IR 05000029/1981006
| ML20009G135 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/09/1981 |
| From: | Foley T, Gallo R, Lazarus W, Varela A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20009G124 | List: |
| References | |
| 50-029-81-06, 50-29-81-6, NUDOCS 8108030360 | |
| Download: ML20009G135 (18) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-29/81-06 50-29/81-05-18 50-29/81-05-19 Docket No.
50-29 50-29/81-05-14 50-29/81-05-02 Category C
License No.
DPR-3 Priority
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Licensee:
Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name:
Yankee Nuclear Power Station (Yankee Rowe)
Inspection at: Rowe, Massachusetts Inspection conducted:
April 1 - May 30, 1981
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Inspectors:
T. foe Giorfrasident Inspector da'te ' signed flAWAmW
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W. La ru e
ns ctor dat'e signed A /a x
7/9/si A. Agrelh, Reactor Inspector dat# signed
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Approved by: /
- <e1 4 a e 9[
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R. gad, Citief, RPS No. dA date signed Div. of Resident and Project Insp.
Inspection Summary:
Inspectica on April 1 - May 30, 1981 (Report No. 50-29/81-06)
Areas Inspected: Routine onsite inspection by the resident and region based inspectors of plant operations including tours of the facility; log and record reviews; operational safety verification; followup of selected IE Bulletins; and circulars; review of Desicn Change Modifications; review of Preparations for Refueling; surveillance t'6 sting; follow up of licensee events; review of fire protection and observation of Physical Security. The inspection involved 151 inspector hours by the resident NRC inspector and two region based inspectors.
Results: Within the areas inspected no items of noncompliance were identified in eight areas:
two apparent items cf noncompliance were identified in two areas, paragraphs 4 b (2) and 12 a (4).
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DETAILS 1.
Persons Contacted H. Autio, Plant Superintendent E. Begiebing, Maintenance Supervisor W. Billings, Chemistry Supervisor T. Danek, Senior Advisor - Operations L. French, Technical Assistant T. Henderson, Plant Reactor Engineer F. Hicks, Training Coordinator J. Hoffman, Manager Engineering Group NSD K. Jurentkuff, Assistant Operations Supervisor J. Kay, Senior Engineer NSD L. Laffond, Technical Assistant W. McGee, Manager of Public Information F. McWilliams, Engineering Assistant R. Mitchell, Technical Assistant L. Reed, Operational Quality Assurance Coordinator N. St. Laurent, Assistant Plant Superintendent J. Staub, Technical Assistant to Plant Superintendent J. Trejo, Health Physicist Supervisor D. Vassar, Operations Supervisor Mercury Company Personnel A. Duda, Construction Superintendent J. Duguay, QA Supervisor J. Leh, Construction Superintendent M. Barla, QC Technician l
M. Trombley, Project Manager Other Personnel M. Cammarano, Inspection Specialist, Hartford Steam Boiler Inspection and Insurance Company The inspector also interviewed other licensee employees during the inspection, l
including members of the Operations, Health Physics, Instrument and Control l
Maintenance, Reactor Engineering, Security and General Office staffs.
2.
Licensee Action on Previous Inspection Items a.
(Closed)
Inspector Follow Item (29/77-30-01): The inspector reviewed the revised licensee procedure OP-4611 Nuclear Instrumentation and l
Reactor Protection System Precritical Check.
Revision 9.
This procedure requires the use of " Instrument in Test" stickers while the instrument l
is out of service. The inspector has witnessed the use of these l
stickers on several occasions during tours of the facility. This item is closed.
b.
(Closed) Unresolved Item (29/78-16-04): The inspector reviewed OP-1221 Movement of Reactor Core Components Within the Spent Fuel Pit.
l Revision 7.
The administrative controls outlined in the precautions of l
this procedure appear to have edequate control over the operation of l
the Spent Fuel Pit elevator. This item is closed.
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(Closed) Unresolved Item (29/79-02-04):
The inspector reviewed OP-4204,,"lionthly Test or Special Operation of the Safety Injectt6n Pumps" and Plant Design Change Request (PDCR 80-18), Installation of SI pump suction Pressure Gages. The licensee has been recording.and evaluating SI pump suction pressures during the preceeding four Surveillances and has revised OP-4204 to require SI pump suction pressure be recorded, however, this revision has not been approved.
PDCR-80-18 has been authorized and approved which will permanently install pressure gages in the suction lines of the SI pumps specifically to comply with ASME Section XI IWP-3100. This item is closed.
d.
(Closed) Unresolved Item (29/79-05-01): This item has been previously addressed in inspection reports 50-29/80-12, 50-29/80-07 and 50-29/79-06. Additionally the inspector has attended training courses providing instruction for the newly revised procedures.
e.
( Closed)
Inspector Follow Item (29/79-05-03): The inspector reviewed OP-2100 Plant Startup from Cold Shutdown and OP-2652 Preparation of the Emergency Core Cooling System for Normal Operation. These procedures have been revised to clarify the operation and positioning of SI-MOV-535. This item is closed, (Closed)
Inspector Follow Item (29/79-05-04 : The inspector reviewed f.
OP-4203, "lionthly Valve Check" and OP-4241, ') Suppler ental Locked Valve L These two procedures constitute the satisfactory control over the safety related locked valves.
This item is closed.
g.
(Closed)
Inspector Follow Item (29/79-05-05): The inspector reviewed OP-4211," Emergency Boiler Feedwater System Operability Test,"and interviewed several operators concerning the operability and purpose of t.ie nitrogen cylinder for the auxiliary trip valve and Nuclear s,aam control valve.
All persons interviewed were knowledgeable l
c the purpose and use of the apparatus. This item is closed.
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(Closed)
Inspector Follow Item (29/79-05-06): The licensee has
revised procedure OP-2652," Preparation of the Emergency Core Cooling
System for Nomal Operation"; to require dual verification and sign offs for those valves which are not otherwise locked and checked on a periodic basis; and are located in the safety injection system direct flow path. This item is closed.
l 1.
(Closed)
Inspector Follow Item (29/81-04-02): The licensee has
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completed the visual inspection and sipping of tht: fuel and has made l
the appropriate notifications to the NRC of the as-found degradation l
of fuel cladding. This item is closed.
l 3.
Shift Logs and Operating Records a.
The inspector reviewed the following plant procedures to determine the licensee established administrative requirements in this area
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l in preparation for review of various logs and records.
i AP-0001, Plant Procedures 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 Cperators
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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 1
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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
" 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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l completed and that selected Technical Specification limits were met.
l Control Room log entries involving abnormal conditions provide
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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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Specification requirements.
l Jum)er (Bypass) log does not contain bypassing discrepancies l
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wit 1 Technical Specification requirements and that jumpers are l
properly approved and installed.
c.
The inspector reviewed, on a sampling basis, the following logs and records for the period April 1, thru May 30, 1981.
l Switching ar.d Tagging Log
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Maintenance Request 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. Lcg Sheets
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Shift Relief Checklist
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Primary Chemistry Log
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Secondary Chemistry Log
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Refueling Coordinators Log
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Radioactive Gaseous Release Pemits
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Radioactive Liquid Release Permits The following observations were noted:
On May 18, 1981, while transporting the reactor head from the Vapor Container to a railroad car for storage, the head brushed against the side of the Vapor Container equipment hatch knocking loose contamination around the railroad car.,a1 thousand disintegrations per minute on and levels resulting in sever The appropriate notification to the NRC was made and the contamination was immediately cleaned up.
Upon subsequent review of the Shift Supervisor's Log, the inspector noted insufficient detail to reconstruct the event, Additionally, on May 19,1981 while filling the Reactor Vessel Cavity, approximately one hundred gallons of borated water from the Safety Injection Tank leaked through an o)en valve from the pressurizer, Thissevent was not recorded in the Slift Supervisor's Log. This matter was discussed with the Plant Superintendent who noted that the Shift Supervisor's Log could contain more detail, however also stated that these events may be recorded in Coordinator's Logs)gs (ie:The inspector noted that the subject informaticn one of the other lo shutdown Maintenance Log and Refueling
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was not in either of these logs. The Plant Superintendent acknowledged that the Shift Supervisor's Log is the official log and this matter would be a topic of discussion at the next staff neeting.
The inspector will continue to review p7 ant logs to verify that they contain sufficient detail.
4.
Plant Tour The inspector conducted a tour of accessible areas of the plant including the Primary Auxiliary Building, Turbine Building, Safety Injection Building, Cable Tray Room, Vapor Container, Switchgear Room, Diesel Rooms, Control Room, Spent Fuel Building, and.HP Control Point Areas.
Details and findings are noted below:
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C a.
Monitoring Instrumentation and Annunciators On several occasions during the inspection, control board annunciators were checked for abnormal alams. The following monitoring instrumenta-tion was checked to verify operability and where applicable, values in-dicated 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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RCS tennerature
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SI tank level
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Boric acid mix tank level
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PWST and DilST levels
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Batteries 1, 2 and 3 bus voltage
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Megawatt electrical output Megawatt thermal output
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Vapor container drain tank level
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Primary vent stack radiation monitors
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Containment air particulate radiation monitor
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No abnomal annunciators were energize (.
b.
Radiological Controls Radiation controls established by the licensee, including posting of radiation areas, radiological surveys, conditions 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 l
Controlled Areas", and OP-8101, " Plant Radiological Surveys".
l (1) During routine tours of the Vapor Container the inspector detected an attitude in which it was perceived that regular plant employees were challenging the authority of health physics contractor techni-cians.
This manifested itself on one occasion when the inspector l
witnessed a maintenance. technician leaving a highly contaminated area within the reactor vessel cavity; while renoving his outer set of paper coveralls, instead of waiting for a health physics tecnni-cian s assistance, who was assisting another, the individual ripped the coveralls off.
Further discussions with both health physics
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contractor personnel and regular plant employees confirmed these -
l feelings of animosity. The inspector discussed this event with l
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the Health Physics Depar. ment head and the Plant Superintendent, emphasizing that activities involving radioactivity must be controlled. The Plant Superintend;nt discussed this at the sub-sequent staff meeting and again with key department supervisors, in the presence of the resident inspector. Subsequcnt tours of the facility have not revealed problems of this nature.
(2)
Technical Specification 6.12 requires, in part, that High Radia-tion Areas greater than 1000 mrem per hour be provided with locked doors to prevent unauthorized entry into such areas; and the keys maintained under the administrative control of the shift supervisor of Plant Health Physicist.
On May 14, 1981 during a routine inspection of the Vapor Container the inspector verified radiation readings in high radiation work areas. The inspector noted several unmarked hot spots greater than 1,000 millirem per hour on the Main Coolant Isolation Valve Stem Leak Off drain line in the area of the " Brass Drain Box".
One hot spot indicated as high as 12,000 millirem on contact and 1,200 milli-rem per hour at one foot from the hot spot and immediately adjacent to a ladder up to the loop No. 2 cubicle. This area was posted as a high radiation area and themost recent survey dated May 12, 1981 indicated maximumegeneral area radiation levels of up to 200 milli-rem per hour. The area was not posted as a high radiation exclusion area, however, the vapot. container is either locked or guarded and and RWP is issued to control entry into such areas. Due to the in-effective surveys of this area, the licensee did not have control within this area.
At the time of this finding, the facility was involved in a refueling outage which resulted in many individuals having unlimited access to all portions of the Vapor Container, including the Main Coolant Isolation Valve Stem Drain Line. While the Vapor Container access is guarded for security reasons, Radiation Work Permits are required, and Health Physics Technicians prov' de surveys cnd radiological. controls, none of these measures:was suffi-ient to preclude possible inadvertent access (un-authorized entry) to this area in which the dose rate was greater than 1000 mrem /hr.
Upon notification, the licensee promptly shielded the drain line sufficient to reduce the general area dose rate to less than 1000 mrem per hour, eleminating the need for a locked door as required by Technical Specification 6.12.
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The inspector informed the licensee that th'is failure to control high radiation areas in accordance with Technical Specification 6.12 constitutes an item of noncompliance (50-29/81-06-01).
c.
Fluid Leaks and Piping Vibrations Systems and equipment in all reas toured were observed for the existence of fluid leaks and abnormal piping vibration.
Except as noted below no inadequacies were noted:
During a routine tour of the Primary Axiliary Building on May 12, 1981 the inspector noted that the Component Cooling pipina appeared to be vibrating excessively. This was brought to.the. attention of the plant staff and subsequently the component cooling flow was varied to reduce the vibration. The inspector also noted that the Safety Injection pumps apparently leaked excessively through the end seals. The inspector also noted however, that maintenance requests had already been issued to correct this problem. This leakage was'noted to % corrected on subsequent tours of the facility.
5.
Surveillance Testing The inspector observed protions 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 thaa 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-4230, Steam Generator Safety Valve Test with the Plant in Hot
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i Standby OP-4215, Surveillace of the Boron Injection Flowpath
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OP-4502, Determination of the Steam Generator Safety Valve Set-
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points Using the Crosby Test Device l
l OP-46?8, Functional Test of PS-SI-14, Low Main Coolant Pressure
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Safety Injection Actuation OP-4519, Station Battery Capacity Test
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OP-5106, Inspection of Mechanical Suppressors
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OP-4226, Testing of Fuel Handling Equipment with the Dummy Fuel
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No inadequacies were identified.
6.
IE Bulletin Followup For the IE Bulletins listed below, the inspector verified the following:
that the written response was within the time period stated in the bulletin, that the written response included the infomation required to be and included adequate corrective action connitnents based on infor-
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nation presented in the bulletin and the licensee's response, that licensee management fomarted copies of the written response to the anpropriate
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onsite management representatives, that infomation discussed in the licensee's written response was accurate, and that corrective actica taken t'y the licensee was as described in the written response. The following bulletins were reviewed:
IEB 81-01, Surveillance of Mechanical Snubbers
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The inspector reviewed the licensee's rerponse to this bulletin dated February 27, 1981, and discussed the adequacy of the response with the NRC Regional Staff and NRC Headqua-ters Staff personnel cognizant of this bullet m It was detemined tnrough these discussions that the licensee's coranitments made in the response were not entirely adequate to satisfy the concerns addressed in the bulletin.
Tne inspector informed the license management on March 30,1981 that the minimum acceptable response to t'le bulletin, that would provide adequate assurance to the NRC, that the licensee's snubbers were operable, would be to perform the stroke test identified in Item 1.a of the bulletin on all snubbers installed on safety-related systems except those which are considered inaccessible and would require an excessive man REM exposure. This position was femally transmitted to the licensee by NRC Letter to Yankee Atomic Company dated May 1, 1981.
Subsequently the licensee satisfactorily tested the mechanical snubbers in accordance with Item 1.0 of the subject bulletin and performed OP-5106 Inspection of Mechanical Suppressors. The inspector reviewed the results of these tests and had no further questions in regard to the above.
-- IEB 80-04, Analysis of A PWR Main Steam Line Break with Continued Feedwater Addition The inspector reviewed the licensee's response dated May 8, 1980 which satisfactorily addresses the concerns of IEB 80-04. The licensee's action in regard to this bulletin has been to install two design changes that will alleviate the two concerns addressed in the bulletin that are applicable to Yankee Rowe; over pres w ization of the contain-ment, and the potential for return to power. The two design changes are; the auto tripping of the condensate pumps on coincidence with high containment pressure and low steam linc pressure, and auto boiler feed pump trip at power levels greecer than 15 MWe. The inspector has witnessed the auto tripping of the boiler feed pumps
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and has reviewed the PORC approved Engineering Design Change Request EDCR 81-09 which installs the condensate pump auto trip. This trip is expected to be completely installed prior to reactor startup.
The inspector identified no inadequacies in this area.
7.
IE Circular Fol10wup For the IE Circulars listed below, the inspeccur verified the circular was received by the licensee n'anagecent, that a review for applicability was perfonned, and that if the circular was applicable to the facility, appropriate corrective actions were taken or were scheduled to be taken.
The following circulars were reviewed:
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-- IEC 81-02, Performance of NRC Licensed Individuals while on Duty IEC 81-03, Inoperable. Seismic Monitoring Instruments
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-- IEC 81-05, Self Aligning Rod End Bushings for Pipe Supports
-- IEC 81-06, Potencial Deficiency Affecting Certain Foxboro 10 to 50 Milliampere Transmitters The inspector identified no inadequacies.
8.
Operational Safety Verification A detailed review of the Safety Injection System Accumulator 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 Emergency Core Cooling System for Nornal Operation. Review of the above system include the following:
verification that each accessible valve in the flow path was in
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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
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for components that are required to actuate upon receipt of a safety injection signal.
visual inspection of major components for leakaSe, proper lubrication,
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cooling water supply, general condition and other factors that might prevent fulfillment of their functional requirements.
verification by observation that the instrumentation essential for
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system actuation and performance was operational.
The inspector identified no inadequacies.
9.
Inspector Followup of Licensee Events
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The inspector responded to events that occurred during the inspection to verify continued safe operation of the reactor in cccordance with the Technical Specifications and regulatory requirements.
The following items, as applicable, were considered during the inspector's review of operational events:
observation 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 material
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and exposure to radioactive material.
verification of correct operation of automatic equipment.
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verification of adherence to plant procedures.
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-- verification of conformance to Technical Specification LC0 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 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 manner which constituted an unreviewed safety question as i
defined in 10 CFR 50.59 (a7 (2), or in such a manner as to represent an unusual hazard to health and safety of the public and environment.
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determination whether the event involved continued operation of
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the facility in violation of regulatory requirements or license conditions; and, evaluation of whether applicable reporting requirements were met.
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The events reviewed during this inspection period are discussed below.
a.
On May 2,1981, during the scheduled plant shut down for refueling, Control Rod No. 17 became inoperable. During the normal insertion and scram testing of the control rods while shutting down, the voltage was removed to the stationary gripper coil (. the subject control rod, and the rod failed to scram. The rod was subsequently driven in by use of a pull down coil. All technical specification requirements applicable to a stuck rod configuration were met.
The licensee subsequently determined by visual examination of the i
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control rod that the control rod follower was bowed; the center being displaced by approximately.5 inches. The inspector observed the control rod and noted the bowing.
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The licensee has subsequently cut the control rod into pieces and transfered the rod to the spent fuel pit for storage. All hafnium control rods with zircalloy followers will be replaced with spare silver, indium and cadmium control rods during this refueling outage.
The licensee submitted Licensee Event Report 50-29/81-05 in regards to this event.
b.
On May 2, 1981 while performing 0P-4502, Determination of the Steam Generator Safety Valve Setpoints Using the Crosby Test Device, the number 1 Steam Generator low set point safety valve (MF-SV-409E) was tested and found to be 5 PSI below the required setting of 935 lbs.
plus or minus three percent. The inspector witnessed the testing of these valves during plant shutdown.
The licensee subsequently reset and tested the valve satisfactorily.
This event was identified in Licensee Event Report 81-06.
The inspector identified no inadequacies in regard to the above event.
10. Preparation for Refueling The inspector reviewed licensee preparations for refueling to ascertain that approved procedures will be available for fuel handling activities and that new fuel had been received and inspected in accordance with approved procedures. The following documents were reviewed.
i OP-7200, Receiving, unloading, and inspecting New Reactor Fuel,
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Rev. 4.
-- OP-7202, Fuel Sipping, Original.
-- OP-7107, Moving Fuel Within the Spent Fuel Pit, Rev. 6.
OP-1206, Transfer of Assemblies in the Spent Fuel Pit Using the
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Manual Tool, Rev. 7.
-- OP-1207, Exchange of Control Rod Drive Shafts, Rev. 5.
OP-1209, Operation of the V.C. Manipulator Crane Handling Fixtures
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and Transfer Equipment, Rev. 6.
OP-1210, Venting and Sampling of Gas from Under the Reactor Vessel
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Head, Rev. 4.
OP-1214, Transfer of New Fuel Between the New Fuel Vault and the Spent
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Fuel Pit, Rev. 7, l
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OP-1221, Movement of the Reactor Core Components Within the Spent
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Fuel Pit, Rev. 6.
OP-1500, Relaxing Reactor Head Studs, Rev. 5.
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OP-1502, Reactor Lower Core Support Structure, Removal and Replacemeat,
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Rev.5.
OP-1507, Reactor Head - Removal, Handling and Storage, Rev. 4.
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OP-1508, Handling and Storage of Reactor Head Studs, Rev. 4.
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OP-1509, Installation of V.C. Manipulator Crane Universal Handling
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Tool, Rev. 4.
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OP-1510, Reactor Upper 3 re Barrel and Plates - Removal and Storage,
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Rev. 4.
OP-1516, Inspection of New Fuel Elevator, Rev. 4.
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OP-1700, Cycle XIV Reactor Refueling and Component Inspection, Rev. 7.
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OP-1100, Dismantling and Reassembly of the Reactor Systems for Core
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XIV Refueling, Rev. 4.
OP-4226, Testing of Fuel Handling with the Dummy Fuel Assembly, Rev. 9.
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-- OP-4239, Setting V.C. Integrity and Operability Check of the V.C. and Spent Fuel Pool Ventilation Systems OP-4505, Inspection and Testing of Fuel Handling Equipment, Rev. 6.
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Execpt as noted below, no inadequacies were identified.
l The inspector noted that reconstitution of fuel assemblies may be necessary during this outage. The licensee plans for Exxon to perform this work.
The required procedures have not been developed / reviewed by the licensee.
This item will be reviewed in a subsequent inspection (50-29/81-06-02).
11.
New Fuel Receipt The inspector reviewed the results of new fuel receipt inspections
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l perfonned by the licensee to verify that inspections had been performed in accordance with OP-7200, " Receiving, Unloading, and Inspecting New Reactor Fuel". The inspection documentation for fuel shipment XN-4 of October 10, 1980, consisting of 36 assemblies was reviewed.
Except as noted below no inadequacies were identified.
OP-7200, Part C is an inspection for fuel assembly bowing. The procedure consists of using two.040 inch spacers to separate a straightedge from the side of an assembly and then measuring the clearance between the straightedge and assembly at various points along the length. The procedure requires recording clearances of less than.04 inches but
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does not specify an acceptance criterion.
Five Assemblies were noted as having clearances less than.040 inches but were found to be acceptable.
Discussions with the Reactor Engineer revealed that any bowing greater than.040 inches was unacceptable. None of thcse assemblies fall into that category. The licensee agreed to revise GN7200 to clearly define the acceptance criteria for fuel bowing. This item is unresolved pending review of the revised procedure (50-29/81-06-03).
12.
Design Change Modifications a.
Reactor Support Structure Modification (1.) Modification Description The modification is designed to increase the seismic resistance capabilities of the RSS. This modification resulted from a seismic evaluation of the support structure at Yankee Rowe, identified in Yankee Atomic letter to the NRC dated October 15, 1980 regarding the Seismic Program at Yankee Rowe. This evaluation identified that the column connection to the foundation base plate and anchor bolt system was deficient in both moment and shear resistance.
Earthquake Engineering Systems Incorporated designed strengthening collars for this connection. These collars are to be installed at the existing foundation.
This modification will provide additional resistance to the seismically induced moments and shears, and increase the ductility of that part of the structure.
(2.)ScopeofInspection l
The inspection effort consisted of a review of the contractor's
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quality control and work procedures required in YA program for Modification of the Yankee Nuclear Power Station Reactor Support Structure Column Bases, cbservation of on-going work performance and review QC records of completed work. The inspector interviewed i
licensee, contractor and craft personnel during this inspection.
The status of this work is estimated to be 20% completed.
(3.) Inspection Details (a)
Procedure Review i
The following QC work procedures were reviewed:
-- External Column Rock Bolt Drilling External Column Rock Bolt Location
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External Column Rock Bolt Setting
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-- Welding Procedure Specification B External Column Rebar, Concrete Form and Concrete Installation
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Procedure l
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No items of non-compliance were idc.1tified.
QC records for work completed under the first three procedures listed above were reviewed.
One item of non-compliance was identified, this is described in subparagraph number 4.
(b) Record Review of Contractor Documentation The inspector performed a record review of licensee procurement control and receipt inspection on purchased material for the reactor column base modification hardware. The licensee's purchase order for Williams Super High Tensile Rock Anchors was observed to be from an approved vendor and to conform to the required specifications. A pre-shipment surveillance was perfomed by licensee.
Receipt inspection records were observed to be adequately identified and complete as to certificates of compliance and certificates of test.
No items of non-compliance were identified.
(c) Work Observed The following work activities were observed:
-- welding of longitudinal #14s rebar in existing foundation, exposed at exterior column lap splices rock anchor bolt setting
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reinforcing steel installation for column pedestals
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-- fitting-up of column pedestal concrete forms grouting of rock anchor bolts
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QC verification of above activities
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Except as identified in subparagraph number 4, no items of non-compliance were identified.
(4.) Noncompliance Item - Proprietary Grout
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YA drawing number SPD-0006-CF-1002, revision 6A and Special Procedure specification N 49855-703 specify grout around rock anchors to be Will-X-cement grout with 4000 pri compressive strength and, the spec'al Procedure specification requires a certificate of compliance certifying the minimum compressive strength of the grout.
a.
contrary to the above the proprietary gr6ut received at the site in purchase order number 66250-49855/ invoice 10871, and used
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around the rock anchors was not accompanied by a certificate of test showing compliance of the manufactured lot to the prescribed strength, b.' contrary to the above the proprietary grout was not user tested at the site to provide assurance of its compressive strength.
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Tnis is -an item of noncompiiance.
(50-29/81-06-05)
(5.) Resolution of Unresolved Item Identified at Exit Interview The inspector expressed at the exit interview this concern regarding the welding of #14S longitudinal reinforcing steel exposed in the
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existing column foundation.
Depending on design requirements the controls imposed ts insure an efficient weld of the lap spliced
- 14S steel at each column base was questioned by the inspector.
In a telephone interview with licensee's mechanical engineering manager April 29, 1981 the inspector was informed that the design of the reactor column base modifications was satisfied without welding the lap splices. The weld is for conservatism to avoid i
possible spall of the foundation concrete under maximum seismic
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load.
Licensee agreed however, to add magnetic particle non-destructive testing to verify tre efficiency of the single-flare-V-groove weld.
The testing of #14S lap-splice rebar weld will be reviewed in a subsequent inspection.
(50-29/81-06-04).
12.
Fire Protection (TI 2515/19)
During the NRR evaluations of fire protection programs at operating plants, various critical areas in some plants were identified where a fire may affect the redundant safe shutdown systems.
In these plants, modifications are being made such as additional fire protection systems or providing an alternate shutdown capability.
Until such modifications are complete, inspectors were requested to inspect these critical fire areas periodically during the routine inspection program. The critical fire areas identified at Yankee Rowe are:
control room
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j switchgear room
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turbine building
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manhole no. 3
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During routine tours of the plant the inspector observed the critical fire areas for the following:
Housekeeping - combustibic material not being stored in a critical
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Temporary use of combustible material in critical areas minimized.
area.
If work is in progress requiring open flares or other ignition sources,
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verify tnat fire watches have been established.
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Locations of fire extinguishers are unobstructed and, if possible,
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check that extinguisher pressure and/or level is adequate.
Fire alarm reporting stations are clearly identified and unobstructed.
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If a Carbon Dioxide or Halon type system is used for suppression, verify
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that the system pressure and general condition of the system is adequate, i.e., nozzles not blocked, valves properly aligned, actuation devices not covered, etc.
Whenever installed fire detection or suppression systems in the critical
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areas were out-of-service, fire watches were established as required by the applicable technical specifications, d
Surveillance tests on systems serving the critical fire areas have
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been accomplished as required by the technical specifications.
Except as noted below the inspector identified no inadequacies:
On May 18, 1981 the inspector noted that the Southwest door to the turbine building and adjacent control room lower hallway door to the turbine building were affixed in the open position. The inspector brought this to the attention of the Fire Protection Supervisor who took immediate corrective action.
The inspector reviewed OP-4563, Surveillance of Penetration Fire Barriers.
Performance of this procedure is the method by which the licensee satisfies the technical specification requirement to verify all Fire Barriers are operable. The turbine building doors are not listed in this procedure as being required to be operable-The Yankee Rowe NRR Safety Evaluation Report (SER) states in paragraph 5.7 Turbine Building, The turbine builcing contains safety-related electrical cables routed from the adjacent switchgear room. An un-mitigated fire in the turbine building could threaten the structural capability of the building which supports the enclosures for the switchgear
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room and control room.
Fires in the areas of cable routings would cause l
the loss of essential service functions.
l It appears to the inspector that a fire in the Station Service Transformers, located outside and approximately five feet from the turbine building and
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adjacent to the Southwest turbine building door could cause an unmitigated i
fire in the turbine building. The licensee laintains that the turbine building doors are not required to be fire barriers. The Yankee Rowe NRR
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SER indicates that the turbine building doors should be included as fire barriers.
This matter is unresolved.
(50-29/81-06-05).
13. Observations of Physical Security The inspector made observations, witnessed and/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.
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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.
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Physical Barriers 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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Access Control Observations of the following items were made:
-- identification, authorization and badging; 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 14. Management Meetings At periodic intervels during the cour:e of the inspection, meetings were held with senior facility manageme.it to discuss the inspection scope and preliminary findings of the resident inspector. A summary of the inspection was also provided at t1e conclusion of the inspection on May 29, 1981.
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