ML20209F629
| ML20209F629 | |
| Person / Time | |
|---|---|
| Site: | Neely Research Reactor |
| Issue date: | 04/09/1987 |
| From: | Jape F, Long A, Menning J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20209F551 | List: |
| References | |
| 50-160-87-01, 50-160-87-1, NUDOCS 8704300335 | |
| Download: ML20209F629 (30) | |
See also: IR 05000160/1987001
Text
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SK880, UNITED STATES
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NUCLEAR RECULATORY COMMISSION
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3 g 101 M ARIETTA STREET, N.W., SulTE 2000
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Report No.: 50-160/87-01
Licensee: Georgia Institute of Technology
225 North Avenue
Atlanta, GA 30332
Docket No.: 50-160 License No.: C-97
Facility Name: Georgia Institute of Technology
Inspection Conducted: February 9-23, 1987
Inspectors: C. 8. b m . Y'7-F)
A. E. Long J Date Signed
h tf.Meng7nglw v'- 9- 2 9-
Q.' Date Signed
Approved by:
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F. gFtpe4 Chief
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Date Signed
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/* Test Programs Section
Engineering Branch
Division of Reactor Safety
SUMMARY
Scope: This routine, unannounced inspection included the areas of organiza-
tion, logs and records, review and audit functions, requalification training,
procedures, surveillance, maintenance, control of experiments, licensee events,
and closeout of open items.
Results: Six violations, three unresolved items, and three inspector followup
items were identified (Paragraph 2).
DESIC HTED ORIGINAL
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8704300335 870414
0 ADOCK 05000160
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REPORT DETAILS
1. Persons Contacted
Licensee Employees
- R. A. Karam, Director, Nuclear Research Center
- L. D. McDowell, Reactor Supervisor
W. H. Downs, Senior Reactor Operator
R. M. Boyd, Radiological Safety Officer
J. A. Mahaffey, Chairman Nuclear Safeguards Committee
M. F. Mercer, Electronics Engineer
D. L. Cox, Reactor Operator Trainee
- S. N. Millspaugh, Deputy Radiological Safety Officer
Other licensee employees contacted included members of the health physics
staff, other Nuclear Safeguards Committee members, faculty members,
students, and office personnel.
Nuclear Regulatory Commission
- D. M. Verrelli, Chief, Projects Branch 1
- A. R. Herdt, Chief Engineering Branch
L. S. Mellen, Project Engineer
- S. J. Vias, Project Engineer
- G. B. Kuzo, Senior Radiation Specialist
- Attended exit interview on February 23, 1987
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l 2. Exit Interview
l The inspection scope and findings were summarized in an exit interview on
February 23, 1987, with those persons indicated in paragraph 1 above and
in a subsequent telephone conversation on April 13, 1987. The inspector
described the areas inspected and discussed in detail the inspection
findings. No dissenting coments were received from the licensee.
The following new items were identified during this inspection:
VIO 160/87-01-01: Failure to Provide or Utilize Procedures (Para-
I graphs 3.a, 5.d, 6.c, 7.a. 7.d)
VIO 160/87-01-02: Failure to Control Experiments per Technical
Specifications (Paragraphs 5.d, 9)
VIO 160/87-01-03: Failure to Perform Weekly Heat Balance Surveil-
lance (Paragraph 7.d)
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VIO 160/87-01-04: Change Made to Facility, Involving Technical
Specification Change, without prior NRC Approval (Paragraphs,5.b and
10 d)
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VIO 160/87-01-05: Failure to Comply with Approved Requalification
Program (Paragraph 11)
VIO 160/87-01-06: Failure to Perform NSC Review'and' Audit Functions
per Technical Specification 6.3 (Paragraphs 8.a,9, 10.f. 10.g. 10.h) x
UNR 160/87-01-07: Formalize and Impicment Methodology to Control
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Temporary Changes to Procedures per Technical Specification 6.4.a
(Paragraph 6.a.1)
IFI 160/87-01-08: Formalize and Implement Methodology to Track
Limiting Conditions for Operation (Paragraph 6.a.2)
UNR 160/87-01-09: Verify that Shim : Blade Positions are Adequate to
Ensure Negative Trip per Technical Specification 3.1.d (Paragraph
7.a) \,
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UNR 160/87-01-10: Verify that Instrunent Calibrations Necessary for
Operability of Safety Equipment in Te:hnical Specification Table 4.1
are Performed (Paragraph 7.c) s
IFI 160/87-01-11: Followup on General State of GTRR equipment
_(Paragraph 13)
IFI 160/87-01-12: Resolve verification of 10 CFR 74 Requirement that
Spent Fuel have Self-Protecting Radiation Levels (Paragraph 12)
The licensee did not identify as proprietary any of the materials provided
to or reviewed by the inspectors during this inspection.
3. Licensee Action on Previous Enforcement Matters -
The inspector reviewed licensee corrective actions on the following
violations:
a. (Closed) VIO 85-02-01 Failure to Provide & Utilize Adequate Surveil-
lance Procedures per Technical Specification 6.4.b
Technical Specification 6.4.b required that written procedures be ' '
provided and utilized for surveillance and testing. Violation
85-02-01 included three . examples of inadequate procedures or failure
to follow procedures.
The licensee had previo; sly committed to the following corrective
actions in a letter dated Septecher 9,1985, from R. A. Karam to
R. Walker:
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1. Revise procedure-2006, " Weekly Reactor Shutdown Checklist," to
verify and record D20 level alarm points.
2. Revise procedure 7203, "ECCS - Monthly Surveillance " to include
verification of actuation of the low level alarm. .
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3. Revise procedure 7202, " Control Rod Drop Time" to allow either a
universal counter or oscilloscope for measuring control rod drop
time. ,
TheinspectorverifiedthattheaboveprocedureshadbeenrevtNedand
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adequately implemented. +
The following additional corrective action requirements were stated
in a letter from R. Walker to R. A. Karam dated October 7, 1985:
1. Interview all licensed operators to identify and correct any
additional examples' of inadequate or inaccurate procedures.
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2. Retrain operators on the methodology for and the necessity of ,
effecting required temporary procedure changes prior to
procedure utilization. / 8
The fonnal interviews with operators committed to by the licensee to
identify procedural inadequacies were not conducted, although a
number of procedure revisions were subsequently made in addition to
those made in response to the violation. These revisions were made
in response to problems identified by operators or through audits.
However, during the inspection it was determined that procedural
inadequacies were known to at least one operator and had not been
corrected (Paragraph 7.a).
To implement retraining on the methodology and necessity of temporary
procedure changes, operators were instructed to have temporary (pen
and ink) changes to procedures incorporated into official revisions.
The inspector identified a lack of provisions for effecting temporary
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i procedure changes prior to procedurequtilization. The licensee
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concurred that their policies on temporary changes should be defined
and formalized (Paragraph 6.a).
- Violation 85-02-01 is being closed on the basis that the corrective
! actions which were not completed will be tracked as part of the
- follow-up of violation 87-01-01 identified during the inspection
g (Paragraphs 5.d,6.c,7.a.7.d).
b. (0 pen) VIO 85-02-02 Failure to Meet Scheduled Surveillance Require-
ment per Technical Specification 4.2.A, Table 4.1.
! This violation involved failure to schedule and perform Procedure
t 7141, "LI-D1 Check," in accordance with the requirements of Technical
l Specification 4.2.a and Table 4.1 which jointly require semiannual
! calibrations.
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The Licensee had modified procedure 7241 to require semiannual
- calibration of the reactor D20 level channels. This revision was
l approved September 13. 1986, and was verified by the inspector to -
! adequately address the issue. The inspector also verified that
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, ' e completion of the surveillance on a semiannual frequency was up to
. date. This aspect of the corrective action was complete.
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E In addition to the commitment to revise procedure 7241, the Licensee
stated in a September 9, 1985, letter to the NRC that the GTRR staff
was reviewing all license requirements towards full compliance. This
review was performed during 1985. However, several licensee findings
were not corrected including failures to verify certain limiting
conditionsforoperation-(Paragraph 7.a).
An additional example of noncompliance was the failure to obtain a
Technical Specification change prior to changing the cover gas froc.i ,
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helium to nitrogen (Paragraph.10.d).
Violation 85-02-02 will remain open until review of the licede '
conditions for compliance has been completed by the licensee and
determined to be adat:ste by the NRC during a future inspection.
c. (0 pen) VIO 85-02-04 Numerous Errors in Facility Drawings in Violation
of Technical Specification 6.5.b.6.
Technical Specification 6.5.b.6 requires the licensee to retain
updated, corrected, and as-built facility drawings for the life ,
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This violation involved errors in system flow
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of the facility.
diagrams, including components appearing on drawings but not actually
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3 installed in the facility, and piping system interconnections not
1 fonsistent with the drawings.
In the September 9,1985, response to this violation, the licensee
connitted to comparing the drawings to the as-built facility, and
correcting any discrepancie:. Additionally, the licensee committed
to developing procedures for documenting authorized modifications to
the facility.
Development of procedures for documenting authorized modifications
to the facility was satisfactorily completed by the licensee.
Procedure 4200, " Changes in Facility Design," was approved by the
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Nuclear Safeguards Committee on April 9,1986. This procedure
1 specifically requires all changes to the facility to be documentede
p including the updating of drawings as appropriate.
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The licensee stated that the GTRR flow diagrams had been updated and
verified through walkdowns, but that the electrical drawings had not
yet been reviewed.
Violation 85-02-04 will remain open until the accuracy of all
categories of facility drawings can be reviewed in more detail during
t a future NRC inspection.
d. (Closed) VIO 85-02-05 Reactor Operator Received Less than 80% on
1984 Requalification Examination and Licensee failed to initiate
retraining per the August 5, 1974, approved program.
This violation involved not ' satisfying the requirement of the 1984
NRC approved requalification program that an operator be retrained in
any areas of the requalification examination in which he failed to
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score 80%.
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Inspection Report 85-02 stated that as soon as the problem was
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identified by the inspector, the licensee initiated corrective
action. The operator successfully passed all areas of the requali-
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fication examination in August 1985 and again in September 1986.
Licensee corrective actions in the areas of annual requalification
examinations and ongoing operator training were determined to be
adequately implemented.
The licensee stated in the September 9,1985 letter to R. Walker that
retraining had been ' initiated on a permanent basis, beginning in
July 1985. Retraining in radiation protection commenced September 9,
1985, and continued for four weeks. The inspector verified that
ongoing weekly training sessions were being conducted (Para-
graph 11.d).
As additional corrective action for VIO 85-02-05, the NRC letter
of October _7,1985, from R. Walker to R. A. Karam instructed the
licensee to continue to follow the approved Requalification Program
per 10 CFR 50.54.1-1, in conjunction with the enhanced retraining
addressed in the licensee's September 9,1985, letter.
The licensee did not satisfy this aspect of corrective action for
VIO 85-02-05, in that reactivity manipulations and performance
evaluations were not being documented as required by the requalifi-
cation program and by 10 CFR 55 Appendix A. (Paragraph 11.c).
Violation 85-02-05 is being closed on the basis that the incomplete
corrective action will be tracked with follow-up of VIO 87-01-05.
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e. (Closed) VIO 85-02-06 Failure to Hold Quarterly Meetings of the
Nuclear Safeguards Committee as required by T. S. 6.2.c.
The inspector reviewed meeting records and verified that the
committee had met at the required frequency during the period
June 1985 and February 1987 (see Paragraph 10.a). The licensee had
supplemented the normal reminders to perform T. S. requirements on
time with the inclusion in the NSC minutes of the date before which
the Committee must next meet to stay in compliance.
f. (Closed) DEV 85-02-07 Failure to Train Operator Candidates to operate
the reactor in a competent and safe manner as certified on the NRC
Form 398.
In response to this deviation, the licensee made the commitment to
develop a formal training program for reactor operator license
applicants. This program was to include (1) Training in the areas
outlined in 10 CFR 55 and (2) Administering a written and oral
examination to each applicant before certifying to the NRC that he or
she was ready to take the NRC's examination. Implementation was
expected to take about a year.
At the time of the inspection, the facility had one operator license
trainee, who had been previously licensed at the facility and was
working under the direction of licensed personnel. This operator
trainee had been attending weekly training sessions, but formal
license candidate training per 10 CFR 55 was not being conducted.
Deviation 85-02-07 was closed on the basis that the licensee will
have conducted the required training for any operator license
candidates before certifying their readiness for the NRC examination.
Records documenting training were to be maintained.
4. Unresolved Items
Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or
deviations.
Three unresolved items identified during this inspection are itemized in
Paragraph 2, accompanied by the paragraph numbers in which they are
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addressed.
5. Organization, Logs and Records (39745)
The objectives of this portion of the inspection were to (1) Ascertain
whether the licensee's organization was as delineated in the Technical
Specifications and (2) Ascertain whether the licensee's records and logs
were maintained according to regulatory requirements.
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a. Organization
The facility organization for management and operation of the reactor
was compared to the requirements of Technical Specification 6.1. No
problems were identified.
Operations console logs from March 1986 through February 1987 were
reviewed to verify that minimum crew composition requirements of
Technical Specification 6.1.d were satisfied. The inspector
discussed responsibilities with operators, and confirmed their
understanding of Technical Specification 6.1.e, which requires a
licensed operator to be present at the controls unless the reactor is
shutdown.
The inspector was informed of plans to reorganize the facility staff.
The licensee stated that they will be requesting a change to the
Administrative Section of Technical Specifications.
b. GTRR Annual Reports
The annual reports for the GTRR covering 1985 and 1986 operations
were reviewed and found to meet the requirements of Technical
Specifications with one exception.
Technical Specification 6.7.a(1) requires that annual reports include
changes to the facility. The annual report covering 1986 did not
report the change of the cover gas from helium to nitrogen as a
change to the facility. This was identified as a portion of
VIO 87-01-04 (Paragraph 10.d).
c. Retention of Records
Technical Specifications 6.5.a and 6.5.b require that certain records
be retained for five years and other records be maintained for the
life of the facility. During the course of the inspection, the
licensee demonstrated that records in each category were on file and
were readily retrievable.
d. Console Logs
The inspector reviewed console logs #29, #30, and #31, which covered
the period October 1985 to February 1987.
The inspector verified that logbook entries for shutdown were being
entered and initialed, and the following Inspector Followup Item was
closed:
(Closed) IFI 85-02-03: Console Log Not Initialed for Significant
Log Entries when Reactor is Shutdown per Licensee Commitment in
IE Report 50-160/84-01.
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The following violations were identified during review of the console
logs:
(1) Missing Initial Condition Stamps - Procedure. 2000, " Reactor
Operation" stated, " Log the following IC data: Each shim blade
position, reg rod position, flux am
settings, and picoammeter readings."pThis readings and gain
was normally pot
done
by filling in an initial condition stamp. The inspection
identified numerous examples where this ' stamp was missing from
the log, including operations on May 19, 1986, May 21, 1986,
and May 23, 1986. The licensee concurred that according to
Procedure 2000, these initial condi.tions should have been
logged.
(2) _ Missing Equilibrium Condition Stamps - Procedure 2000 further
stated, "After thermal equilibrium is established,- log
equilibrium operating date." The licensee stated that this
requirement referred to a logbook stamp, and that the reactor
must generally operate for one half hour to reach equilibrium
conditions. The inspector observed that the stamp was rarely
used during 1986, except when heat balances were performed, even
when the reactor operated for periods of longer than an hour.
This item had been previously b.ought to the attention of the
licensee as UNR 82-01-01, and the licensee's response included a
commitment to use the stamp. The licensee concurred that the
stamp should have been used regularly.
(3) A number of additional missing entries were identified in the
logbooks reviewed. Some examples were:
(a) Failure to log removal of an experiment from experiment
facility V-24 on March 21, 1986.
(b) Failure to completely fill in the critical condition stamp
on April 28, 1986.
(c) Missing Rabbit Experiment number on June 18, 1986.
(d) No check to indicate who filled out the log, page 71 of Log
- 29.
(e) Initial condition stamp not completely filled out, page 133
of Log #29.
The findings listed above were collectively identified as example 1
of the following violation:
(0 pen) VIO 87-01-01: Failure to Provide or Utilize Procedures
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The inspector also noted logbook entries which stated that engineer-
ing class lab experiments were performed, but failed to provide any-
information as to the nature of these experiments (Paragraph 9).
On pages 125,131,139, and 147 of Log #29, no dose rates were -
recorded for the rabbit runs. This was identified as example 1 of
the following violation:
(0 pen) VIO 87-01-02: Failure to Control Experiments per
6. Procedures (42745)
The objectives of the review of GTRR procedures were to ascertain (1)
whether an effective procedure control system had been implemented and
(2) whether the content and scope of the facility procedures were adequate
to control safety-related operations.
a. Administrative Controls
1. Temporary Changes to Procedures
The inspector observed that provisions for making temporary
changes to GTRR procedures have not been formally defined and
documented.
Technical Specification 6.4.a requires all procedures and major
changes thereto be reviewed and approved by the Nuclear
Safeguards Committee prior to being implemented. Changes which
did not alter the original intent of a procedure may be approved
by the Reactor Supervisor. Such changes shall be recorded and
submitted periodically to the Nuclear Safeguards Committee for
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routine review.
No GTRR procedure implements this Technical Specification
requirement or provides details on how the requirement is to be
satisfied.
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Consequently, current and previous violations for inadequate
procedures and failures to follow procedures have resulted at
, least in part from the lack of a defined methodology for making,
j approving, and documenting temporary changes. In addition,
the corrective action commitment for VIO 85-02-01 to retrain
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operators on the methodology for and the necessity of effecting
required temporary procedure changes was not completed (Para-
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graph 3.a).
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The licensee told the inspector that the policy on temporary
changes was that all pen and ink changes were to be incorporated
into permanent revisions. Further questioning revealed that
there was no defined policy for temporary changes which were not
intended to become permanent. The inspector determined that
temporary changes to startup checklists were usually documented
as handwritten notes on the procedure and as comments under
" Abnormal Conditions." The Reactor Supervisor approved these
" temporary changes" by signing the startup approval. Other
situations which possibly amounted to temporary procedure
changes were noted in the console logs. The licensee could not
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identify provisions for documenting deviations from steps or
requirements in procedures which were not completed in checklist
format and filed. When asked what they would do if they were
performing such a procedure and came upon a step which was
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impossible to perform at that time due to unusual circumstances,
several licensee personnel told the inspector that they would
" caucus and do something equivalent."
The inspector determined that pen and ink changes to the daily
and weekly startup checklist had been made for a year and a
half, and in several instances were not made as necessary
(Paragraph 7.d)..~
The licensee agreed to formalize and implement a methodology for
making and documenting temporary changes to procedures. this
methodology is to include measures to document changes made to
procedures which will not be made permanently. Temporary
changes authorized by the Reactor Supervisor per Technical
Specification 6.4.a should be Reviewed by the Safeguards
Committee. Committee approval of significant temporary changes
prior to implementing the change should also be documented.
The thrust in developing the methodology for temporary changes
should be prevention of inadequate procedures or failure to ,
follow procedures. -
- The importance of formalizing the requirement for verbatim
compliance with procedures was also discussed with the GTRR
staff.
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Followup in this area will be identified and tracked as:
(OPEN) UNR 87-01-07: Formalize and Implement Methodology to
Control Temporary Changes to Procedures per Technical Speci-
fication 6.4.a
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2. Limiting Conditions for Operation
The inspector observed that there was no procedure for
documenting when the GTRR was operating in an LCO condition.
Most of the Technical Specification LC0 time clocks were eight
hours. The licensee stated that if a piece of equipment went
out of service during operation, and they had redundancy, they
would continue to operate for the remainder of the day, up to
eight hours. If the equipment was Technical Specification
related, it would be repaired before starting up again.
Normally, equipment failing or being taken out of service was
entered in the console logbook, but the inspector was aware of
at least one case where that was not done.
This item will be tracked as (0 pen) IFI 87-01-08: Formalize and
Implement Methodology to Track Limiting Conditions for Operation
b. Procedure Adequacy
Selected procedures for surveillance, maintenance, and control of
experiments were reviewed for technical and administrative adequacy
(Paragraphs 7, 8, and 9).
c. Procedure Compliance
Several reactor operations procedures were discussed with the
operators to determine whether or not the provisions of the
procedures were being followed.
Review and discussions of the following procedures revealed no
problems or noncompliance. The operator interviewed demonstrated a
knowledge of the provisions and intent of these procedures:
1) Procedure 2250, " Shield Coolant System Operation"
2) Procedure 2300, " Bismuth Coolant System Operation"
Procedure 2250 requires that the shield coolant pump remain on for
eight hours following operation above 1 MW. The inspector verified
that this requirement was followed after operation at 2.3 MW on
2/3/87.
Noncompliance with a third procedure discussed with the operator
was identified. Procedure 2210, " Cooling Tower Special Operation"
requires that water be run through the cooling tower for at least an
hour each week during periods when the reactor was shut down. No
records documented that this was done when the reactor was not
operating between 3/31/86 and 4/14/86, and between 7/24/86 and
8/19/86. This item was identified as Example 2 of VIO 87-01-01.
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The inspector id.entified that as of February 23, 1987, the following
procedures had not been modified to address the fact that the cover
gas had been changed from helium to nitrogen in mid-1986:
Procedure 2350, " Blanket Gas System Operation" 5/2/74
Procedure 2400, " Gas Recombiner System Operation, 5/2/74
It is important to maintain and utilize up-to-date procedures, and a
delay of this length represented inadequate procedures. This finding
was identified as Example 3 of VIO 87-01-01.
d. Control of Revisions
The inspector noticed that copies of Procedure 4200 (Approved 4/9/86)
and Procedure 3100 (Approved 7/11/86) were not included in the
control room set of procedures. Up-to-date versions of three
additional procedures were missing from the set belonging to the
reactor supervisor. This problem was corrected while the inspector
was on site.
The inspector verified that the versions actually being performed
were up to date.
7. Surveillance Testing (61745)
The objective of the surveillance segment of the inspection was to
ascertain whether the licensee's surveillance program was adequate and
conducted in accordance with Technical Specification requirements.
Surveillance was considered to be the verification of set points and
conditions required by the Technical Specification. This included values
or conditions specified or assumed by the Technical Specification that had
no assigned surveillance frequency yet had a definite bearing on safety.
The overall purpose of the surveillance inspection was therefore to verify
that the intent of the Technical Specification surveillance requirements
were met.
a. Review of Technical Specification Requirement Implementation
The purpose of this portion of the inspection was to confirm that
each safety-related item identified in the facility Technical
Specifications was being verified as required.
In the September 9, 1985, response to Violation 85-02-02, the
licensee documented a commitment to review all requirements of
the GTRR license and provisions for compliance. This review was
performed by Mr. William Downs, Senior Reactor Operator, in
September 1985.
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The. inspector reviewed in detail with Mr. Downs the information
.which he had compiled, and found the study to be comprehensive and
thorough. However, not all of the findings of the 1985 study had
been corrected at the time of this inspection.
The Technical Specification study identified the following limiting
conditions for operation which were not being verified at- the GTRR:
(1) Excess Reactivity - Technical Specification 3.1.e required as a
limiting condition for operation that the excess reactivity of
the core be limited to 11.9% delta K/K. The purpose of this
requirement was to ensure that the reactivity worth of any two
of the four available shim blades would be adequate to scram the
reactor. This Technical Specification was added in 1978, but no
verification of excess reactivity was included in GTRR proce-
dures. The licensee confirmed that no specific calculation of
excess reactivity was being performed.
Licensee management verified for the inspector that the excess
reactivity was currently within Technical Specification limits,
and agreed to incorporate verification of the excess reactivity
limits into a procedure.
(2) 02 in helium sweep - Technical Specification 3.6.e required the
D2 concentration in the helium sweep system to be less than 2%
by volume. The licensee confirmed to the inspector that this
requirement was not being verified.
(3) Building Isolation Time - Technical Specification 3.5.b.6
required that the time from initiation of closure to isolation '
valve closure not exceed five seconds.
The licensee told the inspector that this test used to be
performed, but was somehow omitted from the version of Procedure
7200 dated July 1981. Therefore the test was not being
performed. The licensee committed to modifying procedure 7200
to include verification of isolation valve closure time.
These three items have been identified as examples 4, 5, and 6 of VIO
87-01-01.
Additional Technical Specification limiting conditions for operation
which were not addressed in procedures were also identified:
(1) Secondary System Radioisotopes - Technical Specification 3.6.f
required the concentration of radioactive materials in the
secondary coolant system to be less than the values listed in
10 CFR 20, Appendix B, Table II, Column 2.
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. _ _ _ _ _ _ _ - - _ - - - _ _ _ _ _ _ _ . _
a e a
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14
The licensee was only measuring secondary tritium, and explained
to the inspector that they assumed the tritium measurement was
the most sensitive indicator of a primary to secondary leak and
therefore sufficient. The inspector questioned the adequacy of
using tritium to meet the requirements of Technical Specifica-
tion 3.6.f. The adequacy of the licensee's leakage detection
methodology will be reviewed by Region II in conjunction with
IFI 87-02-02, which addressed the adequacy of measurement
methodology for waste stream sampling.
(2) Shim blade minimum position - Technical Specification 3.1.d
stated " Prior to criticality each shim-safety blade which is
withdrawn above full insertion shall be positioned so that a
free fall of the blade towards its full inserted position will
result in a reactor scram activated by a negative period scram."
The licensee could not identify for the inspector any procedure
where this Technical Specification requirement was addressed.
The shim blades were normally positioned above 10 degrees. The
licensee could not identify for the inspector where it had been
established and documented that this position was sufficient to
ensure a negative period scram.
Item 2 was identified as an unresolved item which will be closed when
the licensee verifies and adec;uately documents that the shim blade
worth at 10 degrees is suffic"ent to ensure a negative trip. This
iter. will be tracked as:
(0 pen) UNR 87-01-09: Verify shim blade position ensures
negative trip per Technical Specification 6.4.a
b. Review of Surveillance Procedure Adequacy
Technical Specification 6.4.b.7 requires that surveillance and
testing of safety related systems be conducted in accordance with
procedures which had been appropriately reviewed and approved. This
specification implied adequate procedures. The purpose of this
portion of the inspection was to determine whether or not the GTRR
surveillance procedures were adequate to accomplish the intended
purpose.
The following GTRR procedures were reciewed for technical adequacy
and no problems were identified:
(1) Procedure 7203, September 13, 1985, "ECCS-Monthly Surveillance"
(2) Procedure 7222, July 17, 2982, "ECCS Semiannual and Annual
Surveillance"
,
, - - , - - - . . , - , - - - - - - - - , - - - . , - -
, - - - - , - - - - . ---,.n..._ . - - , -
. . .. _ _. ._ _ - ._. . _ _ _ . .
,
'
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15
(3) Procedure 7202, December 10, 1985, " Control Rod Drop Time"
(4) Procedure 7220. July 17, 1981, " Building Isolation Test"
The reviews of Procedures 7203, 7222, and 7220 included walk-
throughs.
4
The portions of the following procedures concerning the testing of
period trips were reviewed for technical adequacy and walked through:
- (1) Procedure 2002, December 10,1985, " Weekly Precritical Startup
Checklist" ,
i~
(2) Procedure 2003, December 10,1985, " Daily Precritical Startup
.
Checklist"
Technical Specification 3.2.a and Table 3.1 require operable
positive and negative period trip channels with setpoints equal to or
more conservative than ten seconds. Technical Specification 4.2.a
,
and Table 4.1 required these period trips to be tested prior to
startup and calibrated semiannually.
, Period trips were being tested in the daily and weekly startup
checklists by injecting a signal and. observing that a trip occurred.
1
No problems were identified with the adequacy of the positive period
trip test, but the scale of the control board negative period meter
, only went down to thirty seconds. Thus, existing instrumentation was
inadequate to show that the trip actually occurred at a period equal
'
to or more conservative than the Technical Specification limit.
Many of the items on the daily and weekly checklists were only check
3 marked to indicate acceptable results without the actual setpoint
being listed. The operators stated that they know the setpoints and
,
make mental notes that the trips occur at the correct points, or that
<
other parameters being verified are acceptable. The inspectors
discussed with the licensee the possibility of identifying the
-
expected values of additional parameters on the checklists to aid in
! identifying either instrument drift between calibrations or other
degradations. '
c. Instrument Calibrations
i Technical Specification Table 4.2 specifically requires certain
! safety-related instrumentation to be calibrated at specified '
l frequencies. Calibration of various other instruments is included in
the Technical Specification Table 4.1 requirements for calibrations
- of safety systems.
,
'
]
1
- . - - - - - - . . - . - _ . . , . - . . , - . . - _ , . , _ , . - - , - - _ - - .
- , . . - .
.
.,#
-
,
.
16
The inspector reviewed documentation of completed instrument
calibrations and verified that the - period trips, picoammeter
channels, and D20 temperature channels had been calibrated semi--
annually during 1985 and 1986 as required by the Technical Specifica-
tions.
Technical Specification 4.4.b requires the resistivity of the primary
coolant to be measured weekly. The involved sensor CRAD 1, was not
being calibrated. In this case calibration would have been very
difficult to accomplish.
Discussions. with the licensee indicated that calibrations of certain
-senors or other instruments, which are part of the safety systems in
Table 4.1, had been performed at one time but were possibly no longer
being performed.
The licensee agreed to review the adequacy of calibrations of
instruments not specifically addressed in the Technical Specification
Table 4.2 yet required for the functioning of the safety systems in
Table 4.1. This area will be reviewed in more detail in a future
inspection and will be tracked as:
(0 pen) UNR 87-01-10 Verify Instrument Calibrations Necessary for
Operability of Safety Equipment are Performed.
d. Audit of Surveillance Test Results
Records of selected surveillance testing for 1986 were reviewed to
verify that the tests were conducted within the required frequency
-and that the results of the tests were acceptable.
Records reviews for the following procedures identified no problems.
The test were performed at the required frequencies and the results
were acceptable:
(1) Procedure 4000, " Containment Building Pressure Test"
(2) Procedure 7203, "ECCS Monthly Surveillance"
(3) Procedure 7220, " Control Building Isolation System Test"
(4) Procedure 7223, "ECCS Tank Level Calibration Check"
(5) Procedure 7226, " Annual Scram Signal Delay Time"
(6) Procedure 7241, " Instrument and Reactor fank Level Maintenance
and Surveillance and Calibration Check"
The inspection findings were as follows:
!
(1) Procedure 2002, " Weekly Precritical Startup Checklist", and
- Procedure 2003, " Daily Precritical Startup Checklist"
l The inspector noted several omissions on Procedure 2002 in
, recording equipment which was out-of-commission, and equipment
'
substitutions. Except for a few occasions, the Universal
.
. <
.
17 ,
Counter was out of commission from May 1985 through the time
of the inspection, February 1987 (a period of over a year and
a half). This was usually properly noted under " abnormal
conditions," and steps involving the Universal Counter were
marked with an asterisk or "00C." Frequently, a "5216A" was
substituted for the Universal Counter and used to take
discriminator data. This was usually properly noted. However,
instances occurred when the operators failed to write in as
an abnormal condition that the Universal Counter was out of
commission, cr recorded discriminator data without noting that
the 5216A had been substituted. Examples of these errors
occurred on March 3, 1986, February 24, 1986, and June 10, 1986.
The inspector had similar findings in the review of Procedure
2003. The licensee stated that a permanent change to these
procedures was not made because they eventually planned to
restore the Universal Counter to operation.
Because the temporary changes had occurred for over a year and
a half, and because omissions were made in the handwritten
entries, this was deemed an inadequate procedure. This was
identified as example 7 of VIO 87-01-02.
(2) Procedure 2015, " Reactor Power Calibration Data Sheet,"
Technical Specification 4.2.b requires weekly calibration of
reactor power to a heat balance when the reactor was operated at
a power level at or above one megawatt.
The inspector observed that the reactor was operated briefly at
1 MW on February 4,1986, and for approximately nine hours at 1
MW on February 7,1986, without the weekly heat balance being
performed. The console log and completed power calibration data
sheets showed that heat balances had been performed January 22,
1986, and February 11, 1986. The licensee concurred that a
required heat balance had been omitted.
Failure to perform the weekly heat balance surveillance as
required by Technical Specifications was identified as:
(0 pen) VIO 87-01-03: Failure to Perform Weekly Heat
Balance Surveillance
8. Maintenance (39745, 40745, 61745)
The purpose of this portion of the inspection was to verify adequate
performance, control, and documentation of maintenance activities.
, <
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.
18
a. Administrative Controls
Technical Specification 6.4.b.5 requires that preventive or correc-
tive maintenance which could affect safety be conducted by written
procedures.
Technical Specification 6.4.a required all procedures and major
changes thereto to be reviewed and approved by the Nuclear Safeguards
Committee prior to being effective.
The following procedures were reviewed to determine if the licensee
had established adequate procedural measures to control and document
maintenance and facility modification activities:
-
Procedure 4900, February 4,1985, " System Worksheet"
-
Procedure 4901, January 25, 1977, " Preventive / Corrective
Maintenance on Safety Related Equipment"
-
Procedure 4200, April 9, 1986, " Changes in Facility Design"
Procedure 4900 tracked completion of surveillance and maintenance
activities, Procedure 4901 provided specifications for preparing
written " Job Plans" and Procedure 4200 documented changes in Facility
Design and provided guidance with respect to 10 CFR 50.59 evalua-
tions.
Although procedure 4901 contains administrative requirements for
preparing job plans, there was no record that the written job plans
were being rev'ewed or audited by the Safeguards Committee to satisfy
the requirements of Technical Specifications 6.4.a and 6.4.b.5. This
item will be tracked as as example 1 of:
.
(0 pen) VIO 87-01-06: Failure to Perform Adequate NSC Reviews
and Audits per Technical Specification requirements.
.
'
The Inspector also noted that the licensee seldom identified the
specific job plan used for a maintenance activity on the corre-
,
sponding system worksheet. This made records difficult to trace. To
provide more traceable records, the licensee should identify
associatedjobplansonsystemworksheets.
The inspector discussed administrative controls on instrument
setpoints with the licensee. Calibrations were performed per vendor
manuals as allowed by Technical Specifications. The actual setpoints
were frequently not recorded on the system worksheets, which merely
stated that the setpoint was "0K." Most of the setpoints ap] eared in
the Technical Specification but those such as the period trip, which
differed from the Technical Specification setpoints, were not
tabulated in a procedure.
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_ _ _ _ _ _ _ __ _
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19
b. Maintenance Review
Naintenance records on the following items identified in the control
room log were reviewed for compliance with procedural and Technical
Specification requirements. No problems were identified.
(1) Recorder Maintenance, 86-1006-352, October 24, 1986
(2) Repair of Flux Amp, 86-1007-359, October 7, 1986
(3) Repair of Reactor Tank Low Level Channel #2, 86-0212-086,
February 14, 1986
(4) Repair of Gas and Water Recorder, 86-0217-083, February 17, 1986
(5) Kanne Detector Maintenance, 86-0225-141, February 26, 1986
(6) Kanne Detector Maintenance, 86-0507-186, May 8, 1986
(7) D20 Flow Recorder Maintenance, 86-0610-295, June 10, 1986
The following selected job plans were reviewed to verify that the
were prepared in accordance with Procedure 4901, " Preventive / y
Corrective Maintenance on Safety-Related Equipment." No problems
were identified.
(1) " Cooling Tower Mechanical Preventive Maintenance"
(2) " Job Plan for Overhauling Drain Valve Air Actuators"
i
(3) " Job Plan for Regulating Road Maintenance"
- (4) " Replacement of Vent System Low Flow Stack Switch"
'
(5) " Job Plan for Emergency Airlock"
(6) " Removal of Valve Operator on Outlet Reactor Isolation Valve"
l
!
(7) " Job Plan for Truck Door Gasket"
! (8) " Job Plan for Cleaning Bismuth Secondary Flowmeter"
l
! The following System Worksheets were reviewed to determine if the
l analyses of facility design changes were 3erformed as required by
l Procedure 4200, April 9,1986, ' Changes < n Facility Design." No
problems were identified.
l
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20
(1) 85-0831-392, Reactor Isolation Valves
(2) 85-0601-308, Thermal Column Air
(3) 86-0106-061, Containment Exhaust Stack Fan
(4) 86-0623-299, Picoammeter #2
(5) 86-0214-087, Picoammeter #2
(6) 86-0911-357, Magnetic Actuator Amplifier
(7) 86-1007-359, Flux Amplifier #1
(8) 85-0416-142, Airlock Doors
(9) 85-1105-389, Airlock Doors
(10) 85-0327-101, Air Dryers
9. Experiments (69745)
The insaection objective with respect to reactor experiments was to
ascerta" n whether experiments were conducted safely and in accordance with
regulatory requirements. The term " experiments" included such activities
as incore irradiations, beam port irradiations, and class exercises or
demonstrations.
The inspector reviewed the Technical Specification and procedural
requirements and determined that the files for each experiment, should
include: 1) An Experiment Approval Form; 2) An Experiment Schedule Form
for each time the experiment was performed; and 3) An Experimentor's
Checklist for each time the experiment was performed after the procedure
was implemented in mid 1986. In addition, records of each experiment
should have been accurately entered on the data sheet of Procedure 2012,
" Operating Log Experiment Status."
The inspector reviewed the files of the experiments performed during
1986 to verify that the Experiment Approval Forms, Schedule Forms, and
Checklists had been performed as required. The following problems were
identified:
(1) The GTRR form " Request for Minor Experiment Approval," requires a
copy of the calculations of estimated activities of principal
i isotopes to be attached. Numerous copies of this form were on file
,
. for 1986 without attached calculations of estimated activities.
.. .- ---. . - . - _ - , __
- _-__. - . _ _.
. __ __ , __ _ _ - _. __
e
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21
4 .
(2) Procedure 3102 requires that an experiment Schedule Form be completed
and retained in the files each time an experiment is performed.
l Experiment schedule forms were frequently not completed and filed for
runs in the pneumatic facility or for Nuclear Engineering class
j laboratory experiments.
(3) No "Experimentor's Checklist" form, which is required each time an
experiment is performed, was on file for experiment R6512 for the run
, on September 10, 1986.
'
These findings were identified as examples 2, 3, and 4 of VIO 87-01-02.
Technical Specification 6.2.e(5) requires the NSC to audit reactor
<-
operational records for compliance with procedures and Technical Speci-
'
fication requirements. NSC reviews and audits should have included
records of " Request for Minor Experiment Approval" forms, Experiment
Schedule Forms, and " Experimenter's Checklists." This finding will be
,
tracked as example 2 of VIO 87-01-06.
I A Nuclear Safeguards audit of the records for the year 1985 for Procedure
"
2012 identified almost fifty errors, omissions or inconsistencies. The
l Committee responded to these audit findings by revising Procedure 2012 to
provide more guidance on completing the " General Notesd section of this
procedure. This revision was implemented on January 20,1987, and
i
the staff was told to exercise more care in completing the records.
- Insufficient time had elapsed between the implementation of the revision
j and the inspection to assess the effectiveness of the NSC actions.
- Nuclear Engineering class laboratory experiments were frequently being
.
performed at the GTRR, and the Inspector requested access to the
! experiment records for these labs. Some but not all of the Experiment
. Approval Forms for the labs were produced by the licensee. The forms
produced indicated that the lab experiments included such reactivity
manipulations as rod worth measurements and moderator coefficient testing.
As these types of reactivity manipulations were of sufficient safety
i significance to warrant approval of the procedures by the Safeguards
Committee. The licensee could not produce documentation of NSC review and
- approval of the lab procedures.
!
l This item was identified as example 3 of VIO 87-01-06.
4
'.
'
In addition, the inspector observed that console log entries for NE labs
frequently did not specify the nature of the experiment which was
performed. The licensee agreed to provide sufficient information in
j future log entries to ensure traceability.
1
i
i
!
!
l
!
!
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22
10. Review and Audit Function (40745)
The Nuclear Safeguards Committee (NSC) was responsible for the review and
audit of operations conducted at the GTRR to ensure that the reactor was
operated in a manner consistent with public safety and within the terms of
the facility license.
The objectives of inspecting the audit and review functions were to verify
(1) that the NSC was acting in accordance with Technical Specification
Requirements and (2) that follow through on important issues raised at the
review level was effective.
a. OSC Administrative Requirements
No problems were identified in this area.
The minutes of the NSC meetings for the period June 1985, through
February 1987, were reviewed for compliance with administrative
requirements. The composition of the committee, the meeting
attendance and frequency, and the distribution of the meeting minutes
satisfied Technical Specification requirements,
b. Experiment Approval Function
Inspection of the NSC function of experiment approval is discussed in
paragraph 9 of this report.
c. Review of Reportable Occurrences and Unusual Events
No problems were identified in this area.
Technical Specification 6.2.3(2) requires NSC review of reportable
occurrences.
I ' occurrences reportable by the Technical Specification definition
occurred during 1985 or 1986. The inspector observed that the
committee took a conservative approach in deciding when to notify the
NRC.
d. Facility Design Change Review Function
Technical Specification 6.2.e(4) requires that the NSC review and
approve proposed changes to the Technical Specifications and proposed
amendments to the facility license and review proposed changes to the
facility made pursuant to 10 CFR 50.59 (c).
The inspector reviewed NSC committee minutes for the period between
February 1985 and February 1987. The committee discussed the
following proposed changes to the facility: 1) A change in the
______________----__-----_-___-J
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.
23
ionization chamber in one of the two safety power trip channels;
~2) Silencing the siren horn outside Emergency Command Center after
the center is manned; 3) Modification of the "Do Not Enter" Light
actuation circuitry; 4) Redesign of neutron radiography beam H-1 for
better focusing; 5) Switching the cover gas to nitrogen and;
6) Termination of inflating of freight door gasket.
The NSC minutes showed each of these changes to have been adequately
reviewed by the committee. However, although the minutes always
stated whether the modification or change was approved or
disapproved, the results of the 10 CFR 50.59 review were not always
clearly indicated.
The inspector noted that although the cover gas had been changed from
helium to nitrogen during 1986, no Technical Specification amendment
had been made to change the reference in Technical Specification 3.6.e to a helium cover gas. 10 CFR 50.59 only allows facility
changes if a Technical Specification change is not involved. This
failure to obtain prior NRC approval for a facility change involving
a change to Technical Specification was identified as:
(0 pen) VIO 87-01-05: Change made to Facility, Involving Technical
Specification Change, Without Prior NRC Approval.
e. Review and Approval of Proposed Procedures
No problems were identified in this area.
Technical Specification 6.2.e(3) required NSC review and approval of
proposed operating procedures and proposed changes to operating
procedures which change the original intent of the operating -
procedure in a non-conservative manner.
The licensee policy was for all permanent procedure changes to be
approved by the NSC, not just changes which were non-conservative.
The inspector reviewed all the procedure changes approved by the NSC
during 1985 and 1986 and no problems were identified.
f. Audits of Adequacy of Existing Procedures
NSC audits of existing operating procedures for adequacy and
assurance that they achieved their intended purpose are required by
Technical Specification 6.2.e(6).
A formal review of the adequacy of eight selected Reactor Operating
Procedures was conducted by the NSC during 1984. The inspector noted
that the procedures chosen for the review included the checklists
where most of the safety surveillance checks were performed. Results
- -
.
, -
.
.
24
of the procedure reviews were discussed during the February 1985
meeting, and the minutes indicated that the reviews were thorough.
' Subsequent audits of these same procedures, conducted by the NSC
during 1985 and 1986, yielded additional suggestions for procedure
improvements.
The inspector verified that all the review findings were incorporated
into approved procedure revisions.
Because the same eight procedures were reviewed in 1984 and audited
in 1985 and 1986, the inspector questioned whether or not other
safety related GTRR procedures had been sufficiently reviewed by the
committee in recent years to verify that they were still adequate.
This concern is identified as example 4 of VIO 87-01-06.
g. Audits of Operations and Operational Records
Technical Specification 6.2.e(5) requires the NSC to audit reactor
operations and reactor operational records for compliance with
internal rules, procedures, and regulations and with licensed
provisions including Technical Specifications.
'
The NSC performed audits during 1985 and 1986 (of 1984 and 1985 data)
for eight selected procedures and the control room logs.
The inspector reviewed these audits and they appeared to be thorough.
Instances of errors or omissions in completing several procedures
, were identified, and suggestions for upgrading the procedures were
! made. Operations responded to each audit finding, and no items from
these audits remained open. The inspector verified that the agreed
,
upon procedure changes were all implemented.
The NSC requested that operations use added care to ensure that
entries in logs and procedures were not omitted or inaccurate.
'
As previously noted in Paragraph 10f as example 4 of VIO 87-01-06,
- the facility Technical Specifications imply that the NSC should also
i
be systematically reviewing various other records to ensure overall
compliance with requirements and regulations.
h. Audits of Equipment Performance
Technical Specification 6.2.e(7) requires the NSC to audit plant
equipment
anomalies, reportable
performance
with particular attention to operating
occurrences, and the steps taken to identify
and correct their causes.
,
.
o " .
.
25
During 1985 and 1986, equipment problems were addressed by the
committee on a case by case basis. No systematic audit of plant
equipment performance had been undertaken.
The question of Technical Specification required reviews of equipment
function will be example 5 of VIO 87-01-06.
The inspector did note that an NSC review of safety-related
electronics will be performed (Paragraph 13), and will be reviewed as
part of followup on this violation.
The minutes of the Nuclear Safeguards Committee meeting on
September 7, 1984, listed several concerns which were expressed
regarding possible problems at the facility. There was no documenta-
tion in the minutes that these concerns were duly addressed and
resolved. Staff members told the inspector that the concerns had
been addressed, but could not recall the specifics. During the
inspection, several staff members expressed additional concerns to
the inspector about the reliability of the equipment at the facility.
In response to a power excursion event on February 3, 1987, caused by
faulty equipment, the NSC committed to thoroughly evaluating all of
the safety-related GTRR electronics. The overall state of GTRR
equipment will be reviewed in a future inspection as IFI 87-01-11
(Paragraph 13).
11. Requalification Training (41745)
The licensed operator requalifiertion training at the GTRR was inspected
for (1) compliance with 10 CFR and the program approved by the NRC and (2)
adequate documentation to demonstrate compliance,
a. Annual Examinations
No problems were identified in this area.
The approved operator requalification program requires all licensed
personnel to be requalified yearly by either passing or administering
a written examination.
, The reactor supervisor of the facility held a senior reactor operator
(SRO) license. There was on additional SRO, who passed written
requalification examinations in August 1985 and September 1986. The
examinations were administered by the reactor supervisor, satisfying
the procedural requirements for his examination.
1
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26
b. Retraining Program
No problems were identified in this area.
The approved requalification program requires special retraining if
an operator failed to score at least 80% on all sections of the
annual examination. The inspector's review of the grades on each
section of the requalification examination verified no accelerated
operator retraining was required.
The licensee had implemented a permanent retraining program
consisting of weekly training sessions. The inspector reviewed the
records of this training, and attended a portion of the session on
February 9,1987. The meeting agendas covered such topics as fuel
cutting procedures, fireman briefing, respiratory training, hot cell
training, preparation for emergency drills, housekeeping, equipment
repair, approach to criticality procedures, and radiological safety.
c. Reactivity Manipulations and Performance Evaluations
A violation of requirements for documentation of reactivity
manipulations and performance evaluations was identified during the
inspection.
10 CFR 50.54 i-1 requires that "... the licensee shall not ... make a
change in an approved operator requalification program ...."
10 CFR 55 Appendix A requires that requalification include control
manipulations and documented observations and evaluations of the
performance of operators under real or simulated emergency
conditions.
The licensee's requalification program which implements 10 CFR 50.54
and 10 CFR 55 Appendix A, requires each licensed operator to perform
at least five reactivity manipulations each year. A summary of this
information is required to be kept on file to document compliance.
Although it was obvious from console log book entries that each
licensed operator had completed more than the required number of
control manipulations, a summary of manipulations was not documented
as required.
The approved requalification program also requires that yearly
observations and evaluations of the performance of each licensed
operator be documented. The written evaluation must include a
discussion of the operator's performance under either real or
simulated emergency conditions.
The most recent operator performance evaluation in the licensee's
flies was dated December 1983. The reactor supervisor concurred that
the required formal documentation had not recently been done.
..
.
27
This item will be tracked as:
(0 pen) VIO 87-01-05 Failure to comply with approved requali-
fication program
The licensee was previously cited for failure to do performance and
competency evaluations in Inspection Report 81-03.
12. Refueling (60745)
There had been no core loading changes in the GTRR since April 13, 1984.
All but three of the unirradiated fuel elements at the site had been
returned to Oak Ridge for storage in February 1986.
The inspector identified that the radiation levels of spent fuel stored at
the site were not being verified to be self protecting. 10 CFR 73.67
(b)(1)(i) exempted irradiated fuel assemblies from certain physical
safeguards requirements when the external radiation dose rate was in
excess of 100 rems per hour. Usual practice at research reactors is to
recycle spent fuel back tr,to the core to maintain the required 100 r/hr
levels. Both operations and health physics personnel expressed a desire
to verify the self protection of the fuel, but stated that measurements ef
the radiation levels of the spent fuel would cause radiation doses to
personnel and the risk of contamination. The licensee felt this would
contradict the ALARA concept. Operations personnel did once measure the
radiation level of an assembly, and found it to be over 900 R/ hour.
However, Health GTRR personnel expressed a lack of certainty that the
radiation levels of all the spent elements exceeded the 100 R/ hour
requirement. The GTRR staff requested guidance from the NRC in this area.
This was identified as:
(0 pen) IFI 87-01-12: Resolve verification of 10 CFR 73.67 require-
ment that spent fuel have self protecting radiation levels.
13. Reactor Power Excursion (92700)
An unintended power increase from 300 KW to 2.3/MW occurred at the GTRR on
February 3, 1987. The inspector and the NRC project engineer for the GTRR
attended the Nuclear Safeguards Committee meeting on February 10,
regarding the event.
Prior to the incident, the reactor had been started up for a classroom
demonstration and for a neutron radiography experiment using horizontal
beam port H-1. The personnel around the beam port were accompanied by
Health Physics. The reactor power level was at 300 KW when power began to
increase on a positive period of approximately 15 seconds (equivalent to
0.0017 delta k/k). Af ter an estimated thirty seconds of increasing
.-
.
, .. -.
.
'28
reactor power, the area radiation monitors went off and Health Physics
~
neutron meters near the beam port went off scale. At this point the
senior operator on duty manually inserted the shim blades, and the power
increase had been turned around within an estimated 45 seconds after the
start of the excursion. Power had increased to approximately 2.3 W.
No automatic scram on high reactor power level was actuated because
the reactor was being operated in Mode 2, which allows 5 W operation.
Technical Specifications define Mode 1 as power levels up to 1 W , but
standard practice at the facility was to operate with Mode 2 settings even
at low power levels. No positive period trip was actuated because the
trip setting was ten seconds as specified in the Technical Specifications.
No Technical Specifications appeared to have been violated in this
incident.
The power increase was caused by a failure in the automatic rod control
system. The regulating rod took its signal to the servo motor from the
sliding wire in the power level chart recorder. The power increase
occurred when the recorder jammed in such a way as to drive and hold the
regulating rod fully out. The jammed recorder showed no deviation.
The chart recorder was removed and cleaned thoroughly. Possible causes of
the malfunction were identified as a dirty wire, loose contacts, and the
wire being off the pulley and instead of on the post that holds the pulley
in place. The electronics specialist was not certain whether or not the
cleaning process was responsible for the guide wire being out of its
normal position. After the maintenance, the chart recorder was verified
to be operating normally.
The dose to the personnel working around the beam port did not show an'
increase on their TLD's and was estimated to be about 17 mrem.
The Nuclear Safeguards Committee discussed the incident and ways to
prevent recurrences. Concern was expressed by several members that
operator attentiveness and response were not adequate. It was agreed that
sound engineered safeguards must be in place to backup operator actions.
It was brought out that although the GTRR equipment was effectively
designed, most of it was about 25 years old.
The following recommendations were adopted:
'
1. For operation at or below 1 W, the Mode 1 level trip setpoint of
1.25 W would be used.
i
l 2. The positive period trip setting would be 15 seconds instead of the
Technical Specification Limit of 10 seconds.
l 3. The operating staff would devise a design change to incorporate a
l buzzer into the linear picoammeter not used for automatic rod
control, to provide redundancy.
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4. The Chairman of the Committee would evaluate the condition and
operability of all safety related electronic equipment. Most of the
reactor scram functions depend on electronics.
5. The operators would pay close attention to the controls as a matter
of course.
The results of this audit and the follow-up of the results will be
reviewed by the NRC. As this audit was only to address electronic
equipment, the condition and performance of other safety related equipment
will also be reviewed by the NRC in future inspections. This NRC
follow up will be tracked as:
(0 pen) IFI 87-01-11 NRC followup on general state of equipment.
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