ML20113C366

Responds to Request for Addl Info Re Amend Request for FOL R-108

ML20113C366
Person / Time
Site:	Dow Chemical Company
Issue date:	06/18/1996
From:	Rigot W DOW CHEMICAL CO.
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9607010165
Download: ML20113C366 (11)

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The Dow Chemical Company Midland, Michigan 48667 June 18,1996 Mr. Alexander Adams, Jr., Senior Project Manager Non-Power Reactors and Decommissioning Projects Directorate Division of Project Support Office of Nuclear Reactor Regulation Dear Mr. Adams Enclosed is the response to your request for additional information regarding our amendment request for Facility Operating License No. R-108; letter dated May 21,1996.

The enclosure meludes the changes proposed for amendment No. 7 and an explanation  ;

for each question raised in your May 21,1996 letter. If you have any questions regarding this review, please contact me at (517) 636-6584.

Regards G i Ward L. Rigot j Reactor Supervisor 1 i

Dow TRIGA Research Reactor Dow Chemical Company I l

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/g-49-9c, Of/2 o TAMMY A. ECKERD NOTARYPUBLIC BAYCOUNTY MICHGAN ACTNG H MIDLAND COUNTY MICHIGAN MY COMNLSSION EXPIRES APRL 17,2000 S'

9607010165 960618 PDR ADOCK 05000264 P PDR i

l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOW CHEMICAL COMPANY TRIGA RESEARCH REACTOR DOCKET NO. 50-264 l

1. Table 3.3.A has been changed to reflect the proposed change. i l

l 2. Table 3.3.B has been change to reflect a change in the table alignment.

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l 3. TS 4.5 has been changed to reflect correction of a misspelled word " considered"; and l TS 6.2.3 has been changed to reflect the correction to the splitting of the word "that". l i

4. TS 1.29a has been changed to show consistency of the term " safety system". All other references to safety system within the document are of this form. There is no implied difference between " safety system" and " safety-system". Research Manager was changed to " Facility Director" to show consistency with accepted NRC nomenclature.

l I 5. The organization chart in Figure 6.1 has been changed to show reporting lines versus communication lines. As is reflected in this chart, the ROC has responsibility over the reactor operations and communicates to the radiation safety committee (RSC), ,

which has responsibility for the radiation safety program for the entire Michigan site.

Two additional changes has been added. The " Director, Health and Environmental Services" has been changed to " Director, Health and Safety Services". This reflects -

an organizational name change. The " Director, Analytical Sciences Laboratory" has been changed to " Director, Analytical Sciences Laboratories". This reflects an organizational name change.

6. The index should not be included in the Technical Specifications, but should be used as a aid for locating information within TSs.

7. In TS 6.5.2 a. the word "and" has been added to show consistency.

8. In TS 6.6.2 the destination of reports has been changed to the appropriate office.

Additionally, TS 6.61 had the same incorrect destination of operating reports. This has also been changed to the appropriate office.

9. An additional review uncovered an additional inconsistency. In TS 6.52 and TS 6.62 references are made to " Reportable Occurence of the type identified in section 1.28". ^

Due to the addition of an additional definition, Reportable Occurences are now defined in section 1.29. These TS's have been changed to refernce the appropriate definition.

Those pages where changes appear are enclosed.

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1.24. Reactor Shutdown - The reactor is shutdown if it is suberitical by at least one dollar and the reactivity worth of all experiments is accounted for.

1.25. Reactor Ooerations Committee (ROC) - The ROC is charged with direct oversight of the reactor operations, including both review and audit functions.

1.26. Reactor Safety Systems - Reactor Safety Systems are those systems, including associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

1.27. Reference Core Condition - The Reference Core Condition is that condition when the core is at ambient temperature (cold) and the reactivity worth of xenon in the fuel is negligible (less than $.30).

1.28. Research Reactor- A Research Reactor is a device designed to support a self-sustaining nuclear chain reaction for research, development, education, training, or experimental purposes, and which may have provisions for the production of radioisotopes.

1.29. Reportable Occurrence - A Reportable Occurrence is any of the following wb!ch occun during reactor operation:

a) Operation with actual safety system settings for required systems less conservative than the limiting safety system settings specilled ir. Technical Specification 2.2.

b) Operation in violation of limiting conditions for operation established in the Technical Specifications.

c) A reactor safety synem component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.

d) Any unanticipated or uncontrolled change in reactivity greater than one dollar.

Reactor trips resulting from a known cause are excluded.

e) Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary which could result in exceeding prescribed radiation exposure or release limitr.

O An observed inadequacy in the implementation of either administrative or procedural. controls which could result in operation of the reactor outside the limiting conditions for operation.

i g) Release of radioactivity from the site above limits specified in 10CFR20. I i

1.30. Rod. Control - A control rod is a device containing neutron absorbing material which is used to control the nuclear fission chain reaction. The control rods are coupled to the control rod drive systems in a way that allows the control rods to perform a safety function.

Amendment No. 7 3

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TABLE 3.3A BASES FOR REACTOR SAFETY CHANNELS AND INTERLOCKS Scram Channels Scram Channel h R: actor Power Level Provides assurance that the reactor will be shut down automatically before the safety limit can be exceeded Reactor Power Channel Provides assurance that the reactor Detector Power Supplies cannot be operated without power to the neutron detectors which provide input to the NM1000 and NPP1000 power channels Manual Scram Allows the operator to shut the reactor down at any indication of unsafe or abnormal conditions Watchdog Ensules adequate communications between the Data Acquisition Computer (DAC) and the Control System Computer (CSC) units.

Interlocks Interlock / Channel Bases l

Startup Countrate Provides assurance that the signal in the NM1000 channel is adequate to allow reliable indication of the state of the neutron chain reaction.

Rod Drive Control Limits the maximum positive reactivity insertion rate Reactor Period Prevents operation in a regime in which transients could cause the limiting safety system setting to be exceeded Amcodment No. 7 TABLE 3.3B MEASURING CHANNELS I

Measuring channel Minimum Number Operable l l

NMl(XX) 1 I l

NPPlotX) 1 Water Radioactivity 1 Monitor - j i

Water Temperature 1 l Monitor TABLE 3.3B BASES FOR MEASURING C11ANNELS Measuring Channel DASH

. NM1000 Provides assurance that the reactor power level can be adequately monitored.

NPP1000 Provides assurance that the reactor power level can be adequately monitored.

Water Radioactivity Provides assurance that the water i Monitor radioactivity level can be adequately monitored. l Water Temperature Provides assurance that the water Monitor temperature can be adequately monitored. j l

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Amendment No. 7 13 +

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. 4.5. Facilliv Snecific Surveillance Amlicability This specification shall apply to the fuel elements of the Dow TRIGA Research Reactor.

Ohlective The objective of this specification is to ensure that the reactor is not operated with damaged fuel elements.

Snecification Each fuel element shall be examined visually and for chang:s in transverse bend and length at least once each five years, with at least 20 percent of the fuel elements examined each year. If a damaged fuel element is identitled, the entire inventory of fuel elements will be inspected prior to subsequent operations.

The reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal. A TRIGA fuel element shall be considered damaged and removed from the core if:

a) The transverse bend exceeds 0.125 inch over the length of the cladding.

b) The length exceeds the original length by 0.125 inch.

c) A clad defect exists as indicated by release of fission products.

ILuh Visual examination of the fuel elements allows early detection of signs of deterioration of the fuel elements, indicated by signs of changes of corrosion patterns or of swelling, bending, or elongation. Experience at the Dow TRIGA Research reactor and at other TRIGA reactors indicates that examination of a five-year cycle is adequate to detect problems, especially in TRIGA reactors that are not pulsed. A five-year cycle reduces the handling of the fuel elements and thus reduces the risk of accident or damage due to handling.

i Amendment No. 7 26

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Figure 6.1. Administration

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1 1 Director, Health and Director, Radiation Safety Safety Services Committee Analytical Sciences l Laboratories i

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1P l 1r Manager, Industrial 4- I Hygiene Research and -J ---+ Facility j

IIII Director Technology Chair, RSC  ! Chair, ROC j

l Reactor Operations qp l Committee (ROC) qp l

i I Supervisor, l h h Reactor Supervisor Industrial l l

._ _ g Hygiene l g l l

1r I I 1r i 4J l Radiation Safety 4 ,. ,,j Licensed l f SROs l Officer and ROs Line management responsibilities Line Management Reporting

---

Communication Reporting i

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Anwndment No. 7 33 -

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. 6.2.3. Audit Function 1

a. The ROC shall direct an annual audit of the facility operations for conformance to the technical specifications, license, and operating procedures, and for the results of actions taken to correct those deliciencies which may occur in the reactor facility equipment, systems, structures, or methods of operations that affect reactor safety.

This audit may consist of examinations of any facility records, review of procedures, and interviews of licensed Reactor Operators and Senior Reactor Operators.

De audit shall be performed by one or more persons appointed by the ROC. At least one of the auditors shall be familiar with reactor operations. No person directly responsible for any portion of the operation of the facility shall audit that operation.

A written report of the audit shall be submitted to the ROC within three months of the audit.

Deliciencies that affect reactor safety shall be reported to the Facility Director immediately.

b. The ROC shall direct an annual audit of the facility emergency plan, security plan, and the reactor operator requalliication program. This audit may consist of the annual review of these plans for the requallilcation program.

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I Amendment No. 7 37 i

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6.5. Reautred Actions 6.51. In case of Safety Limit violation:

a. the reactor shall be shut down until resumed operations are authorized by the US NRC; and

b. the Safety Limit violation shall be inunediately reported to the Facility Director or to a higher level; and

c. the Safety Limit violation shall be reported to the US NRC in accordance with section 6.6.2.; and

d. a report shall be prepared for the ROC describing the applicable circumstances leading to the violation including, when known, the cause and contributing factors, describing the effect of the violation upon reactor facility components, i systems, or structures and on the health and safety of personnel and the '

public, and describing corrective action taken to prevent recurrence of the violation.

6.5.2. In case of a Reportable Occurrence of the type identified in section 1.29

a. reactor conditions shall be returned to normal or the reactor shall be shut down; if the reactor is shut down operation shall not be resumed unless authorized by the Facility Director or designated alternate; and

b. the occurrence shall be reported to the Facility Director and to the US NRC as l required per section 6.6.2.; and  ;

c. the occurrence shall be reviewed by the ROC at the next scheduled meeting.

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i Amendment No. 7

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. 6.6. Reoorts 6.6.1. Operating Reports A report shall be submitted annually, starting with the first quarter 1991 performance of annual tasks, to the Radiation Safety Committee and to The Document Control Desk US NRC, Washington, DC, with a copy to the Regional Administrator, US NRC Region 111, which shall include the following:

a) status of the facility staff, licenses, and training; b) a narrative summary of reactor operating experience, including the total megawatt-days of operation; c) tabulation of major changes in the reactor facility and procedures, and tabulation of new tests and experiments that are significantly different from those performed previously and are not described in the Safety Analysis Report, including a summary of the analyses leading to the conclusions that no unreviewed safety questions were involved and that 10 CFR 50.59 (a) was applicable; d) the unscheduled shutdowns and reasons for them including, where applicable, corrective action taken to preclude recurrence; e) tabulation of major preventive and corrective maintenance operations having safety significance; f) a summary of the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge (the summary shall include to the extent practicable an estimate ofindividual radionuclides present in the ef0uent; if the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, only a statement to this effect is needed); and g) a summary of the radiation exposures received by facility personnel and visitors where such exposures are greater than 25 % of those allowed or recommended in 10 CFR 20.

l Amendment No. 7

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6.6.2. Special Reports  !

a. There shall be a report to US NRC Region III not later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to The I)ocument Control Desk, US NRC, with a copy to the Regional Administrator, Region 111, US NRC to be followed by a written report that describes the event within 14 days of:

a violation of the Safety Limit; or '

a reportable occurrence (section 1.29), i

b. There shall be a written report presented within 30 days to The Document Control Desk, US NRC, with a copy to the Regional Administrator, Region III, US NRC, of:  !

permanent changes in the facility staffinvolving the reactor supervisor or the facility director; or i significant changes in the transient or accident analysis report as described in the Safety Analysis Report. j

c. A written report shall be submitted to The Document Control Desk, US NRC, with a copy to the Regional Administrator, Region III, US NRC, within 60 days after ,

criticality of the reactor under conditions of a new facility license authorizing an  !

increase in reactor power level, describing the measured values of the operating  !

conditions or characteristics of the reactor under the new conditions.

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I Amendment No. 7 41 l

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