ML19305A036

Responds to NRC 781107 Ltr Re Deviation Noted in Insp Rept 50-170/78-04.Corrective Actions:Reactor Staffing Corrected, Fuel Assembly Displacements Will Be Reported,Operator Requalification Program Instituted & Reactor Room Sealed

ML19305A036
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	11/28/1978
From:	Mcindoe D DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
To:	Brunner E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML19305A034	List: ML19305A036
References
NUDOCS 7901020041
Download: ML19305A036 (13)

Text

U DEFENSE NUCLEAR AGENCY ARMED FORCES RADICBIOLOGY RESEARCH INSTITUTE B ETHESD A, M A RYL AND 20014 DIR 2 8 MOV 1978 Mr. Eldon J. Brunner, Chief Reactor Operations & Nuclear Support Branch Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406

Dear Mr. Brunner:

This refers to the recent inspection of the Armed Forces Radiobiology Research Institute Reactor Facility (License No. R-84) and the results of the inspection as reported by your letter of 7 November 1978, subject:

Inspection 50-170/78-04. Corrective actions are listed below in the format of Appendix A, Notice of Violation, along with comments concern-ing management controls instituted to preclude any similar recurrence.

A. Reactor S+affing: The AFRRI was temporarily in noncompliance with the Technical Specifications in that the Physicist-in-charge was also serving as the Chief Supervisory Operator. This situation was

. corrected as of 6 November 1978 when the presently designated CS0, Mr.

Marcus L. Moore, received his senior operator license, SOP-3370, Docket 55-6874. The present reactor facility stdffing is:

Captain Ronald E. Schaffer, USA, PIC, SR0 Mr. Marcus L. Moore, CSO, SR0 SFC Willard R. Yuna, R0 HM1 Lamar R. Creel, R0 Captain Robert M. Savage, Jr., Trainee In addition, the AFRRI Technical Specifications are being revised as a result of dicussions with the NRC. This revision is addressing staffing and management aspects consistent with the format and content of recent ANSI recommendations and NRC suggestions. A copy of the revised Technical Specifications will be provided to the Commission during the first half of calendar year 1979.

790102004 i

DIR Mr. Eldon J. Brunner B. Fuel Assembly Displacement: This situation was discussed in detail during the inspection and the inspector indicated he felt this was a matter of judgement although his inclination was that a written report should have been submitted. Subsequently, the report lists it as a deficiency This event had not been reported to the Comission because our judgement was that the applicable section of the license was paragraph 4(b) which states that a report , required in thirty days for "...any substantial variance disclosed by operation of the reactor from performance specifications contained in the safety analysis report or in the technical specifications." The change in the excess reactivity was not substantial--it was well within measurement error--

and there were no changes in other reactor characteristics; thus, it was judged that this did not constitute a substantial variance. The judgement on our part was that the paragraph cited by the inspection report was not applicable in that paragraph 4(c) relates to analysis of behavioral and safety aspects of the reactor, not to operational aspects. Nevertheless, in January 1975, a similar event had occurred and was reported to the Comission at that time for informational purposes; we agree a report should have been rendered in this instance for the information of the Comission. Subsequently, the Commission has been formally notified by letter, dated 21 November 1978, of this occurrence.

In the future, if there is any question on our part after proper in-house discussion and safety committee review as to whether a written report is in order, the course will be to consult with the Commission immediately for guidance.

C. Unreviewed Safety Ouestions: The installation of a new reactor console had been discussed at several previous meetings of the Reactor and Radiation Facility Safety Committee. The view of the Committee was that this would enhance the safety aspect of the reactor operation in that greater reliability would be attained with the new equipment and that this procedure was in compliance with SAR and Technical Speci-fications requirements. The Committee concurred in the installation.

However, a written safety evaluation was not presented. Subsequently, a safety evaluation was written and evaluated; the Comittee approved the evaluation on 28 November 1978. Thus, this does not now constitute an unreviewed safety question. A copy of the evaluation is attached.

D. Reactor Operator Requalification Program: All operators review and utilize the documents listed in the requalification plan as a matter of normal operations and on an annual basis as a part of their requalifi-cation procedure under the control and supervision of the Physicist-in-Charge. Although the AFRRI approved " Reactor Operator Requalification Program" requires documentation that this has been accomplished, it is apparent that we had not formally done the necessary documentation.

The requalification checklist has been revised changing Section III which previously read " Additional Training" to now read " Document Review".

Upon completion of the review of a required document, it will be so noted in this checklist.

2

DIR Mr. Eldon J. Brunner E. Facility Procedures Updating: No response to this item is required as indicated in the letter transmitting the inspection report.

F. Reactor Room Door Seals: New gaskets for the doors were on order at the time of the inspection. These were installed on 13 November 1978 but were not satisfactory in that sufficient compression was not attained. Therefore, additional gaskets have been ordered. You will be informed immediately upon their installation which is expected to be shortly.

With the exception of item F above, the AFRRI is in compliance with the NRC requirements. It is expected that item F will be corrected prior to 29 December 1978. Since all the other items pertain to areas of administration or judgemental evaluations, several management pro-cedures have been instituted to preclude similar recurrence. These include weekly meetings between the Institute's Research Program Co-ordinator and the Head of the Radiation Sources Division in which the activities of the reactor facility are reviewed and discussed, a standard procedure of consulting with the NRC when there is any question of interpretation or application of the provisions of various documents, greater utilization of the Reactor and Radiation Facility Committee in routine matters, and a monthly briefing to the Director of the In-stitute by the Physicist-in-Charge. It is felt that these measures are sufficient.

The views of the inspection team are appreciated and their comments have been valuable to us. The Reactor and Radiation Facility Safety Committee has reviewed and approved this response, ncerely, DARRELL W. McIND0E Colonel, USAF, MC Director 3

l

• e

• e 0

'W RE MCES RADI0 BIOLOGY RESEARCH INSTI'lVTE REACTOR INSIRUMENTATION SAFETY ANALYSIS

The installation of new reactor instrunentation at the AFRRI 'IRIGA Mark F reactor facility provides equal or greater operational and safety capabilities with a higher degree of reliability than the previous instrunentation. The most significant operational difference in the instrunentation is the relocation of scrams from the operational channels, linear and log, to two independent power monitoring channels. The instrt:nentation, contained in a single desk-type console, has operati.onal and safety channels. Figure 1 shows a simplified circuit layout of the safety r d control sections of the console that uses signals generated by detectors, physically in or near the core.

Each safety circuit or channel is independent, having an individual power supply, circuit wiring, amplifier, and bistable trip. Although each channel is separate and set for a specific use, the operation of each channel is similar. The following examplifies the operation of the instrunented safety channels.

A detector, ion chamber or thermocouple, is placed in the core environment. Where necessary, a supply voltage is provided. The supply is monitored with a bistable trip such that the loss of voltage results in a scram. The detector output is passed through an amplifier which has a linear output proportional to the imput. The output voltage is placed on the input of a bistable trip circuit. The bistable trip holds a relay in the energized state as long as the input voltage is within a preset range.

With the increase or decrease of the input voltage to a point outside the preset level the relay is released. A set of contacts on the relay is part of the "ANI7' circuit which comprises the scram logic. The scram logic is the series of open-on-failure relay contacts such that any scram signal or component failure results in the loss' of magnet current to the standard control rods and the air supply to the transient rod. Scram signals are the application, or the removal, of voltages from scram circuits such that holding voltages on the coils of the scram logic relt vs are removed. Since these coils are in series, the loss of any one results in a scram. The safety channels are provided with detectors and scram levels as follows:

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1.' Safety Channal Ono-High Flux--Ion chamber d:tector; scram level at 110% of maximun authorized power (1.1 W steady state).

2. Safety Channel Iko-High Flux--Ion chamber detector; scram level 110% of maximum authorized power (1.1 W steady state). Switch to' pulse detector (ion chamber) in pulse mode, scram level 2750 l

• peak, 33 W-sec NVI.

3. Safety Channel, One-Fuel Temperature--Thermocuple detector, embedded in C ring fuel element; scran level 500 C.

4. Safety Channel Two-Fuel Temperature--Thermocouple detector, i embedded in B ring fuel element. Output recorded on console l reocrder in pulse mode; scram level 500 C.

The safety channel levels are indicated en front panel meters in the steady state modes. The nunber one channels, flux and temperature, are indicated on the left hand drawer; the nunber two channels flux and temperature, are on the right hand drawer. In pulse mode, the peak power is read on the blue trace of the console recorder and NVT is read on the meter labled MEGAWATr-SEC, In addition to panel meters, the fuel temperature is read on the red trace of the console recorder in pulse mode.

The operational channels, wide range log and multirange linear, receive a signal from a fission detector placed just above the reactor core. The amplified detector output is split and sent to each channel.

The operation of the channels is similar. The log channel uses a count-rate and campbelling technique to display the reactor power level from source to full power. This output is displayed on a ten decade scale on the console recorder and a front panel meter. The same ten decades are used in nineteen linear ranges by use of a power range switch on the linear power channel. A combinaticn of signals from the linear and log channels provide operational information for the servo action in Mode la and II. Through 2

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usa of a perc;nt demand (of full scale), as seen on the linear channsl, and period information from the log channel, power increases or decreases can  !

be accomplished through automatic servo action.

l The reactor is operated in four modes: Automatic (Mode la); Square Wave (Modo II); and Pulse (Mode III). In Mode la, the steady state power is controlled by servo action of the regulating rod. Manual movement of all other rods is prevented in this mode. In Mode I, manual mode, all rods are manually manipulated. by the operator. In square wave operation, reactor power controlled by servo operation using the transient rod. After a step insertion, the reactor reaches a preset power level, a signal starts the servo, ramps the reg rod to its full up position and power level in maintaied by servo action. In pulse mode, no rod or drive up movement is permitted except the air ejection of the transient rod to a pre-set anvil height. Figure 2 is an operational diagram of reactor control in the various modes of operation. In manual mode, not shown, rod movement is controlled by the operator.

To limit power increases, unless selected operation conditions are met, a set of Rod Withdrawal Prevent (RWP) interlocks act to stop upward motion of reactor control rods until the RWP has been cleared. The following are RWP's: (Refer to Figure 1)

1. 1 M Interlock--prevents ejecting trans rod unless reactor power is below 1 m.

2. Source Level--operating channel must see source level neutrons.

3. Period--power increase must have greater than three second period in Mode 1 and la.

4. Operational H.V.--high voltage must be supplied to fission detector.

5. Bulk H20--pool must be below 60 C.

3

- ' 6. ' OPS-Calibrato--se:n as a sourcs interlock, acts when opsrational channel is in calibrate modes. l i

i Figure 3 diagrams the set of scrams any one cf which can initiste a j scram, causing a shutdown of the reactor. The following explains the action of the listed scrams:

1. Steady State Timer--use to time reactor runs to the nearest hundredth sec. Adjustable to 9999.99 sec.

2. Pulse Mode Timer--use to scram reactor in pulse mode. Adjustable to 9.999 sec. Normally sat to 0.900. ,

3. Manual Scram Bar--located just above rod drive switches.

4. Individual Rod Scram--drops individual rod through single current interuption.

5. Core Movement--scrams with movement of core in any direction.

6. Console Key--Interrupts magnet and air unless key is in the console and in the Operate position or Reset.

7. Lead Doors Between Open and Closed--Doors cannot be between fully open and fully closed.

8. Energency Stop--activated from console and either exposure room.

9. Calibrate Positions--all safety channels must be in operate position.

10. Module Removal--requires all safety channel components to be in place.

11. Loss of A.C. Power--loss of supply voltage to m6gnet and air

- solenoid.

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12. Safe 1 High Flux--greater than 1.1 W.

13. Safe 2 High Flux--greater than 1.1 W.

14. Safe 1 Fuel Temp--greater than 500 C fuel temp.

15. Safe 2 Fuel Temp--greater tipn 500 C fuel temp.

16. Loss of Safe 1 High Voltage--no detector voltage.

17. Loss of Safe 2 High Voltage--no detector voltage.

Figure 4 shows a more detailed circuit including mode and readout switching.

In summary, the newly installed instrtmentation through the use of detectors, amplifying and scram circuit monitors and reacts to satisfy the technical specification requirements contained in Reactor License #R84 for the AFRRI reactor.

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