IR 05000029/1981013
| ML20010H506 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 08/27/1981 |
| From: | Foley T, Gallo R, Raymond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20010H496 | List: |
| References | |
| 50-029-81-13, 50-29-81-13, NUDOCS 8109240550 | |
| Download: ML20010H506 (17) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT 50-29/81-7-6 50-29/81-7-8 Region I 50-29/81-7-11 50-29/81-5-30 Report No. 50-29/81-13 50-29/81-6-3 50-29/81-6-11 Docket No. 50-29 50-29/81-6-2 License No.
DPR - 3 Priority Category C
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Licensee: Yankee Atomic Electric Company 50-29/81-6-16 50-29/81-6/27 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station (Yankee Rowe)
Inspection at:
Rowe, Massachusetts Inspection Conducted: July 1 - August 17, 1981 l/ f6h4
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Inspectors:
T. Foley, Seniv Mtesident Inspector dats signed Qhaan A llamam2)
ekAsi W. Raymond, Sentot Reside $lt Inspector dat~e signed (VermontYankee)
date signe'd Approved by:
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h 37 R. Gallo, Chief, RFS No. lA date signed Div. of Resident and Project Insp.
Inspection L 1ary:
Inspection on July 1 - August 17, 1981 (Report No. 50-29/81-13)
Areas Inspected: Routine onsite inspection by the resident inspector of plant operations including tours of the facility; log and record reviews; operational safety verification; review of operator training and review of physics acceptance testing prior to reactor startup; review of facility modifica;. tons; review of Licensee Event Reports; followup of operational events; surveillance testing and observation of physical security. The inspection involved 96 inspector hours by the resident NRC inspectors.
Results: No items of noncompliance were identified in this inspection.
8109240550 810904 PDR ADOCK 05000029 G
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DETAILS 1.
Persons Contacted H. Autio, Plant Superintendent W. Billings, Chemistry Supervisor E. Chatfield, Training Supervisor T. Danek, Senior Advisor - Operations L. French, Technical Assistant T. Henderson, Plant Reactor Engineer L. Laffond, Technical Assistant W. McGee, Manager of Public Infomation F. McWilliams, Engineering Assistant R. Mitchell, Technical Assistant L. Reed, Operational Quality Assurance Coordinator N. St. Laurent, Assistant Plant Superintendent R. Sedgwick, Security Supervisor J. Staub, Technical Assistant to Plant Superintendent J. Trejo, Health Physicist Supervisor D. Vassar, Operations Supervisor The inspector also interviewed other licensee employees during the inspection.
including members of the Operations, Health Physics, Instrument and Cont d Maintenance, Reactor Engineering, Security and General Office Staffs.
2.
Licensee Action on Previous Inspection Items a.
(closed) Unresolved Item (29/80-13-04) The licensee has revised AP-7008 Pumps and Valves Program to int.crporate the specific re-quirements of IWP 3110, 3210 and 3230 of section XI of the ASME Boiler and Pressure Vessel Code. This item is closed.
3.
Shift Logs and Operating Records a.
The inspector reviewed the following plant procedures to determine the licensee established administrative requirements in this area in preparation for review of various logs and records.
AP-0001, Plant P ocedures and Instructions, Revision 8
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AP-2002, Operations Department Personnel Shift Relief, Revision
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AP-2009, Control Room Area Limits for Control Room Operators
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AP-2010, Control Room Access During Accidents and Operations
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Transients, Original AP-0017, Switching and Tagging of Plant Equipment, Revision 6
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AP-0018, Bypass of Safety Function and Jumper Control Log,
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Revision 8 AP-2007, Maintenance of Operations Department Logs, Revision 7
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AP-0216, Housekeeping and Cleanliness Control, Revision 2
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-- AP-0042, Housekeeping for Maintenance and Modifications, Revision
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Rules Governing In-Plant Tagging Procedures Local Control Rules,
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Revision 3 The above procedures, Technical Specifications, ANSI N18.7-1972
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" Quality Assurance Requirements for Nuclear Power Plants" and 10 CFR 50.59 were used by the inspector to determine the acceptability of the logs and records reviewed.
b.
Shift logs and operating records were reviewed to verify that:
Control Room logs and shift surveillance sheets are properly
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completed and that selected Technical Specification limits were met.
Control Room log entries involving abnormal conditions provide
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t sufficient detail-to canmunicate equipment status, lockout
status, correction, and restoration.
Log book reviews are being conducted by the staff.
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Operating and Special Orders do not conflict with Technical
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Specification requirements.
-- Jumper (Bypass) log does not contain bypassing discrepancies with Technical Specification requirements and that jumpers are properly approved and installed.
c.
The inspector reviewed, on a sampling basis, the following logs and records for the period July 1 to August 17, 1981.
-- Switching and Tagging Log Maintenance Request Log
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Refueling Coordinators Log
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Bypass of Safety Function and Jumper Control Log
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Operating Log
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Rowe Station Log Sheets
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-- Primary A.0. Log Sheets Shift Relief Checklist
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Primary Chemistry Log
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Secondary Chemistry Log
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Radioactive Gaseous Release Permits
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Radioactive Liquid Release Permits
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No inadequacies were identified.
4.
Plant Tour
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The inspector conducted a tour of accessible areas of the plant including the Primary Auxiliary Building, Turbine Building, Safety Injection Buildirg, Vepor Container, Switchgear Room, Diesel ht, oms, Control Room, Spent Fuel Building, and HP control Point Areas. Details and findings are noted below:
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a.
Monitoring Instrumentation and Annunciators On several occasions during the inspection, control board annunciators were checked for abnormal alarms. The following monitoring instrumenta-tion was checked to verify operability and where applicable, values indicated were verified to be in accordance with Technical Specification requirements:
Pressurizer pressure, level and temperature
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Charging flow path (
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l RCS temperature
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-- SI tank level PWST and DWST levels
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-- Vapor Container Drain Tank level l
j Primary Vent Stack radiation monitors
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Containment air particulate radiation monitor
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No abnomal annunciators were energized. No inadequacies were identified.
b.
Radiological Control:
Radiation controls established by the licensee, including posting of radiation areas, radiological surveys, condition of step-off pads, and disposal of protective clothing were observed for conformance with the requirements of 10 CFR 20 and OP-8100, " Establishing and Posting Controlled Areas," and OP-8101, " Plant Radiological Surveys."
No in>.dequacies were identified.
c.
Plant Housekeeping Plant housekeeping conditions, including general cleanliness and storage of materials to prevent fire hazards were observed in all areas toured.
Housekeeping and cleanliness were acceptable.
No inadequacies were identified.
d.
Fluid Leaks and Piping Vibrations Systems and equipment in all areas toured were observed for the existence of fluid leaks and abnormal piping vibration.
No inadequacies were identified except as noted below.
During the perfomance of OP-4211 Emercency Boiler Feed Pump Surveillance, the inspector noted that the turbine seal water supply piping was not supported by pipe hangers or restraints, and during the operation of the Emergency Boiler Feed Pump, the seal water supply piping i
to the Turbine vibrated excessively. This discrepancy was brought to the attention of the Operations Supervisor who initiated Maintenance Request MR-81-769 to correct the discrepancy. The inspector had no i
further questions in regard to the above.
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e.
Pipe Hangers / Seismic Restrair g Pipe hangers and restraints installed on previous piping systems
through the plant were obi.rved for proper installation and tension.
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No inadequacies were identified.
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f.
Control Room Manning / Shift Turnover Control Room manning was reviewed for conformance with the requirements l
of 10 CFR 50.54 (k) and Technical Specifications.
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The inspector verified, several times during the inspection, that appropriate licensed operators were on shift. Manning requirements were met at all times. Several shift turnovers were observed during the course of the inspection. All were noted to be thorough and orderly.
No inadequacies were identified.
5.
Surveillance Testing The inspector observed portions of the following surveillance tests to verify that testing was performed in accordance with technically adequate procadures, that results were ir, conformance with Technical Specifications and procedure requirements, that the results were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by management personnle. The following surveillances were reviewed by the inspector:
-- OP-4211 Emergency Boiler Feed Pump Surveillance
-- OP-4215, Surveillance of Boron Injection Flow Path OP-4205, Safety Injection Operational Check
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OP-4217, Test of No. 2 Charging Pump for Surveillance Operability
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Check OP-4209, Emergency Diesel Generator Test During Refueling Intervals
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OP-4207, Surveillance of the Station Power System and Emergency
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Diesel Generators
-- OP-4611, Nuclear Instrumentation and Reactor Protection System Precritical check OP-4200, Main Coolant System Leak Inspection or 2200 PSIG Pressure
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Test a.
During the performance of OP-4209, Emergency Diesel Generator Test During Refueling Intervals, the inspector noted that the No. 1 diesel overheated during the full load test portion of the test procedure and this portion of the test procedure was declared unsatisfactory.
The licensee subsequently issued Maints: nance Request MR 81-729 which cleaned and flushed the cooling system. The inspector witnessed postions of this process. The licensee subsequently reperformed the
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load test portion of the test satisfactorily. The inspector reviewed the results of this test and also reviewed OP-4207 Station Power and No.1 Diesel Generator Surveillar:ce. OP-4207 is performed on a monthly basis in order to verify the operability of the diesel. generator in accordance with technical specification requiremen+s. The failure of the diesel generator to withstand the full load capacity was reported by the-licensee as Licensee Event Report (LER) 81-16.
b.
The inspector reviewdd OP-2000.99 Emergency Feedwater Pumps P-79-1 and P-79-2 Acceptance Testing and Attachments; and witnessed portions of the 48 Hour Endurance Test on each pump. The inspector noted that during the starting seqJence with a simultaneous loss of offsite power, the Emergency Feedwater Pumps would start with power supplied from tha Emergency Diesel Gener; tors. During the testing, the pumps tripped off on over current while staring each pump when electrically supplied from the Emergency Diesel Generator and while lined up to directly feed the Steam Generators. The inspector cr. viewed emergency. procedure OP-3251 Loss of AC Supply and determined that during loss of AC power the operators are required to shut the pump discharge valve prior to starting the pumps. Discussion with reactor operators confirmed that the ope,ators would have approximicely thirty minutes to shut the discharge valves in the event of a loss of AC and simultaneous need to operate the Emergency-Feedwater Pumps.
The manual initiation of the Emergency Feedwater System is contrary to NUREG 0660; however, the NRC has issued an order to Yankee Atomic Elec-tric Company dated July 10, 1981, which confirms Yankee's commitments to implement TMI related requirements and acknowledges certain proposed exceptions to these TMI requirements which are under further review by the liRC. The order states that the licensee has proposed that automatic initiation of the Emergency Feedwater System is not necessary for Yankee
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Rowe, and that the NRC staff has this matter under re'/iew. The inspector had no further questions at this time.
6.
Operational Safety Verification A detailed review of the Chemical Shutdown System was conducted to verify that the system was properly aligned and fully operational in the standby mode.
This review was performed utilizing system Flow Diagram 9699-FM-83A, OP-4203 Monthly Valve Check, and OP-2652 Preparation of the Emerg :y Core Cooling System for Normal Operation.
Review of the above system include the following:
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verification that each accessible valve in the flow path was in the correct position by either visual observation of the valve or remote position indication.
verification that power supplies and breakers were properly aligned for
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components that are required to actuate upon receipt of a safety injection signal.
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visual inspection of major components for leakage, proper
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lubrication, cooling water supply, general condition and other factors that might prevent fulfillment of their functional requirements.
verification by observation that the instrumentation essential
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for system actuation and performance was operatior.al.
The inspector identified no inadequacies.
7.
Onsite Licensee Event Followup For those LER's selected for the onsite followup, the inspector verified that details of the events were clearly reported including the accuracy of the description of cause and adequacy of corrective action. The inspector also determined whether further information was required from the licensee, whether~ generic implications were identified, and that reporting requirements for Technical Specifications and Regulatory Guide 1.16 had been met, that appropriate corrective action had been taken, that the event was reviewed by the licensee as required, and that continued operation of the facility was conducted within Technical Specification limits. The review included discussions with licensee p(ersonnel, review of PORC meeting minutes, Plant Information Reports in-house reports), and applicable logs. The following LER's were reviewed onsite.
81-09, Primary Vent Stack Noble Gat, Monitor Failure
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81-10, Primary Vent Stack Noble Gas Monitcr Failure
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81-11, Steam Generator No. 2 Tube Degradation
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81-12, Loss of Power to Two Environmental Air Samplers
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81-13, Blister llE Penetrations Failed Type "B" Test
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-- 81-14, Steam Generator Trip Set Point Low 81-15, Pressurizer Code Safety Valve Setpoint Out of Specification
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-- 81-16, No.1 Diesel Generator Overheatir;g 81-17, Inadvertant Dilution of BAMT
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The inspector witnessed portions of the Steam Generator Eddy Current Testing; testing and repair of the Diesel Generator, and reviewed resuits of the corrective actions taken for the above listed LER's.
The inspector identified no inadequacies.
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8.
Inspector Followup of Events The inspector responded to events that occurred during the inspection
to verify continued safe operation of the reactor in accordance with the Technical Specifications and Regulatory requirements. The following items, as applicable, were considered during the inspectoe's review of operational events:
-- observations of plant parameters and systems important to safety to confirm operation within normal operational limits; description of c.;at including cause, systems involved, safety
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significance, facility status and status of engineered safety features equipment; details relating to personnel injury, release of radioactive
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material and exposure to radioactive material;
-- verification of correct operation of automatic equipment; verification of proper manual actions by plant personnel;
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verific3 M, of adherence to plant procedures; t
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verification of conformance to Technical Specification LC0
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requirements; determination that root causal factors were identified and that
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corrective actions, taken or planned, were appropriate to correct the cause; verification that ccrrective action taken was appropriate to
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prevent recurrence; determination whether the event involved operation of the facility
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in a manner which constituted an unreviewed safety question as
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l define in 10 CFR 50.59 (a) (2), or in such a manner as to represent l
an unus al hazard to health and safety of the public and environment; determination whether the event involved continued operation of the
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facility in violation of Regulatory requirements or license conditions; l
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evaluation of whether applicable reporting requirements were met.
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The operational events reviewed during this inspection period are discussed j
below.
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a.
During the current refueling outage the licensee removed several used control rods from the core of the reactor and cut the control rods in half while in the reactor vessel cavity in order to facilitate the transfer of the used control rods to the spent fuel pit for storage. The licensee conducted cleaning operations after the refueling of the core and prior to dewatering the reactor vessel cavity. On July 11, 1981 after completion of dewatering the reactor vessel cavity, several auxiliary operators and a Health Physics (HP)
technician entered the reactor vessel cavity to dewater the reactor vessel stud holes.
During the surveying of the area, the contractor HP identified a snll piece of highly radioactive material. The technicim.'
survey meter went off-scale at 50R/hr. The technician picked up the object with a ball of putty and deposited it in a bucket.
Subsequently, it was determined that the material was a chip from a control rod blade resulting from the cutting of control rods in the reactor vessel cavity as previously stated.
Initial exposure estimates indicated that the technician received 1.1 REM (by dosimetry) to the whole body and 8 REM (by calculation) to the extremities.
Further investigation, re-enactment of the event and use of state of the art survey techniques, determined that the dose rate to the live tissue on contact with the control rod chip (the source) was 2.57 RAD /sec. The calculation performed by the licenses indicate that the contractor received 2.03 RAD's total dose to the extremities and 773 millirem to the whole body during that shift.
The inspector discussed this event with those persons knowledgable of the circumsuances surrounding the event and determined that no federal limits or regulation was exceeded during this event.
The licensee transmitted information to NRC Region I Health Physics Specialist as requested. The inspector had no further concerns in regard to the above matter.
b.
On May 2,1981 during a controlled shutdown, control rod number 17 became inoperable. This subject was addressed in inspection report 50-29/81-06 and reported as Licensee Event Report LER 50-29/81-05.
The licensee's investigation in regard to this event determined that the control rod follower was bent five tenths of a, inch. The licensee's corrective action to prevent re-occurence was to remove the additional control rods of the same vintage and same naterial (hafnium), and replace the rods with spare silver-indium-cadmium
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control rods of a later vintage. Subsequently the licenste visually inspected six additional silver-indium-cadmium control rods as part of the routine refueling inspection program in accordance with OP-1700 Cycle XV Reactor R5 fueling and Component Inspection.
During this inspection the licersee identified additional control rods whose followers were bent as much as three tenths of an inch.
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The licensee replaced all control rods that had apparent bends equal to or greater than three tenths of an inch with new silver-indium-cadmium control rods. This information was reported in inspection report 50-29/81-11. Subsequently, tne licensee performed Centrol Rod Drag. Force Testing in accordance with OP-1700 Cycle XV Reactor Refueling and Component Inspection. The results of this test correlated well with the control rod visual inspection results, however the drag force testing also indicated that as many as three additional control rods may have bends or may be mechanically binding.
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This topic was discussed between the licensee, the NRC Core Performance Branch Personnel, the Resident Inspector and the NRC Licensing Project
Manager..The licensee was made aware that the NRC has concerns in this
area and consideration should be given to replacino additional control i
rods.
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The licensee subsequently plotted the drag force test results against the visual centrol rod inspection (Bowing) results and replaced all control
t rods with any discernable bow or any drag force significantly greater than normal.
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j time testing in accordance with OP-4703 Control Rod Drop Time Measurement, performed prior to reactor start up. No unacceptable conditions were noted.
The inspector subsequently held discussions with the Manager of Operations who stated that the licensee would take measures to incorporate a physical measurement for control rod inspection purposes.in order to alleviate the rather subjective techniques currently in use. The inspector had no further concerns in this area.
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9.
Facility Modifications The inspector reviewed the modifications discussed belott to ascertain that the changes were made in accordance with; 10 CFR 50.59, the
established Quality Assurance program and reviewed in accordance with Technical Specification. The inspector verified that the design
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change was conducted in accordance with approved written procedures
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identifying specifications and codes governing the work; that acceptance l
test procedures which define acceptance values are incorporated, and that the as-built drawings were being changed to reflect the modifications.
a.
Non-Return Valve Modification This change was made to improve the plant's eb411ty to handle a main steam line break on the turbine side of Cie non-return valves l
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(MRV's). The provision for automatic quick closure of the NRV's will ensure that 2 or 3 steam generators will be available for heat
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removal and will minimize the effect of transients imposed on the
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main coolant system in the event of a down stream Main Steam line-break. This was accomplished by implementation of a quick closing (3-5 seconds) stored energy actuator on each of the four main steam NRV's.
The change was performed in accordance with Engineering Design Change Request (EDCR) 79-17. The inspector witnessed portions of the installation procedure and portions of the final acceptance test procedure and reviewed the results of OP-4654, Calibration of the Non-Return Valve Automatic Control System. The inspector identified no inadequacies in regard to the above modification.
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b.
Upgrade of the Emergency Feedwater System (EFW)
This design change increases the capability and the redundancy of the emergency feedwater system, resulting in an increased system reliability. Two new motor driven pumps have been added along with the necessary piping to connect these pumps to the steam driven feed pump discharge header in the normal feedwater piping,
and to the alternate emergency feedwater header in the steam generator blowdown piping. The revised system now consists of three pumps, two motor drived and one steam turbine driven, capAle of feeding the steam generators through two separate feed paths.
AdditionalJy, the capability now exists to manually initiate emergency feed from the control room.
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The inspector witnessed portions of the installation of the
emergency feed pumps. and piping, welding techniques and installation
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of seismic supports. The inspector witnessed portions of the.
testing performed on each pump per OP-2000.99 Emergency Feedwater Pumps P-79-1 and 2, Acceptance Testing and Attachments A thru D.
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The inspector noted that the Emergency Feedwater pumps would trip off
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on overcurrent if lined up to supply water to the steam generators and l
if the Emergency Feedwater Pumps were being electrically supplied l
from the Emergency Diesel Generator. This is further discussed in paragraph 5.b of this report.
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The inspector also examined each of the piping cross connections i
j to other systems and noted a potential problem with the Emergency Feedwater to Steam Generator Blowdown Cross Connection. The
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inspector noted that during normal operation a continuous steam
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generator blowdown is in progress which elevates the steam generator l
blowdown piping to approximately three to four hundred degrees and I
that the emeraency feedwater supply is at times as low as 40 degrees
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C (minimum spec). The inspector questioned the licensee whether the difference in water temperatures when the system is initiated using
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the blowdown flow path had been considered in the safety evaluation for the modification. The licensee referred the inspector to the cognizant engineer to answer the question. The inspector reviewed the Engineering Design Change Request EDCR 79-15 safety evaluation which states in part "These proposed modifications do not increase the possibility of a new type of accident, or reduce the margin of safety as defined in the basis of any Technical Specifications. What it does is improve the emergency feedwater system response to any of these accidents".
The inspector did not note that the safety evaluation addressed thermal stresses or fractures, resulting from the variation in temperature in the blowdown cross connect piping area. This topic is unresolved pending further discussions with the cognizant engineer (50-29/81-13-01).
10.
Preparation for Startup a.
Inspector Witness of Activities The inspector witnessed licensee activities in progress on July 24, 1981 and July 27, 1981 in preparation for reactor startup and physics testing. Licensee activities directly observed by the inspector included:
(1)
Testing and calibration of the reactivity computer on July 24, 1981 in accordance with DP-7106 and OPF-7106.1. The inspector verified that Core XV six-group constants for beta fraction, lambda fraction and beta-effective were entered in the reacti-vity computer and that calculated reactivity values were proper.
(2)
Establishment and completion of prerequisites for reactor criticali ty. Activities in progress and work items completed were reviewed for conformance with Technical Specification re-quirements and the following plant-procedures as applicable:
OP-4611, Nuclear. Instrument and Reactors Protection System
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Precritical Check OP-2103, Reactor Startup and Shutdown
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OP-1701, Core XV Zero Power Physics Tests
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OP-2100, Plant Startup from Cold Shutdown and, OP-2101, Plant Startup from Hot Standby
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The requirements of the fc11owing Technical Specifications were verified to be satisfied on July 27, 1981 by direct inspector observation and/or review of completed procedure check Hsts:
TS Section Title / Requirement 6.1 Shift Staffing 3.1.1.1 Shutdown margin l
3.1.1.3 Main Coolant Flow 3.1.2.6
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Operable Charging Pumps 3.1.2.11 Operable Boration Paths 3.1.1.1.2 Operable Main Coolant Loops 3.4.4 Operable Pressurizer 3.4.5 Leakage Detection Systems 3.4.6 Main Coolant Chemistry 3.4.8 Main Coolant Press / Temp Limits 3.5 Accumulator Operability 3.5.2 ECCS System Operability No inadequacies were identified.
b.
Reload Physics Test Procedure Review Plant prcredures for Core XV Zero Power and Power Ascension Physics Testing were reviewed to verify the procedures were prepared and approved in accordance with Technical Specification and licensee administrative requirements. The review was also conducted to verify the procedures were technically adequate to accomplish the proposed measurements and that procedure content / format included the following:
Statement of objective / purpose;
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-- Identification of required prerequisits; l
-- Identification of suitable precautions; i
Identification of required test equipment;
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Provision of detailed step-by-step
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instructions sufficient to accomplish desired measurement; Identification of data required and data reduction techniques;
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Establishment of suitable acceptance criteria; and
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Requirements for data /results review prior to continued testing / operations
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The following references were used during the review:
(1) OP-1702, Core XV Zero to Full Power Physics Test Procedure (2) OP-7306, Moderator Temperature Coeficient (3) OP-1701, Core XV B.O.L. Zero Power Physicr, Test (4) K worandum NED 81-481, Yankee Rowe Core XV Startup Physics Data, dated 7/20/81 (5) Memorandum NED 81-499, Yankee Rowe XV Fuel Reconstitution (6) OP-7106, Calibration of the Reactivity Computer l
Inspector comments / questions in regard to the above procedures were discussed with the Reactor Engineer during a meeting on July 24, 1981.
Subjects discussed included: general methodologies for test measurements, basis for test acceptance criteria, reactivity computer calibration and operation, definition of conditions to be maintained during testing, and definition and maintenance of core power limits for zero power measurements. All inspector coments were satisfactorly resolved and none required that procedures be revised for proper conduct of testing.
Except as noted below, the inspector had no further comment in this area.
The inspector reviewed the licensee's analysis of the impact on four
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physics parameters expected to result from the fuel changes made during l
the recent refueling outage. This analysis was presented in part by reference (5) above and considered the effects on core total reactivity; l
relative radial core power distribution and radial tilt; peak pin power and peak channel power; and, control rod worth and shutdown margin. No inadequacies were identified in the licensee's analysis.
The licensee also stated that a safety evaluation of the fuel changes had been conducted in accordance with the requirements of 10 CFR 50.59, with the analysis discussed above used as a basis. The safety evaluation determined that the Core XV Performance Analysis as described by YAEC-1240 is bounding and that no unreviewed safet; question existed.
Further, no adverse impact on the applicability of the reference cycle safety analysis corresponding to each accident condition evaluated for core-reload licensing submittals was created.
The inspector had no further comment on this item for the present.
However, the inspector requested and received a copy of the licensee's Safety Evaluation, which was provided in Memorandum NED 81-515 dated 7/29/81.
Subsequent, post-startup NRC Staff review of the Safety Evaluation is considered unresolved.
(50-29/81-13-02)
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11. Operator Pre-Strt-Up Training The inspector attended the pre-start-up training lectures on July 3,1981 and interviewed several reactor operators and Shift Supervisors during the inspection period to determine the following:
that changes / modifications made during the current outage were
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incorporated into the training program that the truning program incorporated a schedule for conducting
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the required lectures that prepared lesson plans or other documents adequately describe
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the scope and depth of the lectures
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that operators were made aware of the operational characteristics
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of the modifications and, that operators were made aware of the applicable emergency procedures
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if required for each modification The inspector noted that at the time of the lecture, emergency procedures were not disseminated to the operators and that some confusion existed in regard to the sequence of actions to be taken in the event that the emergency ft iwater pumps were required. The inspector discussed this with the Pl.t Assistant Superintendent who took the necessary steps to insure t at emergency procedures were in place, and operators-were knowledgable of them prior to startup. The inspector verified this by interviewing several operators during pre-startup testing.
The inspector had no further concerns in regard to the above.
f 12. Observations of Physical Security The inspector made observations, witnessed arid /or verified during the
,
regular and offshift hours that selected aspects of plant physical security were in accordance with regulatory requirements, the physical security plan and approved procedures.
,
i a.
Physical Protection Security Organization inspector observations indicated that a full time member of
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the security organization with authority to direct physical security action was present as required.
!
l manning of all shifts on various days was observed to be as
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l required.
l l
b.
Physical Barriers I
selected barriers in the protected area and vital area were
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observed and random monitoring of isolated zones was performed.
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.
1T
'w^WT1MikF--
..
..
.
,
c.
Access Control Observations of the following items were made; identification, authorization and badging;
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access control searches, including the use of compensatory
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measures during periods when equipment was inoperable; and, escorting
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The inspector identified no inadequacies in this area.
13.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items or items of noncompliance. Unresolved items disclosed during this inspection are described in paragraph 9.b and 10.b.
14.
Management Meetings At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope end preliminary findings of the resident inspector.
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