IR 05000029/1980016
| ML19341D297 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/30/1980 |
| From: | Foley T, Martin T, Raymond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19341D295 | List: |
| References | |
| TASK-1.A.1.1, TASK-1.A.1.2, TASK-1.A.1.3, TASK-1.C.1, TASK-1.C.2, TASK-1.C.4, TASK-2.B.3, TASK-2.D.3, TASK-2.E.1.2, TASK-2.E.3.1, TASK-2.E.4.2, TASK-2.F.2, TASK-2.G.1, TASK-3.A.1.2, TASK-3.D.1.2, TASK-3.D.3.3, TASK-TM 50-029-80-16, NUDOCS 8103050298 | |
| Download: ML19341D297 (22) | |
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DCS 50029-800'916
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
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Region I Report No. _y0-29/80-16 i
Docket No.
50-29 License No. DPR-3 Priority
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Category C
Licensee:
Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name:
Yankee Nuclear Power Station (Yankee Rowe)
Inspectior at:
Rowe, Massachusetts Inspection conducted:
September 16-October 31, 1980 Inspectors:
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/2-30 -#O T. Foley, Reactor Inspec' tor date signed h
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W. J. Raymond, Senior Resident Inspector date signed
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date signed Approved by:
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T. T. Martin, Chief, Reactor Projects date signed Section No. 3. RO&NS Branch Inspection Summary:
Inspection on September 16 - October 31, 1980 (Report No. 50-29/80-16)
Areas Inspected:
Routine, unannounced inspection by the resident inspectors of Plant Operations, including review of logs, records and tours of the facility; fo'ilowup of operational events; witness of surveillance test; review of licensee action on previous inspection findings; review of design change
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modifications; review of potential flooding events; review of solid rad waste onsite storage; review of containment purging; review of Task Action Plan Cate-gory "A" Requirements, and observation of plant physical security.
The inspection involved 95 inspector hours onsite by two resident inspectors.
Results: No items of noncompliance were identified.
61000lb0 h Region I Form 12 (Rev. April 77).
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DETAILS 1.
Persons Contacted D. Army, Technical Assistant H. Autio, Plant Superintendent G. Babineau, Engineeing Assistant W. Billings, Chemistry Supervisor T. Danek, Operations Supervisor B. Drawbridge, Associate Engineer L. French, Engineering Assistant T. Henderson, Technical Assistant K. Jurentkuff, Day Shift Supervisor L. Laffond, Assistant Training Coordinator P. Laird, Maintenance Supervisor N. St. Laurent, Assistant Plant Superintendent R. Randall, Engineering Assistant J. Staub, Technical Assistant to Plant Superintendent J. Trejo, Plant Health Physicist D. Vassar, Assistant Operations Supervisor F. Williams, Engineering Assistant The inspectors also interviewed )ther licensee employees during the inspec-tion, including members of the Ope ations, Health Physics, Instrument and Control, Maintenance, Reactor Engineering, Security and General Office staffs.
2.
Licensee Action on Previous Inspection Findings (Closed) Item of Noncompliance (50-29/80-02-01):
Administrative procedure AP-2005, " Operations Department Surveillance Schedule" has been revised to require a weekly surveillance of the Switching and Tagging Log and the Bypass of Safety Function and Jumper Control Request Log to insure that the subjects records are being maintained properiy.
The 'nspector reviewed the subject logs and identified no inadequacies. This item is closed.
(Closed) Item of Noncompliance (50-29/80-02-02):
Operations Procedure OP-4630, " Accumulator Time Delay Actuation Verification" has been revised to incorporate precautions regarding the polarity sensitive nature of the time l
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Manager of Operations Directive (MOD) 80-03 has been issued I
reinforcing the requirements for proper testing after plant modifications.
The inspector observed that the licensee performed a " full train test" of the Emergency Core Cooling Systems after recent alterations were completed on these systems during the recent shut down period.
This item is closed.
(Closed) Unresolved Item (50-29/80-02-03):
Administrative Procedure AP-7001
" Shift Technical Advisor", has been approved and implemented.
This pro-cedure formally details the responsibilities and authorities of the Shift Technical Advisor.
This item is closed.
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(Closed) Unresolved Item (50-29/80-07-01):
The appropriate emergency procedures have been revised to incorporate additional cautions rer.ommended by the Westinghouse Guidelines.
This item is closed.
(Closed) Unresolved Item (50-29/80-07-02):
The inspector reviewed the applicable procedures and determined that the required steam generator water inventory, necessary for terminating a Safety Injection, is now stated uniformly throughout the procedures.
Additionally a curve of the steam generator water level which assures sufficient decay heat removal has been added to the procedure as an operator guide.
Further, precautions have been added where applicable to limit the Safety Injection pump opera-tion to be consistent with the modified Westinghouse Guidelines.
This item is closed.
(Closed) Unresolved Item (50-29/80-07-03):
The inspector reviewed training attendance sheets for operator training which occurred on October 16, 1980.
The outline for the training, " Seminar in Small Break LOCA and Inadequate Core Cooling Phenomenon in PWR's" was also reviewed.
The outline and discussions with the instructors indicated that the operators had been reinstructed in emergency procedures with particular emphasis on conditions surrounding Natural Circulation, use of the saturation monitor, and use of redundant channels of indication along with the pressurizer level as the primary indication of a small break LOCA.
This item is closed.
(Closed) Unresolved Item (50-29/80-07-04):
The licensee has installed a caution sign near the Saturation Monitor, warning the operator that three rapid successive operations, i.e., changing the meter indication from degrees from saturation to pounds per square inch from saturation or the opposite action will place the instrument into a test mode.
The licensee has installed a sign near the Accident Area Radiation Monitor (AARM) stating the cr;tical limits in scientific notation, at which various categories of emergencitis would be declared.
Additionally, the licensee's emergency procedures have been revised to make the notation consistent with AApt display.
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.
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AP-0001, Plant Procedures and Instructions, Revision 8.
AP-2002, Operations Department Personnel Shift Relief, Revision
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AP-2009, Control Room Area Limits for Control Roor Operators, Original.
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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 5.
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AP-0018, Bypass of Safety Function and Jumper Control Log, Revi-
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sion 7.
AP-2007, Maintenance of Operations Department Logs, Revision 7.
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AP-0216, Housekeeping and Cleanliness Control, Revision 1.
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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.
i The above procedures, Technical Specifications, ANSI N18.7-1972,
" 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:
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Control Room logs and shift surveillance sheets are properly com-pleted and that selected Technical Specification limits were met.
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Control Room log entries involving abnormal conditions provide sufficient detail to communicate 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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Specifications requirements.
Jumper (Bypass) log does not contain bypassing discrepancies with
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Technical Specification requirements and that jumpers are properly approved and installed.
c.
The following plant logs and operating records were reviewed:
Shift Supervisor's Control Room Log:
September 16 through October
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31, 1980.
Special Orders:
446, 461, 465, 466, 469, 470, 471. 472 474, 477,
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478, 479.
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Safety Related Maintenance Request Logs:
80-1007,80-995, 80-
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991,80-990, 80-987,80-979, 80-970.
Switching and Tagging Orders:
all effective orders.
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Bypass of Safety Function and Jumper Control Log Request:
all
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active and inactive requests.
Key Control Log:
September 16 through October 31, 1980.
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Radio Log:
September 16 through October 31, 1980.
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Manager of Operations Directives 80-01 through 80-03.
Plant Information Reports 79-01 through 80-09.
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No inadequacies were identified.
4.
Plant Tour The inspector conducted a tour of accessible areas of the plant including the Primary Auxiliary Building, Turbine Building, Safety Injection Building, Switchgear Room, Diesel Rooms, Control Room, Vapor Containment Spent Fuel Building, Radwaste Building, and HP Control Point Areas.
Details and findings are noted below, a.
Monitoring Instrumentation and Annunciators Control Boaro annunciators were checked for alarms, abnormal for plant conditions, on several occasions during the inspection.
The following monitoring instrumentation was checked to verify that required instru-mentation was operable and that, where applicable, values indicated
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were in accordance with Technical Specifications.
Pressurizer pressure, level and temperature.
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Charging flow path.
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MCS Temperature.
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SI tank level.
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PWST and DWST levels.
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Batteries 1, 2, and 3 bus voltage.
Megawatt electrical output.
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Stack gas radiation monitor.
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Containment air particulate radiation monitor.
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No abnormal annunciators were energized.
No items of noncompliance were identified.
b.
Radiological Controls Radiation controls established by the licensee, including posting of radiation areas, radiological surveys, condition of. step off pads, and the 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 Radiclogical Surveys."
No items of noncompliance 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 items of noncompliance were identified.
d.
Fluid Leaks and Piping Vibrations.
Systems and equipment in all areas toured were observed for the exist-caca of fluid leaks and abnormal piping vibration.
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Pipe hangers / Seismic Restrain'.s Pipe nangers and restraints installed on various piping systems through the plant were observed for proper installation and tension.
No items of noncompliance were identified.
f.
Control Room Manning / Shift Turnover Control Room manning was reviewed for conformance witii the require-ments of 10 CFR 50.54(k) and Technical Specifications.
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 inspec-tion.
All were noted to be thorough and orderly.
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5.
Surveillance Testing a.
The inspector observed portions of the following surveillance tests to verify that testing was performed in accordance with technically adequate procedures, that results were in 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 personnel.
The following surveillances were reviewed by the inspector:
OP-4231, Monthly Waste Gas Leakage Check.
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OP-4252, Security Diesel Surveillance.
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OP-4206, Flow Test of HPSI Pumps on Normal AC Power.
OP-4203, Monthly Valve Check.
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OP4204, Monthly Test or Special Operation of the Safety Injection Pumps.
OP4205, Safety Injection System Operational Check.
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OP-4222, Reactor Rod Control System Precritical Check.
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OP4214, Chemical Shutdown System Operability Check.
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No unacceptable conditions were identified.
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THIS PAGE, CONTAINING 10 CFR 2.790 INFORMATION, NOT FOR PUBLIC DISCLOSURE, IS INTENTIONALLY LEFT BLANK.
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6.
Review of Design Change Modifications Through record review the inspector verified for the design changes listed below that design changes were made in accordance with 10 CFR 50,59; that design changes were reviewed in accordance with Technical Specifications and the established Quality Assurance program; that design changes were conducted.in accordance with written procedures which included identifica-tion of inspections required by codes or standards, and acceptance test procedures which defined acceptance values or acceptance standards; that test records verified performance of equipment modified to Technical Speci-fications/FSAR requirements and performance of modified equipment was reviewed and approved; that operating procedures and modifications were made and approved in accordance with technical specifications; that installa-tion procedures were adequate for the identified function; that as-built drawings were changed to reflect the modifications;and that records of design changes were maintained as described in 10 CFR 50.59(b) and the established QA program.
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PDCR 80-07, HPSI pump individual recirculation lines.
EDCR 80-10, Containment Isolation System Switch Replacement (electrical
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reset).
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EDCR 80-26, Steam Generator Seismic Supports.
With the exception of the item below, the inspector identified no signifi-cant inadequacies.
(1) PDCR 80-07 During the recent shutdown, Plant Design Change Request (PDCR) 80-07 installed individual minimum finw orifices in the recirculation line of each High Pressure Safety Injection (HPSI) pump.
This change was required due to the differences in,the head capacity curves and system flow resistance among the three HPSI pumps which caused a flow imbalance, resulting in the #3 HPSI pump being starved for flow during dead head operation.
The installation of the orifices corrected the flow imbalance problem, however during testing in accordance with OP-2000.85, " Flow Test of two HPSI Pumps on Normal AC Power", following installation of PDCR 80-07, excessive vibration was noted on the #3 HPSI pump.
Investi-gation of this problem lead to disassembling the pump.
Examination of the pump intervals found that a pump disc retaining ring had been
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displaced off the pump shaft.
The retaining ring was replaced, minor adjustments were made with the vendor's representative's assistance and a larger orifice was installed in the #3 HPSI pump recirculation line.
This corrected the vibration problem.
Surveillance Test OP 2000.85 was performed with several field changes incorporated.
The inspector reviewed the documentation, test procedures and Job Order
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.80-125 associated with PDCR 80-07 and verified, in addition to the above, that the field changes made to the procedure were reviewed by the Plant Operations Review Committee.
The inspector had no further questions in regard to the item.
(2) EDCR 80-10 This modification involved replacing existing switches located in the Primary Auxiliary Building (PAB) with remote switches of a similar type located in the Control Room, and replacing the charging pump switches with another type switch.
The reason for the change resulted from an analysis which determined that in the event of a major *ccident the present reset switches (lockout relays) fo the containment isolation system (CIS) valves, which are located in the unshielded PAB, may not be accessible without exposing personnel to excessive radiological hazards.
This chcrgo enables an operator to remotely reset any or all of the CIS valves from a single, shielded point in the Control Room.
After completion of the modification the system was tested in accordance with OP-6000.121, " Installation and Testing of the CIS Switch Replace-ment (Electrici Reset)" and OP-4610, "PS-CI-230 and PS-CI231 Calibra-tion and Containment Isolation System Operability Test".
During this testing a lock out relay (LOR) for one channel was damaged and that channe3 failed to operate.
Investigation of this incident revealed that supervisory lights were installed across the reset lockout relays as an extra operator aid in order to determine when the relays were energized.
The current flow through these lights was tested and found to be of sufficient magnitude to hold a relay open after it was deenergized, thus causing re ays to function out of sequence in the reset log matrices causing a short circut in the system.
The inspector questioned the licensee in regard to the test program and design and engineering evaluation of the added relay supervision lights.
The inspector reviewed the design and engineering evaluation of the supervision lights and the results of the test program which verified that the current flow through the supervision lights was insufficient to energize a lockout relay.
This test program did not however, verify that the supervision light current flow would hold a relay open after it was deenergized.
The licensee removed the supervision lights from the circuit, replaced the relay, and tested both channels satisfactorily.
The inspector reviewed the results of repeated aforementioned tests OP-4610 and OP-6000.121, and found that they had not been reviewed by the cognizant department supervisor.
This matter was brought to the attention of the licensee and will be followed during a future inspec-tion.
(50-29/80-16-01)
7.
Inspector Followup of Events
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The inspector responded to events that occurred during the inspection to ensure continued safe operation of the reactor in accordance with
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the Technical Specifications and regulatory requirements.
The follow-ing items, as applicable, were considered during the inspector's review of operational events:
observations of plant parameters and systems important to safety
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to confirm operation within normal operational limits; description of event, 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 eqtipment;
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verification of proper manual actions by plant personnel;
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verification of adherence to piant procedures; verification of conformance to Technical Specification LCO require-
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ments; determination that root causai factors were identified and that
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corrective actions, taken or planned, were appropriate to correct the cause; verification that corrective 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 manaer which constituted an unreviewed safety question as defined 'n 10 CFR 50.59(a)(2), or in such a manner as to represent an unustal hazard to health and safety of the public and environ-ment;
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determination whether the event involved continued operation of the facility in violation of regulatory requirements or license conditions; and,
evaluation of whether applicable reporting requirements were met.
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The operational event reviewed during this inspection period is discussed
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a.
Reactor Scram on High Flux Level On October 29, 1980 with the reactor at 20% power; a main coolant dilution in progress, steam dump to the Condenser in manual control and the turbine generator at synchronous speed ready to be phased to the transmission lines, the reactor power increased to the low power trip set point of 35% and an automatic scram occurred.
The inspector investigated this occurrence and deter-mined through discussions with operations personnel and review of
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Plant Information Report 80-11 that the cause of this occurrence was attributed to the sluggish operation of the condenser steam dump valve, cotpled with the combined effect of reactivity inser-tion from the Xehon burnup, boron dilution and mod 9tatnr tempera-ture depression.
The moderator temperature depression rese?ted from the steam dump valve hanging up slightly and then being overcompensated in the open direction.
Sluggish operation of the condenser steam dump valve resulted in a rapid increase in the volume of steam being dumped to the condenser.
The higher rate of steam flow caused a cooling of the primary system and attendant insertion of positive reactivity by virtue of the effect of the moderator tempeature coefficient.
The net result was the reactor power level increased to the low power trip setpoint of 35% of full power and the reactor tripped.
The inspector verified that a reactor power level increase occurred simultaneously with a large increase in steam flow.
The inspector had no further questions in regard to this item.
8.
Task Action Plan Category "A" Requirements Certain TMI lessons learned requirements were designated as category
"A" requirements and were required to be completed by January 1,1980.
These items were initially issued in NUREG 0578 then subsequently, rer:.5ered with clarification in TMI Task Action Plan (TAP) (NUREG 0660).
Many of these items have,been previously addressed in Inspec-tion Reports 50-29/80-02 and 50-29/80-07.
Additionally, an onsite visit was made by an NRC team on April 2, 1980 to review and evaluate the licensee's implementation of the Category "A" items.
This evalua-tion and the commitments made by the licensee are documented in a letter from D. Ziemann, Division of Operating Reactors to J. Kay, Yankee Atomic Electric Company dated April 18, 1980.
During this inspection the inspector confirmed the licensee's implementation of the category "A" items as stated in NUREG 0578 and as modified by a Nuclear Reactor Regulation (NRR) letter entitled " Preliminary Clarifi-cation of TMI Action Plan Reosirements", dated September 5, 1980.
The inspector reconfirmed many of the items and verified all the commitments made by the licensee as disci ssed in the April 18, 1980 evaluation by NRR. The following observatians were noted:
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TAP Number I.A.1.1.
This item'was addressed in Inspection Report 50-29/80-02.
The inspector however, reconfirmed that an STA program has been implemented, that the qualifications and training requirements have been met and that formal responsibilities of tre STA have been delineated in approved procedures.
The inspector reviewed the STA training. program examinations and discussed classroom activities with selected STA's during back shifts.
General discussions concerning the STA program, including shift assignments, shift reliefs, and STA functions were held.
The inspector had no further ccacerns in this area.
I.A.1.2.
Shift Supervisor Responsibilities This item was addressed in Ir.spection Report 50-29/80-02.
In addition, the inspector reviewed Manager of Operations Directive (MOD) 79-6 and 80-04 which emphasize the primary management responsibilities and authorities of the Shift Supervisor for safe operation of the plant.
The inspector had no further ccncerns regarding this item.
I.A.1.3.
Shift Manning The inspector verified that the license. received, routed and took adequate corrective action for IE Circular 80-02, Nuclear Power Plant Staff Work Hours.
The corrective action taken included a management directive, Special Order 479, which requires operators to conform with the required maximum work schedules provided in the circular.
The inspector had no further concerns regarding this item.
I.C.1.
Short Term Accident and Procedures Review This item was addressed in Inspection Report 50-29/80-07 and 50-29/80-02 in which sevc-'l concerns were left unresolved.
The inspector has subsequer ily reviewed these items, and the inspector's concerns have been resolved as stated in paragraph 2 of this report.
The inspector had no further concerns in regard to this item.
I.C.2.
Shift and Relief Turnover Procedures This item was addressed in Inspection Report 50-29/80-02, I
which verified that the checklist, logs and procedures in l
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use were adequate.
Further, the inspector witnessed shift
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turnovers on a routine basis and has detennined that the shift relief is performed in a thorough and orderly manner.
The inspector had no further concerns in regard to this item.
I.C.4.
Control Room Access This item was addressed in Inspection Report 50-29/80-02 which verified that adequate administrative controls eeisted to control access to the control rocr. The inspector tours the control room on a routine basis and has not identified any inadequacies in this area.
II.B.3 Post-Accident Sampling The inspector reviewed OP-9450, Post Accident Sampling and Analysis, 0F-2658, Revision 6, Operation of the Post Accident Vapor Containment Hydrogen Control System, Yankee Atomic Electric Company Letter WYR 80-47 from J. Kay to Mr. D.
Crutchfield, Chief Operating Reactors Branch #5, dated May 1,1980 and Yarkee Atomic Electric Ccmpany '.etter WYR 79-163
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from J. Xay to Mr. H. Denton, Director, Nuclear Reactor
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Regulation, dated December _31,1979. The inspector verified that the licensee has implemented interim procedures for obtaining and analyzing reactor coolant samples and contain-ment atmosphere samples with existing equipment while keeping occupational exposures as low as reasonably achiev-able. The licensee's sampling capability at this time, however, is limited to sampling either primary coolant or containment air activity up to.01 curie per milliliter (Ci/ml). The licensee's source term, as defined in Regulatory Guide 1.3 or 1.4 has been calculated to be approximately 2 curies per milliliter, well above the.01 Ci/ml. This area will be evaluated further during a Health Physics Appraisal 50-29/81-01, conducted January 5-16, 1981.
Additionally, the licensee has completed a conceptual design description for reactor coolant and ccotainment atmosphere monitoring in order to meet the category "B" requirements tur this item. Additional modificatiens to the design may be required, as the plant shielding review has concluded that most on-site buildings would be uninhabitable for
several days after the design basis accident.
Further review is required to determine the overall plant shielding requirements.
. tis has been deferred to the Systematic Evaluation Program (SEP). The inspector had no further concerns in regard to this ite.
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II.D.3.
Valve Position Indication for Relief and Safety Valves This item was addressed in Inspection Report 50-29/80-02 in which the inspector witnessed portions of the installation of the acoustic accelerometer system and reviewed OP-6000.109,
" Installation of the Pressurizer PORV, SV's Position Indica-tion System."
In subsequent inspections, the inspector witnessed testing of the completed installation.
No inade-quacies were identified.
II.E.1.2. Auxiliary Feed System Indication and Flow Item 2.1.7.b of NUREG 0578 was addressed in Inspection Report 50-29/80<2 in which the inspector reviewed EDCR-79-17, OP-6000.108, "Special Test," and minutes of the PORC meeting 80-02.
The inspector also witnessed partial installa-tion of the ultrasonic flow transmitters on one emergency feedwater line and later during the inspection discussed the operation of the main control board " flow display" with reactor operators.
No inadequacies were identified.
Item 2.1.7.a of NUREG 0573 was addressed in a letter from G.
Lainas, Safety Assessment Division of Licensing to J. Kay, Yankee Atomic Electric Company dated Juae 16, 1980.
This letter states that the Automatic Initiation and Termination i
of AFW Flow to Steam Generators, will be evaluated under the Systematic Evaluation Program (SEP) and will be addressed in a separate safety evaluation report.
The inspector had no further concerns in regard to this item.
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II.E.3.1 Emergency Power for Pressurizer Heaters This item was discussed in Inspection Report 50-29/80-02 in which
..e inspector reviewed r gineering Design Change a
Request 79-32, " Power Supply Change for PR-MOV-512, "and OP 5000.110, "Special Test," and minutes of PDCR 79-51.
The above documentation indicates that the power supply to motor operated block PRV-MOV-512 on the pressurizer relief line has been removed from the 480V MCCI bus and connected to the
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480V Emergency MCCI bus.
The inspector discussed this l
change with the reactor operator on shift and ascertained that he was knowledgeable of this change.
No inadequacies were identified.
Additionally, the inspector verified, by I
observation of Controlled drawings and discussions with Maintenance and Operations department personnel that four l
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groups of pressurizer heaters normally fed from the 480 volt
.ses can be supplied from emergency power by closing circuit
creakers in the control room.
Each group of heaters has a l
capacity of 37.5 kw per group.
The licensee's representative
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stated that 37.5 kw of heater capacity is sufficient to maintain natural circulation operation.
The inspector had no other concerns in regard to this item.
II.E.4.2 Isolation Dependability
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This item was addressed in Inspection Report 50-29/80-08 in which the inspector reviewed EDCR-7944, EDCR-7955 and OP-6000.106, "Special Test".
The inspector also witnessed portions of the Containment Isolation Reset Features, which ensures that these valves must be individually reset from the control room following their closure.
No inadequacies were identified.
II.F.2.
Instrumentation to Detect Inadequate Core Cooling This item was addressed in Inspection Report 50-29/80-02 in which the inspector reviewed EDCR-79-30 which installed a primary coolant saturation meter which provides continuous indication of the margin from saturated conditions.
Shift reactor operators demonstrated the operation of the device and their familiarity with its capabilities and purposes.
No inadequacies were identified.
II.G.1.
Power Supplies for Pressurizer Polief Valves and Level This item was addressed in Inspection Report 50-29/80-02 and in item II.E.3.1. of this report.
Additionally, the inspec-tor verified by observation of con' trolled drawings and
discussions with the Instrument and Control Supervisor that power supplies for the pressurizer level indication instru-ment channels are supplied by the vital bus, and the solenoid operated relief valve is supplied by an emergency DC power supply.
The inspector had no further concerns in regard to this item.
III.A.1.2. Upgrade Emergency Support Facilities The inspector verified by physical observation that the licensee has established a Technical Support Center (TSC)
adjacent to the control room.
The TSC is equipped with appropriate plant pra dures, drawings, reference material, communications, including dedicated communications with the NRC and a dedicated communication capability with the off site Emergency Control Center (ECC).
The TSC is also equipped j
with a TV screen and controls to operate a TV camera, located j
in the control room, capable of viewing control room instru-l ments with sufficient detail and accuracy to obtain the
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required operating parameters.
The inspector had no other concerns in regard to this item.
III.A.1.2. Onsite Operational Support Center The inspector verified thAt the licensee has established an onsite operational Support Center (OSC) in the cnntrol room back foyer.
The OSC is described in plant proceoures and the site emergency plan.
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III.D.1.2. Primary Coolant Outside Containment The inspector reviewsd AP-7009, Revision 1, System Leak Reduction Program and held discussions with the Reactor Engineering cognizant personnel.
The inspector determined that the licensee has established a leak reduction program for those systems which may contain radioactivity following an accident.
The program includes a periodic visual inspec-tion to identify leakage and take appropriate corrective action, leak rate testing to be performed on a refueling bases, and separate procedures for performing leak checks on the Waste Gas Systems.
The inspector noted that the procedure for performing Quarterly Leak Rate Testing had not been formalized or approved.
The inspector will review this item during a subsequent inspection.
(50-29/80-16-02)
III.D.3.3. Inplant Radiation Monitoring The inspector reviewed OP-8740, " Measurement of Radioactive Airborne Release Rates Under Accident Conditions", OP-8' 1,
" Determination of Noble Gas Release Rates from Main Steau.
Lines Under Accident Conditions," and OP-8701, " Operational
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Source Check of the Eberline SAM-2 for the determination of Radiciodine Airborne Concentration".
The inspector observed that the equipment and operating procedures for the quantifi-cation of noble gas effluents released from the primary vent stacks are available in the event that the existing equipment goes offscale and that approved procedures are available for obtaining release rates from the Steam Dump and Safety Valves.
Noble gas release rates from the Primary Vent Stack (PVS) will be determined using a radiation detector installed on the PVS sampling line with a local read out.
The inspector i
observed this installation.
The radiation readings will then be converted to exhaust concentrations.
The licensee has also provided a description of the method used to deter-mine radiciodine and particulate effluents.
The inspector
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NUREG 0758 Number 2.1.5.C.
Reco1biner Procedures Review and Upgrade The inspector verified by visual observation and by holding discussions with operations department personnel that there is no hydrogen recombiner on site.
The licensee's procedure OP-2658, Operation of the Post Accident Vapor Containment Hydrogen Control System however, adequately addresses the venting of hydrogen frcm containment.
The NRC's oosition in
. regard to this item is that the procedures for use of.the
' hydrogen control system be reviewed considering shielding requirements and personnel exposure limitations.
The Procedure OP-2658 provides for the operation of the Vapor Containment and Hydrogen Control System from behind adequate permanent shielding.
The inspector had no further concerns in this area.
9.
Review of Potential Flooding Events As a result of the Indian Point Unit No. 2 flooding event, each licensed plant was surveyed to determine whether the same conditions exist at each facility.'
The following is a brief synopsis of the conditions that exist at Yankee Rowe in regard to containment sump, levels, indication, pumping and source of water:
There is only one sump associated with the containment at Yankee Rowe.
This sump level has pneumatic level indication readout in the control room with an associated alarmed annunciator.
In addition, the Auxiliary Operator obtains hourly readings off the sump level.
There is no sump pump associated with this system.
The location of the sump in relation to the containment is such that all leakage from the containment drains to the sump, which is external (in another building) and an extension of the contain-ment.
The containment and the sump is pressurized with air under normal operations to approximately 1.5 psig.
Pumping the sump is accomplished by manually opening a valve allowing the pressurized liquid to flow to the Gravity Drain Tank.
Cooling water is supplied to the containment air coolers from the Service Water Systems.
The Service Water System is an open system in that it utilizes river water as its source and leakage could not be detected by an inventory iecrease.
The operating history of the system in regard to leakars has indicated that there has been no SIGNIFICANT leakage.
However, one cooler was replaced due to excessive leakage during the history of the plant.
Additionally, the service water system to the containmen.t coolers can be isolated from outside the containment.
The coolers at Yankee Rowe are only used during the summer months.
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The licensee is presently in process of taking appropriate actions and responding to IE Bulletin 80-24, Prevention of Damage Due to Water Leakage Inside Containment.
The above information was forwarded to Division of Reactor Operations Inspection for review.
The inspector identified no inadequacies in this area.
10.
Review of Solid Waste Processing and Storage During the inspection period each operating reactor site wds surveyed to determine the status of onsite torage of low level radioactive waste.
The following is a brief synopsis of the status of onsite radioactive waste at Yankee Rowe as determined by the inspector, through discussions with cognizant Health Physics personnel:
The licensee is presently operating its radioactive waste facility at one hundred percent capacity.
Any slow down of shipments or other delays, results in backlogs exceeding the design onsite storage capacity for low level waste.
The licensee presently has no plans for expanding the onsite low level waste storage facility.
Waste shipments are regularly shipped to Barnwell, South Carolina.
If waste shipments were terminated, the facility would have one to three months before maximum waste storage capacity would be reached.
Accumulation of waste varies in type with modes of operation, i.e., shutdown or operating, however the amount remains approximately constant except during refuelings, during which the volume triples.
The licensee's present waste in storage meets the January 1981 burial criteria.
This information was forwarded to cognizant Fuel Facilities and Material Safety Inspection personnel for review.
The inspector identified no inadequacies in this area.
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11.
Review of Containment Purging During the inspection period each resident inspector was requested to l
verify the licensee's response to an NRR interim position on Containment i
Purging and to verify the commitments involved.
The inspector reviewed l
the NRR position, the licensee's response dated November 1, 1979, Technical Specifications and OP-2478, " Operation of the Vapor Container (VC) Purge System".
The operation of the VC purge system is adminis-tratively prohibited whenever containment integrity is required, i.e.,
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Main Coolant greater than 200 F or 300 psig.
The procedure OP-2478 also requires that the venting and purging valves will remain locked
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shut when the reactor is not in a cold shutdown or refueling mode.
This information was forwarded to NRC Headquarters for review.
The inspector identified no inadequacies in this area.
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12.
Observations of Physical Security The inspector made observations, witnessed and/or verified during regular and offshift hours that selected aspects of plant physical security were in accordance with regulatory requirements, the physical security plan and approved procedures.
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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 actions was present as required.
manning of all shifts on various days was observed to be as
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required.
b.
Physical Barriers selected barriers in the protected area and vital area were
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observed and random monitoring of isolation zones was per-formed.
Observation of vehicle searches were made.
c.
Access Control (1) Observatt.ns of the following items were made:
identificaticn, authorization and badging;
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access control searches, including the use of compensa-tory measures during periods when equipment was inoper-able; and, escorting.
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Except as noted below, no inadequacies were identified and the inspector had no further comments in this area.
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i THIS PAGE, CONTAINING 10 CFR 2.790 INFORMATION, NTO FOR PUBLIC DISCLOSURE, IS INTENTIONALLY LEFT BLANK.
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13.
Exit Interview At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and findings.
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