ML19305A039
| ML19305A039 | |
| Person / Time | |
|---|---|
| Site: | Armed Forces Radiobiology Research Institute |
| Issue date: | 11/06/1978 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19305A034 | List: |
| References | |
| 50-170-78-04, 50-170-78-4, NUDOCS 7901020047 | |
| Download: ML19305A039 (4) | |
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APPENDIX A NOTICE OF VIOLATION Armed Forces Radiobiology Research Institute Docket No. 50-170 Based on the results of an NRC inspection, conducted on October 11-13, 1978, it appears that certain of your activities were not conducted in full compliance with the conditions of your NRC Facility License No. R-84, as indicated below.
Items C, E and F are categorized as Infractions and items A, B and D are categorized as Deficiencies.
A.
Technical Specification (TS) section III, paragraph A states, in part "The reactor facility shall be under the direct control of the Physicist-in-Charge of the reactor (PIC)....His staff shall include a Chief Supervisory Operator, and at least one reactor operator....The PIC and the Chief Supervisor Operator shall be NRC licensed senior reactor operators...."
Centrary to the above, as of October 13, 1978, the designated Chief Supervisory Operator did not hold a senior operator's license and the PIC was functioning as both the PIC and the Chief Supervisory Operator.
acility License No. R-84, part 4.(c), states, "AFRRI shall report B.
c to the Director, DRL, in writing within thirty (30) days of its occurrence any significant change in the transient or accident analysis as described in the safety analysis report."
Contrary to the above, as of October 13, 1978, a licensee report had not been submitted for the occurrence on August 22, 1978 when two (2) fuel assemblies were discovered misplaced in the core.
The two (2) fuel assemblies, No. 3363 (core location F-16) and No. 3332 (core location F-15) were cocked because the lower locating i
pin was not inserted in the lower core plate alignment hole.
C.
10 CFR 50.59 states in part: "(a)(1) The holder of a license author-izing operation of a production or utilization facility may (i) make changes in the facility as described in the safety analysis report...
without prior Commission appro'al, rless the proposed change, test or experiment involves a change in the technical specifications in-corporated in the license or an unreviewed safety question.
79010200%
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Appendix A 2
(2) A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence 4
or the consequences of an accident or malfunction of equipment im-l portant to safety previously evaluated in the safety analysis report 1
may be increased; or (ii) if a possibility for an accident or mal-l function of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.
(b) The licensee shall maintain records of changes in the facility
...made pursuant to this section, to the extent that such changes constitute changes in the facility as described in the safety analysis report....These records shall include a written safety evaluation which provides the bases for the detennination that the change...does not involve an unreviewed safety question....The records of changes in the facility shall be maintained until the date of termination of the license and records of changes in procedures and records of tests and experiments shall be maintained for a period of five years...."
Contrary to the above a written safety evaluation to determine that a change to the facility did not constitute an unreviewed safety question was not performed when updated power monitoring instrumen-tation was installed and the reactor console was replaced.
The console was replaced in August 1978 and the console and circuitry are described in the Final Safeguards Report.
D.
10 CFR 50.54 states in part, "Whether stated therein or not, the following shall be deemed conditions in every license issued:...
(1-1)... Holders of operating licenses in effect on September 17, 1973 shall implement an operator requalification program...which was submitted for approval by the Atomic Energy Commission." The approved " Reactor Operator Requalification Program," as amended February 5, 1974 states in part: "VI. Document Review...The licensee will review annually the following documents and instructions that are pertinent to the operation of the Reactor Facility: A. Reactor License (R-84). B. Technical Specifications.
C.10 CFR-19, 20, 30, 50, 55 and 70.
D. AFRRI Instructions.
E. Radiation Sources Division Instructions.
F. Safety Department Standard Operating Procedures....
VII. Records...An individual record file will be maintained for each licensee and the record file will contain the following information...
D. The licensee's requalification program progress checklist...." The requalification program progress checklist contains signature blocks j
to document completion of the required annual reading.
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Appendix A 3
Contrary to the above, the approved requalification program was not fully implemented, in that the annual required reading was not docu-mented as having been completed for the licensed operator and licensed senior operator required to participate in the program between January 1,1977 and October 13, 1978.
E.
Technical Specifications section III, paragraph C. states, " Written instructions shall be in effect for, but not limited to:.. 9. Any operation deemed necessary by the Staff Scientist....the PIC of the reactor and the Reactor and Radiation Facility Safety Committee."
All PSD instructions contain a cancellation section (3) which states, "PSD instruction (number) (name of instruction), revised (date) is hereby cancelled." This measure cancels previous editions of issued instructions.
Contrary to the above, on October 12, 1978, facility procedures kept in the control room were not maintained in effect as required; in that current versions were not being maintained.
The following procedure numbers are examples:
No. 5-2, latest copy, pages 1, 3, 4, 5 and 6 dated 18 May,1976 and page 2 dated 20 September, 1978.
Control room copy, page 1 dated 18 May,1978, page 2 dated 20 September,1978, pages 3 through 6 dated 1 October,1974.
No. 5-3, latest copy, dated 14 December,1976.
Control room copy, dated 1 October,1974.
No. 5-4, latest copy, dated 14 December,1976.
Control room copy, dated 1 October,1974.
No. 5-6, latest copy, dated 14 December,1976.
Control room copy, dated 1 October,1974.
F.
Final Safeguards Report, section II, page 5 states, in part, "The door to the corridor behind the reactor control room...is a double door that is sealed with compressible rubber gaskets and latched."
"The double doors at the opposite end of the corridor...is also sealed with a compressible gasket...."
Technical Specifications, section I. A.4. states, "The reactor room shall be designed to restrict air leakage when the positive sealing dampers are closed."
Appendix A 4
i Contrary to the above, as of October 13, 1978, the above doors were not maintained as designed in that gasket material was missing on both doors preventing fulfillment of the design function.
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