IR 05000220/1986009
| ML17055C099 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 07/17/1986 |
| From: | Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17055C094 | List: |
| References | |
| 50-220-86-09, 50-220-86-9, IEIN-86-003, IEIN-86-047, IEIN-86-3, IEIN-86-47, NUDOCS 8607240219 | |
| Download: ML17055C099 (32) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
DCS Nos.50220-860505 50220-860508
Report No. 50-220/86-09 Docket No. 50-220 License No.
DPR-63 Category C
Licensee:
Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility Name:
Nine Mile Point Nuclear Station, Unit
Inspection At:
Scriba, New York Inspection Conducted:
May 19, to July 6, 1986 Inspectors:
W. Cook, Senior Resident Inspector A. Luptak, Senior Resident Inspector C. Marschall, Resident Inspector G. Meyer, Project Engineer R. Struckmeyer, Radiation Specialist Approved:
.C. Lin le, Chief, tor P oject Section No.
,
DRP date Ins ection Summar
Ins ection on Ma 19 to Jul
1986 Re ort No. 50-220/86-09 Areas Ins ected:
Routine inspection by the resident inspectors regarding routine safety reviews, followup of the licensee's actions associated with the withdrawal of control rods with an inoperable Rod Worth Minimizer (RWM) during a startup on June 18, 1986, design changes and modifications, operational safety verification, physical security, plant tours, licensee event reports, unusual event, safety systems operability verification, review of periodic and special reports, review of Static-0-Ring pressure switches, radiological environmental monitoring program, followup on IE Information Notice 86-03, Allegation followup and surveillance testing.
The inspection involved 105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br /> by 5 resident and regional inspectors.
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e Results:
One Violation was identified.
During a startup on June 18, 1986, the Rod Worth Minimizer (RWM) was found to be inoperable after withdrawal of the eighth control rod.
Steps were taken to restore the RWM to service, however, no surveillance test was done to ensure RWM operability.
After moving three more control rods, it was determined that the RWM was not operable.
Details are provided in paragraph f I
DETAILS Persons Contacted A. Barnhart, Reactor Operator H. Barrett, Assistant Operations Supervisor S.
Brown, Chief Shift Operator C. Caroccia, Reactor Analyst M. Laris, Unit Supervisor, Reactor Analyst P,
Mangano, Supervisor, Computer Operations and Maintenance R. Murney, Assistant Supervisor, Computer Operations and Maintenance M. Randall, Chief Shift Operator T.
Roman, Station Superintendent J. Spadafore, Superintendent, Technical Services, Nuclear The inspectors also interviewed other licensee personnel during the course of the inspection including shift supervisors, administrative, operations, health physics, security, instrument and control, and contractor personnel.
Summar of Plant Activities The plant was in the final stages of the 1986 Refueling and Maintenance Outage for the first four weeks of the inspection period.
On June 4,
1986, an Unusual Event was declared when an experimental robot exploded inside a storage area within the site protected area.
On June 15, 1986, while conducting High Pressure Coolant Injection testing prior to startup, reactor water level was inadvertently lowered to the scram setpoint and a
scram resulted.
The reactor was shut down with all rods fully inserted at the time.
The mode switch was placed in startup at the end of the outage on Monday, June 16, 1986, but failure of three of the eight channels of the Intermediate Range Monitoring (IRM) system to clear their downscales caused the licensee to insert a half channel scram and conduct a reactor shutdown.
After correcting IRM connector problems in the three faulty channels, a reactor startup was commenced on June 17, 1986.
The same three IRM channels again failed to clear their downscales causing the licensee to insert a half channel scram and conduct a shut down
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The IRM detectors were replaced and the licensee commenced a reactor startup late on June 17, 1986.
On June 18, 1986, with reactor pressure at approximately 400 psig, failure to start a feedwater pump in a timely manner resulted in a full reactor scram on low level.
During the subse-quent startup on June 18, 1986, the Rod Worth Minimizer was declared inoperable and a reactor shutdown was conducted as required by Technical Specification.
During this event a violation of Technical Specifications was identified by the Resident Inspector.
At the same time, a leak in a recently installed reactor coolant sample line was discovered just outside its containment penetration.
After correcting the computer problem in the RWM and isolating the coolant sample line inside containment a reactor startup began late on June 18, 1986.
During the startup on June 19, 1986, with
bi
reactor pressure at approximately 400 psig, a reactor scram occurred as a result of corrective maintenance to IRM channel 16.
A reactor startup was initiated late on June 19, 1986 and the mode switch was placed in run and the generator synchronized to the grid on Friday, June 20, 1986.
A unit shutdown was conducted on June 21, 1986 to repair a stuck feedwater check valve and the reactor was made critcal later that day.
Full power was reached on June 26, 1986.
3.
Review of an 0 erational Event a.
Introduction On June 18, 1986, the licensee commenced a startup of Unit 1 after repairs were completed to the Intermediate Range Monitor System.
After withdrawing the eighth control rod the licensee discovered that the Rod Worth Minimizer was inoperable.
The Rod Worth Minimizer was restarted and the licensee withdrew three additional control rods without performing a surveillance test to insure the Rod Worth Mini-mizer was operable.
The licensee then determined that the Rod Worth Minimizer was still inoperable.
An inspection was conducted to review the licensee's investigation and to independently review the circum-stances of this event.
b.
Se uence of Events At 10:16 a.m.
on June 18, 1986 the reactor mode switch was placed in startup following completion of corrective maintenance on the Instrument Range Monitoring system.
At 10:27 a.m.
the first rod in rod'roup 1 was selected by the reactor operator and withdrawn to its full out position as recommended by the reactor analyst present during the startup.
Seven more control rods in rod group 1 were selected and withdrawn to their full out position as recommended by the reactor analyst.
After withdrawal of the eighth rod, the reactor analyst noted that the computer monitor displaying rod position indicated two rods at unknown position and one rod at position 28.
The rod displayed at position 28 was in fact fully inserted.
The reactor analyst recommended that the RO select several rods including rods not in the latched group.
Different rods were selected in succession without movement.
The RWM continued to display group 1 as the latched group.
At no time did the RWM indicate a "select error" when a rod not in the latched group was selected.
The reactor analyst concluded that the RWM was inoperable after,verifying with a computer technician in the computer room that the RWM program was not running.
The computer technician restarted the RWM program and the reactor analyst re-initialized the RWM by placing the bypass key in
"NORMAL", depressing the "SYSTEM/INITIALIZE"button and the
"SYSTEM DIAGNOSTIC" button.
The reactor analyst recommended that each of the eight previously withdrawn control rods be selected in the order in which they had been withdrawn.
This was performed and, by the reactor analyst's recommendation, the ninth rod in group 1 was with-drawn to its full out position.
The tenth and el.eventh rods in
group 1 were also fully withdrawn following the same procedure, whereupon the resident inspector observed that no check had been done to verify that the
"SELECT ERROR" function of the RWM was restored to service.
A control rod not in the latched group was selected to determine if the RWM would display a
"SELECT ERROR".
No error light was displayed, and the RWM was declared inoperable by the Reactor Analyst Supervisor.
The Station Shift Supervisor instructed the Reactor Operator to cease moving control rods.
An unrelated problem involving a leak near the containment penetration for the Hydrogen Injection Water Monitoring system was identified at approximately this time, and the decision was made to commence a shutdown in accordance with Technical Specifications.
All rods were inserted following the established rod sequence.
Rod Movement with Ino erable Rod Worth Minimizer When the reactor startup began at 10:16 a.m.
on June 18, 1986, Rod Worth Minimizer operability had been demonstrated by performance of surveillance test N1-ST-V3, Rod Worth minimizer Operability.
The inspector verified that the surveillance was conducted by a licensed operator on June 18, 1986 prior to startup.
When the reactor analyst noted the computer rod position display indicated rods at unknown positions and one rod indicated at an incorrect position, he stopped control rod withdrawal to investigate the RWM trouble.
Investigation revealed the RWM did not indicate an error when a rod not in the latched group was selected.
His phone conversation with the computer technician revealed that the RWM program was not running.
The program was restarted and the RWM re-initialized as stated above.
No surveil-lance test was done to verify RWM operability, nor was a check done to verify the operability of the
"SELECT ERROR" function of the RWM.
Contrary to Technical Specification 3. 1. l.b.(3)(b) which requires that whenever the reactor is below 20% rated thermal power and 12 or fewer control rods have been withdrawn, no control rods shall be moved unless the rod worth minimizer is operable, three control rods were withdrawn with an inoperable Rod Worth Minimizer.
This is a violation.
(50-220/86-09-01)
Ade uac of Control over Rod Worth Minimizer Maintenance The withdrawal of control rods with an inoperable Rod Worth Minimizer (RWM) was due, in part, to an apparent lack of control over the corrective maintenance activity.
When the RWM was found inoperable after withdrawal of the eighth control rod, corrective maintenance was performed without procedural control.
This contributed to the failure to perform post-maintenance testing.
Although no previous
incidents have been identified, implementation of Administrative Procedure 5.0 for computer software maintenance and similar type maintenance activities does not appear to be complete.
The inspectors will follow licensee efforts to resolve this deficiency.
(50-220/86-09-02)
In addition, the Rod Worth Minimizer Operability Test, Nl-ST-V3, performed to insure RWM operability prior to startup, cannot be performed after control rods have been withdrawn because one pre-requisite of the test is to insure that all control rods are fully inserted.
The licensee has nearly completed development of RWM surveillance test to be used in the event of RWM failure after control rods have been withdrawn.
The inspectors will follow the licensee's implementation of this procedure.
(50-220/86-09-03)
Conduct of Control Room 0 erations Durin Startu The inspector noted that a large number of people were present in the control room during reactor startup.
The licensee reduced access to the control room during a subsequent startup by securing one of the doors to the control room and posting a "Keep Out" sign.
The inspector will monitor the presence of unnecessary personnel during future control room activities.
Technical S ecifications Technical Specification 3. 1. l.b.(3)(b) requires that when reactor power is less than 20% no control rods shall be moved without an operable Rod Worth Minimizer, except that it may be bypassed if it fails after the withdrawal of at least 12 control rods.
Technical Specification 3. l.l.e, quoted above, requires that when Technical Specifications 3. 1. l.a through d. are not met, the reactor shall be placed in hot shutdown within ten hours.
The two specifications contradict each other where a failure of the RWM occurs when less than 12 control rods have been withdrawn.
A shutdown is required by 3. 1. l.e, but rod movement is prohibited by 3. 1. 1.b.(3)(b).
It appears that this problem exists at several BWRs with Rod Worth Minimizers that do not have a
The licensee intends to pursue resolution of this conflict with NRR, possibly by means of a Technical Specification change.
Technical Specification 4. l.l.b.(3)(a)(iv) requires that one condi-tion of Rod Worth Minimizer operability is to verify the Rod Block function of the RWM by attempting to withdraw an out-of-sequence control rod beyond the rod block point.
Verification of this operability condition is not prudent if the RWM fails during with-drawal of the first 12 control rods'his check could cause the event that the RWM is designed to prevent; the withdrawal of a high worth rod at low power followed by a rod drop accident.
The result could be fuel enthalpies in excess of 280 calories per gram, causing fuel
e'
damage and dispersal.
The licensee intends to pursue resolution of this conflict with NRR, possibly by means of a Technical Specification change.
The inspectors will follow licensee efforts to resolve both of these problems.
Rod Worth Minimizer Select Block Function One of the windows on the Rod Worth Minimizer Operator Display Console in the Nine Mile Point Unit 1 control room is the
"SELECT BLOCK" window.
The licensee believes that this function was included in the original RWM installation and was removed at a later time at the recommendation of the vendor.
The purpose of the select block function is to prevent control rod motion in the event of a computer hardware or software failure such as that experienced in this event.
The resident inspector will review the licensee's documents in a
future report.
(50-220/86-09-04)
Rod Worth Minimizer Trainin Intervi,ews with several licensee personnel indicate a need for increased training on the operation of the Rod 'Worth Minimizer.
When questioned by the resident inspector, several reactor analysts and operators seemed confused with regard to various functions of the RWM.
lhe licensee has committed to conducting additional RWM training.
Inde endent Yerification of Control Rod Withdrawal Se uence During the star tup on June 18, 1986, the reactor analyst held the rod withdrawal sequence.
The reactor analyst made recommendations to the reactor operator, who then acted on the recommendations.
The reactor operator did not have his own copy of the rod withdrawal sequence and the reactor analyst is not a licensed operator.
It appeared that there was no independent decision-making process on the part of the licensed operator during a normal startup with the RWM operable, since the reactor operator did not have the rod withdrawal sequence so that he could insure a valid recommendation from the reactor analyst.
Responsibility for withdrawing control rods rests with the licensed reactor operator.
Therefore, the licensed operator must have a copy of the withdrawal sequence.
The licensee has committed to conduct training to emphasize the responsibilities of reactor analysts and reactor operators.
Safet Si nificance of 0 erational Event The Bases for Technical Specification section 3. 1. 1 states that control rod withdrawal and insertion sequences are established to assure that individual control rod worth could not be enough to cause
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the core to be more than 0.013 delta k supercritical if the highest worth rod were to drop out of the core in the manner defined for the Rod Drop Accident.
This 0.013 delta k limit, together with the integral rod velocity limiters and the action of the control rod drive system, limits potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximim fuel energy content of 280 cal/gm, the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental data.
The purpose of the RWM is to ensure operator observance of a rod withdrawal sequence designed to limit individual control rod worth.
The RWM accomplishes this by blocking rod motion in the event that it determines that the operator has deviated from the stored rod with-drawal sequence.
Although rods were withdrawn without an operable Rod Worth Minimizer, minimal safety significance is attached to this event since the rod sequence was followed when control rods were withdrawn.
k.
Possible eneric Rod Worth Minimizer software defect During a startup on June 18, 1986, the licensee declared the Rod Worth Minimizer (RWM) inoperable due to a software failure in the Process Computer (see daily report of June 19, 1986).
The model 4400 Rod Worth Minimizer, supplied by General Electric, uses software in the process computer to compare actual rod withdrawal sequence against a sequence stored in memory.
To provide for failed reed switches in the Rod Position Information System, a program step to obtain substitute rod position information from memory was included.
This program step was directed to an incorrect memory address, which caused the RWM program to become caught in a loop.
Although the RWM is designed to apply an insert and withdraw block when a software fai lure is identified, being caught in the loop prevented identifi-cation of a software failure.
The main program eventually locked out the RWM program because it used too computer much time.
The Rod Worth Minimizer appeared to be functioning normally when, in fact, it was inoperable.
The licensee states that this problem has potentially existed for several years, since the Process Computer was first installed and that this same problem could exist at any plant with a Rod Worth Minimizer incorporated into the Process Compute~.
The inspector will review the licensee LER on this issue during a
subsequent inspection.
4.
Desi n Chan es and Modifications On June 18, 1986, while shutting down to repair the Rod Worth Minimizer, the licensee discovered a small leak in a 3/4 inch primary coolant sample line just outside its primary containment penetration.
The line was installed during the recent refueling outage as part of an experimental hydrogen injection/primary water chemistry monitoring system to prevent stress corrosion cracking.
The leak was apparently caused during welding
on the sample line and was not discovered previously because the welding was done after the hydro of the primary system.
This incident will be covered in greater detail in NRC Inspection Report 86-10.
5.
0 erational Safet Verification a
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Control Room Observation Routinely throughout the inspection period, the inspectors independ-ently verified plant parameters and equipment availability of engineered safeguard features.
The following items were observed:
Proper control room manning and access control; Adherence to approved procedures for ongoing activities; Proper valve and breaker alignment of safety systems and emergency power sources; Reactor control panel instrumentation and recorder traces; Reactor protection system instruments to determine that the required channels are operable; Stack gas monitor recorder traces; Core thermal limits; and Shift turnover.
b.
~R"" fL ldG i
R d
The inspectors reviewed the following logs and instructions:
Control Room Log Book Station Shift Supervisor's Log Book Station Shift Supervisor's Instructions Reactor Operation Log Book The logs and instructions were reviewed to:
Obtain information on plant problems and operation; Detect changes and trends in performance; Detect possible conflicts with Technical Specifications or regulatory requirements; Assess the effectiveness of the communications provided by the logs and instructions; and Determine that the reporting requirements of Technical Specifications are met.
No violations were identifie I
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6.
Observation of Ph sical Securit The inspectors made observations to verify that selected aspects of the plant's physical security system were in accordance with regulatory requirements, physical security plan and approved procedures.
The follow-ing observations relating to physical security were made:
The security force was properly manned and appeared capable of performing their assigned functions.
Protected area barriers were intact with gates and doors closed and locked if not attended.
Isolation zones were free of visual obstructions and objects that could aid an intruder in penetrating the protected area.
Persons and packages were checked prior to entry into the protected area.
Vehicles were properly authorized, searched and escorted or controlled within the protected area.
Persons within the protected area displayed photo badges, persons in vital areas were properly authorized, and persons requiring an escort were properly escorted.
Compensatory measures were implemented during periods of equipment failure.
No violations were identified.
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7.
Plant Tours During the inspection period, the inspectors made frequent tours of plant areas to make an independent assessment of equipment conditions, radio-logical conditions, safety and adherence to regulatory requirements.
The following areas were among those inspected:
Turbine Building Auxiliary Control Room Vital Switchgear Rooms Cable Spreading Room Diesel Generator Rooms Reactor Building The following items were observed or verified:
a
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Fire Protection:
Randomly selected fire extinguishers were accessible and inspected on schedule.
Fire doors were unobstructed and in their proper position.
Ignition sources and combustible materials were controlled in accordance with the licensee's approved procedures.
Appropriate fire watches or fire patrols were stationed when equipment was out of servic b.
E ui ment Controls:
Jumper and equipment mark-ups did not conflict with Technical Specification requirements.
Conditions requiring the use of jumpers received prompt licensee attention.
Administrative controls for the use of electrical jumpers and equipment mark-ups were properly implemented.
c.
Vital Instrumentation:
Selected instruments appeared functional and demonstrated parameters were within Technical Specification Limiting Conditions for Operation.
d.
Radioactive Waste S stem Controls:
Gaseous releases were monitored and recorded.
No unexpected gaseous releases occurred.
Plant housekeeping and cleanliness were in accordance with approved licensee programs.
f.
Radiation Protection:
Personnel monitoring was properly conducted.
Randomly selected radiation protection instruments were calibrated and operable.
Radiation Work Permit requirements were being followed.
Area surveys were properly conducted and the Radiation Work Permits were appropriate for the as-found conditions..
No violations were noted.
8.
Review of Licensee Event Re orts LERs The LER's submitted to NRC, Region I were reviewed to determine whether the details were clearly reported, including accuracy of the description of the cause and adequacy of the corrective action.
The inspectors also determined whether the assessment of potential safety consequences had been properly evaluated, whether generic implications were indicated, whether the event warranted on site follow-up and whether the reporting requirements of 10 CFR 50.73 had been me During this inspection period, the following LERs were reviewed:
LER No.
Event Date Subject 86-10 May 5, 1986 Actuation of Reactor Building Ventilation Resulting from Blown Fuse 86-11 May 8, 1986 Contaminated Injury to Contractor Personnel No violations were identified.
9.
Unusual Event On June 4,
1986, the licensee declared an Unusual Event in accordance with their Emergency Response Plan after the licensee's radiological survey robot exploded.
The explosion resulted from the robot's batteries being improperly vented while being recharged.
The robot was located inside the station protected area.
Nine station employees received minor injuries from the explosion.
They were all examined onsite, transported to a local hospital for observation and then released.
The Unusual Event was down-graded immediately after being declared.
The licensee notified the NRC Headquarters Duty Officer of the event via the Emergency Notification System and made a subsequent press release.
On June 5,
1986, Department of Labor investigators conducted an onsite review of the event.
No deficiencies were noted by the resident inspector reviewing the event.
10.
Safet S stem 0 erabilit Verification On a sampling basis, the inspectors directly examined selected safety system trains to verify that the systems were properly aligned in the standby mode.
This examination included:
Standby Liquid Control System Emergency Diesel Generator System Containment Spray System Core Spray System Reactor Building Emergency Ventilation System No violations were note.
Review of Periodic and S ecial Re orts The inspectors reviewed the following periodic and special reports to determine whether the safety significance of each event has been properly evaluated, to monitor plant operations, to determine if appropriate corrective action has'been taken, and to ensure compliance with NRC reporting requirements.
During this inspection period, the following reports were reviewed:
Date Subject Special Special Special Special Special April 11,1986 Fire systems and penetrations.
April 30,1986 Fire systems and penetrations.
May 9, 1986 Fire detection systems.
May 12, 1986 Fire bar rier penetrations.
May 16, 1986 Fire barrier penetrations.
Special May 30, 1986 Fire suppression systems.
Monthly June 2, 1986 Operating Statistics for May, 1986 Special Special June 5,
1986 Fire systems and penetrations.
June 13,1986 Fire barrier penetrations.
The inspector determined that the requirements of Technical Specifications were met and no violations were noted.
= 12.
Review of Static 0-Rin Inc. Pressure Switches During evaluation of a June 1 operational event involving the Reactor Protection System (RPS) at LaSalle 2, differential pressure switches manufactured by Static 0-Ring (SOR), Inc. (series 103) were found to operate erratically, The problem was described in Inspection
& Enforce-ment (IE) Notice 86-47 dated June 10, 1986.
The inspector reviewed the usage of SOR pressure switches at Unit 1 and found no differential pressure switches (including series 103)
used at Unit 1.
The inspector found SOR pressure switches used in the following applications:
the Reactor Building to Torus vacuum breakers, the Post Accident Sampling System, and the pump suction alarms of the Reactor Building Closed Loop Cooling, Shutdown Cooling, Fuel Pool Cooling, and Containment Spray systems.
The inspector reviewed
I.'4
the calibration records of the vacuum breaker and pump suction instruments and found no evidence of erratic behavior or excessive drift.
The inspector confirmed that the procedures did not allow for multiple attempts to achieve acceptable readings.
Based on the above results the inspector concluded that the problems described in Notice 86-47 did not apply to Unit 1, pending any further evaluation results concerning SOR, Inc. switches.
Radiolo ical Environmental Monitorin Pro ram The inspector reviewed the licensee's Radiological Environmnetal Monitor-ing Program Annual Report for 1985.
This report summarizes the results of the sampling and analyses of environmental media to determine the radio-logical impact of station operations.
These environmental media include air, water, vegetation, and aquatic plants and animals.
In addition, direct radiation is monitored by placement of thermoluminescent dosimeters at various locations around the station.
As a result of this review, the inspector determined that the licensee has generally complied with its Technical Specification requirements for sam-pling frequencies, types of measurements, analytical sensitivities, and reporting schedules.
Exceptions to the sampling and analysis program were adequately explained, e.g.,
low air sample volume due to power failure.
The report included summaries of the laboratory quality assurance program and of the land use survey.
The analyses of environmental samples indicated that doses to humans from radionuclides of station origin were negligible.
Followu of IE Information Notice 86-03 IE Information Notice 86-03 identified potential deficiencies in the environmental qualification of Limitorque valve operator motor wiring.
Licensees were expected to review this information for applicablilty to their facility and consider appropriate corrective action if necessary.
The inspector determined that the licensee did review this Notice and during their 1986 Refueling and Maintenance Outage conducted a comprehensive review of all station Limitorque motors via their preventive maintenance and environmental qualification programs.
Nontraceable motor wiring was replaced one for one with environmentally qualified wires.
The inspector had no further questions.
Alle ation Followu On June 13, 1986, the NRC received an allegation (RI-86-A-0073) that a
contractor employee was being required to wear a respirator in a radiation area without first being required to undergo a Whole Body Count and with-out proper training.
The alleger also had concerns about the results of a
Whole Body Count.
Immediate investigation revealed that the contractor was no longer working in the radiation area.
Further investigation revealed that the contractor employee had been previously employed by another contractor company onsite and had received a termination Whole Body Count ten days previous to being re-employed onsite.
Since the employee had not worked at a site other than Nine Mile Point in the ten day period, the licensee used his termination Whole Body Count as his new initial Whole Body Count.
The inspector verified that the employee's respirator and radiation protection training were current.
The inspector also found that the termination Whole Body Count indicated the employee's accumulated dose was very low (less than 2% maximum permissible body burden to the lungs).
No violation was found..
16.'urveillance Testin The inspectors witnessed the performance of selected surveillance to verify that:
Surveillance procedures conform to technical specification require-ments and have been properly approved.
Test instrumentation is calibrated.
Limiting conditions for operation for removing equipment from service are me
.
Surveillance schedule is met.
Test results met Technical Specification requirements.
Appropriate corrective action is initiated, if necessary.
Equipment is properly restored to service following the test.
The following tests were included in this review.
Nl-ISI-HYD-01, "Reactor Pressure Vessel and all ASME Class 1 Systems, Main Steam
& Reactor Vessel Instrumentation, Class 2 Systems" Rev.
0.
The inspector noted that stub tube 37-07 was identified to be leaking 4-8 drops per minute.
The stub tube was rolled.per Nl-MMP-45.99, ".Rolling of Control Rod Drive Housing" Rev.
2.
Nl-RPSP-8,
"Emergency Cooling System Heat Removal Capability Test",
Rev.
1.
Nl-ST-C2,
"Manual Opening of the Solenoid-Actuated Relief Valves and Flow Indication", Rev.
N1-ST-C3,
"Auto Start of High Pressure Coolant Injection", Rev. 2.
No discrepancies were note.
Exit Interview At periodic intervals throughout the reporting period and at the end of the reporting period, the inspectors met with senior station management to discuss the inspection scope and findings.
Based on the NRC Region I review of this report, it was determined that this report does not contain information subject to
CFR 2.790 restriction I