IR 05000220/1982019
| ML20028B365 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/12/1982 |
| From: | Baunack W, Sharon Hudson, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028B357 | List: |
| References | |
| TASK-1.A.2.1, TASK-2.B.4, TASK-TM 50-220-82-19, IEB-79-12, IEB-79-26, IEB-80-06, IEB-80-08, IEB-80-14, IEB-80-17, IEB-80-6, IEB-80-8, NUDOCS 8211300323 | |
| Download: ML20028B365 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I
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82-19 Report No.
Docket No.
50-220 i
DPR-63 C
License No.
Priority Category
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Licensee:
Niagara Mohawk Power Corporation
300 Erie Boulevard West
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Syracuse, New York 13202 I
Facility Name:
Nine Mile Point Nuclear Station, Unit 1 Inspection at:
Scriba, New York Inspection conducted: October 12-31, 1982 Inspectors: 2(b.
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~S. D. Huds'on, Senior Resident Inspector
'datt signed dc.
Sun b
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~W. H. Baunack, Projed Inspector date' signed da e si ned
/ /M[2 Approved by:
R. B. Kisteh-Chief, Reactor Projects date' signed Section 1C Inspection Summary:
Inspection on October 12-31, 1982 (Report No. 50-220/82-19)
Areas Inspected:
Routine, onsite regular and backshift inspections by the resident inspector and one regional based inspector (94 hours0.00109 days <br />0.0261 hours <br />1.554233e-4 weeks <br />3.5767e-5 months <br />). Areas inspected included:
licensee action on previous inspection findings, plant tours, observation of physical security, plant maintenance, quality assurance, IE Bulletins, and review of TMI Action Plan items.
Results:
No violations were identified in the areas inspected.
8211300323 821116 PDR ADOCK 05000220
PDR Region I Form 12 (Rev. April 77)
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DETAILS 1.
Persons Contacted J. Aldrich, Supervisor, Operations T. Breigle, Lead Q. C. Engineer K. Dahlberg, Site Maintenance Superintendent W. Drews, Technical Superintendent J. Duell, Supervisor, Chemistry and Radiation Protection G. Gresock, Safe End Project Manager F. Hawksley, Supervisor, Mech. Maintenance E. Leach, Superintendent of Chemistry and Radiation Management T. Perkins, General Superintendent, Nuclear Generation T. Roman, Station Superintendent The inspector 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.
2.
Licensee Action on Previous Inspection Findings (0 pen)
INSPECTOR FOLLOWUP ITEM (82-01-05):
Close-out of safety-related modifications.
In January 1982, the licensee began a program to review all safety-related modifications dated 1975 to 1979 to verify proper e
close-out of each modification package. Close-out should include updating of drawings and procedures, and a review of documentation to ensure that the work was properly completed. The initial review is being conducted by the Quality Assurance Department. About 40 of the 60 modifications during the period in question have been reviewed to date. The licensee expects to complete the initial review by December 1982.
3.
Plant Tours
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(1) During the inspection period, the inspector made multiple tours of plant areas to make an independent assessment of equiraent conditions, radiological conditions, safety and adherence to regulatory requirements.
The following areas were among those inspected:
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Control Room Turbine Building
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Auxiliary Cc6 trol Room Vital Switchgear Rooms
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Yard Areas
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Radwaste Area
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Diesel Generator Rooms
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Screen House
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Reactor Building
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Drywell
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(2) The following items were observed or verified:
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(a) Radiation Protection:
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Personnel monitoring was properly conducted.
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Randomly selected radiation protection instruments were
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calibrated and operable.
Area surveys were properly conducted and the Radiation
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Work Permits were appropriate for the as-found conditions.
The inspector observed the routine radiation survey of the
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Reactor Building and air sampling while machine cutting
- 14 recirculation loop.
Radiation Work Permit requirements were being followed.
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(b) Fire Protection:
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Randomly selected fire extinguishers were accessible and inspected on schedule.
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Fire doors were unobstructed and in their proper position.
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Ignition sources and combustible materials were controlled in accordance with the licensee's approved procedures.
Fire watches were posted during periods when smoke detection
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equipment was out of service.
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(c) Equipment Controls:
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Jumpers and equipment tagouts did not conflict with Technical Specification requirements.
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The inspector independently verified that the following
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tagouts had been properly conducted by observing the position of breaks and/or valves:
RMU# 134488 on the Centrol Room Emergency Ventilation System RMU# 134486 on the Control Rod Drive System RMU# 134471 on the Fire Detection System
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x (d) Radioactive Waste System Controls:
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On October 19; 1982, the inspector witnessed the discharge
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of #12 Condensate Storage Tank to Lake Ontario. The release consisted of 200,000 gallons demineralized water with an activity of 1.4x10-6 microcuries/ml. The inspector reviewed the Liquid Radioactive Waste Discharge Fom and detemined sthat the tank had been properly recirculated prior to
- sampling, a representative sample had been analyzed, and tiquid effluent discharge monitor alams had been set prior
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- to discharge. The insractor reviewed the valve lineup check
- list and detemined that the discharge was being conducted in accordance with an approved procedure.
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On October 21, 1982, the inspector observed the temporary storage area for radioactive material on the Turbine Building g
elevation 261. The material consisted of scrap recirculation system piping packaged for shipment in wooden boxes. The highest radiation readings on the boxes were 450 mrem /hr.
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These boxes were stored against a wall and_other boxes ready for shipment were used to shield the boxes containing the
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ost highly contaminated piping. The area was properly posted.
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On October 29, 1982, the inspector witnessed the preliminary survey of radioactive waste shipment #NMP-HW-1182-1. The shipnent consisted of scrap recirculation piping packaged in wooden boxes inside a shielded exclusive use van. At one location, the radiation measurement at six fe'et from the side of the van was.10 mrem /hr. The technician performing
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the survey confimed the measurement with another radiation instrument.
49 CFR 173.393 states that a package may not exceed 10 mrem /hr. at any point 2 meters (6 feet) from the outer surface of the vehicle. The licensee added additional shielding to further assure that the radiations level met the required limit. The inspector reviewed Radiation Survey
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- 70235 perfomed on November 1,1982. The maximum radiation measurement had been reduced to,8 mrem /hr at' 6 feet from the
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van.
(e) Review of Logs and Operating Records:
'The inspector reviewed the following logs and instructions for
'he period October 12, 1982 through October 31, 1982:
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Control Rocm Log Book
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Station Shift Supervisor's Log Book
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Station Shift Supervisor's Instructions Safe End Project Log
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The logs and instructions were reviewed to:
Obtain infomation on plant problems and operation;
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Detect changes and trends in perfomance; Detect possible conflicts with technical specifications or
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regulatory requirements; Detemine that records are being maintained and reviewed as
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required, and Determine that the reporting requirements of technical
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specifications are met.
No violations were identified.
4.
Observation of Physical Security The inspector made observations and verified during regular and off-shift hours that selected aspects of the plants physical security system were in accordance with regulatory requirements, physicht security plan and approved procedures. The following observations relating to tre physical security plan were made:
The security force was properly manned and appeared capable of
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perfoming their assigned functions.
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Protected area barriers were intact - gates and doors closed and locked if not attended.
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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
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area.
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Vehicles were properly authorized, searched and escorted or controlled within the protected area.
Persons within the protected area displayed photo-identification
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badges, persons in vital areas were properly authorized, and persons
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requiring an escort were properly escorted.
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On October 26, 1982, the inspector discovered an equipment failure that pemitted access to the control room from the administration building. The inspector detemined that the alam system remained operational during the eqdpment failure and that no unauthorized entries into the control room were made. The inspector verified that the required compens'atory measures were implemented within the time specified in the licensee's security plan.
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No violations were identifie.
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5.
Plant Maintenance The inspector examined portions of various safety-related maintenance activities. Through direct observation and review of records, he determined that:
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These activities did not violate the limiting conditions for operation.
Required administrative approvals and tagouts were obtained prior to
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initiating the work.
Approved procedures were used or the activity was within the " skills
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of the trade".
Appropriate radiciogical controls were properly implemented.
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Quality Control hold points were observed.
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Ignition / fire prevention controls are appropriate.
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During this inspection period, the following maintenance activities were examined:
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Grinding to remove a dye penetrant indication on #15 recirc discharge nozzle.
Weld crown reduction on #11 recirc suction nozzle.
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Machine cutting on #14 recirc pump suction, #13 and #15 recirc pump
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discharges.
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Machining of #14 recirc discharge valve body.
Welding on FW-13D-SEN and FW-13-S-7.
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No violations were identified.
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Quality Assurance of Safety Related Maintenance The inspector examined the licensee's actions on unaccepted field welds of the recirculation (recirc) system piping. Repair of such welds are classified as in-progress repairs and a Quality Assurance Nonconformance Report is not prepared for such welds. The inspector reviewed the
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procedures for repair of defects in welds described in CWI-1399K-W001, l
" Welding for NMPC Recirculation Loop," Revision 7, dated July 21, 1982.
While welding the root pass for #13 recirc system discharge nozzle to the safe end on October 16, 1982, the Argon purge was lost. The inspector
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Jetemined that the unacceptable weld had been properly documented. A l
repair of the weld was perfomed in accordance with an approved procedure.
The inspector witnessed the radiographic examination (RT) of this weld on
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October 20, 1982 and reviewed Radiography Report #NNI-RT-135. The inspector verified that the individuals perfoming the RT were properly certified as Level II technicians.
As a result of this RT, the weld was rejected due to porosity and incomplete fusion. Subsequent repairs were performed and the weld was determined to be acceptable on October 22, 1982 by RT.
7.
Licensee Action on IE Bulletins The inspector reviewed the IE Bulletins listed below to verify that an accurate written response was submitted to the NRC, the response contains adequate corrective action based on the infomation presented in the Bulletin, and the corrective action has been properly implemented:
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IEB 79-12. "Short Period Scrams at BWR Facilities" The inspector reviewed Operating Procedure OP-43, "Startup and Shutdown Procedure," Revision 18, dated August 9, 1982. A step was added to this procedure to prevent continuous rod withdrawal during the approach to criticality and to caution the operators of the high rod worth that is exhibited when withdrawing the first rod in a new group.
The procedure requires one notch step withdrawal of control rods in the power region of the fuel. The power regions of the fuel are within notch position 00 to 30. The Reactor Analyst
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Instruction, " Approach to Criticality" states that one notch step withdrawal is required between notch positions 00 to 30 except for Group I rods. The Technical Superintendent stated that the Reactor Analyst or Reactor Analyst Technician monitoring the startup will instruct the operators when to begin one notch step withdrawal of the rods. Near the end of a fuel cycle, this could be as late as Rod Group IV or V.
He further stated that the above procedures would be revised to reflect this position.
The Bulletin also required an evaluation of the operability of the
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" emergency rod in" switch to perfom its function under prolonged severe use. The inspector reviewed Work Request #10633 dated July 18, 1979 and Quality Control Inspection Report #81-248 dated March 10, 1981. An inspection was performed to determine that the switch was operating properly but an evaluation was not performed to determine its ability to function under prolonged severe use.
The Bulletin further required that the licensee evaluate the use of special rod withdraw sequences during peak xenon conditions.
The Technical Superintendent stated the control rod withdraw sequences are supplied by General Electric and that peak xenon conditions are considered in establishing the sequence. Therefore, no special rod sequences are required for this condition.
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This Bulletin remains open pending the licensee's evaluation of the
" emergency rod in" switch and revision of the procedures on the use of one notch step withdrawal of control rods.
IEB 79-26, Revision 1, " Boron Loss From BWR Control Blades"
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To prevent a loss of reactivity worth of the control blades by more than 10 percent, this Bulletin reconunends the replacement of all blades expected to have greater than 34 percent Boron-10 (B10)
depletion averaged over the upper one-fourth of the blade.
The inspector discussed the licensee's monitoring program of control rod blade exposure and calculation of B10 loss with the Reactor Analyst Supervisor. The exposure used for the upcoming fuel cycle is increased by 20% above the expected exposure to allow an additional l
margin of safety in the calculation of B U depletion. The inspector reviewed the " Recommendation for End of Cycle 6 Control Rod Change-out dated February 28, 1981 and verified that blades were changgd at the end of cycle 6 to ensure that the 34 percent depletion of Bi0 was not reached during the upcoming fuel cycles. Thirty of the control rod blades were replaced in the Spring 1981 outage as a result of these recommendations. This Bulletin is closed.
IEB 80-06, " Engineered Safety Feature (ESF) Reset Control"
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Item 2 of this Bulletin requires the licensee to perfom tests to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of the,~the various isolation or actuation signals. The inspector reviewec following surveillance c.'sts to verify compliance with this requirement:
" Auto Depressurization System Operability Test," ISP-IC-7, Temp.
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Change, Revision 0, dated March 20, 1982, perfomed March 22, 1982.
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" Automatic Securing and Isolation of the Mechanical Vacuum Pumps,"
ST-C7, Temp. Change, Revision 0, dated October 29, 1981, perfomed March 21, 1982.
In each case, the emergency equipment tested remained in its proper position after removal of the initiating signal. This Bulletin is closed.
j IEB 80-08, " Examination of Containment Line Penetration Welds"
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l The licensee's response dated June 9, 1980 stated that for all containment boundary butt welds, radiographic examination was satisfactorily perfomed during the construction phase. No additional testing is required as a result of this Bulletin.
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IEB 80-10, " Containment of Nonradioactive Systems and Potential for an
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Unmonitored Release to the Environment" The inspector reviewed the results of the routine sampling program
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boiler, reactor building cooling water, turbine building cooling water, emergency condensers, service water, stonn sewers, sewage treatment plant, and instrument, service, and breathing air systems.
No unacceptable results have been detected. This Bulletin is closed.
IEB 80-17, " Failure of 76 of 185 Control Rods to Fully Insert During
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a Scram at a BWR" The original Bulletin was addressed in Inspection Report (IR) #80-07.
Supplements 1, 2, and 3 of this Bulletin was addressed in IR #80-19 and in an NRR Generic Safety Evaluation Report of the BWR Scram Discharge System (SER) dated December 9, 1980. Supplement 4 was addressed in IR #81-01. Supplement 5 deleted a special reporting requirement requested in the original Bulletin.
In the Spring 1981 refueling outage, the licensee completed a modification to improve the hydraulic coupling of the scram discharge volume (SDV) and the instrument volume (IV). The modification consisted of a larger diameter vent (2") on the SDV, motor operated vent and drain valves on the SDV and the IV, larger diameter (10")
connecting pipe between the SDV and the IV, (thus increasing the capacity of the SDV) and a separate holding tank to which the SDV would be vented and the IV drained.
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Table 1 of the SER dated December 9, 1980 specifies for inadequate hydraulic coupling, the interim corrective action is to continuously monitor the SDV for water accumulation.
Based on the installed modification to improve the hydraulic coupling, the licensee dis-continued its use of the ultrasonic continuous monitoring system on the SDV.
The inspector noted that the SER (not the Bulletin) addresses the need for diverse and redundant instrumentation on the IV. The licensee's position as stated in NMPC letter T. E. Lempges to T. A.
Ippolito, dated May 14, 1981 is that the existing redundant instrumentation and its current testing requirements are adequate.
Although this aspect of the SER is under review by the Office of Nuclear Reactor Regulation, IE Bulletin 80-17 is closed.
8.
Review of TMI Action Plan Category B Requirements A review was conducted of the licensee's programs in the areas of upgrading of Reactor and Senior Reactor Operator Training and Qualifications (NUREG-0737 Item I. A.2.1), and Training for Mitigating Core Damage (NUREG-0737, Item II.B.4).
The following records were reviewed:
Administrative Procedure No. APN-10b, " Licensed NRC Operator Retraining,"
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Revision 3, dated July 6,1982.
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R0/SR0 Requalification Lecture Curicula/ Schedules 1980, 1981 and 1982.
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Selected Lesson Plans for 1980,1981, and 1982 Lectures.
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Selected individual training records.
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Findings:
The formal Licensed Operator Requalification Training Program consists of one week of training out of each five weeks of work. This includes one week of simulator training (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), and the remainder classroom training (8 weeks in 1981) including examinations.
Records are maintained to show the training received by each individual.
The Licensed Operator Retraining Program during 1981 consisted of classroom lectures in the following areas which are mitigating core damage training or are considered related subjects.
Mitigating Core Damage (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />), which included review of heat
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transfer, heat transfer systems, process instrumentation, recognizing core damage, mitigation of core damage, and site emergency plan procedures.
Reactor Theory (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).
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Instrumentation (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).
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Emergency Systems (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).
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Plant Systems (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).
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Review of plant instrumentation, heat transfer, recognition and mitigation of core damage, reactor theory, emergency procedures, and systems (16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />).
In addition, during 1980 mandatory trcining was conducted in heat transfer and themodynamics (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />), and fluid flow and core themal characteristics (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />) for all STA's, SR0's, and SSS's.
Records show that most R0's also attended these courses. Also, in 1982 courses were conducted in heat transfer and fluid dynamics (11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />).
The total training associated with mitigating core damage and related subjects is clearly in excess of the 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> considered to be adequate for this topic.
All personnel in the operating chain and shift technical advisors have received the training for mitigating core damage through the retraining program.
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To ensure that replacement personnel in non-licensed positions in the operating chain receive the reviewed mitigating core damage training, the licensee will include the requirement for the training in procedure APN-10c, " Training of Non-Licensed Perscnnel."
No violations were identified.
9.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior station renagement to discuss the inspection scope and findings.
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