IR 05000220/1982001
| ML20052H967 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/28/1982 |
| From: | Doerflein L, Sharon Hudson, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20052H960 | List: |
| References | |
| 50-220-82-01, NUDOCS 8205240260 | |
| Download: ML20052H967 (18) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I Report No.
82-01 Docket No.
50-220 C
License No. OPR-63 Priority Category
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Licensee:
Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility Name: Nine Mile Point Nuclear Station, Unit 1 Inspection at: Scriba, New York Inspection conducted: January 1 to February 7, 1982 Inspectors:
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S. D. Hudson, Resident Inspector
/date/ signed b.
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L. T. Doerflein, Resid6nt Inspector
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date signed Approved by: Mk&
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H.'B. KisteMhief, Reactor Projects
'date signed
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Section lc Inspection Summary:
Inspection on January 1 to February 7,1982 (Inspection Report No. 50-220/82-01)
Areas Inspected: Routine, onsite regular and backshift inspections by the resident inspectors (177 hours0.00205 days <br />0.0492 hours <br />2.926587e-4 weeks <br />6.73485e-5 months <br />).
Areas inspected included:
licensee action on previous inspection findings, operational safety verification, physical security, plant tours, surveillance tests, plant maintenance, safety system verification, combustible gas control system, modification activities, and licensee action on IE Circulars.
Results:
No violations were identified in eight of the areas inspected. One violation was identified in each of two areas, (Failure to translate regulatory requirements into procedures and failure to revise drawings and procedures after plantmodifications).
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j Region I Form 12 l
(Rev. April 77)
8205240260 820507
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DETAILS 1.
Persons Contacted J. Aldrich, Supervisor, Operations K. Dahlberg, Site Maintenance Superintendent W. Drews, Technical Superintendent J. Duell, Supervisor, Chemistry and Radiation Protection E. Leach, Superintendent of Chemistry and Radiation Management D. Palmer, Supervisor, Quality Assurance T. Perkins, General Superintendent, Nuclear Generation R. Raymond, Supervisor, Fire Protection T. Roman, Station Superintendent B. Taylor, Supervisor, Instrument and Control K. Zollitsch, Superintendent, Training The inspector also interviewed and talked with 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 (Closed)
INSPECTOR FOLLOW ITEM (78-15-03):
Inadequate lake water analysis.
Starting in July 1978, the licensee sent the lake water samples off site to a contractor's laboratory for analysis. The change in analytical labs eliminated the in-plant background radiation interference which had caused the inadequate lake water analysis.
In January 1981, the licensee began using the Environmental Analytical Laboratory at the James A. FitzPatrick Nuclear Power Plant to analyze the samples. The inspector reviewed the
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lake water sample results for 1978, 1979 and 1980 and found no further
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discrepancies with the data.
(Closed) DEFICIENCY (79-23-03): Three examples of failure to properly document and review plant test procedures.
Procedure N1-RAP-03, " Full Core Con:'rol Rod Scram Timing Sequence,"
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was improperly completed in that a step concerning individual rod speeds was not verified and annotattd as "Not Applicable". The licensee's reply to this item, dated November 5,1979, stated the deficiency was not correct in that T.S. 3.1.1.c.(2) would only be applicable if T.S. 3.1.1.c.(1) was not satisfied. During a subsequent inspection (80-05), the inspector advised the licensee that T.S.
3.1.1.c.(1) specifies maximum limits on average scram insertion times for all rods while T.S. 3.1.1.c.(2) specifies maximum limits on individual rod scram insertion times, and that both limits are applicable and must be met except as noted in T.S. 3.1.1.c.(3). The licensee concurred with the inspector's comments and stated that, in the future, both steps will be considered applicable.
Since inspection 80-05, the licensee has reissued N1-RAP-03 as a startup test procedure. The inspector reviewed N1-RPSTP-2, " Full Core Control
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Rod Scram Timing Sequence," Revision 2, dated September 1981 and verified the acceptance criteria specified in T.S. 3.1.1.c.(1)
through T.S. '..l.l.c.(3) would be satisfied. The inspector also reviewed the data for the last time the test was performed on June 24, 1981 and verified that both T.S. 3.1.1.c.(1) and T.S.
3.1.1.c.(2) were completed.
Procedure N1-RPSP-04, " Reactivity Margin-Incore Loading," requires
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a second verification of the rod selected be made and this second verification was not documented for tests performed on June 21, 1979.
The licensee has since reissued N1-RPSP-04 as a startup test procedure.
The inspector reviewed N1-RPSTP-5, " Reactivity Margin-Core Loading,"
Revision 1, dated June 1981 and verified that now the procedure and data sheet both require a second person perform an independent verification of the control rod selected. The inspector also reviewed the data for the last time the test was perfomed on June 30, 1981 and verified that a second verification of the control rod selected was made and documented on the data sheet.
Procedure N1-RPSP-03, " Reactivity Anomalies," performed during the
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startup test program on July 3, 1979, did not document control rod sequence being used and a step was verified as completed when plant conditions were such that the step could not be completed. The licensee stated the reason for the discrepancy on power level was that N1-RPSP-03, which is nomally done at operating power levels, was used as a substitute for a startup test. To prevent recurrance, the licensee has developed and issued a separate procedure for use during startups. The inspector reviewed this startup test procedure, N1-RPSTP-1, " Cold Critical Comparison," Revision 0, dated July 1980, including the data for the last time the test was performed on July 2, 1981. The inspector verified the completed data sheet documented the control rod sequence being used and no further discrepancies with the procedure or the data were identified.
(Closed)
INFRACTION (80-05-01):
Failure to update and maintain current valve checkoff lists for NSSS and Cardox Systems. The inspector reviewed the following procedures to verify the deficiencies in the valve checkoff lists (VCOL) had been corrected:
N1-0P-01, " Nuclear Steam Supply System," Revision 14, dated June 1981.
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The inspector verified that the VCOL contains the electromatic relief valve blocking valves (MS1-6). The inspector also reviewed the last completed VCOL and verified that MSl-6 were in the locked open position.
i N1-0P-21, " Fire Protection System," Revision 5, dated June 1981. The
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inspector verified that the Cardox System supply valve and the Cardox System supply valve bypass valve were included on the VCOL.
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N1-0P-20, " Service, Instrument and Breathing Air Systems," Revision 8,
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dated September 1981. The inspector verified that valves BA-114-30
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and IA-40 had been added to the VCOL and that valve TCLC97 is properly listed as closed on the VCOL.
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N1-0P-17 " Makeup Demineralizer System," Revision 4, dated
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September 1980. The inspector verified that valve MUD-44 was added to the VCOL.
(Closed) UNRESOLVED ITEM (80-16-03): As-built drawings do not reflect current status. The inspector reviewed DCI-3, " Drawing Control Instruction,"
Revision 4, dated October 1981, and noted the procedure requires that, in addition to the latest available " Final Print" (as-built), any current
" Construction Print" also be maintained. Construction Prints are defined as drawings representing a proposed or interim configuration of the system, equipment or circuit. The procedure notes that although the Construction Print will bear a later revision number than the Final Print, the Construction Print may not necessarily represent the actual as-built configuration of the system. The procedure also makes it clear that Final Prints are the latest existing revisions which may be utilized in normal station maintenance and/or operation. The inspector held discussions with several licensee personnel to confirm that only the Final Prints were being used for normal operation and/or maintenance. Based on these discussions and the review of the Drawing Control Instruction, the inspector determined that it was acceptable to maintain controlled copies of Construction Prints, if applicable, in addition to the Final Prints.
(Closed) UNRESOLVED ITEM (80-05-03): Alarm response book was not controlled.
The inspector verified that the licensee has removed the alarm response book from use. The alarm response procedures are maintained as part of appropriate operating procedures.
(Closed)
INFRACTION (80-05-05):
Failure to follow Radiation Protection procedures. The inspector verified that a meeting was held with the department supervisor and the employees involved to stress the importance of adherence to Radiation Protection procedures. Radiation Protection procedures are also addressed in the annual training for all plant workers.
(0 pen)
INFRACTION (79-21-05):
Failure to lock valve. The inspector verified that the Emergency Cooling Drain Valves #EC701, 703, 705 and 707 and the Containment Spray System Isolation Valves Drain Valves were found to be in position required by the appropriate operating procedure valve lineup. However, CRD-8, inlet to the Control Rod Drive Water Filter #12, which was out of service, was found to be open instead of closed as required by OP-9.
CRD-10, outlet of Control Rod Drive Water Filter #12 was found to be closed vice open. This item remains open pending licensee action to correct the valve lineup.
(0 pen)
INFRACTION (79-21-04):
P&I Drawings not as-built. The inspector reviewed selected portions of the drawings and systems listed below to verify that the drawings were corrected to reflect the as-built condition.
Drawing C-18013-C, Reactor Building Heating, Cooling and Ventilation
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Systems, Revision 7, dated April 1980. The inspector verified that the P&ID was changed to show the four inlet isolation valves to the airborne monitor. This portion of the noncompliance is close i Drawing C-18016-C, Control Rod Drive, Revision 6, dated September 1980.
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The inspector verified that the P&ID was changed to show the two valves between CRD 8 and 9 and their respective Drive Water Filter.
This portion of the noncompliance is closed.
Drawing C-18019-C, Reactor Liquid Poison System, Revision 7, dated
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November 1979. The inspector verified that the drawing was correct in that there is no installed flow switch upstream of valve Lp-12.
This portion of the noncompliance is closed.
Drawing C-18007-C, Reactor Core Spray, Revision 15, dated May 1980.
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The inspector verified that the drawing was changed to remove valves CRS-15 and 16, and to correctly show the vent valve on the reactor side of Isolation Valve 40-15 and the four drain valves installed on valve CRS-ll. However, during the review, the inspector noted that the drawing incorrectly shows a vent valve downstream of valve CRS-12 when in fact the vent valve is installed on the upstream side.
Drawing C-18012-C, Reactor Containment Spray, Revision 13, dated
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December 1980. The inspector verified that the drawing was changed to show the three vent valves installed on the Isolation Valves.
However, during the verification, the inspector noted that there were two vent valves installed downstream of each Isolation Valve which were not shown on the drawing.
Drawing C-18017-C, Emergency Cooling System, Revision 14, dated
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September 1981. The inspector verified that the drawing was changed to show the manual block valves and to remove the check valves on the loop 11 and 12 vent lines to Main Steam. The inspector noted, however, that the drawing was still missing one of two drain valves installed on the reactor side of IV's 39-09 and 10.
This item remains open until drawings C-18007-C, C-18012-C and C-18017-C are corrected to show the as-built condition.
(0 pen)
INFRACTION (79-21-07): Valves not labeled as required. The inspector examined the Liquid Poison, Diesel Generator Air Start, and the Reactor Building Ventilation systems and verified that the valves were labeled and that the labeling was correct. When examining the Emergency Cooling and Containment Spray systems, the inspector found that the Emergency Condenser Makeup Tank Supply valves, FS-44 and FS-45, were incorrectly labeled and that eight Containment Spray vent and drain valves, identified in item 79-21-04 above, were not labeled. Reviewing the VCOL for the Emergency Cooling and Containment Spray systems, the inspector noted that FS-44 and FS-45 appeared twice on the VCOL and each time they were associated with a different P&ID valve number. The eight Containment Spray vent and drain valves were not listed on the VCOL.
This item remains open pending correct labeling of FS-44 and 45 and the Containment Spray vent and drain valves and correcting the respective VCOL's.
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(0 pen)
UNRESOLVED (79-21-08): Revise valve lineup procedures. The inspector examined portions of the systems and VCOL's identified in the unresolved item and determined that the following conditions still exist:
Diesel Generators
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The position of DGA-34, Air Supply Pressure Regulatory Valve, was listed as open on the VCOL. Opening this valve could jeopardize the integrity of the air start motors. Pressure should be regulated by the setting of DGA-34.
Core Spray System
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The requirement to set Pressure Control Valves (PCV) 81-85 and 81-86 was not specified in the VCOL.
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Containment Spray System The power supply for valve 80-21 (No. 112 CTN-SP, Suction IV)
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was not included in Table II.
The VCOL requires the Containment Spray (CTN-SP) and Raw Water
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(RW) flow rate control valves to be locked in a throttled position. When inspected, the RW flow rate control valves were throttled and locked, the CTN-SP flow rate control valves were throttled with the handwheels removed (CTN-SP 5,6,17,20).
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Reactor Shutdown Cooling System Shutdown Cooling Pump (SCP) Seal Water Vent Valve for No.13 SCP (SC-757) was not incorporated on the VCOL.
Control Rod Drive (CRD) System
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The vent on the CRD charging water header has a hose attached vice being capped.
Liquid Poison System
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The power supply for heat tracing was not listed on Table II.
This unresolved item remains open pending appropriate licensee action on the items listed above.
3.
Operation Safety Verification a.
Control Room Observations l
Routinely throughout the inspection period, the inspector independently verified plant parameters and equipment availability of engineered safeguard features against a plant specific checklist to ensure the following items were observed:
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Proper control room manning;
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Availability and proper valve line-up of safety systems;
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Availability and proper alignment of onsite and offsite
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emergency power sources; Reactor control panel indications and bypass switches;
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Core thermal limits;
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Primary containment temperature and pressure;
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Drywell to suppression chamber differential pressure;
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Stack monitor recorder traces; and
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Liquid poison tank level and concentration.
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Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, was being taken.
Shift turnovers were observed to ensure proper control room and shift manning on both day and backshifts. Shift turnover checklists and log review by the oncoming and off-going shifts were also observed by the inspector.
The inspector directly observed portions of routine power operations to ensure adherence to approved procedures, b.
Review of Logs and Operating Records The inspector reviewed the following logs and instructions for the period January 1 through February 7, 1982:
Control Room Log Book
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Station Shift Supervisor's Log Book
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Station Shift Supervisor's Instructions,
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Licensee Event Report Log
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Reactor Operating Log
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The logs and instructions were reviewed to:
Obtain information on plant problems and operation;
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Detect changes and trends in performance;
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Detect possible conflicts with technical specifications or
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regulatory requirements;
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Determine that records are being reviewed as required;
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Assess the effectiveness of the communications provided by
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the logs and instructions; 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, physical security plan and approved procedures. The following observations relating to physical security were made:
The security force on both regular and off-shifts were properly
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manned and appeared capable of performing their assigned functions.
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Protected area barriers were intact - gates and doors closed and locked if not attended.
Comunication checks were conducted.
Proper communication devices
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were available.
Isolation zones were free of visual obstructions and objects that
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could aid an intruder in penetrating the protected area and during periods of darkness, the protected area had sufficient illumination.
Persons and packages were checked prior to entry into the protected
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area.
Vehicles were properly authorized, searched and esco,ted or controlled
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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 l
requiring escort were properly escorted.
Compensatory measures were implemented during periods of equipment
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failure.
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No violations were identified.
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Plant Tours
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During the inspection period, the inspector made multiple tours of plant l
areas to make a independent assessment of equipment conditions, radiological i
conditions, safety and adherence to regulatory requirements. The following
areas were among those inspected:
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Turbine Building
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Auxiliary Control Room
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Vital Switchgear Rooms Yard Areas
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Radwaste Area
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Diesel Generator Rooms Screen House
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Reactor Building
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The following items were observed or verified:
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Radiation Protection:
Personnel monitoring was properly conducted.
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Randomly selected radiation protection instruments were
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calibrated and operable.
Radiation Work Permit require'
vere being followed.
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Area surveys were properly cc..sucted and the Radiation Work
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Permits were appropriate for the as-found conditions. The inspector witnessed the performance of radiation survey
- N61329 on the Reactor Building elevation 281 ft. and survey
- N61562 on the Reactor Building elevation 237 ft.
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Fire Protection:
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Randomly selected fire extinguishers were accessible and
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inspected on schedule.
Fire doors were unobstructed and in their proper position.
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Ignition sources and combustible materials were controlled in
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Equipment Controls:
I Jumpers and equipment tag outs did not conflict with Technical
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Specification requirements.
Conditions requiring the use of jumpers received prompt licensee l
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Vital Instrumentation:
Selected instruments appeared functional and demonstrated
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parameters within Technical Specification Limiting Conditions for Operation.
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Radioactive Waste System Controls:
Gaseous releases were monitored and recorded.
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The radioactive waste system was operated in accordance with
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approved procedures. The inspector witnessed portions of the solidification of concentrated waste shipment #02820113A.
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Housekeeping:
Plant housekeeping and cleanliness were in accordance with
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approved licensee programs.
6.
Surveillance Tests During the inspection period, the inspector witnessed the performance of various surveillance tests. Observations were made to verify that:
Surveillance procedures e.onform to technical specification
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requirements and have been properly approved.
Test instrumentation is calibrated.
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Limiting conditions for operations for removing equipment from
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service are met.
Testing is performed by qualified personnel.
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Surveillance schedule is met.
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Test results met technical specification requirements.
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Appropriate corrective action is initiated, if necessary.
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Equipment is properly restored to service following the test.
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The following tests were included in this review:
ST-sal, " Smoke and Heat Detector Test," Revision 2, dated January 9, 1979.
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The inspector witnessed the testing of heat detectors in #102 Emergency Diesel Generator room on January 14, 1982. The inspector noted that the periodic (every 2 years) review of this procedure by the Site Operations Review Committee had not been performed. The licensee is in the process of revising this procedure to include the smoke detectors recently installed as a part of the 10CFR50, Appendix R fire protection
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modifications. This revision will be completed prior to the next j-scheduled performance of this test.
(220/82-01-01)
ST-Q3, "High Pressure Coolant Injection Fump Operability Test,"
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Revision 2, dated March 2, 1981. The inspector witnessed the testing of #11 feedwater pump on January 7, 1982.
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ISP-RPS-TP, " Reactor Protection System Auto Trip System Instrument
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Trip Channel Test / Calibration," Revision 10, dated December 3, 1981.
The inspector witnessed the testing of reactor water level instrument
- 36-05D on January 29, 1982 and drywell pressure instrument #201.2-476C l
on January 30, 1982. While preparing to perform testing on the drywell pressure instrumentation, the technician inserted a low test signal into the reactor pressure instrument by mistake. This caused a
" half-isolation" signal for the main steam isolation valves. The erroneous signal was quickly cleared by the technician. The inspector
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reviewed the procedure used by the technician and determined that it correctly describes the steps for testing. The inspector discussed the event with the appropriate departmental supervisor who later held a meeting with all his technicians to stress the importance of properly
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following :pproved procedures.
l No violations were identified.
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Plant Maintenance The inspector examined portions of various safety related maintenance and modification activities. Through direct observation and review of records, he determincd that:
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i Redundant components were tested to ensure operability prior to
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starting the work.
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These activities did not violate the limiting conditions for operation.
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i Required administrative approvals and tag outs were obtained prior to
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initiating the work.
Approved procedures were used or the activity was within the " skills t
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of the trade".
Appropriate radiological controls were properly implemented.
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Equipment was properly tested prior to returning it to service.
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During this inspection period, the following activities were examined:
Hydrostatic testing of the Post Accident Sampling System;
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Repair of #12 Feedwater Pump; and
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Replacing packing on #13 Feedwater Booster Pump.
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No violations were identified.
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8.
Safety System Operability Verification On a sampling basis, the inspector directly examined selected safety system trains to verify that the systems were properly aligned in the standby mode.
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Verification that each accessible valve in the flow path is in the
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correct position by either visual observation of the valve or remote position indication.
Verification that power supply breakers are aligned for components
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that must actuate upon receipt of an initiation signal.
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Visual inspection of the major components for leakage, proper
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lubrication, cooling water supply, and other general conditions that might prevent fulfillment of their functional requirements.
Verification by observation that instrumentation essential to system
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actuation or performance is operational.
Motor operated or air operated valves are not mechanically blocked
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and power is available as appropriate.
There are no visual leakage paths between the containment and the
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isolation valves.
During this inspection period, the following systems were examined:
fil Core Spray
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fil and #12 Containment Spray
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Containment Atmosphere Dilution
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Containment Atmosphere Dilution (CAD) System Design and Operation Review 9.
a.
Findings The resident inspector reviewed Operating Procedure OP-9," Nitrogen Monitoring Systems for the Primary Containment
Inerting and H2-02 and Pressure Suppression System," Revision 5, dated February 20, 1981 l
to detemine if the CAD and venting systems could be operated without
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l entry into the Reactor Building since a radiation shielding study l
perfomed for the plant in July 1980 shows that the Reactor Building may not be accessible after a loss of coolant accident.
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It appeared that nitrogen addition (i.e., CAD) can be perfomed remotely from the control room. However, OP-9, Section G.II required entry into the Reactor Building to throttle air-operated blocking valve BV 201.2-08. This is necessary to prevent overpressurizing the filters in the Emergency Ventilation System when venting the containment.
On January 28, 1982 the inspector notified the licensee that OP-9 was inadequate in that it required local operator action in the Reactor Building when it may not be accessible.
On January 29, 1982, the licensee issued a prompt notification, LER 82-02. The licensee also revised OP-9 to allow venting of the containment by a series of " puff" type releases by cycling of existing valves from the control room. On February 1, 1982 as a result of further study by the licensee, it was determined that this method of venting was impractical due to the small volume between the valves. A manual valve in the vent line from the containment was then locked in a throttled position to allow controlled venting at a rate of approximately 4 SCFM. The entire operation could now be perfomed from the control room. On February 4, 1982, OP-9 was again revised to describe this method.
Further review by the inspector also revealed that OP-9 required that venting be conducted if containment pressure approaches 35 psig.
10 CFR 50.44(g) limits repressurization to 50 percent of containment design pressure. The NMP-1 drywell design pressure is 62 psig and the torus, which is the limiting factor, is 35 psig. The licensee issued Revision 6 to OP-9 on February 4,1982 to require containment venting when containment pressure approaches 20 psig. This is 49.7 percent of torus test pressure (40.25 psig) and 57.1 percent of design pressure (35psig).
This change satisfied a commitment made by Niagara Mohawk in their response to an NRC request dated March 1, 1974 and the Safety Evaluation Report for the Full Term License dated July 3, 1974 to limit post-accident repressurization to 20 psig.
The inspector further noted on February 2,1982 that the purging system, as described by the existing procedures, failed to provide the redundancy as required by 10 CFR 50.44(g). Venting from within the drywell or torus was possible but the system as described was not single failure proof because the piping from each location were joined together and passed through a single valve. The licensee has committed to install a pressure control valve on the inlet to the Emergency Ventilation system that will allow an alternate venting path. This has been documented in a letter to D. B. Vassallo, dated January 29, 1982.
The licensee stated that this modification will be completed during the Spring 1983 refueling outage.
For the interim, the licensee has issued Revision 7 to OP-9 dated February 5,1982 wnich describes an alternate venting method involving cycling of the drywell nitrogen vent and purge isolation valves.
The drywell nitrogen vent and purge inboard and outboard isolation valves are both 24" butterfly type valves. First, the outboard isolation valve is checked closed. Then the inboard isolation valve is cycled
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3 to allow gases from the drywell to fill the 5 ft volume between the inboard and outboard isolation valves.
Next, the outboard isolation is opened to discharge the fixed volume of gases to the Emergency Ventilation System. This prevents overpressurizing the filters in the Emergency Ventilation System. The steps are repeated as necessary until the containment pressure is reduced to less than 20 psig. The operating time for each of the isolation valves is less than 30 seconds.
b.
Background The original design of the station, as described in the Final Safety Analysis Report, dated June 1967, included a containment inerting system. This system was to maintain a nitrogen inerted atmosphere in the containment to prevent the combustion of any hydrogen generated by a zirc-water reaction following a loss of coolant accident. The initial nitrogen inerting after a reactor start-up was supplied by a tank trailer. Any make-up needed during power operations was supplied by high pressure gas storage bottles.
In the Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License dated July 1972, the licensee committed to the installation of a containment atmosphere dilution (CAD) system by 1974. The system was designed to meet the
" Regulatory Position" of the Atomic Energy Commission Safety Guide No. 7. " Control of Combustible Gas Concentration in Containment Following a Loss of Coolant Accident," dated March 10, 1971.
(Now known as Regulatory Guide (R.G.) 1.7 which was finally issued in 1976).
The CAD system would be used to add nitrogen to the containment to dilute the oxygen concentration to below the oxygen flammability limit.
A containment venting system discharging thru the emergency ventilation system would also be provided a back-up in the event of a failure of the CAD system. Safety Guide No. 7 did not require redundant containment venting capability.
In Amendment No. 2 to Application to Convert Provisional Operating License to Full-Term Operating License dated March 14, 1974, the licensee submitted a drawing showing the redundant CAD system and the venting system.
The licensee also responded to an AEC request dated March 1, 1974 for the following additional information related to the CAD system:
Ic, " Provide confirmation that your emergency operating procedures include limiting the post-accident repressurization of the containment to less than 20 psig."
The licensee's response stated:
"The following will be added to Nine Mile Point Special Procedure No.1,
' Loss of Coolant and 115 KV Power' when the CAD system becomes operational. When placing the CAD system in operation the maximum pressurization level of the primary containment will be limited to 20 psig.
It would take 38 days to reach this pressure with zero
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containment leakage. At this time venting should be initiated at a rate of 4.4 SCFM.
However, since containment leakage is not zero but around 0.5 percent per day, the 20 psig pressure will probably never be reached."
The Safety Evaluation Report for the Full-Term License dated July 1974 concluded that the CAD system as described would meet the intent of General Design Criteria 41 and the guidelines of Draft Regulatory Guide 1.7 and was considered acceptable.
The CAD and venting systems became fully operational in November 1977 10 CFR 50.44 was issued for coment as a proposed rule on October 21, 1976.
It became effective as a final rule on November 27, 1978.
It defines a purging system as a system for the controlled release of the containment atmosphere to the environment through filters, if needed.
A repressurization system is defined as a system used to dilute the concentration of combustible gas by adding an inert gas or air to the containment. A purging system is nonnally part of the repressurization system.
10 CFR 50.44 requires that a combustible gas control system (a repressurization system is acceptable) which meets the requirements of Criteria 41, 42 and 43 of Appendix A of 10 CFR 50 be provided.
If a purge system is used as part of the repressurization system, it shall also conform to the above criteria.
Criterion 41 requires that each system have suitable redundancy to assure that its safety function can be accomplished, assuming a single failure.
c.
Conclusions The licensee and NRC Region I independently conducted a record search to determine whether there was any further correspondence in the docket relating to the new 10 CFR 50.44 requirements subsequent to issuance of the full-term license. None was identified. The licensee stated that either the new requirements had been inadvertently overlooked or they had determined at the time that the system described in their application for full-tenn licensee was acceptable.
In conclusion, even though the system as constructed provided the means
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for a single containment purge path, and the licensee was subsequently i
able to procedurally provide an additional interim alternate purge path, the procedures, as issued and approved at the time of the inspectors review, did not provide the capability for combustible gas control that is required by 10 CFR 50.44(g) in that:
l (1) To perform the purging operation the procedure (0P-9) required entry into the reactor building which, in accordance with a 1980 shielding study, may not be accessible after a loss of coolant l
accident.
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(2) The procedure (OP-9) permitted repressurization to 35 psig, a value greater than the 20 psig comitted to in the licensee's application for full-tem license and greater than the 50 percent of design pressure limit imposed by 10 CFR 50.44(g).
(3) A single failure proof redundant purge path was not procedurally available.
10 CFR 50, Appendix B states in part, "III. Measures shall be established to assure that applicable regulatory requirements... for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures and instructions." The procedural inadequacies described above and the failure to provide, procedurally and design wise, a redundant purge path in ac"ordance with the requirements of 10 CFR 50.44(g) is considered a viola.fon.
10.
Modification Activities The inspector reviewed the documentation for the installation of the drywell to torus differential pressure instrumentation, modification no. N1-77.34, to ensure that modifications meet the requirements of Technical Specifications, 10 CFR 50.59, and 10 CFR 50, Appendix B.
The inspector reviewed the ifcensee's Safety Evaluation dated April 26, 1977 to determine that the licensee had considered the effects of this modification on plant safety. The Safety Evaluation also indicated that the modification had been reviewed by the Site Operations Review Comittee on April 29, 1977. The Safety Evaluation stated that the valves and piping which will be used for this mcdification will be manufactured in accordance with the~ appropriate requirements of 10 CFR 50.
The instrumentation to be installed was not manufactured in accordance with 10 CFR 50, Appendix B due to the short lead time before the Spring 1977 outage. However, during the Fall 1978 refueling outage, this equipment was to have been replaced with equioment which meets the quality requirements of 10 CFR 50, Appendix B.
The licensee is in the process of determining if the qualified instrumentation has been installed. This item is unresolved pending NRC review of the licensee's findings.
(220/82-01-03)
The inspector also verified that hydrostatic tests had been perfomed on April 29 and May 2,1977 and the instruments were calibrated on June 18, 1977.
Instrument Surveillance Procedure ISP-201.2DP, "Drywell and Torus Pressure /Drywell Level," Revision 1, dated October 30, 1980 has been revised to ensure that a periodic calibration of the instruments is performed.
The inspector noted that P&ID #C-18014-C had not been correctly revised since it does not show the instrument sensing line's isolation valves from either the drywell or torus.
ISP-IC-21, " Pre-Startup Valve Lineup Check,"
Revision 4, dated June 29, 1981 and ISP-IC-23, " Integrated Leak Rate Test of the Primary Containment," Revision 9, dated June 8,1981 have not been revised to include the same set of isolation valves. The drywell to torus differential pressure instrument, which indicates in the control room, was observed to be in servic.
10 CFR 50, Appendix B, states in part, "VI. Document Control - Measures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes which prescribe all activities affecting quality. These measures shall assure that documents, including changes, are reviewed for adequacy..." These measures were not effective in that P&ID #C-18014-C was not accurately revised and ISP-IC-21 and 23 were not revised to reflect the installation of instrumentation important to safety. This is a violation.
(220/82-01-04)
The inspector reviewed Quality Control-0perations Surveillance Report
- SR 81-023 dated September 25, 1981 and Nonconfomance Report #81-35 dated October 2, 1981. There currently appears to be active involvement of the licensee's Quality Control Department in plant modifications. However, there appears to have been some weaknesses in the proper "close-out" of modifications during the period from 1974 to 1979. This "close-out" should include verification that the following activities have been perfomed, as necessary; Safety Evaluation as required by 10 CFR 50.59;
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Site Operations Review Committee and Safety Review and Audit Board
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Review'
Installation in accordance with approved plans and procedures;
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Pre-operational testing;
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Revision of drawings;
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Revision of operating, surveillance, and maintenance procedures;
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Training in use of new equipment or systems.
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The licensee acknowledged the inspector's concerns and has begun a review
of safety-related modifications proposed from 1974 to 1979. The resident inspector will monitor the licensee's activities in this area.
(220/82-01-05)
j 11.
Licensee Action on IE Circulars For the IE Circulars listed below, the inspector verified that the Circulars were received by licensee management, that a review for applicability was performed and that appropriate corrective actions were taken. The following l
IE Circulars are closed:
IEC 80-11, Emergency Diesel Generator Lube Oil Cooler Failures.
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IEC 81-01, Design Problems Involving Indicating Pushbutton Switches
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Manufactured by Honeywell Incorporated.
IEC 81-03, Inoperable Seismic Monitoring Instrumentation.
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IEC 81-06, Potential Deficiency Affecting Certain Foxboro 10 to 50
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Milliampere Transmitters.
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12.
Unresolved Item An item about which more information is required to determine its acceptability is considered unresolved. Paragraph 10 contains an unresolved item.
13.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and findings.
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