ML20138C511
| ML20138C511 | |
| Person / Time | |
|---|---|
| Site: | 05000083 |
| Issue date: | 03/10/1986 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20138C489 | List: |
| References | |
| 50-083-86-01, 50-83-86-1, NUDOCS 8604020516 | |
| Download: ML20138C511 (6) | |
See also: IR 05000083/1986001
Text
{{#Wiki_filter:__ ., i A rth UNITED STATES -[ q'o NUCLEAR REGULATORY COMMISSION .g* REGloN il
_ n y j 101 MARIETTA STREET,N.W.
- I
e ATL ANTA, GEORGI A 30323 \\...../ , Report No.: 50-83/86-01 Licensee: University of Florida 202 Nuclear Sciences Center Gainesville, FL 32601 Docket No.: 50-83 License No.:.R-56 Facility Name: University of Florida Training Reactor . Ir,spection Conducted: February 18 - 21, 1986
Inspector: dh mo ' g P. T. Burnett 6/ / Date Signed " ' ^ Appro ed by: _/7A 8/4[[d F. Jape, Sect' ion Chief (/ / Date Signed Engineering Branch Division of Reactor Safety SUMMARY Scope: This routine, unannounced inspection entailed 26 inspector-hours at the site during normal duty hours, in the areas of maintenance, modifications, and surveillance. Results: One violation was identified: Failure to docisment the safety review required by 10 CFR 50.59 paragraph 5. 8604020516 860327 PDR ADOCK 05000083 O PDR L L a
- . REPORT DETAILS 1. Persons Contacted Licensee Employees
- M. J. 0hanian, Chairman of RSRS and Associate Dean for Research,
College of Engineering
- W. G. Vernetson, Acting Director of Nuclear Facilities
- P. M. Whaley, Acting Reactor Manager
Other licensee employees contacted included operators and office personnel.
- Attended exit interview
2. Exit Interview The inspection scope and findings were summarized on February 21, 1986, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings. No dissen- ting comments were received from the licensee. Ine licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection. One violation and licensee commitments are listed below: a. VIO 083/86-01-01: Failure to maintain records of facility changes as required by 10 CFR 50.59(b) paragraph 5. b. IFI 083/86-01-02: Review SAR paragraph 7.2.3, and revise as necessary by August 31, 1986 - paragraph 6. c. IFI 083/86-01-03: Review CORA calculations for proper identification of materials, and revise SAR as needed by August 31, 19S5 paragraph 6. d. IFI 083/86-01-04: Evaluate storage and release of Wigner energy by May 31, 1987 - paragraph 6. 3. Licensee Action or Previous Enforcement Matters (Closed) Deviation 083/85-01-01: Failure to comply with Section 17 of the Safety Analysis Report and ANSI N402-1976. The licensee has written and approved a series of procedures to implement the requirements on ANSI N402, Quality Assurance Program Requirements for Research Reactors. The organiza- tion 'of the procedures is different from that described in the licensee's letter of April 19, 1985, but it is consistent with later discussions between the licensee and Region II supervision. Each of the 17 program requirements of the standard is addressed in at least one of the new proce- dures. - - - .
' . 2 -4. Unresolved Items No unresolved items were identified in this inspection. . ' 5. Reactor Maintenance and Modification (40750) On September 3,1985, the licensee reported that safety blade 3 failed to insert fully when dropped from a partially withdrawn position. A similar event had occurred on January 26, 1985. The licensee then shutdown the reactor for an exhaustive. investigation of the ca'uses of the blade failures and to implement the corrective action dictated by the results of the investigation. -This review and inspection of those activities were facili- tated by the licensee's practice of issuing internal progress reports to mark milestone events during the five-month outage. Ultimately, the investigation led to the complete defueling of the reactor, partial removal of the graphite moderator, and complete disassembly of the control blade drives. The cause of the failure to insert was finally traced to a metallized graphite bushing that was frozen to shaft coupling assembly =and had to be pried off for disassembly. The interior of the bushing was found to be rippled with rough wear. patterns. All other shaft bushings on drive 3 as well as the other drives slid off the shafts with ease. A decision was made to replace all similar bushings on all drives with new ones of the same original design and materials. The shaft coupling to safety blade 3 was found to be rusted and scarred, and, in the opinion of the licensee, not serviceable. The decision was made to replace the AISI 1040 steel blade couplings on all four drive units with locally-fabricated couplings of 304 stainless steel. To assure that all proposed modifications are evaluated for unreviewed safety question considerations, as required by 10 CFR 50.59, the licensee uses UFTR Form' SOP-0.4A, Unreviewed Safety Question and Determination, to guide the evaluation. Supporting documents are attached to the form as necessary. The inspector reviewed about a dozen of the forms completed during the outage, including that for the modification discussed above, and discussed selected cases with the licensee. In all but one case, the review and documentation were found to be acceptable. The exception was the modification of the control blade shrouds by cutting , away part of the top of each shroud to facilitate viewing blade operation. " This modification was made early in the outage before the decision to remove the blades and drives totally 'was made. The Safety Analysis Report in paragraph 4.1.1 and Technical Specification 3.2.1(1) both state that the shrouds protect the control blades. However, none of the documentation of the safety review mention the protective function of the shrouds or how that function would be affected by the modification. Discussions with the reactor staff and members of the Reactor Safety Review Subcommittee (RSRS) i . . - . . . . . . - . - - - . - - - . -
m ' 3 confirmed that the protective function was discussed during an RSRS meeting and that a conclusion that it was not an unreviewed safety question was reached. The. inspector.found the oral arguments convincing with respect to the safety issue. However, the lack of documentation has been identified as a potential violation: VIO 083/86-01-01: Failure to maintain records of facility changes as required by 10 CFR 50.59(b). Following the completion of the maintenance and modification program, the control / safety blade drop times were measured repeatedly under a variety of conditions: 'a . With no magnetic clutch following reconnecting of all drive components, b. With the magnetic clutch operating prior to fuel loading, Following fuel loading and the replacement of th'e first layer of shield c. blocks, and d. Following restacking of all concrete shield blocks. The maximum average drop time for any rod for any condition was 0.475 seconds and the fastest average was 0.400 seconds. These numbers are well within the one second limit of Technical Specifications, and in the licensee's judgement reflect a return to the as-new co'ndition. 6. Review of the Safety Analysis Report (SAR) The SAR was reviewed to gain familiarity with the facility. In that review, two items were identified ig which the SAR description of the facility did not appear to be accurate: a. Paragraph 7.2.3 describes operation of the control rod inhibit system and automatic control system, which is different from the performance described in Technical Specification 3.2.1. Surveillance procedures confirm performance in conformance to the requirements of Technical Specifications. b. Figure 4-16, which describes _ material areas used in the CORA computer program analysis of reactor neutronics, labels an area as water which properly should be graphite. At the exit interview, the licensee made two commitments, which will be tracked as inspector followup items: c. IFI 083/86-01-02: Review SAR paragraph 7.2.3, and revise as necessary by August 31, 1986, d. IFI 083/86-01-03: Raview CORA calculations for proper identification of materials, and revise SAR as needed by August 31, 1986.
V . - 4 The SAR review also revealed .that the storage of Wigner energy in the graphite moderator and the potential for autorelease of that' energy had not been considered. At the exit interview, the licensee made a commitment to complete an analysis of the potential lifetime storage of Wigner energy and associated risks by May 31, 1987. This will be tracked as IFI 083/86-01-04: Evaluate storage and release ~ of Wigner energy. 7. Review of Surveillance Procedures The following surveillance procedures were reviewed: a. SOP-A.7 (Revision 1), Determination of Control Blade Integral or Differential Reactivity Worth. The method of worth determination is that traditionally used on research reactors with semaphore control and safety blades. Individual blades are dropped from the critical condi- tion at low po'wer, in this case 100 watts. The subcritical decay of flux is then analyzed to infer rod worth. The inspector witnessed portions of the worth determination for safety blade 2, including drops from 100, 90, 80, and 70% withdrawn. Later the results and analysis were discussed with members of the staff. The operators performing the surveillance appeared to be well versed in the procedure requirements and the methodology in use. b .~ SOP-0.5 (Revision 1), UFTR Nuclsar Instrumentation Calibration Check and Heat Balance. The procedure reflects a technically adequate method of performing a heat balance and determining the thermal power of the reactor. However, vague definitions of the symbols used in the equa- tions' appear to make it difficult for less-experienced, newly-licensed, personnel to use. c. 50P-E.7 (Revision 0), Measurement of Temperature Coefficient of Reac- tivity. This procedure is adequate to perform the annual -surveillance required by Technical Specification 4.2.1 to assure that the coolant temperature coefficient is negative. This surveillance has been overdue since November 1985; performance is not possible with the reactor shutdown. It is scheduled for performance after the rod worth determination. The test completed on November 29, 1984, was acceptable. d. SOP-E.8 (Revision 0), Verification of the UFTR Negative Void Coeffi- cient of Reactivity. This procedure is responsive to the requirement of Technical Specification 4.2.1(3) to verify biennially that the coolant void coefficient is negative. The last performance of this procedure, on October 26, 1984, was revieved and found acceptable. Special Procedure for Sequencing Fuel Load Increments to Load UFTR Core e. was written spec fically to guide the- post-maintenance, post-modifica- tion refueling. Review of the completed procedure and discussions with facility personnel confirmed that the reloading had been accomplished in a safe and controlled manner. t 4
, , - . 5 8. Requalification of Operators In the course of the extended outage none of the licensed operators were able to maintain their proficiency by operating the reactor. In a letter to NRC Region II, dated January 6,1986, the licensee proposed an operator requalification process. That process was found acceptable by the. region, and that finding was corresponded to the licensee in a letter . dated January 28, 1986. Satisfactory completion of the requalification process for all current operators was confirmed by review of completed tests and records of performance. The final step in requalification of one operator, ~ a startup to one Watt, was witnessed. No violations, other than the one identified in paragraph 5, or deviations were identified. a }}