ML20217D541
| ML20217D541 | |
| Person / Time | |
|---|---|
| Site: | University of Michigan |
| Issue date: | 03/24/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217D535 | List: |
| References | |
| 50-002-98-202, 50-2-98-202, NUDOCS 9803300058 | |
| Download: ML20217D541 (11) | |
See also: IR 05000002/1998202
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U.S. NUCLEAR REGULATORY COMMISSION
Docket No:
50-002
License No:
R-28
Report No:
50-002/98202 (DRPM)
Licensee:
University of Michigan
Facility Name:
Ford Nuclear Reactor
Location:
Ann Arbor, Michigan
Dates:
February 23-27,1998
Inspector:
T. M. Burdick, Reactor inspector
Approved by:
Marvin M. Mendonca, Acting Director
Non-Power Reactors and Decommissioning
Project Directorate
Division of Reactor Program Management /NRR
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9803300058 980324
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EXECUTIVE SUMMARY
. Ford Nuclear Reactor
Report No. 50-002/98202(DRPM)
' This routine, announced inspection included aspects of organization, operations, and
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maintenance (39745); review and audit (40745); requalification training (69003); and
procedures (42745).
' Oraanhation. Onarations. and Maintenance
The licensee recently posted a vacancy announcement for the position of Reactor ~
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Manager. This vacancy was open since 1996 pending the University's review of the
Ford Reactor's future status,
e - Operations were temporarily interrupted periodically during the period due to
maintenance, refueling, and unplanned scrams,
Operating problems led the licensee to incorporate new preventive maintenance and
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surveillance for the pneumatic cylinder actuator for the ventilation dampers and its air
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supply into the regular schedule. (Section 1.0)
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Review. Audit. and Desian Chance Functions
e . The inspector identified what appeared to be an inadequate 10 CFR 50.59 review of one
design change affecting Technical Specifications (TS) and the Safety Analysis Report
(SAR) assumptions. This was an apparent violation. Also, the NRC was not notified of
this 10 CFR 50.59 problem as required by TS 6.2.2.b.2. This was also an apparent
violation. (Section 2.0)
Ooerator Reaunlification
e Operators were evaluated in accordance with the requalification program requirements.
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Procedures .
- e Operators were familiar with procedures, and ' associated equipment and limits. An
inspector follow up item was identified on the consistency between operating procedure
guidance and a recently installed modification. (Section 4.0)
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REPORT DETAILS
Plant Summary
The Ford Nuclear Reactor at the University of Michigan continued to operate in support of
tescaing in the graduate and undergraduate programs and of various research and irradiation
services. The University concluded this past summer that future operation of the reactor
would be supported for the duration of the current license, such that a vacancy
announcement for the Reactor Manager position has been issued.
1.0 Organization, Operations, and Maintenance
a.
Insoection Scone (39745)
The inspection was to verify whether the licensee had maintained its organization
as required and whether operations and maintenance were consistent with
requirements.
b.
Finding.s_and Observations
The licensee has not had a permanently assigned Reactor Manager since 1996.
The Assistant Manager, Reactor Operations continued as the acting Reactor
Manager. The licensee recently posted a vacancy announcement for the position
of Reactor Manager as previously mentioned.
The reactor had been operated continuously for several hours during the svening
shift of October 4,1996, within 1.2"F of the reactor safety limit of 11VF reactor
inlet temperature as discussed further in Section 2.0. The licensed shift
supervisor responsible for that shift stated that operation that close to the safety
limit was within the procedural guidance and that it was done to avoid starting
and stopping the cooling tower fans, which causes the reactor pool temperature
to vary widely. This problem has recently been addressed by the installation of
one variable speed fan motor. When asked by the inspector, one of the more
recently licensed operators felt that operating so close to the safety limit was
inadvisable and preferred a larger margin to ensure it was not exceeded. Base'i on
the licensee's calculations the reactor would increase the pool temperature about
1*F every 3.5 minutes at full power without the secondary cooling system in
operation. Reactor pool and inlet temperature was logged by the operators every
two hours. A recent rod rundown was installed to prevent exceed.ng the limit
which is discussed in detailin Section 2.0, and associated operator procedure
guidance is discussed in Section 4.0. Observation of shift operations and
dircussion with shift operators indicated that they continued to be capable of
safely operating within prescribed guidelines.
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Preventive maintenance measures were recently added to the reactor maintenance
schedule and the building checklist involving the Reactor Building ventilation
damper air cylinder and air supply. Ventilation dampers were required to close
automatically on high radiation. The licensee had discovered that the cylinder
stroke was erratic during the freezing months and determined that excessive
moisture in the compressed air line to the cylinder was freezing. The air lines and
cylinder were covered with insulation and electric strip heaters were applied to the
cylinder. Air line moisture removal was added to the checklist at regular intervals.
Also, lubrication of the cylinder at regular intervals was added to the maintenance
schedule.
The reactor building experienced flooding to the beam port floor and the basement
when storm water from the adjacent campus grounds could not be
accommodated by the storm sewer system. The licensee concluded that blockage
that was found in the storm sewer caused the problem. Construction of multiple
new projects surrounding the Phoenix Memorial Lab had also reduced the soil area
to absorb runoff. A modification was approved to cap the storm drain off and to
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reroute nonradioactive drainage to the sanitary sewer. The licensee normally has
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not discharged any radioactive material to the sewer but in this case they did
because all of their storage capacity was filled by the flood water. Operation was
curtailed for about 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Further review of this event will take place during
the next inspection.
c.
Conclusion
The licensee was in compliance with the organization and staffing requirements.
Operators were aware of operating limits prescribed by the licensee. Operators
had varying responses to the procedure guidance discussed.
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2.0 Review, Audit, and Design Change Functions
a.
Insoection Scoce (40745)
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To verify that the licensee had established and conducted review and audit
functions required by the technical specifications and whether modifications to
the facility were consistent with 10 CFR 50.59 and technical specifications (TS).
b.
Observations and Findinos
The inspector reviewed modifications that were safety related.10 CFR 50.59
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required that the licensee perform a safety review before making changes to the
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facility. Any changes that affected the TS were required to be presented to the
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NRC before they were made to obtain necessary TS changes and modification
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approval as appropriate.
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(1) ~ 10 CFR 50.59 Review
The licensee had observed periodic primary cooling flow reductions. In response,-
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Ford Nuclear Reactor (FNR) Modification Request No.120ia primary cooling pump
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and motor replacement oombined with the pump discharge check valve internals
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removal was performed between April 8 and April 18,1996. It was presented to
the Safety Review Committee after completion since the licensee considered it a .
non-substantive modification.
- At the time of the installation the licensee determined in their 10 CFR 50.59
safety review that no TS were affected and no prior NRC approval or TS changes
were sought. The modification resulted in an increase in primary cooling flow rate
- of about 150 gallons per minuta (GPM). The licensee did not identify any adverse
effects to the reactor facility and normal operations resumed.
Operating Procedure 103 required that reactor inlet temperatures be maintained
within the prescribed band of greater than 90 and less than 116'F. On
October 4,1996, a shift crew observed that before automatic rundown actuation -
of the limiting safety system settings (LSSS) (required by TS 3.2) would occur at
a core outlet temperature of 129'F, reactor inlet temperature would have already
exceeded the TS safety limit (SL) of 116'F (all other SL and LSSS were satisfied).
On October 6,1996, the licensee documented the shift crew's observation and on
October 8,1996, confirmed that the LSSS may not prevent exceeding the SL
because of the increased flow rate and the resultant decrease in the temperatu're
rise across the core.
The inspector confirmed this assessment through review of the operating logs for
the period following the pump and valve modification. Reactor inlet temperriture
on October 4,19a6, was at 114.8'F while reactor outlet temperature was at
126.2'F. ' From this data the inspector concluded that, had the reactor pool
temperature been permitted to increase, the 129'F LSSS rod rundown would not
have prevented the 116*F SL from being exceeded. The licensee's safety analysis
- for the pump and valve modification failed to identify the modification's affects on
the TS LSSS rendering it ineffective in oreventing the reactor inlet SL from being
exceeded.
One root cause appears to have been the licensee's inadequate evalut!. tion of the
correlation between core flow and the reactor's safety limit of 116'F reactor
coolant inlet temperature and the 129 F LSSS core outlet temperatwe. Licensee
modification documentation did not address possible changes or idt/ntify any
changes in reactor cooling water temperatures. Another root causa may have
been that the SRC did not review and approve this modification prior to it being
completed.
The reactor has a SL and a LSSS of 900 GPM on the primary coolant flow rate.
Other applicable safety limits include: the 116*F reactor coolarst inlet
temperature,4.68 megawatts (Mw) reactor thermal power, and 18 feet of water
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above the top of the core. The basis of tlw safety limits was "that the calculated -
maximum cladding temperature in the bottom of the, hot channel of the most -
compact FNR core (25 elements) will not reach the boiling point of the water
' coolant."
As described in the Safety Analysis Report (SAR) Chapter 13, page 46, the LSSS'
set point was conservatively based upon a 2.0 Mw power level with a reactor
core temperature delta-T of 15'F, a 131'F reactor core coolant outlet
temperature, and an assumed 2'F instrument error.
For perspective, page 46 of the SAR conservatively concludes that "the reactor
can be safely operated at a power level of 2 Mw and a primary coolant flow rate -
as low as 164 gpm without exceeding the saturation temperature anywhere on
the fuel' plate cladding surfaces." Further, this section of the safety analysis
shows that calculated core outlet temperature could be as high as 151'F (more
than 20*F margin) before cladding' temperature would reach saturation
temperature with all other SL values at their TS limits.-
The risk of exceeding the reactor inlet temperature safety limit during the period
from April 18,1996, until corrective action to reduce the LSSS for reactor outlet
temperature was not high becau.se of operator procedure compliance. Procedure
guidance did require operation with pool temperature greater than 90 but less
than 116'F. Bulk pool temperature would not have changed rapidly. At 2.0 Mw
without secondary cooling, it would require about 3.5 minutes to increase 1*F.
Although the reactor was operated for several hours at greater than 114*F on
October 4,1996, the potential for exceeding the SL was minimal based on the
margins in the analyses and equipment setpoint conservatisms, procedural
requirements and operator training. However, the lack of full understanding of
regulatory requirements in the implementation of 10 CFR 50.59, and the
incomplete evaluation of the effects of the modification on tha system and the TS
are a concern. The licensee missed opportunities to identify the need for a TS
change during the original modification 10 CFR 50.59 evaluation and review, in
the SRC review, in post installation testing, and in five months of operation. This
is an apparent violation of 10 CFR 50.59 (a)(1) (VIO 50-002/98202-01).
(2)- Corrective Action
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On October 6,1996, the licensee documented a conversation with a member of
the shift crew regarding the possibility that the reactor outlet temperature rod
rundown LSSS may not have prevented the reactor inlet temperature from
exceeding the SL of 116*F. On October 8,1996, the licensee issued a temporary
operating instruction to shift crews instructing them that the LSSS for reactcr
outlet temperature must be reduced to 126"F to prevent.the 116*F SL from being
exceeded. The licenses stated that the delay between discussing the concern and
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taking corrective measures was due to evaluation of the operator's concern.
When the licensee finally discovered that the LSSS could not have performed as
designed, the reactor had been in that condition for more than five months. The
reactor had been operated continuously except for periodic maintenance and
refueling shutdowns.
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During an SRC meeting on April 29,1997, that discussed potential additional
considerations, the licensee discussed the possibility of restoring primary flow to
the original valuo which would have eliminated the problem with the LSSS set
point, but this was not considered further to date.
To permanently address the problem, FNR Modification Request No.123, was
designed to add a new rod rundown signal from the reactor core inlet instrument.
The TS SL for reactor inlet temperature is 116*F. The set point for the new
rundown was established at 114*F to avoid potentially exceeding the 116*F TS
SL considering an instrument error of 2*F. The modification was approved by the
SRC on June 30,1997, and installed on September 30,1997. At the time of
installation, the rod rundown set point for reactor outlet temperature was returned
to 129'F. Although this modification was meant to correct the problem with its
SL and the LSSS, the licensee safety review had determined no TS change was
required and NRC approval was not sought.
The licensee missed opportunities to identify the need for a TS change in several
subsequent design reviews after identification of the technical problem in October
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of 1996. However, it should be noted that the licenses took prompt and technical
sound corrective actions once the problem was identified in October of 1996.
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(3)
Reporting Requirernents
TS 6.6.2.b.2 requires that the licensee prepare a written report to the NRC and
forward it within 30 days when they discover any substantial variance from
performance specifications contained in the TS and SAR. As previously
described, the licensee identified the problem and took corrective action on or
about October 8,1996. The licenseo never sought to obtain from the NRC any
change to the TS LSSS nor was NRC notified of the problem. Failure to report the
observed condition after the October 1996 identification of the problem is a
Violation of TS 6.6.2.b.2 (VIO 50-002/98202-02).
A root cause was the apparent lack of understanding the licensee had of the
significance of the SL re!ationship to the LSSS as described in the SAR and
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required in the TS. The inspector determined by discussion with operators and
management that they did not fully understand the regulatory implications. The
Ass!stant Manager of Operations said he originally had the understanding that the
license safety limits needed to be met collectively but not individually. The FNR
has four safety limits: primary cooling flow rate, reactor coolant inlet
temperature, pool water level, and reactor power as previously discussed. His
understanding was based on the safety analysis and limits from the previous
licenso safety limits at the facility.
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Further, during conversations with three licensed operators, the inspector
identified that two of them were not aware of the actions required by 10 CFR 50.36 if a safety limit were exceeded although they were sensitive that it was a
license limit. The licensee committed to prepare a copy of 10 CFR 50.36 for
required reading by alllicensed operators.
(4) Other System Observations
The inspector observed that the pump and motor unit installation was not
consistent with the original design. The unit was strapped to the original bed
plate using metal banding material rather than with a bolted fastener arrangement
as originally installed. The motor electrical connector box cover was also missing.
Specifications for the original design were not available for review. The licensee
committed to evaluate these differences. This will be an inspector follow up item
(IFl 50-002/98202-03),
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Conclusions
The licensee failed to conduct a comprehensive safety review as required by
10 CFR 50.59 and f ailed to notify the NRC as required by TS when it identified an
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inadequate LSSS in accordance with the SAR analysis,
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3.0 Operator Requalification
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a.
Insoection Scoce (69003)
To determine that operator requalification activities and training were conducted
as required.
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b.
Observations and Findinos
All operators were actively enrolled in the requall'ication program according to the
licensee's approved program. Annual operating tests and written examinations
were acceptable and graded accurately.
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Conclusio_r1
Operators were evaluated in accordance with the requalification program
requirements.
4.0 Procedures
a.
Insoection Scoos (42745)
The inspection is to determine whether facility procedures met regulatory
requirements.
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' b.
Observations and Findings
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- The licensee operating procedure OP-103 " Reactor Operation, Maintenance, .
Systems, and Components," R32,05/09/95, step .12.10.5, permits bulk pool
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temperature operation at greater than 90*F but less than 116*F. ' The licensee
installed a modification which initiates a rod rundown at 114*F on September 30, .
1997. Operators questioned by the inspector regarding procedure guidance on
pool temperature were aware of the limits and the new actuation set point. The
inconsistency between the procedure and modified system is an inspector follow-
up item (IFl 50-002/98202-04).
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Conclusions
Operators were familiar with selected procedures, and associate'd equipment and
limits. An inspector follow up item on the consistency between equipment and
procedures was identified.
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5.0 Exit Meeting Summary
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.The inspector presented the inspection results to members of the licensee
management at an exit meeting on February 27,1998. The licensee acknowledged
the findings presented. : The inspector asked the licensee whether any material
examined during the inspection should be considered proprietary. None were
identified.
A second exit was conducted telephonically with the licensee on March 9,1998.
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Persons Contacted
University of Michiaan
+ 'R. Fleming
Director, Phoenix Memorial Laboratory
'B. Ducamp
Assistant Manager, Reactor Operations
'J.
Lee
Chair, Safety Review Committee
The inspector also contacted other technical and administrative staff personnel during the
inspection.
- Denotes those attending the exit meeting on February 27,1998.
+ Denotes those on telephone exit meeting on March 9,1998,
insoection Procedures Used
IP39745 Class i Non-Power Reactor OrgarJ2ation, Operation, and Maintenance
IP40745 Class 1 Non-Power Review, Audit, and Design Change Function
IP69003 Class 1 Non-Power Reactor Operator Requalification Training
IP42745 Class i Non-Power Reactor Procedures
items Ooened and Closed
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50-002/98202-01
10 CFR 50.59 review failed to identify affected TS
50-002/98202-02
Licensee failed to notify NRC of TS and SAR variances
50-002/98202-03
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Primary cooling pump not installed as original design
50-002/98202-04
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Licensee procedure guidance inconsistent with modification
List of Documents Reviewed
Administrative Procedures
Logs and Records
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Maintenance Procedures
Modification Records
Operating Procedures
Reactor Operating Licenses
Requalification Records
Safety Analysis Report
Safety Review Committee Minutes
Surveillance Procedures and Records
Technical Specifications
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List of Acronyms used
CFR
Code of. Federal Rogulations
DRPM
Division of Reactor Project Manrgement
Umiting safety system setting-
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NRC
Nuclear Regulatory Commission
Public Document Room
Safety limit
Safety Review Committee
TS
Technical Specification
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