ML20024H256

Forwards Response to NRC 910501 Request for Addl Info on Proposed Tech Spec Changes for License R-84

ML20024H256
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	05/17/1991
From:	Irving G DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9105310052
Download: ML20024H256 (34)

Text

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ARME D f OHCl 8 HADion! OLOGY h[ S[ Af tCH INS 1tTU1L B[1Hf SDA M ARYL ANO 20t\14 5145 DIR May 17,1991 SilllJI!CT: Additional Infortnation on Projued Technical Specifications Changes for 1.icense No. R 84.

Document Control Desk l).S. Nuclear Regulatory Conuniwion Washington, D.C. 20555 Gentlernen:

please find enclosed nine (9) attachments that address the issues of your request for additional infortnation dated 1 May,1990. If you have any questions or conunents, please contact the Reactor I;acility Director, Mr. Mark Moore, or the Reactor lixecutive Officer,1st 1 t Matt 1;orsbacka, at (301) 2951290. Your prompt reply will be greatly appreciated.

Sincerely, .

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Colonel, LISAli, ilSC Director linclosures:

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hir. Al Adams USNRC, Mail Stop 11820 Washington, D.C. 20555 hir. Marvin hiendonca USNRC, Mail Stop llH2O Washington, D.C. 20555 Mr. Thomas Dragoun USNRC, Region i King of Prussia, PA 19406 1

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i Attachinent i

Comment i I In your amendment request you have added a new definition for Reactoi l'acility Director (RI D) and have amended Section 0.1.2 of the Technical Specifications ('i'S) to include additional

. information concerning the responsibilities of the Rl D The two 'IS Sections allected by you request repeat the same information to a large extent. Ilowever, the list of responsibilities for  !

the RI D dillers between the defmilion and Section 6.l.2 of the 'IS. Please consolidate your l requested changes concerning the RiiD to one place in the ~IS. We can find no Section 6.1..).l. l Please Correct. Section 6.1.2 of the 'IS also discusses brief absences of the Rl;D and his designee. What time interval do you mean by 'briel"7 Justify the time interval chosen.

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Reply to Comment i To consolidate the changes concerning the information pertaining to the RI D. please delete the proposed definition 1.17 IEACLQlLl%CIL11.LDilECLOR. The proposed Section 6.1.2 should be changed as follows:

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'ihe Director, Al RRI, shall have license resimmibility for the reactor f acility. The Reactor Facility Director (RFD) shall be responsible fer administration and operation of the Reactor l'acility and for determination of applicability of procedures, experiment authorizations, maintenance, and operationt The RI D may designate an individual who meets the requirements of Section 6.1.3.1.a to discharge the RFD's responsibilities in the j RFD's absence. During brief absences (periods less than four hours) of the Reactor Facility Director and his designee, the Reactor Operatiom Supervisor shall discharge these responsibilities.

The four hour time interval was chosen to allow the RI:D and his designee to be absent f or short periods of time to have lunch, run errands, or attend short meetings outside of AFRRI. ,

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Attachinent 2 Comment 2 Amendment No. 20 iwued on October 4,19'X) amended Section 4.2 5 of the 'lS 'Ihis Section new dif fers f rom your request of April 30. I'FM). Please provide a resised Section 4.2.5 to reflect the addition of fuel tollowed control rods (1 i CR4 to the teactor. Also, we note that on the second line of the basis of the TS, we intioduced a ty pographical citor by spelling cladding,

" clawing" l' lease coricet this error in yout request.

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Reply to Comment 2 The request of April 30, 1990 included information penair. g to the 1 FCRs. The original i request is reproduced here for your convenience: l l

Page 22. Section 4.2.5: Replace the Specifications in its entirety as follows to clarify the l requirement for fuel element surveillance and to accommodate the fuel follower control rals:

All the fuel elements present in the reactor core, to include fuel follower control rmis, shall be insrcted for damage or deterioration, and measured for length and how at intervals separated by not more than 500 pulses of insertion greater than $2,00 or annually (not to exceed 15 months), whichever occur first. Fuel elements in long-term storage need not be measured until returned to core; however fuel elements routinely  :

moved to temporary storage shall be measured every 500 pulses ofinsertion greater than

$2.00 or annually (not to exceed 15 nianths), whichever occurs first.

Safety Analysis: Fuel it, long-term storage is not subject to the rigors of the fuel used in the reactor core rhus damage to the fuel which may manifest itself as elongation or lateral bow is not a possibility for fuel in long-term storage. The high degree of purity maintained in the AFRRI TRIGA pool assures that deterioration or corrosion of the cladding will not be a problem.

Page 22.3ccliolL1D: Replace the flasis in its entirety for clarification as follows:

The frequency of inspection and measurement is based on the parameters most likely to affect the fuel cladding of a pulse reactor, and the utilintion of fuel elements whose characteristics are well known.

The limit of transverse bend has been shown to result in no dif ficulty in disassembling the core. Analysis of a worst case scenario in which two adjacent fuel elements suffer sufficiently severe transverse bends to result in the touching of the fuel elements has shown that no damage to the fuel elements will result via a hot spot or any other known rnechanism.

Safety Analysis: This is an administrative change in the wording to clarify the passage. There are no safety implications.

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i Attachment 3 Comment 3 You have requested an amendment to Section 4.4 of the 'IS concerning verification of operation of the ventilation systems. You give as justification for the change conformance with ANS Standard 15.1. liowever, the wording you have chosen does not match ANS 15.1 l'or monthly inspections, thelatenal shall not exceed six weeks. Please amend you wording to conform with ANS 15.1.

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Reply to Comment 3 l> lease replace the Specifications of Section 4,4 in its entirety as follows to confoim with the

' latitude recommended in ANSI 15.1.4.f; The operating mechanism of the tusitive scaling dampers in the reactor room ventilation systeri shall be verified to be operable and visually inspected at least monthly (interval not to exceed six weeks),

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! Attachment 4 i Cornment 4 I

l' You have requested to add information to the basis of Section 5.2.1 of the 'lS to justify addition of the Fl:CRs to the reactor. In light of the additional information you have provided in

! response to requests for additional information from NRC concerning the 1:1 CRs, please confirm .

I the changes to this Section requested on April 4,1990 are still valid.

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Reply to Cornment 4 Please replace the second paragraph los the basis of Section $.2.1 ol' the 'lS with the following:

The local power density of a 12.0 weight percent tuel follower is 21% greater than an 8.5 weight percent standard TRili A l'uel eleinent in the 1)-Ring. The volume ot' luel in a f uel f ollowed ital is 56% of the voluine of a standard TRiliA f uel clernent. 'Iherchite, the actuall ower prisluced in the f uel f ollowed rod is .0% less than the [mwer prisluced in a standard 'l Rlli A f uel clernent in the D-Ring.

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Attach.1:ent 5  !

Comment 5 In your submittal of April 4,1990, you describe changes to Section 6.1.1 of the TS. liowever, there are additional changes to I;igure I that you do not describe or justify. I' lease explain and justify all of the changes mode on liigure 1.

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I Reply to Comment 5 All changes to I i:ure 1 in Section 6.1.1 are administratis e changes to reticct changes in esisting terminology. There are no safety implications. I' lease replace

• Chief, Radiation Sources Division" with " Chief, Reactor Division" and

• Reactor Stall

• with

• Reactor Operations Stall

• Attachment 6 Comment 6 You have requested amendment of Section 6.1.3.1.b of the 'IS concerning the Reactor Operations Supervisor (ROS). You discun the requirement for the ROS to have one year l experience before appointment as the ROS. What type of eywrience is this? 1.icensed as a l Senior Reactor Operation at Al:RRI? Please clarify and justify this change.

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I Reply to Comment 6 Replace the requirements of Reactor Operations Supervisor in its entirety with the following:

At the time of appointment to this gusition, the ROS shall hase 3 years nuclear experience. Iligher education in a science or nuclear engineering licld may fulfill up to 2 years of experience on a one-for-one basis. The ROS shall hold a USNRC Senior Reactor Operator license on the Al:RRI reactor. f u addition, tne ROS shall have 1 year of experience as a USNRC licensed Senior Reactor Operator at Al:RR1 or at a similar facility before the appointment to this position.

Attachment 7 Comment 7 la light of yor request to increau authorized reactor power to 1100 Kw, do you desire changes to section 3.1.1 of the TS?

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Reply to Comment 7 No. It is our stated intention to limit normal steady-state power operations to 1(XX) LW. The purpose of the request to increase the maxirnum steady-state power level allowed by the 1:acility Operating License is to resolve the conflict with TS 3.1.1 which allows a maximum steady-state power level of 1.1 MW for the purposes of testing and calibration.

Attachment 8 l Comment 8 in our Raluest for Additional Information dated November 14, 1990, we requested additional I information on your request to increase maximum steady-state power level. Your reply was limited to fil:CRs. Please provide a safety analysis that discusses operation of the reactor at i100 kW. I:or reference, we have enclosed a Safety Analysis for Oregon State University that  !

supponed an amendment raising their licensed maximum power to 1100 kW.

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i Reply to Comment 8 sal l!TY AN A1,YSIS Oli INCRl!ASl! 013 M AXIMUM Al.l OWi!D STl!ADY STATl! l'OWi!R LliVI!L l' ROM 1(XX) LW to ll(X) LW

Introduction 1

'Ihe purpose of this safety analysis is to resolve the inconsistency in the maximum allowed steady- state power level reported in the Technical Specifications and the I acility Operating I License No. R 84. Technical Specification (TS) 3.1.1 allows a maximum steady-state power level of 1100 kW for the purposes of testing end calibration. Amendment No.18 of the Facility Operating License, however, states that the maximum steady-state power level shall be 1000 kW.

As stated in TS 3.1.1, the normal steady state operating power limit of the reactor shall be 1000 kW, so it is our intent to simply amend the Facility Operating License to reflect the maximum power limit allowed in the Technical Specifications.

The maximum steady-state power level of i100 kW allows for the following processes:

When the AFRR1 TRIGA reactor is operating in the automatic mode, it is normal for the feedback elfects of a properly functioning servo system to result in small (1-2% )

power oscillations around the mean power. When the reactor is at 1.0 MW it is possible to briefly operate at Imwers slightly greater than 1.0 MW without violating the Facility Operating License.

Table 2 of the AFRRI Technical Specifications indicates that the maximum set point of the high flux safety channel may be set at 1.1 MW. Thus if the reactor is shut l down by a high power scram, there would be no violation of the Technical Specifications provided the power did not exceed 1.1 MW. Additionally, the provision to physically rather than electronically test the high flux safety channel exists.

As stated earlier, the intent of this license modification is not to allow fm prolonged steady state operations at power levels of 1100 kW. Any steady-state operations that exceed 1.0 MW will be brief in nature (tens of seconds or less), so there will be no significant change in the fission product inventories reported in the design basis accidents described in the Safety Analysis Report for AFRRI TRIGA Mark-F Reactor.

Safety Limitations The safety limit for a TRIGA reactor is the fuel element temperature. If the fuel temperature exceeds its safety limit, the disassociation of hydrogen and zirconium in the fuel moderator could cause a hydrogen overpressure condition between the fuel moderator and the cladding which could result in a loss of cladding integrity. The safety limit for the standard TRIGA fuel is based on data that includes a great deal of experimental evidence obtained during high-performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pressure from the disassociation of zirconium hydride will remain belew the ultimate stress, provided that the temperature of the fuel does not exceed 1000 C while immersed in water.

For steady-state operation, fuel temperatures are dependent up(m the heat transfer characteristics of the fuel element and coolant. As established in " Maximum Temperature Calculation and Operational Characteristics of Fuel Follower Control Rods for the AFRRI TRIGA Reactor

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Facility" (See correspondence to USNRC dated March 5,1991) the maximum powible power density in the AFRR1 TRIGA reactor at a steady-state power level of 1.1 MW is 72.4 W/cm'.

Equation 17 in the referenced report will allow us to determine the maximum fuel temperature for a fuel element in the centu of the core:

R R R R+ g T,' Tf

+ q"'R,'

(( j)2 -21n( *)- 1) + q"'R(((j)2-1)( In , c)4 ))

f i i t t e o

\\'here T, = matimum fuel temperature T, - bulk coolant temperature R, - 1.82 cm (radius of randard fuel element)

R, = 0.23 cm (radius of Zr icxl) c= 0.051 cm (cladding thickness) k, = 0.18 \\1cm *C (thermal conductivity of U Zr H) k, - 0.138117cm *C (thermal conductivity of SS3M) h = 1.339 ff7cm' *C (free convective heat trarufer coeDicient of water) q"' = 72.4 li7cm 3 Solving the above equation with a bulk water temperature of 48.6 C (the condition at which h was determined) yields a maximum fuel temperature of 425.1 C. It should be noted that the reported value is hypothetical as the central thimble in the AFRR1 TRIGA is occupied by a control rod rather than a fuel element. This does, however, represent the most conservative value of the maximum attainable temperature in any case of steady-state operation. As shown by this calculation, the maximum temperature of this conservative limiting case falls nearly 575 C short of the technical specifications limit of 1(XX) C. Furthermore, the limiting safety system setting of 600 C will remain exactly the same.

Operational Impact The proposed change to the Facility Operating License will have no impact on either pulse or steady-state operations. No modification of operating procedures, maintenance procedures, or technical specifications will be required if the proposed change is approved. In addition, there is no impact on the DBA as reported in the SAR as the DBA assumes 100 continuous hours of operation at the muimum allowed steady state power before a fuel element is mptured due to a mechanical impact. Because the 1.1 MW steady-state power limit is intended only for short durations of time, the saturated fission product inventory will be unchanged. Thus no changes

in the facility SAR or emergency plan or procedures will be required.

The basis for Technical Specilication 3.1.1 states: ' Thermal and hydraulic calculations and operation experience indicate that TRIGA fuel may be safely operated up to power levels of at least 1.5 megawatts with natural convective cooling." in addition, the AFRRI heat exchanger l is rated at 1.5 MW. Since it is our stated intention to limit routine steady-state operations to 1.0 I MW, we anticipate no changes in ecoling capabilities.

Conclusion There are many TRIGA reactor-years of experience which attest to the safety of this reacmr type at power levels measurably in excess of 1.1 MW. We would like to reference the NRC's Safety Evaluation Reports (SERs) for the University of Texas and Oregon State University TRIGA reactors which determined a 1.1 MW peak power was found to be completely safe and I

acceptable. In addition, the basis for the cunent Technical Specification 3.1.1 establishes that l

the maximum safe steady state power for a natural convection cooled TRIGA reator is 1.5 MW.

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Attachment 9 i

Proposed Replacement Pages for the Technical Specifications for the AFRRI Reactor Facility Armed Forces Radiobiolgy Research Institute Bethesda, MD 20889-5145 May,1991

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1- l TECilNICAL SPECIFICATIONS FOlt Tile  !

AFitill itEACTOlt FACILITY LICENSE NO. It-84

. _ . DOCKET #50-170 TAllLE OF CONTENTS Parc 1.0 DEFINITIONS 1.1 ALArtA i 1.2 Cb snnel Calibration 1 1.3 Caannel Check 1 1,4 Uhannel Test i 1.5 Cold Critical 1 1.0 Core Crid Position 1 1.7 Experiment .1 1.8 Experimental Facilities 1 1.9 Fuel Element 2 1.10 Instrumented Element 2 1.11 Limiting Safety System Setting ' 2 1.12 hicasured Value 2 1.13- hieasuring Channel 2 1.14 On Call 2 1.15 Operable : 2 1.16 Pulse hiode 3 1.17 Ileactor Operation 3 1.18 ' fleactne Safety Systems 3 1.19 Ileactor Secured 3 1.20 lleactor Shutdown 3 1.21 Iteportable Occurrence 3 1.22. Safety Channel 4 1.23 Safety Lim:t 4 1.24 Shutdown hiargin .

4 1.25 Standard Control Rod 4

.- l .26 . Steady State blode . 4 1.27 Transient Ilod 4 2.0 -SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2,1 Safety Limit - Fuel Element Temperature 5 2.2 Limiting Safety System Settings for Fuel Temperature 5 3.0 LIMITING CONDITIONS FOll OPERATIONS 3.1 - Reactor Core Parameters 7 3.1.1 Steady State - Operation - 7

3.1.2 Pulse Mode Operation- 7 3.1.3 Reactivity Limitations a 3.1.4 Scram Time - 8 i

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- c. -Ileactor Pool i

- d, - Core Experiment Tube

e. Portable Deam Tubes

f. Pneumatic Transfer System

g. Incore Locations 1.9 FUl'L El EMENT A fuel element is a single TitlG A fuel rod, or the fuel portion of a fuel follower control rod.

1.10 INSTRUMENTEli El EMENT An instrumented element is a special fuel element in which sheathed chromal/alumel or equivalent thermocouples are embedded in the fuel.

1.11 LIMITING S AFFTY SYSTEM SETTING limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions.

1.12 MEASURED VALUE A measured value is the magnitude of a variable an it appears on the output of a measuring channel.

1.13 MEASURING Cil ANNEb A measuring channel is that combination of sensor, interconnecting cables or lines, amplifiers, and output device that are connected for the purpose of measuring the value of a variable, 1.14 ON CALL A person is considered on call if

a. The individual has been specifically designated and the operator knows of the designation; .

b. The individual keeps the operator posted as to his/her whereabouts and telephone number; and

c. The individual is capable of getting to the reactor facility within 30 minutes under normal circumstances.

1.15 OPERABLE A system channel, device, or component shall be considered operable when it is capable of performing its intended functien(s) in a normal manner, 2

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1.16 PULSE MODE Operation in the pulse mode shall mean that the reactor is intentionally placed on a prompt critical excursion by making a step insertion of reactivity above critical with the transient rod, utilising the appropriate scrams in Table 2 and the appropriate interlocks in Table 3.

The reactor may be pulsed from a critical or suberitical state.

1.17 REACTOR OPER ATION Reactor operation is any condition wherein the reactor is not shut down, or any core maintenance is being performed, or there is movement of any control rod.

1.18- FACTOR SAFETY SYSTEMS Reactor safety systems are those systems, including their associated input circuits, that are designated- to initiate a reactor scram for the primary purpose of protecting the remetor or to l provide information that may require manual protective action to be initiated.

l 1.19 REACTOR SECURED The reactor is secured when all the following conditions are satisfied:

a. The reactor is shut down.

b The console key switch is in the "off' position, and the key is removed form the console and is under the control of a licensed operator, or is stored in a locked storage area. i l

c. No work is in progress inv.siving in-core fuel handling or refueling operations, l l maintenance of the reactor or its control mechanisms, or insertion .or withdrawal of in- l l- core experiments, unless sufficient reactivity is removed to insure a $0.50 (or greater) l shutdown margin with the most reactive control rod removed. J i

1.20 REACTOR SHUTDOWN l

The reactor is shut down when the reactor is suberitical by at least $0.50 of reactivity. '

1.21 REPORTARLE OCCURRENCE A reportable occurrence is any of the following that occurs during reactor operation:

a. Operation with any safety system setting less conservative than specified in Section 2.2, Limiting Safety System Settings.

b. ~ Operation in violation of any Limiting Condition for Operation, Section 3.

c. Malfunction of a required reactor or experiment safety system component that could l- render the system incapable of performing its intended safety function unless the

j. malfunction is discovered during tests.

d. Any unanticipated,or uncontrolled positive change in reactivity greater than $1.00.

e. An observed inadequacy in the implementation of either administrative or procedural l controls, so that- the inadequacy could have caused the existence or development of a l- condition that could result in operation of the reactor in a manner less safe than l

conditions covered in' the Safety Analysis Report (S AR).

l f. The release of fission products from a fuel element through degradation of the fuel cladding. Possible degradation may be determined through an increase in the background activity level of the reactor pool water.

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g. An unplanned or uncontrolled rel,ase of radioactivity that exceeds or could have eneceded the limits allowed by Title 10, l' art 20 of the Code of l'ederal llegulations (10 CFil 20), or these technical specifications.

1.22 SAFFTY Cil ANNIX A safety channel is a measuring channel in the reactor safety system that provides a reactor protective function.

1.23 S AITTY IIMIT Safety limits are limits on important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

1.24 SilliTDOWN t f AltGIN Shutdown margin shall mean the minimum shutdown reactivity considered necessary to provide ec. 9dence that the renetor can be made subcritical by means of the control and safety systems, starting from any permissible operating c,nditions, and that the reactor will remain suheritical without further operator action.

1.25 ST AND Altl) CONTitol it OD A standard control rod is a control rod having an electro-mechanical drive and scram capabi' is. It is withdrawn by an electromagnet / armature system.

1.26 STl' Al> r' ST ATT' MODI' Operation in the steady state mode shah mean the steady state operation of the reactor either by manual operation of the control rods or by automatic operation of one or more control rod (servocontrol) at power lesels not exceeding 1.1 megawatts, utilizing the appropriate scrarns in Table 2 and the appropriate interlocks in Table 3.

1.27 Til ANSIENT ItOD The transient rod is a control rod with scram capabilities that can be rapidly ejected from the reactor core to produce a pulse. It is activated by applying compressed air to a piston.

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Specification

- Functional checks . hall be made annually, but not to exceed 15 months, to insure the following:

a. With the lead shield doors open, neither exposure room plug door can be electrically opened.

b. The core dolly cannot be mmed into position 2 with the lead shield doors closed.

e. The warning horn shall sound in the exposure room before opening the lead shield door, which allows the core to move to that exposure room unless cleared by two licensed operators.

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These functional checks will verify operation of the interlock system. I'.xperience at AFiliti indicates that this is adequate to insure operability.

4.2 5 REACToll FUEI l't,EM ENTS Aimlic abilit y This specification applies to the surveillance requirements for the fuel elements.

Objective The objective is to verify the integrity of the fuel element cladding.

Specific atima All the fuel elements present in the reactor core, to include fuel follower control rods, shall be inspected for darr. age or dettrioration, and measured for length and bow at intervals se; arsted by not more than 500 pulsen of insertion greater than

$2.00 or annually (not to exceed 15 months), whichever occur first. Fuel elements in long-term storage need not be measured until returned to core; however fuel elements routinely moved to temporary storage shall be measured every 500 pulses of insertion greater than $2.00 or annually (not to exceed 15 months), whichever occurs first.

Basis The frequency of inspection and measurement is based on the parameters most likely to affect the fuel cladding of a pulse reactor, and the utilisation of fuel elements whose characteristics are well known.

The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of a werst case scenario in which two adjacent fuel elements suffer sufficiently severe transverse bends to result in the touching of the fuel elements has shown that no damage to the fuel elements will result via a hot spot or any other known mechanism.

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4.3 COOL, ANT SYSTENIS A t+Dlic abilit y This tipecification applies to the surveillance re<iuirements for moniitoring tb 901 w ater and the water-conditioning system.

Obiective The objective is to assure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, eliminating possible radiation harards from activated impurities in the water system, and limiting the potential corrosion of fuel cladding and other components in the primary water system.

Spacifications

a. The pool water temperature, as measured near the f t to the water purification system, shall be measured daily, whenever operation ce planned.

b. The conductivity of the water at the output of the purification system shall be measured weekly, w henever operations are planned.

Ilaia liased on experience, observation at these intervals provides acceptable surveillance of limits that assure that fuel clad corrosion and neutron activation of dissolved materials will not occur.

4.4 VENTil, ATION SYSTEh!

A tinlic abilit v This specification applies to the facility ventilation system isciation.

Objective The objective is to assure the proper operation of. the ventilation system in controlling the release of radioactive material into the unrestricted environment.

S pecificat ion The operating mechanism of the positive sealing dampers in the reactor room ventilation system shall be verified to be operable and visually inspected at least monthly (interval not to exceed six weeks).

Ilasia Experience accumulated over years of operation has demonstrated that the tests of the ventilation system on a monthly basis are sufficient to assure proper operation of the system and control of the release of radinactive material.

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- 5.0 - DESIGN FEATURES

$.1 SITE AND FACll,ITY DESCHil' TION At inlicability This specification applies to the building that houses the reactor.

Obiective -

The objective is to restrict the amount of radioactivity released into the environment.

Snecifications

a. The reactor building, as a structurally independent building in the AFRiti complex, shall have its own ventilation system branch. The effluent from the reactor ventilation system shall exhaust through absolute filters to a stack having a minimum elevation that is 18 fe-t above the roof of the highest building in the AFRRI complex. -t

b. The reactor room shall contain a minimum free volume of 22,000 cubic feet.

c, The ventilation system air ducts to the reactor room shall be equig.ed with positive sealing dampers that are activated by fail safe controls, which will automatically clor off ventilation to the reactor room upon a signal from the reactor room ali particulate monitor.

d. The reactor room shall be designed to restrict air leakage when the positive sealing dampers are closed.

. Basis The facility'is designed so that the ventilation will normally maintain a negative pressure with respect to the atmosphere, so that there will be no uncontrolled leakage to the environment; The free air volume within the reactor building is confined when there is an emergency shutdown of- the ventilation system. Building construction and gaskets around doorways help restrict -leakage of air into or out of the reactor room. The stack height insures an adequate dilution of effluents well above ground level, The separate ventilation system branch insures a dedicated air flow system for reactor effluents.

- 5,2 HEACTOR CORE AND FUEL 5.2.1 IIEACTOR FUEL Aoclicniijity These specifications apply to the fuel elements, to include fuel follower control rods, used in the reactor core.

Obiective These objectives are to (1) assure that the fuel elements are designed and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear-characteristics, and (2) assure that the fuel elements used in the core are substantially those analysed in the Safety Analysis Report, Snecifications The individual nonitradiated standard TitlG A fuel elements shall have the following characteristics:

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n. Uraniunt content: Maximuru of 9.0 weight percent enriched to less than i

20% uranium 235. In the fuel follower, the maximum uranium content will 0

I be 12.0 weight percent enriched to less than 201 uranium.235

b. llydrogen to-rirconium ratio (in the Zril ). Nominal 1.711 atoms to 1.0 Zr atorns with a range between 1.0 and I

c. Cladding: 308 stainless steel, nominal 0.020 inch thick d, Any burnable poison used for the specific purpose of compensating for fuel burnup or long term reactivity adjustments shall be an integral part of the manufactured fuel elements.

.Ilasis n maximum uranium content of 9 weight perec ut in a standard TillG A element is greater than the design value of 8.5 weight percent, and encompanes the maximum probable variation in individual elements. Such an increase in loading would result in an increase in power density of less than 6%. An increase in iower density of 0% in individual fuel element reduces the safety margin local by 10 g3, at most. The hydrogen-to-zirconium ratio of 1.7 will produce a maximum pressure within the cladding well below the rupture strength of the cladding.

The local power density of a 12.0 weight percent fuel follower is 21*i greater than an 8.5 weight percent standard TitlG A fuel element in the D Iting.

llecause the volume of the fuel in a fuel follower fuel element is 50% of the volume of a standard TitlG A fuel element, however, the actual power produced in a fuel follower fuel element is 33% less that the power produdced in a standard TillG A fuel element in the D ring 5.2.2 IlEACToll CollE Annlicability These specifications apply to the configuration of fuel and in-core experiments.

Obiective The objective is to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.

S pecific at ions

n. The reactor core shall consist of standard TillG A reactor fuel elements in a close packed array and a minimum of two thermocouple instrumented TillG A reactor fuel elements,

b. There shall be four single core positions occupied by the three standard control rods and transient rod, a neutron start-up source with holder, and positions for possible in-core experiments.

c. The core shall be cooled by natural convection water flow.

d. In-core experiments shall not b placed in reijarent fuel positions of the Il-ring and/or C-ring 26 I

c. l'uel elements indicating an elongation greater than 0.100 inch, a lateral bending greater than 0 0025 inch, or significant visible damage

- shall be considered damaged, and shall not be used in the reactor core.

Itaain Standard TitlG A cores have been in use for years, and their safe operational characteristics are well documented.1:xperience with TillG A reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to (a) assure that the cladding material will not be :.ubjected to stresses that could cause a loss of integrity in the fuel containment, and (b) assure ad*quate coolant flow.

5.2.3 CONTitOL llODS A ntlic abiliti These specifications apply to the control rods used in the recctor core.

Obiective The objective is to assure that the control rods are designed to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

S pecific at ions

n. The standard control rods shall have scram capability, and shall contain borated graphite,114C powder, or boron and its compounds in solid form as a poison in aluminum or stainless-steel cladding. These rods may have an aluminum, air, or fuel follower. If fuel followed, the fuel region will conform to the Specifications of 5.2.1.

b. The transient control rod shall have scram capability, and shall contain borated graphite, ll,,C powder, or boron and its compounds in solid form as a poison in aluminum or stainless-steel cladding. This rod may incorporate an alun.inum, poison, or air follower.

Ilasis The poison requirements for the control rods are satisfied by using neutron-absorbing borated graphite, 11 C powder, or boron and its compounds. These materials must be contained 4in a suitable cladding material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment. Scram capabilities are provided by the rapid insertion of the control rods, which is the primary operational safety feature of the reactor. The transient control rod is designed for use in a pulsing TilIG A reactor.

5.3 SPECIAl NUCI E All M ATERI AL STOlt AGE A onlicabilit y This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Obiert is c The objective is to assure that stored fuel will not b~mme criti< al and will not reach an unsafe temperature.

27

S ticcific at ion All fuel elements not in the reactor core shall be stored and handled in accordance with applicable regulations. Irradiated fuel clernents and fueled devices shall be stored in an array that will permit sufficient natural convective cooling by water or air, so that the fuel element or fueled device temperature will not caceed design values Storage shall be such that groups of stored fuel elements will remain subtritical under all conditions of mod er r.t io n.

Ita=ia The limits imposed by this specification are conservative and assure safe storage and handling. I'xperience shows that approximately 67 fuel elementa are required, of the deaign used at AFilllt, in a closely packed array to achieve criticality. Calculations show that in the event of a full storage rack failure with all 12 elements falling in the inont reactive nucleonic configuration, the inass would be less than that required for criticality.

Therefore, under normal rtorage conditions, criticality cannot lic reached, 28

.~

e 6.(} ADMIN 14Ti( ATIVE CONTi1OIS 6.1 ORG AN17 ATION i

6.1.1 STitUCTUlt E The organization of personnel for the managernent and operation of the AFilitt reactor facility is shown in Figure 1. Organization changes inny occur, based on Institute requirernents, and they will be depicted on internal documents, llowever, no changes tuny be made in the Operation, Safety, and Emergency Control Chain in which the lleactor Fncility Director has direct responsibility to the Director, AFItiti.

Director, AFRRI UP*""*l 8*ty. Reector and Radiation Chvrman. Amme '8"I Safety and coramt Fecility Safety Committee t

IIcalth Dept.

. l  !.

Chairman,  ;

' Radiation '

{ Soureca Dept. j i

a i,

Adytsor7 Advisory I Chief, Rextor j l Division i I

l i I 'I t

i Reector Farihty Di1ector ....

........ .j l

Recctor Opeintions Supervisor l

Reactor Operations Staff' Tigure 1. Organisation of Per o. .

d emr . . . . . . . . :)peration of the AFRR1 Reactor Fa ility.

Any reactor staff member has access to the Director for matters of safety.

29 I

4 6.1.2 RFSPONSIBIblTY The Director, AFRRI, shall have license responsibility for the reactor facility. The Reactor Facility Director (ItPD) shall be responsible for administration and operation of the Reactor Farility and for determination of applicability of procedures, experiment authorisations, maintenance, and operations. The RFD may designate an individual who meets the requirements of Section 6.1.3.1.a to discharge the RFD's responsibilities in the RFD's absence. During brief absences (periods less than four hours) of

[' the Reactor Facility Director and his designee, the Reactor Operations

. Supervisor shall discharge these responsibilities.

6.1.3 STAFFING 6.1.3.1 Selection of Per=onnel

a. Reactor Facility Director At the time of appointment to this position, the Reactor Facility Director shall have 6 or more years of nuclear experience. liigher education in a scientific or nuclear engineering field may fulfill up to 4

[

years of experience on a one-for-one basis. The Facility Director must l have held a USNRC Senior Reactor Operator license on the AFRRI reactor for at least 1 year before appointment to this position.

b. Reactor Operations Supervisor (ROS)

At the time of appointment to this position, the ROS shall have 3 years nuclear experience. liigher education in a science or nuclear

[ engineering field may fulfill up to 2 years of experience on a one-for one basis. The ROS shall hold a USNRC Senior Reactor Operator license on the AFRRI reactor, in addition, the ROS shall have 1 year L of experience as a USNRC licensed Senior Reactor Operator at AFRill t

or at a similar facility before the appointment to this position,

c. Reactor Operators / Senior Reactor Operators At the time of appointment to this position, an individual shall have a high school diploma or equivalent, and shall possess the appropriate USNRC license.

d. Additional staff as required for support and training. At the time of appointment to the remetor staff, an individur.i shall possess a high school diploma or equivalent.

I 6.1.3.2 Operations l

l a. hiinimum staff when the reactor is not secured shall include:

l 1. A licensed Senior Reactor Operator (SRO) on call but not necessarily on site

2. Radiation control technician on call

3. At least one licensed Reactor Operator (RO) o Senior Reactor Operator (SRO) present in the control room  ;

4. Another pernos within the AFRRI complex who is able to carry out written emergency procedures. instructions of the operator, or to summon help in case the operator becomes incapacitated.

30 l 9

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