IR 05000244/1987004
| ML17261A466 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/16/1987 |
| From: | Kenny T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17261A462 | List: |
| References | |
| 50-244-87-04, 50-244-87-4, IEIN-85-045, IEIN-85-45, NUDOCS 8704290236 | |
| Download: ML17261A466 (32) | |
Text
U. S, NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-244/87-04 Docket No. 50-244 Licensee No.
DPR-18 Priority Category C
Licensee:
Rochester Gas and Electric Corporation 49 East Avenue Rochester, New York Facility Name:
R.
E. Ginna Nuclear Power Plant Inspection at:
Ontario, New York Inspection Conducted:
February 8, 1987 through March 14, 1987 Inspectors:
T. J. Polich, Senior Resident Inspector, Ginna N.
F. Dudley, Lead Reactor Engineer (Examiner)
Approved by:
P
/
T..
enny c
ng Chief, Reactor Da e P
je Se No.
2A, DRP Ins ection Summar
Ins ection on Februar
1987 throu h March
1987 Re ort No.
50-244/87-04. R...
b*l b<<
l
b regional and resident inspectors
'(181 hours0.00209 days <br />0.0503 hours <br />2.992725e-4 weeks <br />6.88705e-5 months <br />).
Areas inspected included:
licensee action on previous findings; review of plant operations; operational safety verification; surveillance testing; plant maintenance; inspector follow-up items; requalification examinations; review of peziodic and special reports; main control board modifications, and temporary instruction 2500/16.
Results:
In the ten areas inspected, two violations were observed.
One violation involved failure to perform a source check of radiation monitor R-18 (paragraph 2.a).
The second violation involved a failure to perform an adequate
CFR 50.59 review of lead shielding installations (paragraph 4.d).
Additionally, during this inspection period the licensee exhibited an informal approach to certain activities, such as surveillance testing, procedure changes and control of tags (paragraphs 4.c and S.c).
8704~90236 870420 PDR ADOCK 05000244
DETAILS 1.
Persons Contacted During this inspection period, the inspectors held discussions with and interviewed operators, technicians, engineers and supervisory level personnel.
J.
C. Bodine, Nuclear Assurance Manager
"D. L. Filkins, Chemistry
& Health Physics Manager R.
W. Kober, Vice President, Electric and Steam Production
"R. A. Marchionda, Training Manager
"T. A. Marlow, Maintenance Manager
"T. A. Meyer, Superintendent Ginna Support Services T.
R. Schuler, Operations Manager
"M. T.
Shaw. Administrative Services Manager B. A. Snow, Superintendent Nuclear Production
"S.
M. Spector, Superintendent Ginna Production
"R.
W. Vanderweel, Ginna Modifications Project Manager
"J. A. Widay, Technical Manager
"R.
E.
Wood, Supervisor Nuclear Security
"Denotes persons present at Exit Meeting on March 18, 1987.
2.
Licensee Action on Previous Ins ection Findin s
~
~
a
~
(Closed)
Unresolved Item (87-02-01)
Failure to perform monthly source check of R-18 since 1976.
On February 2,
1987, while performing Periodic Test (PT)-17.2,
"Process Radiation Monitors R-ll-R-22,", the control room operator noted Liquid Rad Waste Monitor, R-18 did not respond when the front panel switch was placed in the check source position.
The licensee performs PT-17.2, monthly to comply with the Technical Specification (TS) requirements to source check process radiation monitors.
The licensee.determined that during the 1976 refueling outage R-18 was replaced with an instrument that did not contain a check source or a check source solenoid, used to expose the source to the detector.
Additionally, it was determined that the check source position of the front panel switch did not perform any function.
Since the 1976 replacement of R-18 it was impossible to perform a source check by using the PT"17.2 procedure.
Failure to perform a monthly source check of R-18 is a violation of Technical Specification 4.1.4.
The inspector conducted a review of PT-17.2 revisions, Procedure Change Notices (PCNs),
completed procedures and trouble cards to establish a 'chronology of events from the replacement of R-18 in February 1976 until May 1976.
This review identified the following deficiencies:
No procedure change or revision to reflect modification changes (i.e., lack of check source or check source solenoid)
~
No procedure caution to reflect the Technical Specification note on the use of an external source in high background areas.
Inadequate root cause investigation when high background conditions existed in 1976.
Multiple opportunities existed for licensee management to identify equipment and procedure deficiencies since 1976 because each of the procedures, PCNs, and revisions were approved by the Plant Operations Review Committee.
While some aspects of this issue were identified by the licensee, a
violation is being.issued because multiple opportunities to identify and correct the problem existed.
This unresolved item is closed and will be upgraded to a violation (87-04-01).
(Open) Inspector Follow-up Item (82-21-02)
Residual Heat Removal (RHR) subbasement flood pr'otection.
During a previous inspection it was noted that when the drain line to the auxiliary building sump became clogged water would backup into the basement level pipe chase and spill into the subbasement.
The inspector also noted standing water one to two inches deep in the northeast corner of the room.
These conditions were noted to potentially reduce the overall flood protection for the RHR pumps.
The licensee identified the source of the blockage to be a one inch pancake type orifice installed in the auxiliary building floor drain line.
In mid-1986 the licensee installed a one inch pipe with valve in the drain line above the orifice to allow draining. of the six inch floor drain line when it clogs.
The inspector observed water dripping from the ceiling and standing water in the subbasement during Inspection 50-244/86-18 and noted the condition had also been reported to licensee management by the licensee's (}C inspectors in reports dated November 14, 1985 and August 19, 1986.
The licensee determined that=the water dripping from the ceiling was from an abandoned conduit.
The licensee plugged the conduit as described in IR 50-244/86-21.
On March 1, 1987 the inspector, accompanied by an auxiliary operator, observed water on the floor and northeast corner wall when making a tour of the subbasement.
The inspecto~ also observed water at the bottom of the RHR line in the pipe chase on the basement level .
On March 2, 1987 at 9:55 A.H. the licensee discovered approximately 10 inches of water in the subbasement.
An auxiliary operator who discovered the flooding lifted the sump pump floats and the pumps started but cavitated because the sump was empty.
The subbasement floor drain to the sump was found clogged allowing water to accum-ulate in the subbasement.
The source of water was determined to be flushing water draining to the floor drain system, which backed up into the pipe chase and flowed down the northeast corner wall.
Site management plans to turn the subbasement flooding problem over to corporate engineering for resolutions This item remains open.
3.
Review of Plant 0 erations a ~
b.
Throughout the reporting period, the inspector reviewed refueling outage and start-up testing activities.
The reactor was shut down on February 6, 1987 and remained shutdown until March 9, 1987 when the reactor was brought critical at 7:00 P.M..
The generator was paralleled to the grid at 2:59 P.M.
on March 10, 1987.
The reactor power was increased as secondary water chemistry limits permitted.
The major activities observed by the inspector were: reactor vessel head removal; fuel shuffle; reactor reassembly and head stud tension; Microprocessor Rod Position Indication (MRPI) system installation; reactor trip breaker testing light modification; steam generator snubber maintenance; steam generator repair; dilut'ion to criticality; low power physics testing; and power ascension.
On February 20, 1987, while in cold shutdown with all normal off-site power isolated an Emergency Diesel Generator (EDG)
1B panel alarm was received on the Hain Control Board (MCB).
The alarm was due to a low fuel oil day tank level caused by a clogged strainer on the fuel oil transfer pump suction line.
The licensee had started the "A" and "B" EDGs at approximately 6:00 A.M. to allow the ¹12 transformer (normal off-site supply) to be isolated for maintenance and testing.
At 11:58 A.M. the EDG 1B panel alarm was received on the MCB.
An Auxiliary Operator (AO) was dispatched and found an actual low level in the day tank.
The AO crosstied the diesel fuel oil system to clear the alarm condition, but the alarm did not clear.
At approximately 12:30 P.M. the 1A EDG panel alarm was received on the MCB.
A low level condition existed in that day tank also.
The licensee formed a bucket brigade and rigged portable pumps with hoses to fill the diesel day tanks from the diesel storage tanks.
At 1:50 P,M.
a 1,000 gallon fuel oil
truck arrived on-site.
The licensee notified the NRC at 2:30 P.M.
of a four hour non-emergency event.
Off-site power was restored at approximately 1:00 A.M. on February 21, 1987, and the diesel generators which remained running during this event were secured.
Maintenance and troubleshooting on the "A" EDG fuel oil problem began immediately and was completed at 8:30 P.M.
on February 22, 1987 when the "A" EDG was declared operable.
The "B" EDG maintenance and troubleshooting began at 11:00 P.M.
on February 22, 1987 and this diesel was declared operable on February 24, 1987.
The licensee plans to submit a Licensee Event Report on this event, This item remain open pending review of the licensee's corrective action.
(87-04-02)
4.
0 erational Safet Verification
'a ~
General During the inspection period, the inspectors observed and examined activities to verify the operational safety of the licensee's facility.
The observations and examinations of those activities were conducted on a daily, weekly or monthly basis.
On a daily basis, the inspectors observed control room activities to verify compliance with selected Limiting Condition for Operations (LCOs) as prescribed in the facility Technical Specifications (TS).
Logs, instrumentation, recorder traces, plant conditions, and trends were -reviewed for compliance with regulatory requirements.
~ Shift turnovers were observed on a sample basis to verify that all pertinent information of plant status was relayed.
During each week, the inspectors toured the accessible areas of the facility to observe the following:'eneral plant and equipment conditions Fire hazards and fire fighting equipment Radiation protection controls Conduct of selected activities for compliance with licensee's administrative controls and approved procedures Interiors of electrical and control panels Implementation, of selected portions of the licensee's physical security plan Plant housekeeping and cleanliness Essential safety feature equipment alignment and conditions The inspectors talked with operators in the control room, and other personnel.
The discussions centered on pertinent topics of general plant conditions, procedures, security, training, and other aspects of the involved work activitie Radiation Controls The inspector noted several instances of licensee personnel and contractors at the auxiliary building radiation work permit (RMP)
sign-in desk with anti-contamination clothing lab coats unbuttoned.
A lab coat, gloves, hat and shoe covers is the minimum anti-contamination clothing required ior access to this. area on any licensee RMP.
This was identified to licensee management in a previous inspection.
Also the security guards stationed at the containment personnel hatch and RWP sign-in desk personnel were frequently observed reading books, 'newspapers, and magazines.
Ta Control When the licensee was performing Refueling Shutdown Surveillance Procedure (RSSP)-2.3, the inspector and licensee gC personnel noticed a conditional release tag attached to the "A" Diesel Gen-erator local control panel.
This tag should have been removed in early 1986.
On March 13, 1987, the inspector observed two test tags on safety injection valves 890A and 890B.
The test tag control log indicated the test requiring these tags had been completed on March 8, 1987.
The inspector informed the shift supervisor who had the tags removed.
Additionally,'uring this inspection period the inspector observed the following types of tags adrift -on the floor of the auxiliary building.
Active Hold Tags Inactive Temporary Cable Tags Inactive Temporary Fluid System Provision Tags The inactive Temporary Fluid System Provision Tags were on the floot near the safety injection pumps for over a week.
Tem orar Lead Shieldin The inspectors questioned the licensee's use and installation of lead shielding around radiation monitors R-17 and R-20 after a
recent tour of the auxiliary building.
The R-17 installation con-sisted of approximately one ton of lead bricks surrounding seismic class 1 Component Cooling Water (CCW)
pump suction line (Reference UFSAR Table 3.2-1 Chapter 3) and was partially supported by a pipe clamp and spring canisters which supports the CCW line.
The R-20 installation consisted of approximately one ton of lead bricks and plate surrounding seismic class 1 Service Mater (SW) piping (Reference UFSAR Table 3.2-1 Chapter 3).
Also, approximately four square feet of 3/8 inch lead sheet was 'secured by nylon rope to valve 786A (3/4 inch pressure gage isolation valve in the spent fuel pool cooling system)
and to flow indicator 2020 in the SW system.
The inspector
determined that these installations existed for many years (since the 1970-1971 time frame) with no formal documentation or evaluation of the adequacy of the piping installations.
The licensee developed administrative procedures for control of temporary structural features and control of temporary lead shielding in late 1985 and implemented these procedures in early 1986.
Mhen these procedures were implemented, preexisting lead shielding was included in the program.
The inspector reviewed the lead shield authorization forms for the R-17 and R-20 lead installations.
The R-17 package contained a handwritten Preliminary Evaluation dated April 10, 1986 which stated the support was adequate but seismic upgrading was advised.
The R-17 package also contained an Inter-Office Correspondence dated June 24, 1986, which stated an operating basis earthquake would render the pipe restraint inoperable.
The inspector discussed the Inter-Office Correspondence with the author and discovered no calculations were performed and it was the engineer's opinion that the shielding arrangement was inadequate.
The licensee removed approximately one ton of lead bricks from the platform around R-17.
The platform consists of two one-legged sections'ach section has an edge that rests on the pipe clamp supported by 'spring canisters.
The licensee has also removed the lead sheet from around the service water pipe above R-20 which was secured by nylon rope to a valve and a flow indicator.
The lead bricks around R-20 supported on a metal platform are still in place.
The licensee failed to perform a safety evaluation at the time the, lead shielding was installed.
Additionally, the licensee failed to perform an adequate safety evaluation when temporary lead shielding controls were implemented in 1986.
Failure to perform and maintain safety evaluations is an apparent violation of 10 CFR 50.59.
(87-04-03)
S rin Canisters During the investigation of the temporary lead shielding around R-17 the inspector observed three of four spring canisters on the Chemical and Volume Control System (CVCS) divert line to the CVCS holdup tanks.
The inspector determined that the line was grouted in place where it passed through a wall between two spring canisters.
The licensee is currently investigating the required spring canister settings and the support drawings for this line.
This item is unresolved pending resolution by the licensee.
(87-04-04)
Containment Stora e
After the licensee completed their final inspection of containment per administrative procedure A-3. 1, "Containment Storage Inspection",
the inspector observed a cardboard, box containing reactor cavity filters stored in the "caged area" on the upper level of containment.
The inspector also noted equipment, ladders, lockers and tool boxes stored on all levels of containment.
These items are secured with nylon rope or chained per A-3. 1 however, the licensee does not per-form a formal analysis or maintain an inventory of equipment weight to insure the ropes or chains are adequate to restrain equipment during a seismic event.
Licensee management has committed to further assessment of containment storage.
The inspector will follow the licensee's assessment in a future report.
One violation was identified.
5.
Surveillance Testin a
~
The inspector witnessed the performance of surveillance testing of selected components to verify that:
the test procedure was properly approved and adequately detailed to assure performance of a
'atisfactory surveillance test; test instrumentation required by the procedure was calibrated and in use; the test was performed by qualified personnel; and the test results satisfied Technical Specifications and procedural acceptance criteria, or were properly resolved.
During this inspection period, the inspectors witnessed the performance of selected portions of the following tests:
Periodic Test (PT)- 23', "Containment Isolation Valve Leak Rate Testing Nitrogen Supply to Pressurizer Relief Tank", effective November 6, 1986.
PT-23.3,
"Containment Isolation Valve Leak Rate Testing Makeup Mater to Pressurizer Relief Tank", effective September 26, 1986.
PT-32, "Reactor Trip Logic Test A or B Train", effective February 24, 1987.
RSSP-2.3,
"Emergency Diesel Generator Trip Testing", effective June 27, 1986.
The inspector has noted extensive use of PCNs to "fix" plant procedures.
In particular, the inspector noted three PCNs to PT-32, performed on February 24-25, 1987.
Two PCNs changed the expected test light indications for the Shunt Trip Assembly (STA) and Under Voltage Trip Assembly (UVTA) the third PCN added new STA and UVTA test light indications to be observed.
These PCNs were required although revision 24 to PT-32 was effective February 24, 198 Licensee personnel followed the procedure until the incorrect test light indications were observed, then changed the procedure to reflect the actual indications received due to Station Modification (SM)-3698.
The licensee was using portions of PT-32 to perform parts of SM-3698.6, "Testing of Reactor Trip Breakers Test Light Addition".
When PT-32 was stopped, without completion, on the evening of February 24 and current simulators and resistor jumpers were removed the licensee did not note the delay in the comments section of PT-32.
On February 2S when testing was resumed the initial conditions were not formally re-verified and initialled although some simulators and resistors were removed.
The inspector did observe the technicians check the simulators and resistors after a steam pressure instrument was found not properly simulated.
The inspector questioned the technicians regarding the documenting of delays and and removal of test equipment.
While no formal documenta-tion was available one technician said he wrote down the simulators and resistors that were removed the evening of the delay and had the information on his desk.
This information was not used when testing resumed.
The inspector discussed these findings with the Technical Manager and the Results and Tests Supervisor.
This item will remain unresolved pending licensee evaluation of the need to document removal of test equipment when te'sts are delayed.
(87-04-05)
No violations were identified.
6.
.Plant Maintenance a 0 b.
During the inspection period, the inspector observed maintenance and problem investigation activities to verify:
compliance with regulatory requirements, including those stated in the Technical Specifications; compliance with administrative and maintenance procedures; required gA/gC involvement; proper use of safety tags; qualifications; and reportabi lity as required by Technical Specifications.
The inspector witnessed selected portions of the following maintenance activities:
M-40.7, "Steam Generator Snubber Inspection and Maintenance",
effective June 20, 1986.
SM-4068. 1, "Radiation Monitoring System Upgrade", effective February 27, 1986.
SM-4068.3,
"Replacement of the Radiation Monitoring System (RMS)
Process Monitors R10A, Rll, and R12 (Electrical Portion)", effective February 9, 198 SM-4068.4,
"Replacement of the RMS Process Monitors R10A, Rll and R12 (Mechanical Portion)" effective February 13, 1987.
SM-3698.6,
"Testing of Reactor -Trip Breakers Test Light Addition",
effective February 10, 1987.
Calibration Procedure (CP)-35, "Intermediate Range N-35 Calibration and/or Maintenance",
effective November 1,
1985.
No violations were identified.
7.
Ins ector Follow-u From Ins ection Re ort 50-244/87-02 On February 6,
1987, the licensee found Residual Heat Removal (RHR) refueling water return flow indicator FI-931B isolated.
The licensee determined the flow indicator was valved out in January 1987 while performing Calibration Procedure (CP)-931B, "Calibration and/or Maintenance of RHR Refueling Water Return Flow F-931B".
The two Instrumentation and Control (I&C) technicians who performed this CP and signed off that the transmitter was unisolated and vented did not return the system to an operating condition because of an inade-quately written procedure.
b.
The licensee retrained all 18C personnel regarding proper realignment of instruments for operability.
Also the licensee has revised CP-931B to move the sign-off for restoration of the system to a more logical point in the procedure.
On February 7, 1987, while cooling down the reactor plant for a refueling outage the "B" Main Steam Isolation Valve (MSIV) failed to close.
The licensee made the appropriate notification to the NRC.
The licensee methodically investigated the reason for the failure by examining the pneumatic valve operator, the valve packing and the valve internals.
The MSIVs are 30 inch self-aligning, swing disc, inclined seat check valves manufactured by Atwood and Morrill.
The licensee modified the valves in 1975 to minimize steam flow restrictions.
The modification raised the disc stop to allow the disc to remain horizontal when fully open.
At the time the modification was made the old disc stop was notched out and a
new valve disc stop was added.
The clearance between the disc and the old disc stop material was very close.
The licensee determined the cause of the failure was contact of the valve disc with the old disc stop material.
The valve disc made contact with the disc stop material as evidenced by clean shiny metal on the valve disc and old disc stop.
The licensee removed the old valve disc stop material and has inspected the "A" MSIV for a similar problem.
The "A" MSIV did not have any such disc stop material remaining when inspecte The licensee submitted a letter dated March 9, 1987 to the NRC as follow-up to the four hour notification.
No violations were identified.
8.
Re uglification Pro ram a.'Sco e
1.
The written examinations prepared by the facility for the first four weeks of a six week annual requalification examination cycle were reviewed by NRC examiners, and NRC prepared questions were substituted for facility prepared questions in the Reactor
'perator (RO) and Senior Reactor Operator (SRO) examinations administered fhe fourth week.
The substitution of questions was based on the general guidance for the scope of written examin-ations contained in NUREG-1021, Chapter ES-202, Section E; the guidance concerning depth of knowledge contained in NUREG-1021, Chapter ES-203, Section F; the importance ratings of knowledges and abilities contained in NUREG-1122; the facility prepared learning objectives for the previous requalification cycle; and examiner judgement.
Both the facility and the NRC graded the NRC prepared examinations and grades were compared.
2.
The facility's evaluation of an unscheduled plant trip on November 28, 1986 caused by an incorrect operator action was reviewed.
The scope of the operator's retraining program and
'he facility planned follow up evaluations of the operator involved in the unschedu1ed plant trip were also reviewed.
b.
~Find<n s
1.
The facility is in the process of accrediting their training programs with the Institute of Nuclear Power Operations (INPO).
A separate organization has been established to complete the development and accreditation of the training program.
As a
result the training department is provided with learning'objec-tives to which they train and evaluate the licensed operators.
The training department requires 205 new questions on each requalification examination.
Approximately 75/~ of the RO and SRO examinations given the fourth week of the evaluation cycle contained questions from the three previous examinations.
The examinations given the second and third week of the cycle con-tained over 50% repetitious questions.
The facility maintains all copies of the examinations during the examination cycle.
However, the facility takes no other steps to ensure the security of an examination once it has been administere As a result of an NRC review of the examinations administered during the fourth week of the evaluation cycle approximately 57%
of the questions in the RO and 70% of the questions in the SRO examinations were replaced.
The following items were considered to be weaknesses in the facility prepared requalification examinations:
Section 1 was 50% repetitive from the previous three examinations.
Section 1 of the RO and SRO examinations were identical Section 2 was 66% repetitive from the previous three examinations'ection 3 was 100% repetitive from the previous three examinations.
Section 3 did not provide sufficient coverage of the secondary plant.
Three of the five questions on the secondary plant covered the steam generator level control system.
S'ection 4 contained a question on radiological controls and a question on the usage of the Critical Safety Function Status Trees.
Both questions had appeared on all three previous examinations.
No other questions on these areas were included in the examination.
Section 6 was 72% repetitive from the previous three examinations.
Section 6 of the SRO examination was comprised entirely of questions that appeared in section 2 and section 3 of the RO examination.
Section 7 was 91% repetitive from the previous three examinations.
Section 7 of the SRO examination was comprised almost entirely of questions that appeared in section 4 of the RO examination.
Section 7 did not examine for a sufficient depth of knowledge.
Five of the nine questions required only recall of information and comprised about 53% of the point value of the sectio Section 8 was 88% repetitive of the previous three examinations.
Four of the nine questions in the section had appeared on all three previous examinations and com-prised 53% of the point value of the section.
Section 8 did not examine to a sufficient depth of knowledge.
Eight of the nine questions required only recall information and comprised about 87% of the point value of the section.
Section 8 was too heavily weighted towards the Emergency Plan/ Event Reporting.
Three of the nine questions in this section addressed this area and comprised approximately 31%
of the point value of the section.
Overall the RO examination contained 41'ecall type questions and the SRO examination contained approximately 50% recall type questions.
The higher percentage of recall type'questions on the SRO examination compared to the RO examination was considered inappropriate.
Section 8, which is the only section unique to the SRO exam-ination, contained 88% recall type questions.
Due to the large percentage of recall type questions in this section the depth of knowledge of the operators in terms of analysis and synthesis of information was not adequately evaluated.
Therefore, it was determined that section 8 was inadequate to evaluate individual and generic weaknesses at an SRO level of knowledge.
After the administration of the NRC modified examinations, the facility reviewed the answer key provided by the NRC and appro-priate changes were made.
The NRC and the facility indep'endently graded the examinations.
The facility assigned all operators passing grades on the written examinations.
The NRC assigned two operators less than passing grades.
Due to problems asso-ciated with reproducing the examination for parallel grading, the facility changes made to the answer key and the subjective judgement of the different graders the pass fail decisions made by the facility were acceptable.
The results of the parallel grading are presented in attachment l.
All facility overall grades were within 5% of the grades assigned by the NRC.
Facility sectional grades were within 10% of the grades assigned by the NRC with the exception of section 8 of the SRO examination.
However, the grades on the NRC prepared questions were over 10% lower than the grades on the'acility prepared question The information being examined on NRC prepared questions, had high importance factors as determined from NUREG 1122.
The large disparity between the performance on NRC and facility prepared questions is noted but has not been analyzed, 2.
The evaluation of the results of the requalification examina-tions is conducted by a facility evaluation group independent of the training department.
The findings and recommendations of this evaluation were not available for review by the NRC since the evaluation had not been completed.
L The. facility determined that the root cause of an unscheduled
.
reactor trip which occurred while conducting a surveillance test on November 28, 1986 was performance based.
In an effort to identify other operator performance problems, training has been provided to supervisory personnel on the evaluation of indi-vidual performance indicators.
The facility determined that the operator had been adequately trained in the last two years, as part of his initial license training program, to properly con-duct the surveillance.
Weaknesses had been identified in the operator's performance on the latest simulator evaluation but the facility had determined that no additional training was required.
Following the trip a comprehensive training and evaluation program was established for the =operator involved in the event.
The program is extensive and long range.
The operator was provided additional training on the simulator and was assigned directed self study prior to being assigned selected licensed responsibilities.
An ongoing evaluation of the operator's per-formance is being conducted by management and will be closely monitored by the resident inspector.
c.
Conclusions 2.
Overall the training department's implementation of the requal-ification program was adequate.
Security for the facility prepared annual requalification examinations is questionable due to the large percentage of questions'sed from previously admin-istered examinations.
Section 8 of the facility prepared SRO examination did not evaluate an adequate depth of knowledge.
The facility's initial evaluation of the unscheduled plant trip was adequate.
The retraining program for and the continuing evaluation of the operator involved in the event is considered to be appropriate.
No violations were identifie Review of Periodic and S ecial Re orts Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9. 1 and 6.9.3 were reviewed by the inspector.
This review included the following considerations:
the reports contained the information required to be reported by NRC requirements; test results and/or supporting information were consistent with design predictions 'and performance specifications; and the reported information was valid.
Within this scope, the following report was reviewed by the inspector:
Monthly Operating Report for January 1987.
Main Control Board Modifications Verified NRC letter dated February-5, 1987 (DiIanni to Kober) addressed Systematic Evaluation Program (SEP)
NUREG 0821 Supplement 1 Sections 2.7.2, 2.7.3, 2.7.4, and 2.9.
Issues raised in these sections were resolved subject to confirming that six modifications to the main control board identi-fied in the licensee's letter dated January 9,
1984 had been adequately implemented.
The inspector verified these six modifications were
.
installed during this inspection period.
The inspector reviewed Engineering Work Request (EWR)-3575 "Stiffener Modification to the Main Control Board" to insure the siR modifications identified in the licensee's report wei e included in the EWR.
Drawing 33013-771
"Main control Board Stiffener Supports Details" generated by EWR-3575 was used to perform a walkdown of the main control board to verify the drawing reflected the "as-built" condition.
The inspector identified no unacceptable conditions.
Tem orar Instruction 2500/16 As a result of IE Information Notice 85-45, "Potential Seismic Interaction Involving The Movable In-Core Flux Mapping System Used in Westinghouse Designed Plants",
the licensee began evaluating their.in-core flux mapping system during the 1986 refueling outage.
The licensee began work to upgrade the system during the 1987 refueling outage.
Portions of the seismic upgrade work were observed by the inspector during.the inspection period.
The licensee plans to complete the remaining portion of the upgrade during the 1988 refueling outage.
This temporary instruction is considered closed based on the work completed to date and the proposed schedul At periodic intervals during the inspection, meetings were held with senior facility management to discuss the inspection scope and findings.
On March 18, 1987, an exit meeting was held to review the details of this inspection report with licensee management.
Based on the NRC Region I review of this report and discussion held with licensee representatives, it was determined that this report does not contain information subject to
CFR 2.790 restrictions.