ML20138C511
Jump to navigation
Jump to search
| 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*
_
*
n REGloN il
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
F. Jape, Sect' ion Chief (/ /
8/4[[d
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,
c. Following fuel loading and the replacement of th'e first layer of shield
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.
e. Special Procedure for Sequencing Fuel Load Increments to Load UFTR Core
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
}}