNRC that the modifications have been completed. The AGN reactor has been redesignated AGN-201H and an increase in the amount of Special~ Nuclear Material (SNM) for possession and use has been authorized. This latter authorization places the facility into the " low" category .for strategic nuclear materials as aefined by 10CFR'73.67. Accordingly, an. upgraded physical security plan has been submitted to the USNRC and is being evaluated. Modifications to the ~ reactor and. facility,' however, have not been made and additional SNM has not been received. Thus, the reactor has not-been operated

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beyond the limits specified in .the license prior. to amendment.

Section K of this report containsla statistical summary of personnel exposures pursuant to' Chapter 1, Part 20.407, 10CFR.

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TABLE OF CONTENTS Page Nu:::ber

~ AB STPA CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . i

' C0'iT E flTS ~ . . . . . . . . . . . . . . . . . . . . . . . . . . ii A. - REACTOR OPEPATIfiG EXPERIENCE . . . . . . . . . . . . . . 1 B. UNSCHEDULED' REACTOR SHUTDOWNS- ..... .... .. .. 2 C. PREVEriTIVE- AND CORRECTIVE FAINTE!!AfiCE .. .... .. . 2 5

1. Major Safety-related- Corrective Maintenance

2. Resul.ts of Major Surveillance Tests

-and: Inspections

~D. ' CHANGES IN FACILITY DESIGN, PERFORFANCE-

- CHARACTERISTICS, OR PROCEDURES RELATED TO -

LREACTOR' SAFETY ... ... .'. . . ...,. ... . . . . . . . .

- 4 E. CHANGES WHICH. AFFECT THE FACILITY'S

-DESCRIPTIO1 .. . . . . . . ... .-. . . . . .. ...... 4:

F. : CHANGES T0 PROCEDURES . .,.-. . . . . . . . . . . .-. .-

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G. NEW.0R UNTRIED EXPERIMEfiTS.... ... . . . . ... . . . . .

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IH. 'RADI0 ACTIVE!EFFLUEfiTS. . . .:. . . .-. . . . .r. ......- 5-

-I. LENVIRONMEfiTAL SURVEYSSPERFORMED--

> .:0UTSIDE THE~ FACILITY-... . . ... .-. . . ... . . .:.-. . =- 5 '

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Page Number J. RADIATION EXPOSURES GREATER THAN 100 MILLIREM , . . . . . . . . . . . . . . . . . . . . . 6 K. PERS0flNEL EXPOSURE AND MONITORING FOR 10 CFR 20 ... . ........ ......... 6

1. Total Number of Personnel Monitored

2. Statistical Summary L. AUDITS AND INSPECTIONS . . . . . . . . . . . . . . . . . 6 APPENDIX A. AMEflDMENT N0. 1 TO FACILITY OPERATING LICEtlSE R-127 iii w

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A. REACTOR OPERATIllG EXPERIEriCE

1. Operator Training Programs Six power plant employees participated in programs designed by Memphis State University (MSU) to provide research reactor startup experience for cold license candidates.

.In addition, 29 students from two classes of MSU's specialized fluclear Skills Related Training Program performed reactor startups and training exercises as part of that program's normal curriculum. A total of-106 reactor startups were conducted during the operations exercises that comprised these training programes.

2. Staff Operator' Training Several startups were conducted for the purpose of main-taining and evaluating operator proficiency. As'of

-December 31, 1980,- the MSU Center for Nuclear Studies staff held two Senior. Operator licenses and one Operator

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~ license for the AGN-201 Reactor.

3. - AdditionalJ0perations :and Total Startups '

r Additional operations were conducted for the purpose. of-

. satisfying surveillance requirements and routine tests or calibrations. A total of 131 successful reactor start-ups were performed 'during the period of -this report.

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4. Total Hours -of Critical ~ 0peration During 1980:

61.'46

. Max. Power Max.' Power Month . Hours" (Milliwatts)  : Month ~ Hours (Milliwatts)

~ Ja'n ' 10.9 56. Jul '2.18 50 Feb' f0' 0: Aug: _.'O O

Mar - 1. 0 - 50 Sep; 2.45- .'4 9 Apr- . 0. 65._ 50 Oct. .16.38- 54-May ' O.38 ' 50 Nov. 9.73 50-Jun '3.72 48- Dec 14.07 .60

B. UNScilEDULED REACTOR SHbTDOWNS None

C. PREVENTIVE AND CORRECTIVE MAINTENANCE

-1. Major Safety-related Corrective Maintenance None required.

2. Results of Major Surveillance Tests and Inspections

a. ' Safety and Control Rod Assemblies Fuel Inspection:

This bi-annual surveillance procedure was completed on October 14' 1980. A small quantity of loose fuel particles (<.I milligram U-235 content) was observed in the fuel capsules for the Coarse Control Rod and both Safety Rods.__ The quantity and circumstances were similar to those reported during the 1978 inspection and are not considered to be the result of abnormal deterioration of

.the AGN-201. design fission product barriers.

b'. ' Reactor Shiel.d' Tank Visual Inspection: This bi-annual surveillance procedure was completed on October 13, 1980.

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' The shield tank water-was sampled (gross activity less tha~n minimum detectable, where MDA = 5 x 10 uCi/;al referenced to Cobalt-60) andLdrained to permit entry.

~The shield water level interlock microswitch and actuating . lever were replaced' as a: preventive measure,

-and minor areas of spot-rust' were cleaned and repainted.

The shield tank was refilled with. reactor' grade pure water.

c. Control L Rod: Drive. Assembly Inspection and Lubrication:

This-annual' surveillance procedure was. completed on.

.0ctober113, 1980. . The' drive assemblies were found in

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satisfactory condition with no evidence of.: abnormal wear

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or deterioration.

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d. Measurement of Safety and Control Rod Scram and Insertion Times: This annual surveillance procedure was completed on January 14, 1981. The results are included in this 1980 report to provide continuity of raported data.

Insertion (cm/sec) Scram (millisec.)

Safety No. 1 0.421 166 Safety No. 2 0.469 143 Coarse Control 0.367 149.6 Fine Control 0.564 N/A

e. Reactivity Measurements: This annual surveillance require-ment was. completed on February 20, 1981. The results are included.in the 1980 report to provide continuity of reporte'd. data.

Parameter '% Reactivity Control Rod Integral Worth:

. Fine 0.32 Coarse- 1.21 f Reactivity! Insertion Rate:

Safety No. :1 .032/sec Safety. No. 2 .~035/sec

. Coarse .028/sec

-Fine .01/sec Excess- Reactivity -(Glory Hole' Empty, 200C,.all Rods'IN) 0.196'

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Shutdown Ma,rgin;(Most Reactive Rod.IN)- '2.66 e

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D. CHANGES IN FACILITY DESIGN, PERFORMANCE CHARACTERISTICS, OR PROCEDURES RELATED TO REACTOR SAFETY No changes in the existing facility design or performance charac-teristics were made during 1980. However, such changes have been authorized pursuant to- Amendment No. I to the facility operating license. This amendment was issuec by the U.S. Nuclear Regulatory Commission (USNRC) on March 28, 1980, and is described in Section E of this report.

s 'E. CHANGES WHICH WOULD AFFECT THE FACILITY'S DESCRIPTION dmendment No. 1 to the Facility Operating License No. R-127 has re-i designated the reactor as AGN-201H. It authorizes, upon completion of all modifications-described in the application for amendment and written notification thereof, steady state ope ation not to exceed 20 watts' thermal and _ intermittent operation not to. exceed 1000 watts

' thermal. _The amendment further incorporates revised Technical Specifications for.the authorized power levels. A copy of this amendment is included in this report as Appendix A.

Amendment No. I to the facility operating license also authorizes MSU to: increase possession limits of special nuclear materials (SNM) to 1400 grams of contained U-235 enriched equal to or less than 20%.

This amount of SNM places MSU in the " low" category for strategic nuclear materials as defined by 20 CFR 73.67. An upgraded physical security plan designed-to meet the requirements of. this new class-ification-proposes an implementation date predicated on the time of I

actual receipt of the amount of SNM.that increases the SNM on site to the quantity of. Low Strategic Significance-defined by 10 CFR 73. This

-plan was submitted toithe_USNRC on July 23, 1980 and is being evaluated.

Due to unanticipated circumstances, modifications pursuant to Amendment Ntn '1 of the' facility license- have not been started and the reactor has not been operated at power levels greater than 60 milli-t watts thermal as of the date of this report.

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F. CHANGES TO PROCEDURES Aside from the license amendment described in Secticn E of :nis report, no char,ges to facility procedures were irpiecented during 1930.

G. fiEW OP. U'iTRIED EXPERI!ENTS Nor.e H. RADI0 ACTIVE EFFLUENT 5

1. Liquid: None

2. Airborne: None

3. Solid: None I. ENVIRONMENTAL RADIOLOGICAL SURVEYS PEFIOPJED OlJTSIDE THE FACILIT(

Areas of unrestricted access begin at the outside walls of the Reactor Room. A general area radiation survey conducted December 30, 1930,

-revealed the maximum detectable level of direct radiation to be 0.22 =R/hr (gamma) teasured upon contact with the outside east wall.

The maximum level of neutron radiation reasured 0.1 crea/hr (minimum sensitivity of the instrument) 'at the same location. The reactor was teing operated at a steady power level of 60 milliwatts' (caximum obtainable with existing instrumentation) for the duration of this survey.

Random wipes /scears of surfaces both inside and outside the AGi-201 Facility (nor= ally. performed on a routine weekly basis throughout the year) ~ have not detected any loose surface activity above natural background.

The reactor was not operated above 60 cilliwatts during 1980,

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therefore, no full power radiation survey results are available.

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J. RADIATION EXPOSURES GREATER THAN 100 MILLIREM (50 MREM FOR PERSONS UNDER 13 YEARS OF AGE)

None

- K. PERSONNEL EXPOSURE AND' MONITORING 10 CFR 20, Part 407, (a)(2) and Part 407(b)

1. Personnel monitoring'was provided for a total of 66 persons during 1980.

2. Statistical . Summary: .

-Estimated Whole~ Body Number of Individuals Exposure (Rems) in Each Range No measureable exposure. . . . . . . . . . . . . . . 43 Measureable-exposure less than 0.1 . . . . . . . . . 23

'0.1 to 0.25 . . .................. 0 0.25 to-0.5 . . .................. 0 0.5'to 0.75 . . ..................

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0 0.75 to 1. . . . . . . . . . . . . . . . . . . . . . 0 1 to 2 . . . . . . . . . . . . . . . . . . . . . . . 0 2 to 3 . . . . . . . . . . . . . . . . . . . . . . . 0 3 to 4 . . . . . . . . . . . . . . . . . . . . . . . 0

'4 to 5 . . . . . . . .-. . . . . . . . . . . . . . . 0 5.to 6 . . . . . . . . . . . . . . . . . . . . . . . O i 6 to 7 . . . . . . . . . . . . . . . . . . . . . . . 0 7 to 8 . . . . . . . . . . . . . . . . . . . . . . . 0

,8 to 9L. . . . ... . . . . . . . . . . . . . . . . . 0-

-9 to 10 . . . . ....:.............. 0

-10~to.11.. . . . . . . . . . . . . . . . . . . . . . 0-11.to 12 . . . . . . . . ... . . . . . . . . . . . . 0 12+. . . . . . . . .. .................

O L .' AUDITS AND INSPECTIONS 1 An audit of the performance, training, and qualifications of the facility staff was performed by. the. Reactor Safety Committee on.

September 18,11980. .The performance, training, and qualifications werelfound . acceptable. :

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APrENDIX A AMEN 0 MENT NO. I to FACILITY OPERATING LICENSE R-127 i P00R ORIGINAL i

i APPENDIX A (21 pps)

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/en sec,,,#'o UfJITED STATES 8 D .,., ",{ NUCLEAR REGULATORY COMMISSION e-

,j j ya WASHINGTON. D. C. 20555

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Docket tio. 50-538 <[ 4

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l_ Mr. D. W. Jones g '

Director, Center for fluclear Studies cp Memphis State University Memphis, Tennessee 38152 @#

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Dear Mr. Jones:

The Comission has completed the review of your application dated March 23, 1979 and supplements dated August 3 and 28,~1979, and has issued the

. enclosed Amendment flo.1.

This amendment pertains to your AGil-type reactor redesignated Model AGil-20lH.

It authorizes steady state power not to exceed 20 watts thermal and inter-mittent operation not to exceed 1000 watts thermal. The amendment further incorporates revised Technical Specifications for the authorized power levels.

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This amendment also authorf zes Memphis State University (MSU) to increase possession _ limits of special nuclear materials (Stim) in connection with reactor operation to 1400' grams of contained U-235 enriched equal to or less

(_ than 207.. This amount of S!!M places the MSU in' the " low" or " moderate"

. categories for strategic nuclear materials, as defined by 10 CFR 73.67 (formerly 73.47). -

In'accordance with your request to ship and receive 700 grams of AGil-201 fuel from Oak Ridge flational Labs,10.CFR 70 and 71 already provide the necessary authority as long as MSU maintains the procedures delineated in the license application.

This new classification requires that MSU submit an upgraded Physical Security

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-Plan prior to August 22, 1980. Background information on physical protaccion

.for flon-power Reactor Licensees was mailed to you in a packet dated January 7, 1980.

In or' der,to operate your reactor'above 0.1 watt, all modifications described in-your application and_ supplemental. information shall be comoieted anc the

~ Commission so . notified by letter.

Copies of .the Safety- Evaluation / Environmental. Impact Appraisal and the flotice/

flegative -Declaration are also enclosed.

Sincerely,

.&V Robert 1W. Reid, Chief Operating Reactors Branch 14 Division of Operating ~ Reactors-

Enclosures and cc

See next page

Mr. D. W. Jones s

Enclosures:

1. Amendment !;o. 1

2. Safety Evaluation / Environmental Icpact Appraisai

3. Notice / Negative Declaration cc w/ enclosures & inccming dated:

3/23/79, 8/3 & 28/79 Director, Office of Urban & Federal Affairs 103 Parkway Towers 404 James Robertson Way flashville, Tennessee 37219 Mayor of the City of Menphis -

City Hall - Room 700 125 North Main Memphis, Tennessee 38103 O

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UNITED STATES >

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'% ; / MEMPHIS STATE UNIVERSITY DOCKET N0. 50-538 AMEt!DMENT TO FACILITY OPERATING LICENSE Amendment Noc 1 License Nc. R-127

1. The Nuclear Regulatory Commission (the Comission) has found that:

A. The application for amendment by Memphis State University (the licensee) dated March 23, 1979, as supplemented-August 3, 1979, and August 28, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regu-lations set -forth in 10 CFR Chapter I; B. The facility will operate in canformity with the application, the pro-visions of the Act, and the rules and regulations of the Comission;

.C.

There is reasonable assurance _(1) that the activities authorized by this amendment can be conducted without endangering the health and

( safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will r.ot be inimical to the cormon defense and -security or to the health and safety of the public; and E. The issuance of this ~ amendment' is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications

-as indicated in the attachment to this license amendment, and paragrapns 2 2.A, 2.B.(2),- 2.8.(3), 2.C.(1) and 2.C.(2) of Facility Operating License No.

R-127 are hereby revised as follows:

2.A This license applies to the Model AGN-201H, Serial No.108, nuclear research reactor and associated equipment (the facility) owned by Memphis State University. The facility is located on its campus in Memphis, Tennessee, and described in the licensee's application for construction permit and operating license dated April 11, 1975 and amendments thereto.

'2.B.(2) Delete 2.B.(3)-Pursuant to the Act and ~10 CFR Part 70 "Special Nuclear Material,"

f-: to receive,-possess, and use up to 1400 grams of contained uranium

(.- 235 enriched equal to or less than 20% in connection with operation of facility, and

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2.C.(1) 11aximum Power Level The licensee is authorized to operate the reactor at s::eady stateO power levels not in excess of 20 watts (thermal).

mittent power levels not in excess of 1000 watts (thermal) for any seven consecutive day period and not in excess of 3.36 kilowatt-hours is also authorized.

2.C.(2) Technical Soecifications

, -The Technical Specifications contained in Appendix A, as revised through Amendment flo. 1, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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"3. -This' license amendment is effective as of the date of its issuance.

(_ FOR THE tiUCLEAR REGULATORY C0KiISSIOil W. P.E &mill, . ting Assistant Director for Operating Reaccor Projects Division of Operating Reactors Attachsents

Changes.to the' Technical -

-_ Specifications

.Da'te of Issuance: March 28, 1980 s

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l' ATTACHMErlT TO LICErlSE AMEllDMENT NO.1

l ' FACILITY OPERATING LICEllSE fi0. R-127 DOCKET NO. 50-538

> Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

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6 SAFETY LIMITS AND LIMITING SAFETY SYSTEtt SETTINGS 2.0 2.1 Safety Limits Apolicability These specifications apply to the maximum core temperature and minimum shield water tem;erature and level during steady state or transient operation.

10biective To assure that the integrity of the fuel material is caintained and that essentially all fission fragments are retained in the core matrix.

2.1.1 Soecification I The maximum core temperature shall not exceed 200*C during either steady state or transient operation.'

Bases The polyethylene core material does not melt below 200*C and therefore assures integrity of the core and retention of essentially all fission fragments at temperatures below 200*C.

l 2.1.2 Speci_fication_

' (, The reactor shield tank water temperature shall be maintained above 10*C, and the water level in the tank shall not be mora than 12 inches below the top of the reactor shield tank.

Sases

. Low reactor shield tank water temperature may result in freezing of the water. The result of expansion due to freezing of the water may This condition damage the shield tank and other reactor components. A safety would degrade core coatainment and shielding capability. .

limit of 10 C provides a margin for confidence that the reactor will

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not be operated with frozen shield water.

The shield tank water level of 12 inches below the top of the tank provides an adeouate medium for continuous neutron flux monitoring-during reactor operation and ensurea adequacy of the facility secon-dary radiation shield for operations greater than 100 milliwatts.

1 Amendment-No.'l b.

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2.2 Linitine Safety System Settincs

(}' Acclicability Tnis specification applies to the parts of the reactor safety '

. syste which will limit maxirun ccre temperature.

Objectiy_e To assure that automatic protective action is initiated to prevent a safety limit frc= being exceeded during steady state or transient operation.  ;

Scecification i The core thermal fuse shall r.elt when heated to a temperature of 120*C or less resulting in core separation and a reactivity loss of greater than 5% ak. , ,

Bases I

In the event of failure of the reactor to scram, the self-limiting characteristics due to the high negative terperature coefficient, and the celting of the thereal fuse at a terperature below 120*C will I assure safe shutdown without ex'ceedir.g a core temperature of 200'C.

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% dndment No.'1 ~ .,4 _

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3.0 LIMITING CONDITIONS __ FOR OPERATIO?!

3.1 Reactivity Limits Apolicability This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.

Obiective To assure that the reactor c tn be shut down at all times and that

. the safety limits will not b. - exceeded.

Specifica_ tion

a. The available excess reactivity with all control and safety rods fully inserted and including the potential reactivity worth of all experiments shall not exceed 0.65% ak/k. l

b. The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 2% ak/k, referenced to 20*C. l

c. The reactivity worth of the control and safety rods shall ensure subcriticality on the withdrawal of the coarse control rod or any one safety rod.

Bases.

(c _ The limitations on total core excess reactivity assure reactor periods of sufficient. length so that the' reactor protection system and/or operator action will be able to shut the reactor down without exceed-

• ing any. safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive position.

3.2 . Control and Sa fetv Systems Acolicability These specifications apply to tha reactor control and safety systems.

s Amendment:No. 1- -

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Obiective To specify lowest acceptable level of performance, instrument set

- points, and the minimum number of operable components for the reactor control and safety systems.

Specification The reactor shall not be made critical unless tne follcwing speci-fications are met: .

a. The total scram withdrawal time of the safet', rods and coarse control rod shall be less than 200 milliseconds.

b. The maximum reactivity addition rate far each rod shall not exceed 0.04% Ak/k/sec..

c. The safety rods and coarse control rod shall be interlocked such that: ,

1. Reactor startup cannot commence 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 reactor to scram.

The reactor source level measured by the Nuclear Safety Channel

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No.1 countrate instru=ent is more than 120 CFM.

f. The reactor core tank is pressurized with dry nitrogen to at least 1 psig.

g. All reactor safety system instrumentation shall be operable in accordance-with Table 3.1 with the following allowable exception:

Nuclear Safety Channe1~ No.1 may be bypassed provided Nuclear Safety Channel Hos. 2 and 3 are verified to be operable.

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T TABLE 3.1 i

Set P_oint Function .

Safety Channel _

fluclear Safety #1 2.12% full scale Scram at source levels

<1E% of full scale Low Countrate Scram at power > 2000 watts fluclear Safety lC (Log) 1 2000 watts High Power

> 5 see Scram at periods < 5 sec Reactor Period fluclear Safety #3 (Linear)

• Scram at >95% of full scale High Power < 95% full scal'e

> 5% full scale Scram at source levels Low Power ~

<5% of full scale

> 15"C Scram at temperature < 15*C Shield Water Temperature

< 10.5 inches Scram at water levels >10.5

- Shield Water Level ~

inches below top of shield water tank Scram at displacements Seismic Displacement < 1/16"

> 1/16" 1 5 psig Scram at core tank pressure Core Tank Pressure > 5 psig Scram at operator option Manual Scram

-- . Alarm at or below level set

- Radiation Monitor _ to meet. requirements of 10 CFR Part 20 Alarm' at or below level set Air Particulate --

to meet requirements of TO CFR Monitor Part 20 l Ac.endment- flo. I ' -

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I Bases

[ The specifications on scram reactivity rate in conjunction with the safety system instrumentation 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 ample time will be available for manual or automatic scram durina any operating conditions. Interlocks on control and safety rods assure an orderly approach to criticality and an adequate shutdown capability. .

The minimum 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.

Pressurizing the core tank with dry nitrogen to at least 1 psig assures that evolved hydrogen from high power operations cannot aggicmerate into hazardous concentrations. ~

The neutron detector channels (Nuclear Safety Channels #1 through #3) assure that reactor power levels are adequately monitored d" ring reactor startup and operation. Requirements on minimum neutron levels will prevent reactor startup unless the startup channels (Nuclear Safety Channels !1 and #3) are cperable and responding, and will cause a scram in the event of instrumenta-tion failure. .

In order to provide assurance that at least two Nuclear Safety Channels are operative for all ranges of reactor operation and to prevent overranging the channel 1 startup instrument, Nuclear Safety Channel #1 is allowed to

{ be bypassed for operations above 40 milliwatts only if the remaining two channels are verified to be operable.

Since the AGN-201 core negative temperature coefficient of reactivity in conjunction with the maximum potential excess reactivity specified in 3.1 prevents reactor operation at high power levels for time intervals necessary to approach established safety limits or limiting safety system settings, the high power level scrams are established to provide redundant automatic pro-tective 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 the rate of rise of reactor- power to periods which are manually controllable and will automatically scram the reactor in the event of large reactivity additions.

- The AGN-201's negative temperature coefficient of reactivity causes a reac-tivity increase with decreasing core temperature. The shield watar tempera-ture safety channel will prevent reactor operation at temperatures below 15'C therecy limiting potential reactivity additions associated with tempera ure decreases.

Water in the shield tank is an important component of the reactor shield and operation without the water may produce excessive radiation levels and inade-quate neutron flux monitoring capabilities. The shield tank water level safety channel will prevent reactor operation without adequate water levels

, in the shield tank. '

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Amendment No. 1 .

The reactor is designed to withstand 0.69 accelerations and 6 cm dis-placements. A seismic instrument causes a reactor scram whenever the

( instrument receives a horizontal acceleration that causes a horizontal 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 seismic disturbance that may cause damage to the reactor or its components.

The manual scram allows the operator to manually shut down the reactor if an unsafe or otherwise abnormal condition occurs that does not other-wise scrmn the reactor. A loss of electrical power de-energizes the safety and coarse control rod holding magnets causing a reactor scram thus assuring safe and immediate shutdown in case of a power outage.

A radiation monitor must always be .<allable to operating personnel to provide an indication of any abnormally high radiation levels and an air particulate monitor must be available to warn operating personnel of a degradation in core tank or gas monitoring system integrity so that appro-priate action can be taken to shut the reactor down and assess the hazards to personnel.

3.3 Limitations on Experiments Acolicability This specification applies to experiments installed in the reactor and its experimental facilities.

Objective To prevent damage to the reactor or excessive release of radioactive materials

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in the event of an experimental failure.

Soecification

a. Experiments containing materials corrosive to reactor components or which contain liquid or gaseous, fissionable materials shall be. double encapsulated.

b. Explosive materials shall not be inserted into experimental facilities of the reactor.

c. The radioactive material content, including fission products of any experi-ment shall be limited so that the complete release of all gaseous, parti-culate, or volatile components from the experiment will not result in doses-in excess of 10% of the equivalent annual doses stated in 10 CFR Part 20 for persons occupying (1) unrestricted areas continuously for two hours starting at time of release or (2) restricted areas during the length of time required to evacuate the restricted area.

d. The radioactive material' content, including fission products of any doubly encapsulated experiment shall be limited so that the complete release of all gaseous, particulate, or volatile components of the experiment shall J( Amendment No.;1l _g. -

• .%,,---w.q. g ny, %, _.9 g

not result in exposures in excess of 0.5 Rem whole body or 1.5 Rem thyroid to persons occupying an unrestricted area continuously for a period of two hours starting at the time of release or exposure in excess of 5 Ren whole body or 30 Rem thyroid to persons occupying a restricted area during the length of time required to evacuate the restricted area.

Bases These specifications are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting frem an experi-ment failure and to protect operating personnel and the public from excessive radiation doses in the event of an experiment failure.

3.4 Shielding .

Applicability

.These specifications apply to reactor shielding required during reactnr operation.

Obiactive The objective is to protect facility personnel and the public from

.{ 3.4.1 radiation exposure.

Specification The following shielding requirements shall be fulfilled prior to reactor startup and during reactor operation:

a. The reactor shield tank shall be filled with water to a height within 10 inches of the highest point on the manhole opening. .

b. The thermal column shall be filled with water or graphite. Access to the reactor building roof area above the reactor shall be restricted during reactor operation.

c. The faci'.ity secondary shield shall be in place with removable shield plugt installed.

3.4.2 Specification Access to'the reactor room shall be prohibited, except for radiation surveys, during operations above 20 watts.

Bases The "acility shielding in conjunction with designated restricted radiation f

. areas .is designed to limit radiation doses to facility personnel and to the

( public to a level below 10 CFR 20 limits under operating conditions, and to a level below criterion 19, Appendix A,10 CFR 50 reconnendations under acci-

. dent conditions.

-Amendment No. I .

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5.0 DESIGil FEATURSS

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5.1 Reactor _

a. The reactor care, including control and safety rods, contains approximately 560 grams of U-235 in the form of 20% enriched UOTneIbwer dispersed in approximately 11 kilograms of polyethylene.

section of the core is supported by an aluminun rod and a thermal fuse. The fuse melts at temperatures below 120 C causing the lower core section to fall away from the upper saction reducing reactivity by at least 5% ak/k. Sufficient clearance between core and reflec-tors is penvided to ensure free fall of the bottom half of the core l during the most scvere transient.

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b. The core is surrounded by a 20 cm thick high density (1.75 gm/cm )

graphite reflector followed by a 10 cm thick lead gamma 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 might leak from the core. A valved gas handling system is pemanently con-nected to the core tank assembly to pemit monitoring and disposal of gases which may accumulate in the tank from high power operations.

c. The core, reflector, and lead shielding are enclosed in and supported by a fluid-tight steel reactor tank. An upper or "themal column tank" may serve .as a chield tank when filled with water or a thermal column when filled with graphite.

d. The 61/2 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 water. A 44 inch thick concrete block shield supporting a top cover that con-tains approximately 18 inch thick borated paraffin encloses the 61/2 foot diameter shield tank to provide a secondary neutron and gamma shield for operations above 100 milliwatts. The complete reaccor ,

shield shall limit doses to operating personnel in restricted and unrestricted areas to levels less than permitted in 10 CTR 20 under l operating conditions.

e. Two safety rods and one control rod (identical in size) contain up to 20 grams of U-235 each in the same form as the core materi,1. These rods are lifted into the core by electromagnets, driven by reversible DC motors through lead screw ascemblies. De-energizing the magnets causes a spring-driven, gravity-as! 'sted 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.

5.2 Fuel Storace

- Fuel, including fueled experiments and fuel devices, not in the reactor shall- be stored' in locked rooms in the reactor building. The storage array shall.be such that K is no greater than 0.8 for all conditions of moder-ation and reflectioneff Amendment flo. l' ' '

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8 .w..% UNITED STATES

• , f g,s.,d gQ v g NUCLEAR REGULATORY COMMISSION WASHING TON, D. C. 20555

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SAFETY EVALUATIO*1 Af!D EtiVIRONMENTAL IMPACT APPPAISAL BY THE

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OFFICE OF_ NUCLEAR REACTOR P.EGULATICA SUPPORTING A".END"E:!T NO.1 TO FACILITY OPERATIf;G LICE?iSE NO. _ R-127 MEMPHIS STATE UtilVERSITY DOCKET NO. 50-738,

'ii[$0D0CTI0ft By letter dated March 23, 1979, as supplemented August 3 and 28,1979, Memphis State University (MSU or the licensee) requested that Facility Operating License tio. R-127 be amended to permit:

1. Operation of the Model AGil-201, Serial 108 tiuclear Reactor at continuous power levels up to and including 20 Watts (thermal) and inter-mittently at power levels up to and including 1000 Watts (thermal),

2. Transportation of up to 700 grams of contained U-235 between Oak Ridae flat-ional Laboratories, Oak Ridge, TN and the MSU, South Campus, Memphis, TN as supplemental fuel loading for the reactor, and ,

Receiving and possession of up to 1400 grams of contained U-235 in connection

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with operation of the facility.

Operation at these power levels would require modification of the reactor instrumentation and control system, installation of additional chf elding, installation of a gas handling system, and additional physical and admin-istrative controls.

Upon completion of the modifications for steady state 20 Watt operation and intermittent 1000 Watt operation, the reactor would be designated Model AGil-20lH, Serial tio. 108.

9 DISCtSS U1 The AGN-201 reactor is a portable, self-contained reactor using a homogenous fuel mixture of ' uranium oxide and polyethylene enriched with U-235. The reactor is. presently authori::ed to operate at a maximum power level of 100 mil 1iwatts-(thermal).

Tne increased power level is needed for the MSU scheduled research prngrams that require higher neutron flux levels than current power levels can attain. This will, in general, significantly enhance the overall educaticnal

- and research capabilities of- the- AGri-201 reactor. , _ _ , ,

Oak Ridge National Laboratories (ORill) is currently storing 700 grams of-(- contained U-235 which was removed from a previously scrapped AGN-201. This

. fuel can serve as replacement fuel for the MSU reactor. MSU will supervise the shipment of the fuel. from DRNL which is proposed to be accomplished in

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two separate shipments. This fuel wi M -Mred onsite in the shipping centainers which will.be locked in a secure area protected by locked doors, r ~ an electronic burglar alarm, and a patrol security force.

Modi fi cations _ .

A. Reactor Instrumentation Control Syste s ,

Instrumentation and circuitry will be modified to provide required operational

.centrol and to meet safety considerations. The three nuclear safety systems will ,

be modified with new instruments ano ion chameers; a nign core tank pressure scran circuit and instruments wili be provided; appropriate core te perature in- 1 struments will be installed.

B. Shielding Additional shielding to that provided by the water-filled reactor tank is required.

It will consist of two concentric cylinders, one cylinder of barated concrete and the other of ordinary concrete block. They'will be arranged so that the seams will overlap to prevent radiatien streaming and to provide a totalwall thickness

'of_44 inches. The cylindrical shield will have a top support and shield assembly consisting ef.18 inches of barated paraffin.

C. _ Monitoring-Gas Handling System A gas . handling connection will be installed to interface with a new gas handling systen that will- penetrate the core tank te permit monitoring of gas and gaseous products.in the core void areas.

Administrative' Controls

- Areas exposed to higher levels of radioactivity will be posted as exclusion or . .

restricted creas during- reactor operation. Areas inside the building will be controlled

, by facility precedures; areas outside the building that may experience radiation during operation _will be restricted by.use of a security fence and gate.

I. SAFETY: EVALUATIO:1 The modified MSU AG?i reactor.is- similar in design and operation to the U. S.

fiaval Post-Graduate. School (UStiPGS) reactor which ' operated at a steacy state pcwer of 20 watts and intermittent power level of 1000 watts. The UStiPGS reactor was an AGil-201 Serial 100' reactor (Docket tio. 50-43). -Modifications .similar to

those proposed by MSU were evaluated _for .the USilPGS reactor and were approved byLthe'Ccemission on June 15, 1961. The reactor was operated at 20 watts

. steady state and 1000 watts -intermittent for approximately eight:(8) years with no apparent. fuel deterioration and no -danger to the, public health and safety.

The- reactor was' then transferreu to_CaliforniaLState Polytechnic' University at

' San. Luis .0bispo,. California. . Following : receipt of its license in 1973 (Occket -

_ tioL S0-394), the1(USilPGS). reactor has operated satisfactorily.

Inf addition, AG?i-201, Serial 108 reactor design feature characteristics and oper-

'ating conaitions had:been previously evaluated in support of Facility Operating 1 License No._R.-127fissued December 10 1976.

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[ Core N ificaitens:

The1:nodifications'to the. core tank ~ and shield water tank assembly.and:the oper -

..ating procedures; assure 1that the reactor _will be operated with the maximum degree -

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of core tank integrity and safety similar to that previously evaluated for the AGN-201 reactor design (Decket No. 50-538, December 10,1976).

Instrumentation modifications will increase the monitoring and associated reactor safety component response capabilities over those included in the previously approved design (Docket No. 5G-533).

Radiation Calculations _

The licensee provided the following evaluation with regard to the proposed shielding calculations. We have reviewed these calculations and concur with them.

a. Continuous Operation at 20 Watts Calculations for 44" concrete shielding around the AGN-201 reactor indicate that dose rates from gama radiation at the outer surface of the concrete shield will be c 0.8 mR/hr. Neutron radiation will not be transmitted through the shield. Calculations for the surface of the top shield consisting of 18" of borated paraffin show that the maximum dose rates would be <68 mR/hr gama and

<0.2 mrem /hr fast neutrons for a water-filled thermal column, and <612 mR/hr At gamma and <6 mrem /hr fast neutrons for a graphite-filled thermal cFlumn.

10 feet above the shield (height of roof), these dose rates decrease to 7 mR/hr gamma and <0.1 mrem /hr fast neutrons (water filled), and 63 mR/hr gamma and <0.6 mrem /hr fast neutrons (graphite filled). Thus, the radiation dose rates outside the reactor room will only exceed limits specified for unre-( stricted access (10 CFR 20) on the roof directly above the . reactor. Existing procedures prohibit access to this area during reactor operation.

b. Intermittent Operat_ ion at Power Levels Greater than 20 Watts Calculations for the proposed shielding at 1000 watt operation indicate that dose rates at the surface of the concrete shield would be <38 mR/hr gamma and <.06 mrem /hr neutrons. The borated paraffir. top shield will limit dose rates to <3.4 R/hr gamma and <9 mrem /hr neutron for a water-filled thermal column, and <31 R/hr gamma ans <270 mrem /hr neutrors for a g-aphite filled thermal coluEn. These values would decrease to <340 n.R/hr gama and <1 mrem /hr neutrons (water filled thermal column) or <3.1 RThe gamma and <27 mreiii/hr

-neutrons (graphite filled thermal column) at a height of 10 feift above the reactor (heightof. roof).

Assuming the reactor could be operated at 1000_ watts for 15 minutes (approx-imately 50% longer than expected), the highest dose available would be <8 rem on top of the polyshield, 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 graphite filled thermal column.

This ' dose would be <780 mR gan=a and <7 mrem neutrons. The dose at the surface of.the ' concrete shield would be <10 mR gamma and <.02 mrem neutrons. Since not more than ene high power operation could be conducted within a one-hour time interva?, and since access to the roof is prohibited during all reactor

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[' operations, the administrative controls governing access to posted restr tec areas ex osu m ,

Fuei Intecrity The fuel consists of polyethylene material with uranium dicxide (enriched poly-to 19.95 in U-235) unifomly dispersed throughout the polyethylene, ethy ene is an organic material that can sustain radiation damage when ex:osed to fissicn product bcebardment. Test data was provided by Aerofet- Nati General Nucleonics of samples of core caterial exposed in the ArgonPd n/crgnal The CP-5 reacter is a 5 cegawatt (flux-10 I.aboratory LP-5 reactor. Tests included exposures at full power fcr pericds up to one week sec) reactor.

continuous operation. Analyses of these tests revealed that radiation damage was evident in a reduced density and there was some loss of hydrogen frca the polyethylene. An extrapolation of these results, assuming that the inte: rated flux-time (nyt) is responsible for the damage, for centinuous operation a~t At 100 watts equates to a core life of six years prior to any damage cccurring.

20 watts continucus operation the core life would be approximately 30 years.

As the normal operating cycla is less than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, orless

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than Intermittent 24% of the total time, the projected life approaches 125 yea"0 n/cm# sec for 1

operation at 1000 watts (thermal) at flux levels of 4.2 x 10 short periods of time (less than 20 minutes) will not greatly affect core life. From this analysis it is reasonable to conclude 3 that +he AGN-201 coren/c8 sec) operating 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week at 20 watts (flux - 9 x 10 mittent operation at 1000 watts would sustain no radiation damage over the remaining 16 years of reactor operaticn authorized by MSU's license.

Fuel Storape Three hundred fif ty grams of the fuel will be stored in each of two Depart entEacn c of Transportation (DOT) specified 6J type centainers. Calculations for AGN-201 fuel ind with vermiculite and sealed. Accordingly, it is not possible critical cass is greater than 650 grams of U-235. devised.

for critical mass to be achieved with the fuel containerizatica Desien Basis Accident (DEAL Ue have postulated a DBA for an AGM-201 operatinc at 20 watts to be the failure of the polyethylene moderator cladding 5f a single fuei disc in air. If all ele ents ruptured and all fission products were released, the activity w.uld be approxicately 200 curies, principally Iodine-131, whento This equates operatir.c.at 1000 watts and 600 millicuries at 20 watts.

6.9 x 10" uCi/ml if uniformly discersed throughout the reactor reca (10* liters).

The release from the postulated DBA wnen uniformly dispersed thrcughout the reacter rocm is 1.27 x 10 pCi/mi which is less than the Part 20 limiting cose for an unre-stricted area. Any release to the outside at:xsphere would be Cniy a scall Restric-fraction of this dose and therefore presents no hazard to the Dublic.,-~bEA ting 10C0 watt oceration to less than 20 minutes vill ensure that any occurring when at this power level will not significantly affect the fore-going assu ptions which are conservative by a factor of 100.

L Although the DEA is only rerotely possible, MSU is desianating the reactor roca a centrolled area; and should it occur, the additi5nal crecautions of cvacuating the reactor rocm will be acccmplished and emergency procedures implemented .

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Gas Handline Syste. (GHS1 h w 'iThe GHS modifications are described in detail in Section II B4 (p.9)

Ficure II-4 and Section IV A4 (p.29) of the acclication for amendment dated The proposed system and faci'iity operating procecures will

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x Maren 23,1979.

~hne a degree of performance and reliability equal to or greater than pre-vious'.y evaluated systems used in AGN-201 Serial 100 reactors.

As stated above, the core material will sustain no damage over the 16 years of operation remaining in the authorized MSU license. The GHS, hov:ever, is designed to identify radiation damage by being able to measure fission products that escape frr~m radiation damage to the fuel discs. MSU cen-cludes and we agree that the proposed GHS, the existine air particulate -

activity monitor and alarm system and associated procedures will ensure that  !

personnel protection recuirements of 10 CFR 20 can be adequately met.

1 CCNCLUSION ON SAFETY Moreover, due to the fact that: (1) no unusual problems have arisen with this reactor during over 20 years of authorized operation at 0.1 watts at both l Argonne National Laboratory and MSU, (2) the revised TS require surveillance and periodic testing of safety related equipment to i assure continued safe operation of the reactor at 20 watts and to assure that any significant component degradation will be detected in a timely manner, and (3) the USNPGS operated a similar AGN-201 reactor at 20 and 1000 watts without. evidence of any unusual problems over a period of eight (8) years, we have concluded that the mfd AGNd01 reactor can be r operated in a safe manner at 20 watts steady state and intermittently at

[' 1003 watts. Furthermore, based on the foregoing censiderations, we have

!L concluded that the estimated life of the facility will extend far beyond the end of the current license period. Therefore, from a reactor safety stancpoint, the a= enc.en is acceptaole . .

a'e have concluded, based on the considerations discussed above, that: (1)

ere i', reasonable assurance that the health and safety of the public will not be endangered by operation at 20 watts steady state and intemittently at 1000 watts in the proposed manner, and (2) such activities will be con-ducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the co=on defense and security or to the health and safety of the public.

t II. ENVIRONMENTAL IMPACT APPRAISAL .

' Inasmuch as the power level of this research reactor facility remains below 2 MWt, the environmental impact remains as stated in the license '

granted to MSU in 1976 (Docket No. 50-538). The 1976 Environmental Impact Appnisal: was based upon a generic evaluation for research.

reactors of 2 MWt or less which was approved January 28, 1974 -

Based on the foregoing analysis, we. have concluded that there will be no significant environmental impact attributed .to this proposed license power increase. Having made this conclusion, we have further ccncluded that~ no environmental impact statement for the proposcd action need be

{ -prepared and that a negative declaration to this effect is appropriate.

Dated: . March 28,-1980 -

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7590-01 U.':ITED STATES NUCLEAR REGULATORY CO."14ISSION

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DOCKET "0. 50-533 ME"PHIS STATE l'aI'!ERSITY N0TICE OF ISSUANCE CF AMEND.*;E';T TO FACILITY OPERATI"G LICENSE AND NEGATJVE DECLARATION The U. S. Nuclear Regulatory Commission (the Commission) has issued Jr.endment No. I to Facility Operatin' License No. R-127, issued to Memphis State Cniversity, which revised the license and Technical Specifications for operation cf the AG"-201 nuclear reactor (the facility) located on the licensee's campus in Me.phis, Tennessee. The amendment is effective as of its date of issuance.

The amendment authorizes: 1) an increase in the facility's licensed maximum pcwer level form 100 milliwatts (thermal) to 20 watts (thermal) for c:ntinucus operation and authcrizes intermittent operation at pcwer levels nc: inJexcess of .1000 watts (thermal); ar.2 2) an increase in the amount, fr:c 700 grams to 1400 grams, of contained U-235 enriched equal to or less thin 205 for use in connection with the facility. .

i The applicaticn for the amendment complies with the standards and requirements of _':he Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set fce:h in the license amendment.

Notice of-Proposed Issuance of Amendment to Facility Operating License in connection with this action was published in the FEDERAL REGISTER on September 7,

- I373 (~? FR 52339). :o requer: for a hearing or petition for laave to intervene was fi!ad foiicwing.natice of the proposed action.

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• 7590 01 .

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I he C .issica has orar: red Tn enrironrantal irpact a:.;raisal for this 1::'::n and has con:!uded that an environmental in::ac: state en: is no: .arrantad be:ause there will be no envirenrental incact attributable ta the action.

Fcr further details with respect to this action, see (1) the applica.ica for arer.dment dated March 23, 1979, as supplenented August 3,1979, and Auguc* 28, 1979, (2) k:endment ?b. I to License ?;o. E-127, f 3) the Ccmission's related Safety Evaluation /Envirenrental Icpact Apprais'al. All of these items are available for public inspecticn at the CO nissien's Public Document RO:m,1717 H Street, fl. W. ,

.'.'ashingt n, D. C. A c py of items (2) and (3) may be obtained upon request addresssd to the U. S. !!uclear Regulatcry Cor:-issicn, Washington, D. C. 20555, Attantion: Director, Division of Operating Reactors.

Dated at Bethesda, Maryland this 28th day of l'. arch 1980.

FOR THE ?iUCLEAR REGULATORY CO."MISSIO?!

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R bert 'd. Reid, Chief Ocerating Raactors Bran;h !4 Division of Operating Reacters

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