ML20134N702
| ML20134N702 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 10/15/1996 |
| From: | Keller R AFFILIATION NOT ASSIGNED |
| To: | NRC OFFICE OF ADMINISTRATION (ADM) |
| Shared Package | |
| ML20134N672 | List: |
| References | |
| FOIA-96-423 NUDOCS 9611270084 | |
| Download: ML20134N702 (2) | |
Text
l Information Focus on Energy, Inc.
20608 Gleaning Court, Suite 102 GAITHERSBURG, MD 20882
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~e F0lA/PA REQUEST
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Case No:
9 M ZJ l
i Date Rec'&
to -1/ - T 6 Action Off; h
i.X: C 50-October 15, 1996 l
Freedom of Information Officer U.
S. Nuclear Regulatory Commission Washington, D.
C.
20555 l
Re: Freedom of Information Act Request - N109 l
l
Dear Sir /Ms.:
This is a request under the Freedom of Information Act.
I request that a copy of the following records and documents be l
provided to me:
A copy of all Site Resident reports, Region Inspections, and Headquarters inspections / site visits for the Maine Yankee Nuclear Power Plant ' Docket Number 50-309) for the period l
from January 1, 1996, through September 30, 1996.
l l
Electronic copies of the records and documents are preferred, if available.
If not available, paper copies are acceptable.
In order to help determine my status to assess fees, you should l
know that I am a representative of the news media as defined in l
10 CFR 1004, Section 1004.2(n).
My Company publishes a newsletter on icsues related to commercial and governmental uses of nuclear materials.
The information requested is of current l
interest to our readers.
l I also request that, if the material requested contains classified or otherwise exempt information, that reasonably segregated portions be provided after deletion of portions which are exempt under the FOIA.
KELLER96-423 PDR l
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I I request a waiver of all fees for this request.
This request meets the criteria of 10 CFR 1004.10(a) (8) (i) for waiving of fees because (1) the subject of this request concerns the operations and activities of the Department of Energy (DOE - a j
government agency), (2) the information will contribute to an understanding of safety at a key DOE facility not presently l
readily available to the public and, (3) the information can l
contribute significantly to a public understanding of government
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I operations.
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l The INFORMATION FOCUS ON ENERGY NEWSLETTER is distributed 'a minimal cost to persons throughout the world, including members of scientific, public interest and academic organizations through l
subscriptions via the INTERNET.
Payment for copying fees would l
add significantly to the costs of production, thus possibly l
reducing the range of distribution.
er President i
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U.S. NUCLEAR REGULATORY COMMISSION j
REGION I i
REPORT NUMBER:
50-309/95-26 j
DOCKET NUMBER:
50-309 I
LICENSEE NUMBER:
DRP-36 i
l LICENSEE:
Maine Yankee Atomic Power Company i
329 Bath Road i
Brunswick, Maine 04011 FACILITY:
Maine Yankee Atomic Power Station INSPECTION DATES:
November 14, to December 31, 1995 INSPECTORS:
J. Yorokun, Senior Resident Inspector I
W. Olsen, Resident Inspector li E. Conner, Project Engineer A. Cerne. Reactor Engineer
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G. Dentel, Peactor Engineer C. Beardslee, Reactor Engineer l
J. Williams, Operations Examiner i
J. Collins, NRR l
C. Dodd, NRR APPROVED BY:
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F. Rogge, Chief L/ V Ddte j
eactor Projects Branch 8 l
Scope:
Resident inspection and safety assessment of plant activities including operations, maintenance, engineering, and plant support.
In addition, two other inspections and a meeting as follows are
- attached, i :
RADIOLOGICAL SAFETY INSPECTION i :
SECURITY INSPECTION MEETING REPORT FOR NRC/ MAINE YANKEE NOVEMBER 21, j
1995.
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Overview:
See executive sumary, i
I 96020601'23~V60130 l
PDR ADOCK 05000309
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e, EXECUTIVE
SUMMARY
Plant Ooerations There was good management attention and presence during fuel loading. Once corrective actions for previously identified fuel loading problems were implemented, operators completed fuel loading activities safely. The improvement in fuel loading activities was indicative of an effective problem resolution and corrective action implementation process. Other plant restart evolutions such as lowering of cavity level for vessel head installation were completed without any incidence.
Plant operators performed well and showed good safety awareness. Noteworthy was a senior reactor operator's quick decisive action to put the plant in a safe and predetermined condition during diagnostic activities on a reactor coolant pump when high pump vibration required that the pump be immediately secured.
The plant continued to maintain a good chemistry program that provided excellent support to safe plant operations as demonstrated by efforts to address issues created by steam generator tube sleeving.
The licensed senior reactor operators demonstrated an ac table understanding of technical specifications. Licensed reactor operators are proficient in locating and manipulating controls and instrumentation. The crews generally functioned as a team, demonstrated an acceptable level of performance in the simulator providing reasonable assurance that plant operational emergencies could be effectively addressed.
The recent training given to licensed operators provided a sound basis for them to startup and operate the plant in a safe and effective manner. Modeling of steam generator characteristics and resultant plant behavior were realistic for the changes associated with sleeving the steam generator tubes.
However, a violation of NRC requirements was identified when the combination of the inadequacy of procedure 1-104-14.3, and an inadequate control of plant configuration caused a potentially radioactive significant spill in the containment.
The activity was not prescribed by appropriate sequence (steps) in the procedure and operators failed to maintain proper configuration control when they started filling a generator with " tagged" open drain valves.
Maintenance Maine Yankee personnel performed well at analyzing and repairing the reactor coolant pump in a safe manner.
There was excellent supervisory oversight and good support from the engineering department.
Your staff indicated that actions were been taken to conduct an overall review and assessment of the diesel driven fire water pump's reliability in accordance with the new maintenance rule being implemented.
Inspectors had expressed concern on the reliability of the pump in light of the problems that your staff has experienced with the pump in the past couple of years.
Surveillance tests were conducted safely by very knowledgeable individuals.
Exceptional performances were noted during the conduct of the engineered a
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safety features integrated surveillance tests prior to plant restart. Pretest briefings were good and detailed. Test controls were clearly stated and reiterated.
The test processes were discussed and made clear to all personnel. The on shift designated managers and the test directors preformed l
well conducting the briefings.
Enaineerina i
The engineering department continued to provide good support to the plant.
Good safety perspective was evident. Corporate engineering department i
- i personnel demonstrated excellent technical capabilities and management oversight with the steam generator sleeving activities. 1hese activities were 4
i completed safely.
Plant engineering department personnel continued to provide good support to the plant in addressing engineering issues such as with the j
secondary component cooling system valves; and the demineralized water storage tank.
Engineerii:9 duigr. changes and installation controls were adequately
- I implemented, with good records of quality programs department activities, as-i built walkdowns, and functic,nal verification and testing.
Engineering backlog was being controlled well and safety issues received the highest priorities.
Reactor engineering and Yankee Nuclear Service Olvision personnel provided excellent support for fuel loading activities and performed the inspections of control element assemblies that were imparted by the refueling machine during 4
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fuel loading, safely, i
Plant Succort The material condition of radiological waste systems, including the domineralizer and filter pipe tunnel, high radiation storage bunker, and current radiological waste processing equipment, were in good condition.
Radiological controls to support putage work were also good. Your self-assessments of the radiological controls program, including quality program department surveillance and radiological incident reports (RIRs) included excellent observations, identified program areas for improvement, and were being properly communicated to management. However, a violation of NRC requirement was identified involving three examples of failure to maintain administrative control of access to high radiation areas.
The physical security program was being appropriately directed towards assuring public health and safety.
Three previously identified open items were closed. Management support was evident. Required audits were thorough, in depth and comprehensive in scope. Security training was being performed in accordance with the NRC approved training and qualification plan and vital area access control was being limited to only authorized personnel who needed access to perform their duties.
Protected Area detection equipment was installed and operated as committed to in the physical security plan and timely repairs were being completed on security equipment.
In Fire Protection area, personnel performed well at identifying and correcting the potential pressure locking problem with the fire protection 111 m
system. There were prompt corrective actions taken to resolve a potentially serious condition which may have prevented portions of the fire protection system from operating as needed.
Safety Assessment /Ouality Verification Quality Programs Department (QPD) personnel involvement in the fir,a1 oversight and independent review of the steam generator sleeving project acilvities was very comprehensive and provided good backup to the extensive in-p ocess
- urveillance controls that were in place. There was an excellent initiative by QPD in the use of statistical analysis to monitor various sieraing processes which enhanced various process re11 abilities.
Management provided an excellent oversight of outage activities, fa rgent issues were well discussed with good focus on safety. Upper manag9 ment involvement in the steam gencrator closeout program controls (e.g., NCR 95N-024) was good.
Plant Operations Review Comittee and Nuclear Safety and Audit Review Comittee members exhibited excellent questioning attitude during their meetings. The comittee members demonstrated excellent safety perspective during thet. discussions of issues.
Maine Yankee's implementation of the interim corrective actions process was progressing well. Good consideration of nuclear safety was being made in determining what comitments were accomplished or terminated. The Learning Process Re-Engineering Team appeared to be well organized.
Overall, personnel efforts were good at ensuring that the plant would be restarted safely from the extended shutdown. The Restart Readiness Team's effort was comprehensive and well received by the licensee's upper management.
There was good management support and resolution of identified issues. QPD personnel provided good and independent oversight of restart ac'Ivities.
Maine Yankee demonstrated excellent safety perspective in this aces.
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TABLE OF CONTENTS
' EX EC UT I VE S UPfMR Y,..................,.....,,.
11 1.0 PLANT OPERATIONS I
1.1 Fuel Loading I
1.2 Operation at Lowered Cavity Level..............
3 1.3 (VIO 50-309/95-26-01) Water Spill in Containment Durin Steam Generator No.3 Fill.............. g 4
1.4 Reactor Coolant System Chemistry 5
1.5 Operations Depertment Restart Readiness...........
6
=1.5.1 Licensed Operator Readiness for Startup and Power Operation.......................
6 1.5.2 Simulator Modeling of Steam Generator Behavior after Tube Sleeving....................
7 1.6 S a fe ty Sys t ems Wal kdown..................
8 1.6.1 Containment......................
8 1.6.2 Emergency Diesel Generators..............
9 1.6.3 Primary Component Cooling Water System 9
1.6.4 125 Volt DC Olstribution System............
10 1.6.5 Chemical and Volume Control System 10 1.7 Reactor Coolant Pump (RCP) No.2 Diagnostics.........
11 2.0 MAINTENANCE............................
12 2.1 Maintenance Observation...................
12 2.1.1 Control Element Assembles (CLAs) CN, E6, and 03 Inspection 13 2.1.2 Overtorqued Steam Generator No.3 Primary Manway Studs.
13 2.1.3 Reactor Coolant Pump No.2 Repairs 14 2.1.4 Diesel Driven Fire Pump (P-5) Repairs.........
14 2.1.5 RC-M-25 Repairs....................
15 2.2 Surveillance Observations..................
15 2.2.1 Safety Injection Actuation with loss of A/C Test 15 2.2.2 ECCS Operational Test Recirculation Actuation System 16 2.2.3 Results Review, ECCS Operational Pump and Flow Check Valve Testing.....................
16 3.0 ENGINEERING............................
16 3.1 Steam Generator Tube Sleeving Completion 17 3.1.1 Completed Non-Conformance Reports...........
18 3.1.2 Quality Program Department Verification Activities 18 3.1.3 Eddy Current Testing Review..............
19 3.2 Engineering Design Change Package and Records Review 20 3.2.1 EDCR 95-41, Reactor Coolant Pump Differential Pressure
!nstrumentation Upgrade - Reactor Coolant S measurement................ystem flow 20 3.2.2 EDCR 95-45, Containment Spray System Pressure Locking Modifications.....................
21 3.3 Secondary Component Cooling Water Yalves SCC-A-460 1 461 22 3.4 Demineralized Water Storage Tank Leakage 23 3.5 Reactor Coolant Pump Vibration Moritoring System Upgrade 23 y
EMw
TABLE OF CONTENTS (Cont) 3.6 Engineering Backlog.....................
24 4.0 PLANT SUPPORT....................,......
24 4.1 Radiological Controls....................
24 4.2 Security 24 4.3 Emergency Preparedness 25 4.4 Fire Protection.......................
25 4.4.1 Fire Protection System Piping Modification 25 4.4.2 Fire Damper Fusible Link Surveillance.........
25 5.0 SAFETY ASSESSMENT / QUALITY VERIFICATION 26 5.1 Plant Operation Review Committee (P]RC)...........
26 5.2 Nuclear Safety and Audit Review Co'.nittee (NSARC)......
27 5.3 Corrective Action Program and "!r.terim Corrective Action Program - Task Prioritization Frocess" 27 5.4 Restart Readiness Programs 28 5.5 Previously identified issur.s 28 5.5.1 (Closed), URI 95-002-01, Containment Spray Pumps Flow and NPSH 28 5.5.2 (Closed) URI 95-08-01, Demineralized Water Storage Tank Tell-Tale Leakage 29 5.5.3 (Closed) VIO 95-022-01, Inadvertent Removal of HPSI Pump Recirculation 29 6.0 ADMINISTRATIVE 30 6.1 Persons Contacted......................
30 6.2 Summary of Facility Activities 30 6.3 Interface with the State of Maine..............
30 6.4 Exit Meeting 30 :
RADIOLOGICAL SAFETY INSPECTION :
SECURITY INSFlCTION :
MEETING REPORT FOR NRC/ MAINE YANKEE NOVEMBER 21, 1995 vi
DETAILS 1.0 PLANT OPERATIONS The plant was maintained safely in shutdown while steam generator tube sleeving and other cycle 14/15 refueling outage activities were completed.
Following the completion of the planned corrective actions for earlier fuel loading problems the plant resumed fuel loading on November 19, 1995. The last fuel bundle was loaded into the reactor on November 23, 1995 without any significant incidence. Plant heat up was delayed when reactor coolant pump 2 indicated high vibration during operation. Otagnosis of the problem was ongoing at the end of the inspection period. Plant restart was also further delayed by issues involving the ew rgency core cooling system (ECCS) and containment capabilities brought forth in an anonymous letter fontarded to the NRC by the State of Maine. The resolution'of those issues was ongoing at the end of the inspection period.
The inspectors verified operability of selected Engineered Safety Features (EST) trains and assessed the condition of plant equipment, radiological controls, security and safety. The inspectors evaluated plant housekeeping and cleanliness to ascertain that they were maintained well and had no detrimental effect on plant safety.
Once corrective actions for previously identified fuel loading problems nre -
implemented, operators completed fuel loading activities safely.
Communication was excellent. Management involvement was evident. The inspector noted that management improvement in fuel loading activities was indicative of an effective problem resolution and corrective action implementation process. Details of inspection findings in the operations area are discussed later in this section.
On December 18, 1995, the inspector observed operations personnel draw a steam bubble in the pressurizer in accordance with station procedure 1-9-1, Reactor Coolant system Fill and Vent. During the process, a dedicatad operator was stationed at the primary side of control room board to monitor plant pressure while the reactor coolant system was solid. The nuclear plant operators were knowledgeable of the procedure requirements and performed satisfactory. The bubble was formed at 9:38 a.m. and reactor plant pressure was maintained at approximately 150 pounds per square inch in preparation for reactor coolant pump runs.
1.1 Fuel Loading Fuel loading resumed on November 19, 1995 and the last fuvi bundle was loaded on November 23, 1995. Then, the three control element assemblies (C[A), which had been impacted by the refueling machine spreader during fuel loading, were removed to the spent fuel pool (SFP) for inspection. Prior to the resumption of fuel loading activities, Maine Yankee had established a " management fuel load inspection expectations" program, which addressed the increased
' management attention needed to be placed on fuel loading activities. Managers were scheduled to provide coverage of fuel loading activities during all the shifts. The coverage involved the control room, spent fuel pool, and refueling cavity area visits for assessment of refueling activities. The inspector observed good management attention and presence during fuel loading.
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2 The inspectors verified that the Itcensee had satisfactorily completed the prerequisites for fuel movement required by procedure 10-1, Core Reloading, prior to comencing fuel loading. The inspectors verified the following:
Equipment hatch closed as required.
Personnel hatch operable (both doors could remain open with contingency plans in place for closure if required).
Containment vent and purge valves handswitches positioned in the
" Refueling" mode.
Manipulator Crane RMS (R!-6104) and Containment Low Range Area Monitor RMS (RI-6105) operable as indicated on the control room RMS panel.
Four Wide Range 1.og Channels operable (minimum two required).
RHR Train B in service with Train A availaMe if required.
Cavity level at about 44 ft. 6 in., required level was at least 31 ft.
i.e. 23 f t above top of core.
Comunication established between the Control Room, Refueling Bridge, and the Spent fuel Building.
A boration path established per procedure 1-7-2, Operations at Cold shutdown, Attachment 1.
The inspectors attended licensee briefings in the control room and on the refueling floor prior to comencing fuel movement. The briefings were detailed, clear and comprehensive. The lessons learned from a previous "cumy" fuel bundle move were properly discussed. The inspectors observed fuel movement activities from the spent fuel pool area, the refueling floor, and bridge, Activities were conducted smoothly and safely. Comunication was clear and controlled. However, during the second shift of fuel loading, the inspector noted what appeared to be a comunication problem when several questions, such as: when to send upender back to SFP; when to grapple /ungrapple fuel in the SFP; and when to obtain another fuel bundle were being asked of the refueling supervisor at the same time.
In addition, two problems occurred while the crew was attempting to lower a fuel bundle into the reactor. First, a foreign material was observed in the core at the location that the fuel was about to be placed. Then, the upender hydraulic pressure indicated 170 psig, which was less than the minimum 200 psig specified in the procedure. The refueling supervisor appropriately halted refueling operation to discuss the issues. With operations halted, and the fuel bundle safely placed in the upender machine, the inspector expressed his concerns with comunication to the refueling crew members and the reactor engineering personnel. Following a detailed discussion that included clartfication of expectations for comunication, the crew resumed fuel loading. The inspectors observed that comunication was improved and good and that the refueling floor supervisor showed good comand and control of refueling activities. Personnel performed knowledgeably and well. To address the upender hydraulic pressure issue, a temporary procedure change (TPC) was implemented to change the minimum acceptable pressure from 200 psig to 150
" pstg. The concurrence of the machine manufacturer was obtained for the change. What had appeared to be a foreign material in the core turned out to be a discoloration on a fuel bundle.
During fuel movement on Novenicer 21, 1995, another control element assembly 1
3 (CEA) was inadvertently lifted out of the core by the spreader for about an inch.
Personnel spotted the lift and lowered the CEA. Then, they adjusted the machine, and withdrew tL spreader as previously planned for this type of occurrence. This made the number of CEAs inadvertently pulled out during this refueling, three (two for a few inches and one for a' Jut four feet).
After all fuel had been loaded, Maine Yankee unloaded the three fuel bundles, whose CEAs (GN, 03, & E6) had been impacted by the spreader during fuel loading activities, into the spent fuel pool for inspection by Maine Yankee and Yankee Nuclear Services Division engineers. No damage was noted and the three bundles were returned to their respective locations in the core. The inspector reviewed the licensee's inspection plans and li W ned to the pre-inspection discussions / briefing. The inspection activities were video taped for record purposes. The inspector later watched the inspection videos and observed no discrepancy with the licensee's conclusion. Each CEA was pulled out to about 5 feet and inspected from all directions (North, South, East and West). Some crud patterns and scruffs were noted, but no excessive scratchc5 were observed.
The inspector, concluded that and the root causes of previous fuel loading problems had been identified, the licensee took effective corrective action and completed fuel loading activities safely and without any other significant problems.
1.2 Operation at Lowered Cavity Level Using some of the guidance contained in NRC Temporary Instruction (TI) 2515/103, Loss of Decay Heat Removal, the inspector ascertained that the licensee was properly prepared to prevent ar.d if necessary, respond to loss of decay heat removal when the cavity level was reduced to just below the vessel.
flange (about 19 feet, 4 inches) in accordance with procedure 1-17-4, Cavity Draining, for reactor vessel head installation. Three reactor vessel level indications were operable in the control room. These were the reactor vessel narrow range level indicators LIA-104X and Y, and cavity wide range level indicator LIA-105. Control room " cavity low level alarm" was operable (window RH 3-10) with an alarm setpoint of 19 feet. Reactor Coolant System (RCS) temperature indications were available in the control room from core exit temperature (CET) thermocouple and the residual heat removal (RllR) inlet temperature. The high temperature alarm was set at 110 degrees fahrenheit
(*F) on the plant computer. The CET indication was 104 'F and RHR inlet temperature was 106 'F.
Both trains of RHR were operable with the B train in operation. The high pressure safety injection (HPSI) system was available as a source of cooling and inventory if required. A detailed " time to boil" analysis (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) had been completed and properly considered in making contingency plans.
The evolution was completed without any incidence. Plant operators performed well and showed excellent safety awareness while lowering the cavity level and raising the vessel level.
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4 1.3 (VIO 50-309/95-26-01) Water Spill in Containment During Steam Generator
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o No.3 Fill 1
On December 7,1995, operators commenced filling the secondary side of steam generator (S/G) No.3 with water using procedure 1-104-14.3, Placing S/G No.3 l
In Wet Lay-Up and S/G No.3 Low Pressure Tube Leak Test. During the process, water was spilled onto the -2 foot elevation of the containment. Apparently,-
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water had drained from the generator into the containment sump and the containment sump had filled and overflown onto the floor of the containment in l
the reactor vessel head storage area. The reactor vessel head was installed j
on the reactor vessel.
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Investigation revealed that three valves in the drain line from the steam generator to the containment sump had been inadvertently left open during the j
steam generator fill evolution. During the steam generator sleeving i
activities, the valves had been administratively " tagged" open to aid in the l
ventilation process required for sleeving activities. However, after the tags were turned over to the operations department, operators had relied on the 1
i fill procedure to get the proper valve line-up and had failed to close the valves prior to commencing filling of the generator. Contributing to the 4
j problem was that the steps in the procedure that verif ted the alignment of the drain valves came after the generator filling was initiated. The inspector i
determined that due to what appeared to be a combination of inadequacy of the
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procedure and operators not maintaining the proper configuration control, about 1,200 gallons of water was spilled from the steam generator. After i
about 100 gallons of this filled the containment sump, about 600 gallons backed up into the containment trench and the remaining 500 gallons or so i
flooded onto the containment floor. The spill onto the floor was less than an inch and was restricted to the reactor vessel head storage area.
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l The inspector reviewed the white tagging procedure (0-14-1) and the outstanding white tags for the drain valves that had been left open to determine if any inadequacies in the process had contributed to the problem.
The inspector noted that operators had not performed the monthly review of outstanding white tags in December as required by step 6.12.2 for the first i
Sunday of every month. When nottfled of this discrepancy, the plant shift i
supervisor immediately initiated the required reviews.
In addition, a temporary procedure change was made to procedure 1-104-14.3 to verify the position of the steam generator drain valves prior to initiation of steam generator fill.
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The inspector determined that while the spill in the containment caused no significant radiological conditions, it resulted from inadequate configuration control by the operators.
It appeared that the outstanding white tags for the drain valves had not been properly reviewed and evaluated against the steps of the fill procedure. A possible contributing factor was that operators had not i
^ ~ performed monthly reviews of outstanding white tags required by procedure 0-j 14-1, White Tagging Procedure. The combination of the inadequacy of procedure 1-104-14.3, and an inadequate control of plant configuration caused a j
potentially radioactive significant spill in the containment. This appears to be a violation of 10 CFR 50, Appendix B, Criterton V, instructions, i]
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5 Procedures, and Drawings, which requires that activities affecting quality be
+i: : prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.
1.4 Reactor Coolant System Chemistry During the steam generator tube sleeving process, Afrollene pneumatic oil was used within the steam generator bowls. This type oil had not been classified for use on nuclear safety applications by Maine Yankee and was restricted from use in safety related nuclear systems that had chemistry control. Upon discovery of this problem the station chemistry department sampled the
-Airo11ene oil to determine what chemicals were ) resent. The salapling results showed that significant levels of sulphur and p1osphorus were present in excess of those specified for unrestricted use in the reactor coolant system (RCS).
Maine Yankee chemistry department personnel developed a plan to clean the RCS after' steam generator sleeving activities and prior to resumption of power oper'ations.
In addition to removal of the sulphur and phosphorus, the plan included removal of the following: any residue from the reactor coolant loop decontamination effort that occurred early in the outsoes any laser welding sout any remnants from a bruken heat treatment bulb glassi steam generator vent lation dehumidifier lithium; and any paint chtp left in the steam generator. Several hold points were created in the outage critical path for monitoring RCS chemistry parameters.
The inspector reviewed Maine Yankee's RCS cleanup plan, station procedure 3 4-1, Primary System Chemistry Surveillance and technical specifications, sections 1.1, 3.2, 3.5, 3.6, 3.7, 5.8 and 4.2-1.
The inspector determined that the plan was appropriate with sampling / testing at the proper stages of the plant restart effort to verify proper chemistry. The normal reactor coolant cleanup path is through the letdown system purification train and this system will be used to accomplish the required cleanup activities.
On December 18, 1995, inittel RCS chemistry sampling indicated a very low amount of sulfur and phosphorus present after initial cleanup activities. The samaling continued on a daily basis until the end of the inspection period wit) no abnormal conditions noted. The licensee will continue to monitor the RCS chemistry until a determination is made that all of the contaminants are removed or at target levels that are safe for plant operation.
The inspector concluded that Maine Yankee demonstrated good safety perspective at investigating and promptly resolving this issue. Plant personnel developed a very comprehensive action plan to address RCS chemistry prior to plant startup. Thelicenseecontinuedtomaintainagoodchemistryprogramthat Sf0ditilElllGnl140E00110 lifts 1601BSlfillBnit
6 1.5 Operations Department Restart Readiness 1.5.1 Licensed Operator Readiness for Startup and Power Operation The inspector assessed operator readiness for startup and power operation by observation of crew performance in the simulator and review of training materials to support plant startup and power operation. NRC inspectors had previously evaluated the licensed operator requalification training (LORT) program during the week of July 17, 1995 (Inspection Report No. 50-309/95-18) and concluded that the LORT program was satisfactory.
The inspector observed two operating crews in the simulator handling various equipment malfunctions and plant events. Each crew was evaluated on one simulator scenario, which ran for about one hour. The inspector also observed and evaluated the Maine Yankee assessments of crew performance.
The inspector concluded that operator understanding of plant and system response and interpretation of alarms was good. Crew diagnostic skills were good. The licensed senior reactor operators (SR0s) demonstrated an acceptable understanding of technical specifications. Licensed reactor operators (R0s) were proficient in locating and manipulating controls and instrumentation.
The crews generally functioned as a team and handled events well. The inspector concluded that the crews demonstrated an acceptable level of performance in the simulator and provided assurance that plant operational emergencies could be effectively addressed.
The inspector noted that communications among crew members were acceptable but were not always in accordance with Maine Yankee standards as described in procedure 1-300-6, " Operational Communications", Rev. 6, dated March 1993.
For example, operators often failed to use repeat backs and acknowledgments when directions were given. Also, there were instances where auxiliary operators (A0s) were given direction by only procedure number without defining the page and step number.
Maine Yankee evaluations of crew performance were determined to be effective overall.
The inspector reviewed recent operator training and on-shift reading materials that supported plant startup and operation. LORT block 6 (1 week for each licensed operator) conducted over the October-November time period included 1.5 days on emergency operating procedure (EOP)ded: changes and 2.5 days on plant startup. Training on plant modifications inclu
- RCP vibration monitoring upgrade e Zinc addition and monitoring system o Bus 8 load reductions e New Westinghouse (W) fuel e Steam Generator repair by sleeving all tubes
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Required reading assignments completed by licensed operators included:
1 i
l e Procedure changes i
e Study results of the effects of steam generator sleeving on plant l
behavior (nuclear and balance of plant) t e E0P review
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e Summary of differences between Combustion Engineering and H fuel e Reload core design 1
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t The inspector concluded that.no recent training given to licensed operators
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provided a sound basis for them to startup and operate in a safe and effective manner.
1.5.2 Simulator Modeling of Steam Generator Behavior after Tube 51eeving i
The inspector evaluated simulator modifications related to steam generator i
tube sleeving and the resultant decrease in primary coolant flow through the steam generators. The evaluation comprised an adaptation of NUREG 1258,
" Evaluation Procedure for Simulator Facilities Certified Under 10 CFR 55."
Simulator software modifications and associated testing wero implemented in accordance with facility procedures and ANSI /ANS 3.5-1985. The facility demonstrated an effective modification of the simulator software consistent with best-estimate engineering data and predicated reference plant j
performance.
The simulator modifications were implemented prior to the availability of reference plant actual data, which is not the normal case. ANS!/ANS 3.5-1985 j
requires the simulator to be updated only after the changes are installed in i
the reference plant. The facility recognized that the modifications will require followup evaluation when actual plant data becomes available.
followup evaluations will be accomplished as part of the routine periodic test 1
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- program, s
The simulator modifications involved a reduction in primary coolant flow through the steam generators as a result of the steam generator tube sleeving.
The facility provided engineering data predicting primary coolant flow j
reductions directly related to tube sleeving that ranged from 0.8% (best-estimate) to 4.4% (conservative case). The simulator model used 3% total 3
primary-coolant steam generator flow reduction, i
The primary coolant model did not require modification in the sense of adding or deleting flow paths or calculational nodes. The simulation methodology already considered losses associated with entry into and exit from the steam generator tubes. The effort, therefore, was of derivation of increased flow i
loss coefficients, alteration and recompliation of the computer code and testing. New coefficient values were derived off-line to effect a 3% flow 4
I nduction at full power. The changes were tested by comparison with previous
, + documented performance and applicable design basis performance documentation 4 <
4 i
using RETRAN, RELAP, and GOTHIC. Both the most recent (1994) annual simulator j
- - performance test data and transient benchmark (1991) data were used to evaluate the required changes.
3 1
(
8 Plant parameters that were specifically evaluated by the facility with respect to best-estimate predictions included:
- m <-
e plant efficiency e primary coolant hot / cold leg temperatures e primary coolant average and delta temperatures e reactor coolant pump electrical current (amps) e reactor protection system calculated power.
j Other plant parameters were compared to previous test data for consistency.*
The following transient tests were evaluated by the facility in accordance l
with Annual Simulator Test procedure 18-343-1:
e trip all feedwater pumps l
e trip all reactor coolant pumps i
e stuck open PORY with loss of all SIAS l
e main steam line break in containment e design steam generator tube rupture.
Operator training with the modified simulator model began in September 1995.
l The inspector concluded that modeling of steam generator characteristics and resultant plant behavior was realistic for the changes associated with sleeving the steam generator tubes.
1.6 Safety Systems Walkdown The inspectors conducted walkdowns of selected safety related systems to ascertain that the systems were properly aligned for operation and to verify system readiness. The inspectors checked valves positions against the l
required positions, inspected equipment condition, and verified correct level, pressure and temperature indications. The following systems were inspected:
1.6.1 Containment Throughout the inspection period, the inspectors conducted routine inspections of the reactor containment. On December 22, 1995, the end of outage containment cleanup effort was in progress and station radiation protection personnel were observed to provide good radiation controls over the removal of l
equipment from the containment. However, the inspector noted that the reactor coolant loop areas were still in need of cleaning, that several pipe lines in the loops had damaged insulation and many pieces of equipment were not seismically restrained. The inspector noted that valve RC-M-25 in loop 2 had l
a packing leak after a new type of packing had been installed. This issue is l
discussed in section 2.0 of this report.
I i
Tho inspector noted that the contamination from a radioactive spill on the -2 4
foot elevation of the reactor containment located in the tre.ich area by the containment sump was properly posted and shielded. While no significant safety related system discrepancy was noted, the inspector concluded thi.
since several cleaning activities were still needed, more containment
9 walkdowns would be needed prior to plant restart. The containment conditions was determined to be unacceptable. The inspector informed the licensee of this conclusion.
1.6.2 Emergency Diesel Generators The inspector conducted a walkdown of both emergency diesel generator (EDG) rooms during preparation for startup. No operability concerns were identified.
Some of the minor issues identified included two abnormal fuel oil annunciators, a bucket half-full of lube oil, paint drop plastic under air accumulators, old operator aid tags, and four deficiency tags. The fuel oil annunciators were for EDG 1A day tank high/ low level and EDG 1B integral tank high/ low level. Both alarms were for levels slightly above normal.
In response to the other discrepancies, maintenance personnel removed the lube oil bucket and plastic, and provided feedback on the four deficiency tags (D's). DT 95-1358, to replace EDG 18 flexible air hoses, had been volded but the tag had not been removed. The other DTs, 93-4384 (EDG 1A), 94-2697 and 95-2649 (for EDG 18) were for minor air, lube oil, and fuel oil leaks scheduled to be repaired during a future di:!sel outage. The inspector concluded that these issues were of low safety importance and that appropriate corrective actions were taken or planned.
1.6.3 Primary Component Cooling Water System The inspector conducted a field inspection of the primary component cooling (PCC) system with a walkdown of the PCC piping and components in five separate areas of the plant. The inspector used the PCC piping and instrumentation diagram (P&ID ll550-FM-94A) and the system procedure no. 1-15-1, Primary Component Cooling System, to provide the valve lineup and component configuration criteria. All the identified equipment discrepancy tags were cross-checked against the computer-based work control program to verify the current operational status of the tagged items. The inspector found all valves correctly positioned in accordance with the PCC operations procedure, as modified by some field tagging orders, and identified no major system operational concerns. However, minor configuration discrepancies (e.g.,
missing caps on some vent and drain valve piping) and other " housekeeping" issues were raised for discussion with the licensee prior to plant startup.
Also, the inspector noted that the discrepancy tag on the positioner for the temperature control valve (TCV), PCC-T-20, that bypasses the PCC heat exchangers was over two years old. While the work order remains open and the TCY has been determined to be operable, the status of such repairs, as well as the status of other discrepancies over a year old with the repair work not yet started, were deemed to represent evidence of a PCC maintenance backlog that the licensee work priorttles have not yet addressed.
Overall, the partial PCC walkdown verified system functionality with no component operability concerns or Technical Specification issues identified.
The inspector confirmed PCC flow to one emergency diesel generator (DG) and the requisite valve lineup to support DG-1A operability. Procedure 1-15-1 precautions (e.g., a continuous PCC system vent path) were spot-checked for j
compliance durt% the walkdown inspection. Other than the minor discrepancies and equipment repair status items discussed above, the inspector had no m
'J' 10
. further questions regarding the licensee's configuration control of the PCC m r m system and its readiness to support plant full power operations.
1.6.4 125 Volt DC Distribution System The inspector inspected the 125 volt de bus distribution system including battery buses 1, 2, 3 and 4 in the protective switchgear room, non-nuclear control (NNS) battery bus sections 1 and 2 in the protective cable tray room, safety room distribution cabinets, and other remote instrument. panels.
Several breakers were tagged out for work being performed and inconsistancies.
-between the distribution panels and the plant procedure tables were
-. identified.
For 125 volt de Distribution Cabinet DC/CE-2, on the control room north wall, Switch 16, labeled, "DC to Indicating Lights on HELB Valves ESF l
' Panel B," by Table 115 - Revision 6, was off. The licensee stated that this was made a spare breaker by EDCR 95-2001, dated December 14, 1995.
' Replacement Table 408 - Revision 7 was hand marked that Breaker 16 was a spare. For NNS Battery Bus Section 1125 volt de (Protected Cable Tray),
Breaker 2 was marked as " Spare" on Table 111 - Revision 3 and as "125 volt dc
+ Control Power to MCC-9C" on Table 404 - Revision 4.
Breaker 2 was off when l
reviewed. This was not consistent with the fact that MCC-9C was powered.
l Further investigation revealed that the correct supply for MCC-9C was Breaker 4.
Operation's personnel submitted a document change request (DCR No.95-272) l to correct the problem. The inspector concluded that, except for the j
discrepancies mentioned above, the 125 volt de Power System was adequately controlled by the Itcensee.
1.6.5 Chemical and Volume Control System The inspector inspected the charging sub-system of chemical and volume control system (CVCS) in the primary auxiliary building (PAB) and in the containment
- m ~
- during preparation for startup. The alignment for this system was specified.
in different attachments to Procedure 1-11-6, Chemical and Volume Control System Operation. The inspector reviewed Attachment A, CVCS Valve Alignment and Attachment D P-14A, "A" Train Charging Pump Valve Alignment. The inspector found some minor discrepancies as follows: in the PAB, valve CH-192, P-14A Suction Drain, had no valve tag and valve PAP-63 was lock-wired but not held in position. An operator working in the area corrected these two problems.
Other noted discrepancies were: valve CH-99, P-14A Discharge i
Cross-connect Header Vent, was leaking (had water bag to catch leak) and deficiency tags (DT) 93-4294 and 4295 for boron buildup on body gaskets of TR-75 and LS-4007, respectively, DT 94-3013 for pressure regulator blowing air on HV-1, DT 95-1877 for pump P-14A failure to auto start when breaker was shut, and 95-3038 for P-14S leak on inboard gland lube oil coll.
Issues identifie<1 in the containment included: valves CH-173 and 174 covered with lagging (no tags visible); CH-73 danger tagged "Do Not Operate" by Tagging Order (TO) i 2108, dated December 22, 1995; CH-A-33, shut by TO 2108; CH-60, P-7 Suction from Primary Water, lagged - no tag; could not locate the removed Manual Speed
- Control Lever; valves CH-135,141,142,143, and 144, all for FL-73, not on valve list; and, CH-198, Zinc Injection Skid Outlet Header, closed and tagged
> - Do Not Operate". The inspector discussed these problems with operations 6 >
r personnel. A procedure note allowed for not checking the high radiation area i
valves CH-173 and 174. Tagging order 2108 was for other outage work and the l
O 11 Zinc addition system was isolated pending review of the system performance at
'another plant.
The inspector also inspected the boric acid (BA) sub-system of CVCS using
' ~
Procedure 1-11-5, RCS Boron Control and Chemical Addition.
Issues identified included an unsecured BA drum lift, loose valve operating handle on BA-12 (P-6A recirculation to BAST), long duct tape between drain line and piping support, loose contamination sign by BA pumps, and three deficiency tags. The deficiencies tagged were: BA-24 stem loose - flats not true indication of valve position, (DT 95-2533) dated October 18, 1995; BD-T-12 and 22 have possible leak through (DT 95-2143 and 2144); BA-7 leak under installation, (DT 95-1206) dated May 1,1995; P-81 acid leak on the floor when pump BA-81 runs, (DT 95-290) dated January 28, 1995; and, BA-33 reset open/ shut stops and reattach handle, (DT 95-221) dated January 21, 1995. The licensee reviewed the status of DT 95-290, 2143, and 2144 and noted that DT 95-290 had been assigned a normal operating priority of 7 and will be repaired accordingly.
Valves BD-T-12 and 22 recently passed the local leak rate test and the tags will be removed. The other problems were still under review by the licensee.
The inspector was satisfied that in light of the relatively low safety significance of these discrepancies, continuing reviews by the licensee was acceptable and that the system operability was not impacted. The inspector concluded that the identified discrepancies were minor and once identified, adequately curiacted er being addressed and that the licensee's alignment of the system was adequate.
1.7 Reactor Coolant Pump (RCP) No.2 Diagnostics On December 20, 1995, the inspectors observed Operations and Engineering personnel conduct a series of reactor coolant pump (RCP) No.2 runs in an attempt to determine the reason for the high pump vibration. Earlier, during the initial startup of the pump, the licensee had observed significant vibration indications and subsequently shut down the pump.
In order to determine the extent of the problem and the accuracy of the new vibration monitoring equipment, the licensee decided to restart the pump and closely monitor the control room indications and use local vibration monitoring equipment set up at the pump. The control room personnel were in constant communication with the local pump personnel to verify the accuracy of the new system. The pump start was to be performed using existing plant procedures, however overall trouble shooting was to be accomplished in accordance with a flow chart developed by the licensee.
The inspector noted that reactor operators and control personnel involved in this evolution were briefed extensively on the planned flow chart, however the precautions and limitations normally associated with a planned evolution were not present in the plan.
There was no formal documented process for the troubleshooting effort.
For example, all applicable test precautions, limitations and termination criteria (except for vibration " Danger" alarm)
~ were not clearly listed specifically for the diagnostics. The licensee indicated that the precautions were part of several procedures being used and 1q -
that.the sequence of events was listed on paper to facilitate troubleshooting activities.
For example, RCP operation was in accordance with procedure 1 7. Reactor Coolant Pump Operation and two or more pump operation was in W.
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. J
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~
12 accordance with procedure 1-1, Plant Heatup. Operators indicated that they knew all the limitations and by using the applicable approved procedures, they were appropriately covered.
In fact, all concerns expressed by the inspector were properly covered in approved plant procedures that operators were using to conduct pump runs.
At approximately ninety seconds after starting the pump, a "RCP VIBRATION DANGER" alarm annunciated on window L4-4 in the control room. The engineer monitoring the vibration panel noted that the alarm was due to a phase imbalance for which a baseline number had not been established. The engineer-comunicated that the pump did not need to be tripped because the amplitude of the vibration had not exceeded the danger limit. However, the senior reactor operator instructed the reactor operator to trip the pump as pre-agreed upon i
and discussed during the pre-briefing. The inspector considered the senior reactor operator's quick decisive action to put the plant in a safe and predetermined condition to be a strength. This was indicative of a very safety oriented and focused operations personnel. According to the licensee, the input from the phase imbalance to the danger alarm should have been disabled until the proper magnitude was established. Within an hour, the licensee disabled the phase imbalance and several other inputs to the danger alarm. After further testing and reviews, Maine Yankee determined that maintenance diagnostics would be required. The plant was cooled down and reactor coolant loop 2 was isolated and drained for RCP No.2 inspection. At
'ke end of the inspection period, this diagnostic activity was still ongoing.
2.0 MAINTENANCE 4
The inspectors ascertained that activities were performed safely and in accordance with approved plant procedures. On November 16, 1995, the inspectors met with managers of the maintenance department to discuss how previously identified issues were being addressed and corrected and the future initiatives planned by the department. The meeting also provided an opportunity for the inspectors to assess management awareness of and involvement with maintenance issues.
Overall, natntenance and surveillance activities continue to be performed well. The inspectors ascertained that activities were performed safety and in accordance with approved plant procedures.
2.1 Maintenance Observation The inspectors cbserved and reviewed selected maintenance activities to assure that the activities were conducted safely; complied with technical specifications and work order (WO) requirements; that required approvals and releases were obtained prior to commencing work; that the work procedures were appropriately detailed and followed; and that equipment was properly tested
13 and returned to service. The inspectors observed portions of the following
-, emaintenance activities:
e WO 95-2715, inspect CEAs GN, E6 and D3 e
WO 95-3177. RC-H 25 Determ MOV and Retest e
WO 94-02871, Check S/G Ho.3 Hanway Studs Torque
.e-WO 95-3138-00, Reactor Coolant Pump No. 2 Uncouple and Inspect e
WO 95-3143-01, Perform Test Run of No. 2 RCP e
WO 95-3138-04, Uncoupled Motor Run (RCP-2) e WO 95-2029, Diesel Fire Pump (P-5) Replace Inboard / Outboard Pump Bearings i
2.1.1 Control Element Assembles (CEAs) GN, E6, and D3 Inspection Work order number No. 95-02715 was initiated to control work activities involved with moving three CEAs from the core to the spent fuel pool for inspection. During fuel loading, these three CEAs had been mistakenly lifted out of their fuel bundles from a couple of inches to about four feet when the spreader in the refueling machine had interacted with the CEA spider / hub areas. The inspector reviewed the work order prior to the work being performed and noted that in combination with the generated " move sheets," it properly established the procets and method for the inspection. The inspections were accomplished safely, no damages were noted and the CEAs were returned to their proper locations (016, 54 and R15 respectively) in the core.
2.1.2 Overtorqued Steam Generator No.3 Primary Manway Studs During the installation of the cold leg primary manway in steam generator (S/G) Number 3, the licensee identified that the stud nuts had potentially been overtorqued during the cold torquing. This was identified after the
. torquing had been completed, when a discrepancy with the torque wrench was identified. The torque wrench had failed when it was being used on another manway. Using work order (WO) No. 94-02871-01, maintenance personnel checked the torque on the installed nuts on the No.3 S/G manway and found that five of the nuts had been torqued to values in excess of the specified 1200 f t-Lbs Stud numbers 9,13,17,18 and 19 had torque values of-about 1,700 Ft-tbs each. An engineering technical evaluation (TE Ho. 347-95) was generated to address this issue.
It was determined that torquing the studs to 1,800 ft.
Lbs would not cause their allowable limits to be exceeded, thereby not degrading the studs. Plans were implemented to restore the stud nuts to their appropriate torque valves. The potential deformation of the manway gasket material was also reviewed. The gasket material was designed to sit in a recessed area of the primary manway thereby leaving a metal to metal (manway to manway cover) contact for the sealing area. With this design, the excessive torque on the manway studs could not cause any excessive compression of the gasket material. The inspector reviewed the licensee's determination regardir,g the gasket and also reviewed drawing SM-94-03, revision 1 to verify
. that the engineering determination was in accordance with the tolerances in the design of the manway and gasket installation. When the inspector
~ ~' questioned the adequacy of the metal to metal interaction between the manway and manway cover and the absence of any documentation indicating that the licensee had evaluated this configuration, the licensee revised TE 347-95 to W
yp. [
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14
. address the issue. The inspector reviewed the evaluation and found that it w W 4 properly addressed the issue.
- - c wihe inspector was satisfied that Maine Yankee had properly dealt with this issue. No discrepancy was identified with the engineering analysis performed.
,aThe studs were properly " cold" torqued and the " hot" torque was scheduled to
.be performed during plant heat up.
2.1.3 Reactor Coolant Pump No.2 Repairs 4
-On December 10, 1995, Maine Yankee operation's personnel conducted a test run of No.2 Reactor Coolant Pump (RCP) with the motor coupled to the pump. The vibration readings were at about -30 mils which was above the maximum limit of -
1 25 mils. The pump was secured for inspection. Plant engineering personnel determhed from analysis of the vibration data that the source of the l
excessive vibration appeared to be in the pump coupling. The plant shift e supervisor initiated a maintenance work order to uncouple and inspect the pump 1
and motor.
f
)-
On December 22, the inspector observed Maine Yankee maintenance personnel, l
with assistance from Yankee Atomic engineers, uncouple No. 2 RCP to determine i
the source of excessive vibration. The repair effort was accomplished in accordance with station work order (WO) number 95-3138-00 and the appitcable-portions of station procedure 5-13-2, Removal of a Model N-9000 R.ractor Coolant Pump Seal Cartridge. The licensee identified several coup 1tng bolts i
with improperly fabricated threads which appeared galled. This thread galling l
appeared to have prevented the bolts from being properly torqued and
-subsequently led to improper mating of the coupling surfaces. These bolts were supplied by the pump vendor as part of a package of repair equipment that replaced the pump rotating assembly, pump bearing, spool piece and pump to motor coupling. Plant engineering department personnel determined that the < -
original coupling bolts RCP No. I were still available and after inspection i
these were reinstalled in the No. 2 RCP coupling. The pump coupling was reassembled in accordance with station procedure 5-16-2, Installation of a j
j Model N-9000 Reactor Coolant Pump Seal Cartridge in a RCP. At the close of i
the inspection period Maine Yankee was preparing to conduct test runs on the pump to determine if the repairs were successful.
The inspector determined that Maine Yankee Personnr1 had performed well at analyzing and repairing the complex equipment in a safe manner. There was excellent supervisory oversight and good support from the engineering department.
2.1.4 Diesel Driven Fire Pump (P-5) Repairs During a weekly surveillance test of the diesel driven fire pump, plant personnel observed vibration levels to be increasing. A work order (WO No.
m 95-2029) was developed to investigate and repair the problem which aproared to be in the inboard pump bearing. After disassembly, the inboard and outboard
~. - bearing were found to be worn on the outer race. The bearings and bearing housings were replaced and the pump reassembled. The pump was repacked in
<accordance with station procedure 5-34-2 Pump Packing. The pump was later
15 tested successfully.
l The inspector noted that this fire pump appears to have had a considerable
-amount of problems in the past and expressed concern over its reliability.
The licensee indicated that actions were been taken to conduct an overall
}
review and assessment of the pump's reliability in accordance with the new j
maintenance rule being implemented.
i 2.1.5 RC-N-25 Repairs On December 22, 1995, a packing leak was identified from reactor coolant loop' 2 valve, RC-M-25. This valve is the fill valve for the loop and is normally closed during plant operation. The valve was electrically disconnected, the leakoff line removed and the valve bonnet and internals were removed to the
" Hot Shop" for inspection and repair. The valve stem was found to have a circumferential indentation approximately one and half inches up the stem from the back seat. A new stem and stem nut were reinstalled in the valve. A i
technical evaluation to disposition minor cracks in the valve wedge was developed and approved. The valve was reassembled. The leakoff line was rewelded and quality control personnel conducted final visual examination of the welds in accordance with station procedure YA-VT-12.
The inspector noted that Maine Yankee personnel performed the repairs on the safety related equipment well. Activities were conducted in a controlled manner with good supervisory oversight. There was good support by radiation controls personnel and proper independent verification of weld quality by the 4
station quality department.
4 2.2 Surveillance Observations i
i
+The inspector observed and reviewed selected surveillance activities'to assure that the activities satisfied technical specification requirements; that l'
personnel adhered to administrative and surveillance procedures; that test instruments were calibrated; and that test results satisfied the acceptance criteria and when they did not, that the licensee took appropriate actions.
The inspectors observed portions of the folicwing:
2.2.1 Safety Injection Actuation with Loss of A/C Test On December 3,1995, the inspectors observed testing activities of the engineered safety features (ESF) systems required by technical specifications 4.1 and 4.6.
Two system tests were observed. The first one was conducted i
with ;.rocedure 3.1.14A, "A" Train EDG/ECCS Cold Shutdown Test. The intent was to test the emergency diesel generator (EDG) automatic start, output breaker closure, and emergency Bus 5 load shedding and loading function. The test also verified operation of all automatic ECCS "A" train valves in conjunction with actuation signal testing of SIAS "A" ahd CSAS "A".
The test was
.. performed concurrently with procedure 3.1.15.1A, Containment isolation Train "A" Surveillance Test. This test verified proper operation of containment
-isolation valves in response to a high containment pressure initiating a containment isolation signal. The inspector verified correct alignment of plant components (pumps, valves, breakers, switches, tank levels (refueling
- i j.l,+.'
j
- l.
t 16 4
i i
and demineralized water storage tanks)) in the control room, EDG room, l
we.ar:* emergency switchgear room, containment spray building, and the primary auxiliary building. The inspector also inspected test instrument hook up in i
a--
the control room and at the 4160 volts circuit breakers. The inspector l
identified some minor discrepancies.that were also identifb d by the licensee, i
j
'The licensee initiated timely and well thought actions (such as procedure j
improvement) to rectify the discrepancies.
l Prior to conduct of' the tests, the inspector attended the licensee's briefing in the control room. The briefing was good and detailed. The applicable test controls were clearly reiterated. The test process was discussed and made i
clear to all personnel. The on shift designated manager and the test i
directors did an excellent work of conducting the briefing. The following
[
plant initial conditions were noted: RWST Tank level at 325,000 gallons; j
Pressurizer level at 35%; and CET L-20 reading 114 *F.
2.2.2 ECCS Operational Test Recirculation Actuation System I
The inspector observed portions of test activities to verify the recirculation j
actuation system (RAS) portion of the engineered safety feature systems. The test was conducted in accordance with procedure 3-1-15.2, ECCS Operational a
j Test Recirculation Actuation System. The test was performed well and test personnel showed good knowledge and capability. No discrepancy was noted.
I 2.2.3 Results Review, ECCS Operational Pump and Flow Check Valve Testing The inspector reviewed the results of the Emergency Core Cooling System (ECCS) pumps and check valves 1995 refueling testing. The test was performed in accordance M ch procedure 3.1.15.3, ECCS Operational Pump and Flow Check Valve Testing, to satisfy the requirements of technical specifications 4.6, thereby demonstrating that the pumps would perform as designed in an emergency situation. The test acceptance criteria were fully met. The high pressure safety injection (HPSI), low pressure safety injection (LPSI), and containment spray (CS) pumps delivered their required flows at the test pressures. The inspector noted that a temporary procedure change (TPC 95-343) had been incorporated into the procedure to substitute a test instrument for a permanent plant instrument for measuring flow during the test. However, no provisions had been included in the procedure for recording the test instrument number. While this oversight had no relative safety significance, recording instrument numbers in the procedure would facilitate instrument traceability to the procedure in the future. This minor oversight was acknowledged by the itcensee's operations staff who indicated that it would be corrected. No other discrepancies were identified.
3.0 ENGINEERING The engineering department continued to provide good support to the plant.
- Good safety perspective was evident. Corporate engineering department personnel demonstrated excellent technical capabilities and management w.c. oversight with the steam generator sleeving activities. These activities were
't completed safely.
Plant engineering department personnel continued to provide good support to the plant in addressing engineering issues such as with the
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'Iu secondary component cooling system valves; and the demineralized water storage
' tank.
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3.1 Steam Generator Tube Sleeving Completion i
Steam generator sleeving activities were completed in December. The total
' number of tubes sleeved and available for service in each steam generator (S/G) were 5,467 in S/G Ho. 1, 5, 492 in 2, and 5,509 in 3.
The licensee had i
projected 5,373 in each generator. While it appeared that the sleeving l
campaign had been successful, the effectiveness would not be determined until' i
[
the reactor coolant system is operational during plant heatup.
i j
As documented in NRC inspection report number 50-309/95-22, the licensee's programmatic controls and Quality Programs Department (QPD) ject were evaluated verification j
activities related to the steam generator tube sleeving pro 4
i i
while in progress. During the current inspection, the inspector reviewed updated process controls for laser "rewold" options, the justification for j
io eliminating the separate heat treatment of the upper hydraulic expansion (HE) area and the repaired tube pressure test criteria. The inspector also J
i examined a sample of completed non-conformance reports (NR) and anomaly I
tracking sheets (ATS) and checked the status of Westinghouse (W) licensee'sopen items awaiting final resolution. Finally, the inspector assessed the j
i i
plan for the accountability of the completed process activities and the 4
J acceptability of repairs in all three S/Gs. The opea technical items were confirmed to represent issues involving documentatie': or future inspection i
plans with no impact upon plant startup. The recoros of final testing and 1
1-inspections were reviewed to verify project closecut in accordance with work -
order controls and independent QPD checks and sampling activities.
4 With respect to the classification of "rewelds." the inspector reviewed Revision 2 of the welding procedure specification (WPS 74374) authorizing upu i
- " ~
to five laser rewelds without the need for an outboard sleeve reweld option.
j The r'Isposition of " double-weld" NRs was checked against the established i
. welding acceptance criteria and'for heat treatment considerations. A sample of H final status reports was also examined to verify that heat treatment of the upper HE area was performed, where the dimensional tolerances for the inclusion of the HE in the laser weld heat treatment could not be met. The inspector verified the controlled sequence of heat treatment activities where j
multiple appiteations were necessary.
4 For the NR and ATS sample review, the inspector assessed the licensee handling i
and disposition of cases of dual and mis-positioned welds, heat treatment i
i issues, and sleew/ tube sheet hard-roll torque problems. The inspector 4
discussed with cognizant licensee engineers the status of three NRs (95-020,95-056, & 95-061), which required engineering analyses to resolve the identified concerns. The relevant M " design and analysis" reports were reviewed for consistency with the proposed final NR recommendations.
In one v
.
- we case (NR 95-061). the licensee concurred with an inspector observation that a j
documented ASNT Level III assessment of the flaw detection capabilities of the k
reddy current testing (ECT) using a motorized rotating pancake coil would provide additional support to the NR disposition.
In another case (NR 95-056), the inspector checked with plant personnel that the required actions 1
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.(e.g., RCS chemistry sampling activities) were conducted and were being j
%e u apprt.ortately evaluated. The inspector also confirmed that the licensee was i
V track),79 these NRs, as well as other open technical issues, on a S/G project
- M '~ status report with specific action assignments and resolution references.
a l
E 3.1.1 Completed Non-Conformance Reports As part of the sleeving project closecut inspection, the inspector reviewed i
1 the heat treatment prncess, critical characteristics, and associated in-process documents, including non-conformance reports (NCR). Critical i
?
j 1
characteristics associated with heat treatment were being monitored, recorded, and independently verified to be within acceptable ranges. All NCRs associated with heat treatment had been addressed and closed out at the time of this inspection. The inspector reviewed three NCRs in detail (NCR MY !
038,058,059) for technical adequacy and verified that the recommended i
dispositions, additional heat treatments and baseline eddy current testing 1
-(ECT), were performed.
In addition, numerous sleeves required rewelding and
.y additional heat treatments. The inspector randomly selected several of these F
" tubes, and independently verified that the additional heat treatments were completed.
I 3.1.2 Quality Program Department Verification Activities
.l The inspector reviewed the initial Quality Program Department (QPD) audit of Westinghouse sleeving activities. The inspector vertfled that all findings and observations, except one, had been addressed and closed. MY indicated that the one remaining open item would become part of the startup document checklist to assure that it was addressed prior to startup. The inspector noted a good initiative by QPD in the use of statistical analysis to monitor various sleeving processes which enhanced various process reliabilities. The inspector selected three QPD closed "Open items" (01-95-094, 145, 146) associated with the sleeving project for review. The inspector reviewed the recommendations and verified that enrrective actions were impicmented.
Overall, QPD involvement and oversQht of the sleeving project was good.
In assessing the final S/G tube sleevmg project closecut process, the inspector reviewed several closed work order documents (e.g., 95-01031 for S/G of 1 tube repairs), verifying satisfactory nondestructive examination (NDE) leak the completed welds, acceptability of the low pressure (secondary side) testing, and QPD involvement in the reconciliation of the status and records for all plugged tubes..The licensee clso conducted an independent review of the process controls and quality records for thirteen randomly selected sleeves per S/G, as well as a sample of 24 plugged tubes, to confirm adequate material traceability, consistent application of the individual processes, and satisfactory resolution of the applicable NR/ATS reports. As a result of this independent review, all processes were found to be properly completed with no deficiencies or open items identified. The inspector reviewed the resulting issued to
,.~-~ report and an additional licensee non-conformance report (95N-024)losure of control the acceptability of the completed work, restoration and c a un <
the S/Gs, and turnover of the repaired components as operable equipment. QPD involvement in the final oversight and independent project review activities was found to be comprehensive, providing good backup to the extensive in-p i.
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T 19 process surveillance controls that were in evidence.
Finally, the inspector noted that while ultrasonic testing (UT) served as the j
~ e-code examination for the acceptance of the laser welding of the S/G tube i
sleeves, a baseline ECT of the tubes was also required.
In addition to the confirmation of the tube and sleeve quality, this ECT also provided another check of the M process controls; as any anomalies (e.g., mispositioned welds, improper heat treatment) would be identified by the ECT acceptance criteria.
- Thus, the final tube sleeving NDE, supplemented by the code-required leakage test at nominal operating pressure, not only provided evidence of the repaired S/G tube integrity, but also complemented the M quality assurance and QPD i
independent oversight activities in establishing the requisite accountability for the sleeving project closeout. The inspector verified upper licensee management involvement in the S/G closecut program controls (e.g., NCR 95N-024) and identified no unresolved issues or safety concerns related to the 5
completion of the S/G tube sleeving project.
3.1.3 Eddy Current Testing Review 2
As previously stated, baseline eddy current testing (ECT) was performed on all S/G tube sleeves. ECT was utilized to confirm tube and sleeve quality, and provided another check of H process controls. The inspector reviewed the eddy current data analysis guidelines, independently analyzed a sampilng of eddy current data for S/G tube sleeves, and reviewed MY's informal response to NRC staff questions regarding the S/G inspection program.
In the review of the ECT analysis guidelines, the inspector noted the following:
- The analysis guidelines did not provide clear guidance on the expectations for analysis of the straight portion (non-welded, non-expanded) of the sleeve. Additional discussions did indicated that this-
-C
- section of sleeve was adequately reviewed during the baseline ECT.
Overall, no deficiencies were identified which would invalidate the current ECT analysis results. Eight sleeved tubes were randomly selected, and independent ECT analysis was performed by the inspector. The inspector's final analysis supported the W analysis. MY's informal response to NRC staff questions regarding the S/G inspection program was reviewed by the inspector.
No concerns were identified that would affect plant startup.
Following completion of sleeving activities, each steam generator was subjected to several post repair tests including a tube leak test. This test was performed by filling the secondary side of the steam generator with water and maintaining a test pressure of about 165 psig while observing for any leakage out of the tubes under the hot and cold leg tube sheet areas. The inspector observed portions of test activities on steam generator 2.
Test procedure 1-104-14.2, S/G 2 in Wet Lay Up and S/G 2 Low Pressure Tube Leak Test was used for the test. The procedure specified a maximum test pressure
- of 230 psig.- The. specific section of the procedure used for conducting the test was section 6.3, Low Pressure Tube Leak Test. A pressure of 165 psig was
-attained. The low pressure tube leak test was part of the post repair tests prescribed for the sleeving activity and was specified in work order 95-01032.
The inspector concluded that the sleeving campaign had been accompitshed by
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20 M yery technically competent Mngineers. The key stages of the campafg (sleeve
%@ dinsertion, expansion, laser elding, hard rolling, heat treatment, u$trasonic
'3 1r testing, eddy current testing and leak testing) were accomplished
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- Li dsatisfactorily without any significant outstanding issues. This effort was.
j
, noted as a significant strength in the engineering organization. There was o
good support from maintenance and quality programs departments, j
-3.2 Engineering Design Change Package and Records Review f
'The inspector reviewed the following Engineering Design Change Requests (EDCRs) and associated design basis packages, to include the applicable
)
'A engineering change notices (ECNs), the 10 CFR 50.59 determinations, and_any-t
- relevant vendor information:
i.)
3.2.1 EDCR 95-41, Reactor Coolant Pump Differential Pressure Instrumentation l
Upgrade - Reactor Coolant System flow measurement.
For EDCR 95-41, the inspector also checked the completed work order records i
for evidence of welding process controls, proper NDE and hydrostatic testing of the modified' reactor coolant pressure boundary, and component calibration of the new differential pressure transmitters. The inspector reviewed the QPD s'
records for the receipt inspection of the pre-established " critical characteristics" for the Rosemount transmitters. Based upon identified weight i
differences in the supplied transmitters, because of the special stainless steel versus normal aluminum housings, the inspector requested and received-from the Yankee Nuclear Services Division the as-built review of EDCR 95-41 (Calculation No. MYC-1792, Revision 1). The inspector reviewed this calculation set to check the pipe stress analyses of the tubing runs, the changes to the tubing supports, and the transmitter rack connections,
}
including consideration of the heavier housings.
In accordance with the EDCR.
I installition requirements, field walkdowns were confirmed to have been I
conducted by seismic capability engineers to ensure acceptable configuration support and mounting dotails.
' The inspector verified the completion of all EDCR 95-41 open items, but raised the following issues to be addressed by the licensee engineering staff:
f The inspector questioned the EDCR calculations for the radiation induced f
drift errors of the Rosemount transritters. Because of the stainless steel (SS) housings specific to the transmitters supplied to Maine Yankee, the tost results of different models were used in the EDCR e
calculations. On December 21, 1995, the licensee documented a "TELECON"
't with the Rosemount Principal Application Engineer concurring with the EDCR calculational assumptions, based upon the use of the SS housings.
j As a result of the 10 CFR 50.59 review, the inspector noted that the Plant Heatup procedure., no.1-1, had not been revised, as spectfled in the EDCR package, to ensure the proper instrument valve alignments for o
bypass valves. Mis-positioning of the bypass valves can adversely
< affect a reactor protection system signal provided by the S/G channel "A" low flow transmitter. The cognizant licensee design engineer explained that the correct bypass valve alignment is accomplished by
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21 means of a supporting procedure to " Plant Heatup" and that the correct m e
- reference in the 10 CFR 50.59 evaluation would be added accordingly by an ECN.
The inspector discussed with the licensee staff the adequacy of Itcensed operator training relative to this EDCR, which adds some computer alarms and enhances the data for RCS flow rates available in the control room.
Since this design change was implemented to support reactor engineering functions, not normal operations, the operator training package provides more of an overview than specific operational impact details. However, based upon the " failure modes and effects" (FME) analysis, the inspector
-}
noted that a postulated leak on the newly installed low-pressure side t
instrument tubing would send conflicting data to the control room; i.e.,
i high RCS flow across the S/G delta-P transmitter and low RCS flow across the new reactor coolant pump delta-P instrumentation in the same loop.
The inspector indicated that such FME information, while not intended f
for direct operator use, would be appropriate to highlight in enhanced training to the operations shifts.
Overall, the inspector determined that the EDCR 95-41 design and installation I
controls were adequately implemented, with good records of QPD activities, as-
+ built walkdowns, and functional verification and testing. The licensee took-appropriate actions on the issues raised by the inspector; and no further questions or unresolved safety concerns were identified.
f 3.2.2 EDCR 95-45, Containment Spray System Pressure Locking Modifications With respect to EDCR 95-45, which had received previous NRC inspection as documented in inspection report 50-309/95-22, the inspector evaluated the
. design details against the provisions of 10 CFR 50, Apper: dix A. GDC 56, as
". - allowed on "some other defined basis" for containment isolation adequacy; and:
also against the FSAR section 5.1.2.1 criteria for Class IV piping penetrating the containment. The inspector reviewed the revised P&lD, 11550-FM-92A, to check the consistency of the component configuration with the EDCR details.
In this regard, the inspector noted that two newly installed drain valves, CS-137 & 138, while listed as locked closed containment isolation (LCCI) valves in the appitcable procedure, were not so indicated as LCCI valves on drawing number ll550-FM-92A. Furthermore, an EDCR 95-45 enclosure illustrating the valve manufacturer (Anchor / Darling) details for bonnet venting differed from the final modifications, in that the bypass line was connected to the downstream (containment) side of the valve (versus the Anchor / Darling upstream
. recommendations) for containment isolation reasons. These inconsistencies.
while not affecting the adequacy or functionality of the valve pressure locking modifications, were discussed with cognizant design engineers, who indicated that an ECN would be issued to clarify the design detail acceptability, s m.-Additionally.sthe inspector reviewed the EDCR 95-45 work order records and functional test instructions. Welding, NDE, pressure testing, and stress
- analyses were verified to be in compitance with code (ANS!/ASME B31.1)
-provisions. Valve Type "C" leak testing was also evaluated in line with
, criteria delineated in 10 CFR 50, Appendix J.
While the inspector noted that m
t 22 the modified containment spray isolation valves were acceptably Appendix J
- ~ w leak tested as a unit, the functional test instructions did not provide for-pressure testing of each individual modification components i.e., drain m yalves, CS-137 & 138, were not leak tested. This was dis assed with a
- licensee engineer who indicated that the need for such functional testing would be further evaluated. Given the licensee's designatica of valves CS-137.
& 138 as locked closed and capped containment isolation valves, the " leak tight" function of the valves is unclear, since a leak-tested check valve is positioned between each drain valve and the containment boundary.
In any case, the proposed ECN for EDCR 95-45 should also address the rationale for not conducting a functional leak test of the bypass line drain valves associated with this modification.
Otherwise, the inspector had no further questions regarding the implementation of EDCR 95-45. The modification details adequately addressed the pressure locking concerns documented in NRC Generic Letter 95-07, which were identified by the licenses to potentially affect the containment spray system isolation
_ valves that were modified by this design change.
3.3 Secondary Component Cooling Water Valves SCC-A-460 & 461 The inspectors reviewed engineering actions to resolve the problems with the -
secondary component cooling (SCC) water valves, SCC-A-460 and SCC-A-461. The problems with these contromatics butterfly valves had first been identified in March, 1995, when the valves had failed their seat leakage tests. During subsequent repair activities, the licensee identified that the disc in SCC-A-l 461 was mis-oriented 180 degrees. A licensee event report (LER 95-006) was submitted to the NRC as required. Later on, in September,1995, after the valves failed another leakage test, they were disassembled for inspection.
j This inspection revealed that the seat ring in SCC-A-461 was not properly in
.a place. While the ring in SCC-A-460 appeared to be properly in place, an j
increase in leakage had occurred since the last time it was re-assembled.
Further tests after re-assemble still indicated excessive seat leakages. The valves were then sent to a research laboratory for a more detailed test to try j
to identify the cause of the problem.
i Maine Yankee was finally able to determine that the root cause of the problem was the valve disc interference with the seat ring during valve travel. The licensee had initially thought that the seat ring material was in question and j
had substituted rubber material rings for the original teflon material ones, i
When the problem continued, the licensee was able to identify that the disc interference was the root cause.
In conjunction with the valve manufacturer, Anchor Darling (A/D), the licensee machined the valves' disc edges such that no damages were inflicted on the seat rings during valve travel. Proper engineering analysis was in place to address the impact of this modification on the valve design.
Basically, the design of the valve was unaffected by the modification. The rubber material rings were replaced with the original
~ material teflon types.
-Following the final installatten in November,1995, the valves were successfully leak tested. The testing involved several strokes of the valves and an extended closure period. The licensee concluded that the problems with
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23
' these contromatics butterfly valves had been resolved. The inspector reviewed 4 the licensee's actions and discussed them in detail with engineering personnel.
It appeared that the actions taken were appropriate.
Plant engi.aring department personnel showed good technical capabilities at
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resolving this issue. The inspector expressed no other concern.
3.4 Dominera11 zed Water Storage Tank Leakage
. Earlier in 1995, licensee personnel had observed some water dripping from the demineralized water storage tank (DWST) tell tale pipes indicating leakage from the tank. The tank was opened for inspection and the results showed localized corrosion areas. Repairs were implemented under work order 95-937 to floor and floor to wall joint areas. The repairs also included a plate i
installation to cover floor to wall areas. Following the completion of these j
repairs, however, the tell tale leakage was again noticed. Subsequent i
. inspection revealed some cracks in the plate weld and some corrosion above the l
plate on the weld.
Further repairs were implemented undar work order 95-2703-
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01 in November, 1995.
But after the tank was closed and filled, the tell j
% tales were still dripping. The licensee again opened the tank for a 100%
i floor and wall areas inspection and found no defects. The tank was then filled and about 135 ml of rhodamine injected into the tank to conduct further aleakage testing and also to attempt to determine the source of the tell tale -
leakage. No indication of tank contents were observed in the tell tale leakage. The licensee monitored the tank volume and analyzed the tell tale water and saw no indication of tank leakage.
It was then deduced that the leakage from the tell tales was from the area between the tank and the missile shield around the tank. This area could drain into the sand bed beneath the tank and leak through tell tale pipes. Maine Yankee plant engineering department personnel initiated a technical evaluation (TE 325-95) to address this problem. Their efforts were well documented and appeared comprehensive.7 j
3.5 Reactor Coolant Pump Vibration Monitoring System tipgrade
-The inspector reviewed engineering design change request (EDCR 94-47) package used for implementing the upgrade to the reactor coolant pump vibration monitoring system and observed control room implementation and usage of the l
system. The package was a major modification to upgrade the monitoring i
capabilities to provide effective predictive maintenance or root cause analysis of reactor coolant pump (RCP) failures. The system has been installed and the functional tests have been partially completed, j
I The inspector reviewed the engineering design change package and did not find any significant discrepancies. The Itcensee considers the RCP vibration 3
monitoring system a Non-Safety related System (NSS) providing alarms and diagnostic functions only. A 10 CFR 50.59 determination by the licensee concluded that no unreviewed safety questions were involved. Although the system is NSS and non-seismic, all components were seismically mounted to
" m preventsconcerns with having non-seismic equipment over seismic category I equipment. The inspector completed a brief walkdown of the system verifying u 6therseismic restraints. The inspector noted from the control room observations and the review of the EOCR that some issues such as the following still needed to be evaluated: the completion and effectiveness of functional l
24
, _ ; tests for the system; the adequacy of the quality assurance review of the M4" system (especially the computer program review); and the apparent lack of c. guidance on when and how the system will be baselined.
In addition, the
~ licensee was still to determine the effects of temperature and low vibration magnitude on baselining. The inspector determined that the new system had the capability of enhancing the licensee's ability to improve root cause analysis.
j The inspector concluded that the licensee did a good job of reviewing safety system impact.
3.6 Engineering Backlog The inspector reviewed the engineering backlogs for the current outage, for cycle 15 capital program as of January 1,1995, for the next operating period, for cycle 15 capital expenditure timing projection, and for the future, long term capital program cycle 16 project ranking. The inspector discussed the j
setting of priorities with engineering management. The current outage i
projects not completed or not scheduled for completion before startup were captured in the next operating period listing. The future listing included all known requests for work. Priorities are set by the Maine Yankee Special Project Review Committee, made up of four senior managers. Each manager assigns a priority for all major projects. The projects are ranked in order.
of importance, and the discussion made as to what projects will be funded for-every outage and/or operating cycle.
Priority 1 projects are defined as those, " required to ensure that a structure, system or component described in.
the Technical Spc:ifications or the Final Safety Analysis Report is capable of performing its intended safety function." The inspector concluded that the licensee was adequately controlling the engineering backlog and that safety issues received the highest priorities.
4.0 PLAh7 SUPPORT Plant support activities in the areas of radiological controls, security, and emergency preparedness were conducted safely during this period. The inspectors monitored work practices, and conformance to reautrements and procedures.
1 4.1 Radiological Controls Inspectors routinely reviewed radiological controls including Organization and Management, external radiation exposure control and contamination control.
The inspectors also monitored standard industry radiological work practices, and conformance to radiological control procedures and 10 CFR 20 requirements.
4.2 Security The inspectors verified that security conditions met regulatory requirements, the requirements of the physical security plan, and complied with approved
- a procedures. The inspectors observed security staffing; protected and vital area barriers; vehicle searches and personnel identification; access contro'.,
' S badging; and to assure that they were in accordance with requirements and that appropriate compensatory measures were used when required.
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3The: inspectors verified that the emergency response facilities were well-N
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4.4 Fire Protection i
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'4.4.1 Fire Protection System Piping Modification I
,During this inspection period Maine Yankee fire protection personnel modified '
- four fire protection system deluge valves by adding new relief valves and flow i
limiting orifices. The post modification testing verified that the relief valve was capable of operation and that the restricting orifice limited flow 4
stc prevent inadvertent deluge operation. The testing was evaluated as 3
.' successful and the valves were returned to normal operation. This modification was made to correct a previously identified problem with the i
(system. The problem was that pressure locking in the system had, at times, iprevented some solenoid valves from operating properly. Maine Yankee
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W H developed a technical evaluation to address the problem and determine a resolution for it.
It was concluded that installation of a relief valve at a
. location on the actuation piping with a restricting orifice to limit flow when the relief lifted would resolve the problem.
, The inspector determined that Maine Yankee personnel performed well at identifying and correcting the problem. There were prompt corrective actions taken to resolve a potentially serious condition which may have prevented
-portions of the Fire Protection system from operating as needed.
L4.'4.2 Fire Damper Fusible Link Surveillance The inspector reviewed Maine Yankee's control of fire damper fusible links.
These links are designed to melt at set temperatures and cause fire dampers to close to prevent the spread of fire. Maine Yankee has fusible links that have
,. melting temperatures ranging from 165 'F to 212 'F.
In early 1995, Yankee Atomic Electric Company personnel conducted an audit
.(Audit 95-10) of the fire protection system. As a result, several technical evaluations were written to evaluate current fire damper installations and to provide proper guidance to correct several identified deficiencies. During the corrective actions for these deficiencies the fire damper fusible link:.
l were verified to be correctly installed.
j 4
The inspector reviewed Maine Yankee's surveillance program for fire dampers
+
-described in station procedure 3-19-8.. Fire Damper Surveillance, and determined that the proper in field checks were being performed.
Fire dampers zwere inspected for proper installation and material condition and the fusible links were verified to be installed. However, the temperature rating was not-m.-. routinely vertfled during each refueling surveillance because Maine Yankee had detennined that< this was not necessary.. Butithe inspector was concerned that
--vno technical justification was available to support not verifying the temperature ratings.' Maine Yankee indicated that one would be available for reviews during subsequent inspections. The inspector was satisfied with the
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(W kN);hk 26 em Cr;?m ; ; licensee's actions. The inspe-tor concluded that the current program was qncomprehensive and being performed by qualified personnel, Lt j
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5.0 SAFETY ASSESSMENT / QUALITY VERIFICATION.
)
- i On November 28,-1995, the inspector met with the Acting Manager of the Quality l
- .p:
Programs Department (QPD) and other members of the department to discuss the 1
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$jgd QA Restart Quality Verification and QC planned process for closecut of the S/G pg -
' sleeving project. The inspector noted that management was knowledgeable of WQ $ ongoing issues.
It appeared that QPD was providing excellent independent f4 assessment of plant activities. A program for Quality Assurance (QA) i 69" verification of restart activities was developed and appeared to be
- ]/*,[1 comprehensive. A comprehensive list of issues and processes to be accomplished for the closecut of the S/G sleeving activities was denlooed, l
+
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On December 6, 1995, the inspectors attended an Outage Management Councti i
(OMC) meeting during which special startup tests required to verify sleeving y m.
sactivities were discussed. Expectations relative to RCS flow and temperatures
~
(Flow down about 1%; Tc up about 6.2 degrees and Th up about 0.7) were a
presented. The processes for verifying the newly installed RCS flow instrumentation (Loop DP) were also discussed. The processes were controlled!
l under sub-work order fer sleeving as a functional test. The inspectors E
i observed that management was providing an excellent oversight of outage i
+
activities. The issues were well discussed with an evident fy,'us on safety.,
L I
S.1 Plant Operation Review Committee (PORC) l t
lThe inspector attended a special PORC meeting on November 29, 1995 which was O
convened to address the issue of what appeared to be improperly heat treated
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k steam generator tube sleeves. Thirty five 30" sleeved tubes had been heat h
treated with the heater positioned between the initial weld and reweld l;
locations and subsequently had not received a second planned heat treatment at i
the hydraulic expansion area. When the error was discovered, the tubes had already been hard rolled into the tube /tubesheet.
A Westinghouse Field Change Request (FCR 044, STD-FP-1995-7382, ROSA Ili La N r i
Welded Sleeving) recommended stress reliving of the transitions and tf l
possible to conduct the heat treatment at a target temperatura of omea 1100 i
i-and 1200 degrees fahrenheit, s
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The PORC members discussed the technical issues of this change and determined e
that it was acceptable. The PORC members exhibited excellent questioning attitude during the discussion process. The inspector determined that Maine R.j-.
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Yankee properly reviewed and approved the requested change to the sleeving p acess and expressed no concerns.
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27 5.2 Nuclear Safety and Audit Review Committee (NSARC) r f @[d's
'On November 28, 1995, the resident attended a portion of a special NSARC u meeting." The special meeting was held to discuss the status of the steas Mi generator sleeving campaign; three proposed technical specification changes; jeg radiation protection issues; and the progress of the restart readiness M
assessment team. Plant personnel were available to present the various issues H d['
to the committee. The committee members demonstrated excellent safety
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perspective during their discussions of the issues.
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5.3 Corrective Action Program and ' Interim Corrective Action Program - Task
.[
Prioritization Process' M]
following an NRC inspection performed in early 1995 and documented in NRC inspection report number 50-309/95-09, Maine Yankee undertook a comprehensive 4
re-engineering of its corrective action program to address problems identified 1
in the report.
Basically, the inspectors identified that lower priority Mlwa corrective actions had often remained open for a long period of time and the
,f, corrective action lists were long. This presented a potential for repeat i
occurrences while long term corrective actions for an event were being J:
disposition. While the re-engineering was on-going, the licensee instituted 7.
an interim corrective action program to direct increased attention on issues of relatively greater importance,
- a. A The interim corrective action program constituted of a method of determining i,
the relative importance of assigned tasks and prioritization process for
- l getting tasks accomplished. The program did not modify any existing systems 3
but was simply focused on three existing processes: Unusual Occurrence Reports e
1-(UOR) & Radiological Incidence Reports (RIR); Corrective Action Requests l~
'(CAR); and Maine Yankee Task Tracking System (MYTTS).
All managers were required to inventory their commitments in the three process a.
on a quarterly basis. The licensee developed a process for working off the commitments or terminating them provided a reasonable justification could be
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provided. All justifications were signed by the responsible department managers and records of such maintained. The prioritization of the items continued to be done in accordance with procedure 0-16-1, Problem Identification and Classification.
To develop a new and comprehensive corrective action program, the Itcensee established a 7-person team (The Maine Yankee Learning Process Re-Engineering Team) to review Maine Yankee's current corrective action process and j.
developing a new end improved process.
10 November, the inspectors met with the team leader and discussed the team's charter, goals and objectives, and to evaluate the team's efforts and progrets. The inspectors noted that the teaa appeared to be well established and constituted of dedicated individuals from
@% u.the various departments.(quality progran.s. operations, technical support, j.
maintenance, and engineering). Most importantly, senior manaeement's support
.hb. " ~ of the team was endent. The team's goals and the process foi implementing g j-the team's results were clearly stated.
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A 28 The inspector reviewed the licensee's implementation of the interim process
~ % h and noted that it was being well. implemented. Proper consideration was provided to nuclear safety in determining what commitments were accomplished i
. ~ e...op. terminated. 'The Learning Process Re-Engineering Team appeared to be well organized.
5.4 Restart Readiness Programs The inspectors reviewed some of the actions taken by various departments in preparation for plant restart. This review was necessitated by the fact that the plant had been in an extended outage of almost a year in duration. Maine 2
Yankee had previously established a Restart Readiness Team to identify and track the various items that each department would be responsible for during plant restart. This team made a final presentation to the Outage Management Committee (OMC) on December 12, 1995. The inspector attended the meeting to assess the team's success and also to assess management's reception of the team's findings and their follow up resolution to any identified problems.
The team concluded that the plant and plant personnel were ready to proceed N"
with plant restart in accordance with established procedures. While the team found that the various departments had satisfactorily completed their assigned tasks for plant restart, some areas of possible enhancements were noted.
^
These areas were presented to the OMC as recomended actions during the t
meeting.
i For an independent verification process, Quality Programs Department (QPD) personnel developed a QPD Restart Readiness and Start-up Quality Verification Plan.
The plan was to verify that activities associated with restart readiness and plant start-up were conducted in a safe manner. The process involved conduct of assessments, audits, surveillance and inspections to measure the effectiveness of restart activities.
Issues identified by the group were presented to the OMC on a periodic basis.
Overall, the inspector concluded that the licensee's efforts were good at ensuring that the plant would be restarted safely from the extended shutdown.
+
The Restart Readiness Team's effort was comprehensive and well received by the Itcensee's upper management. There was good management support and resolution of identified issues. QPD personnel provided good and independent oversight of restart activities. The inspector concluded that the licensee demonstrated excellent safety perspective in this area.
5.5 Previously identified issues I
The inspectors reviewed licensee's actions to address previously identified issues such as corrective actions for violations (VIO), and resolutions of unresolved items (URI). An unresolved item is one that requires further reviews to determine if it is acceptable or a violation or deviation. The issues reviewed are discussed below:
5.5.1 (Closed). URI 95-002-01, Containment Spray Pumps Flow and NPSH During a previous NRC inspection, documented in HRC inspection report number 50-309/95-02, the inspectors identified an issue concerning the testing of the
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29 h ? containment spray (CS) pumps. The issue was that during prevfous full flow i
tr@Ntesting of the pumps,,Haine Yankee had tested the pumps in accordance with i
mt technical specification 4.6 but at inadequate design flow parameters.
In the
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- W test procedure (3.1.15.3), at 100 psi, each spray pump was required to deliver j
i a flow of at least 3,700 gallons per minute (gpm). However, the CS sumps' 1
curve indicated that at a discharge pressurs of 100 psi, a flow of a)out 5,200 gpm should be attaint.ble. Therefore, the previous test acceptance criterion j
of 3,700 gpm at 100 psi had been inadequate to demonstrate that the pumps would perform as intended.
A review of previous test data showed that the pumps' flows measured at a i
discharge pressure of 100 psi were always more than the acceptance criterlon of 3,700 gpm, but none appeared to be up to 5,200 gpm. While the results of previous tests had been satisfactory with regard to the acceptance criteria, j
the results had not demonstrated proper pump operability. Maine Yankee properly notified the NRC in accordance with 10 CFR 50.72 when this issue was i
identified. This item was left open pending completion of Maine Yankee's retest activities and completion of NRC's reviews of the test results and engineering evaluation.
In November, 1995, Maine Yankee satisfactorily
,mt retested the CS pumps to the appropriate acceptance criteria. The results of this test was reviewed by the inspector and documented in Section 2.0 of this
< report. The inspector also reviewed the engineering evaluation addressing previous test results and status of the pumps. Although being tested to inadequate criterton, it appeared that the pumps would have performed to the pump curves if test conditions and instrumentation had been properly 1
established. The inspector noted no other discrepancy and concluded that the j
CS pumps had been properly tested to demonstrate that the would perform as i
designed in an accident. This item is closed.
5.5.2 (Closed) URI 95-08-01, Demineralh'ed Water Storage Tank Tell-Tale L
Leakage i
During a previous NRC inspection that was documented in NRC inspection report number 50-309/95-08, the inspector identified leakage from the demineralized I
water storage tank (DWST)'s tell-tale pipes at a rate of about 38 drops per minute. The source and significance of this leakage had not been determined l
at that time. This issue was left unresolved pending completion of Itcensee's j
efforts to determine the source of the leakage and the relative safety significance. The source of the leakage has been determined and repaired.
Details of the Itcensee's resolution of this issue are contained in section 3.0 of this report. The safety significance was minimal since the leakage was very small and at no time was the quantity of water in the tank less than the minimum required by technical specifications for availability of the emergency feedwater system. This item is closed.
5.5.3 (closed) VIO 95-022-01, Inadvertent Removal of HPSI Pump Recirculation Line Orsfice During a previous NRC inspection that was documented in NRC inspection report m W number 50-309/95-22, the inspectors identified a violation of HRC requirement.
The violation involved failure of Maine Yankee's maintenance personnel to perform work en a safety related check valve in accordance with procedure 0-
- F, Li; 30 16-3 Work Order Process. This resulted in an orifice being mistakenly 6'MWremov,ed from the high pressure safety injection pump recirculation line.
MC' The inspectors reviewed Maine Yankee's corrective actions to address this issue. The orifice was properly reinstalled and tested. Workers were briefed on the event. Members of the engineering organization completed a field
'o verification of all safety related pumps' recirculation orifices to ensure that they were in place. This was done because part of the reason for the
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inadvertent removal of the HPS! orifice was that it was not properly differentiated from the piping. As documented in an engineering memorandum l
(WES-95-012), Maine Yankee plans to install labels on all safety related
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orifices. The inspector was satisfied that the implemented and planned j
corrective actions sufficiently addressed this issue. This item is closed.
6.0 ADMINISTRATIVE 6.1 Persons Contacted j
During this report period, inspectors conducted interviews and discussions with various licensee personnel, including plant operators, maintenance technicians and the licensee management.
6.2 Summary of Facility Activities In response to an allegation raised concerning the adequacy of the containment and the emergency core cooling systems, a NRC S-person team conducted inspection activities at the Bolton, Massachusetts office of Yankee Nuclear Services Division.
The inspection was conducted on December 11 - 19, 1995.
The results will be documented in a separate correspondence.
During the inspection period the inspectors conducted backshift inspection on November 17, December 5, 8, 11, 18, 20, 21, 22, and 29, 1995, and deep backshift inspection on November 19, and 23, and December 3, 14 and 17, 1995.
6.3 Interface with the State of Maine Periodically, the resident inspectors and the onsite representetive of the State of Maine discussed findings and activities of their corresponding organizations.
6.4 Exit Meeting Inspectors periodically held meetings with senior facility management to discuss the inspection scope and findings. At the conclusion of the inspection, the inspectors also presented a summary of findings for the report period.
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U. S. NUCLEAR RECULATORY Com !SSION l',J, REGION !
I kb' DOCKET /REPORTNOS:
50-309/95-26 LICENSEE:
. Maine Yankee Atomic Power Company I
., Brunswick, Maine 04011 I aine Yankee Nuclear Generating Station' M
FACILITY:
LOCATED AT:
Wiscasset, Maine 4
i INSPECTION DATES:
December 11-14, 1995
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INSPECTOR:
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8 n 4L R. C. Ragland] Jr., R&dfation Specialist Date' Radiation Saf'ety BraDCh 4
Division of Rear TSafity
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Joh R. White, Chief V
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lation Safety Branch (vision of Reactor Safety T.
~ Areas Insoected: The inspection was performed to review the material condition of radioactive waste processing systems, occupational radiation exposure for outage activities, and to perform a followup inspection to evaluate the licensee's performance witi respect to the violations and weaknesses identified in NRC Inspection Report No. 50-309/95-06.
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1.0 PURPOSE y,
i s @@v' An announced radiological safety inspection was perfor t
M Huclear Generating Station from December 11-14, 1995. The inspection was
. performed to review the material condition of radioactive waste processing i
,M;y systems, occupational radiation exposure for outage activities, and to perform
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$5, a followup inspection to evaluate the licensee's performance with respect to
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the violations and weaknesses identified in NRC Inspection Report No.
50-309/95-06.
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2.0 REVIEW OF RADWASTE SYSTEMS FOR MATERIAL CONDITION The inspector performed tours of radwaste systems primarily to evaluate i
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material conditions. Areas inspected included the demineralizer and filter l
pipe tunnel, waste resin storage tank, waste evaporator bottom tank, high rad l
storage bunker, and current radwaste processing equipment. Overall material l
j conditions were good. None of the resin tanks or radwaste rooms were in a e
condition that could be categorized as " abandoned in-place." There was no j
WP significant accumulation of debris or trash, and ground water intrusion did not appear to be a problem. A more complete review of radwaste processing and radwaste equipment surveillance will be performed during the radwaste/
transportation inspection planned for the current SALP inspection cycle.
I 3.0 REVIEW 0F RADIOLOGICAL CONTROLS FOR OUTAGE WORK il The inspector reviewed radiological controls implemented for outage work, including as low as is reasonably achievable (ALARA) reviews for radiologically-complex work; radiation work permit and ALARA briefings; radiological boundaries, surveys and contamination monitoring; and plans for removing equipment, tools, and high radiation sources from the containment.
3.1 ALARA Reviews for Radiologically-Complex Work i
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- The inspector reviewed ALARA controls implemented for the decontamination of l
the fuel upender pit, repair of upender pit limit switches, and handling high j
radiation sources within the Containment Building. The inspector's evaluation was based on a review of ALARA evaluations, attendance at an ALARA briefing for the upender limit switch repair, and discussions with cognizant personnel.
Evaluations included detailed job plans, provisions for radiation dosimetry, i
limits on stay times, and comunications with health physics personnel. Based on this review, the inspector concluded that a sufficient level of planning and attention was given to radiologically-complex work.
No violations of NRC requirements were identified.
3.2 Radiation Work Permit (RWP) and ALARA Briefings The inspector received and discussed RWP and ALAPA briefings with health N p.
physics technicians and the ALARA staff RWP briefings provided information on work area dose rates and conditions, allowable stay times, and RWP requirements.
Individuals performing the briefings were knowledgeable, 1
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)rovided opportunities for questions, and verified that individuals receiving 3
f** 4 t w ariefl9gs clearly understood RWP and ALARA requirements. The inspector did 4
D( 'j~ point out to licensee staff that briefings conducted in the ALARA office were subje:t to distractions (e.g., telephone calls, foot traffic, and office q
noiso)lans were being made to establish a separate area / room specifically forLic i
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)lanning meetings / briefings. The inspector noted that overall RWP and ALARA I
0;' i' 3riefings were good.
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No violations of NRC requirements were identified.
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3.3 Radiological Boundaries, Surveys and Contamination Monitoring q.
k The inspector performed a review of bounrfarles and postings established for i
T radiological controls. Licensee performance was evaluated by a review of I
radiological postings observed during tours through the Containment and 1
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. Auxiliary buildings. Radiological boundaries, including radioactive materials area, radiation area, high radiation area, and lock high radiation area 4
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' o ~ boundaries, were well defined, correctly posted, and were being routinely maintained. Continuing improvement in this program area was noted.
j The inspector reviewed radiological survey data for the Auxiliary and Containment buildings. This review was performed by an evaluation of o
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radiological survey data posted at the entrance to the Auxiliary Building and by a review of survey data maintained by the Containment Building health -
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physics control points. Radiological surveys were clear and legible, were
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performed on a routine basis in accordance with job and schedule requirements i
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narrative comments useful for interpreting survey data (e.g., water level in 1
the upender pit at time of survey).
Improvement in this program area was y
j noted.
1 The inspector reviewed licensee controls for contamination and contamination a monitoring.
Licensee performance was evaluated by observing personnel use of 3
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contamination-monitoring equipment and the performance of equipment j
contamination surveys by health physics technicians at the Containment Building personnel hatch.- The licensee had positioned swinging gates at outside exit points from the auxiliary and fuel buildings; to remind plant 1
j personnel to perform contamination monitoring prior to exiting. The inspector observed personnel using contamination-monitoring equipment in accordance with j
posted requirements and noted that the addition of swinging gates and postings at building exit points were effective. The inspector also observed health physics technicians performing contamination surveys on cables and equipment 3
at the contaminated-clean area boundary located at the Containment Building a
personnel hatch. Removable contamination surveys were thorough; however, the inspector noted that only removable contamination surveys were performed at j
.1 the Containment Butiding personnel hatch; fixed contamination surveys were not
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M.4% performed at this location. Licensee staff informed the inspector that fixed 2.1 and removable contamination surveys are performed when materials are surveyed waw r for unconditional release from the restricted area. The inspector noted that VP the licensee's performance in this area was adequate.
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(@Idi3.4 Demobilization of the Containment Building
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The inspector reviewed licensee plans for demobilizing the Containment F 4-in WMM p; ;: Building (removing tools, equipment, and high radiatio s
preparation for plant operation. The inspector performed tours t
- Containcent Building and noted that a significant amount of miscellaneous i
W 6 materials, tools and equipment, and radioactive-sources were present.. The ZRhifinspector expressed a concern that the task of establishing, comunicating,-
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vd and implementing a containment demobilization plan, in a short period of time,-
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while ensuring aroper control of contamination, packaging of radioactive -
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materials, and landling of high radiation sources, presented a significant i
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g li and informed the ins)ector that a " Radiological Control Plan for Containment-l 1~
' Demobilization" had seen drafted.:.This plan was developed to establish
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the radiological controls organization. The inspector reviewed the i
4 - containment demobilization plan and found it to be very good. However, the i i
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A~. 1 demobilization of the Containment Building.
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4.0 ACCESS CONTROLS 70 HIGH RADIATION AREAS M h sThe inspector reviewed licensee controls for access to high radiation areas. "
The inspectors evaluation was based on a selected walkdown of high radiation i
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QQN discussions with cognizant personnel.
l 4.1 High Radiation Area Walkdown 1
X The inspector performed a selected walkdown of high radiation areas in 1he i
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Auxiliary and Containment buildings to evaluate radiological postings, j
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barriers, and controls for access to high radiation areas.
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,d;g.lligh radiation area postings were clear, legible, and, according to d
radiological survey data, correctly posted. High radiation area barriers, j
/T" including ropes, fencing, doors, and locks, were noted as very good. Health M'
physics monitoring of locked high radiation areas in containment, including i
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- constant video monitoring of selected locked high radiation areas, locked high i
Q radiation area key controls, and radiation work permit controls, was very y f f igood,
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M 4.2 Radiological Incident Reports (RIRs) e.
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W radiation areasn'Three RIRs related to access controls to high radiation i
$@ Won November 21 pC areas were generated since the last review of this program area conducted l
M "during the NRC management meeting held at the Maine Yankee Information Center, i
@w 'iinvolved a temporary loss of administrative control over access to high
, 1995. Each of these incidents were licensee identified and i
,1, ~ radiation areas in which the major portion of the body could receive a dose in excess of 1000 mrem per hour.
i (4.2.1 RIR 95-038
% RIR 95-038 was written to address a temporary administrative loss of control
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over_ access to a high radiation area. On December 1, 1995, at approximately L
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9:15 p.m., the key to a locked door for a high radiation area in which the i
radiation intensity to a major portion of the body could have exceeded 1000 l
o N mrom per hour was not maintained under the administrative control of the 4
MdRadiation Protection Manager, when a health physics technician left the key
~ t T uncontrolled at the exit point to a contaminated area on the 21 foot elevation W%#in'theprimaryAuxiliaryBuildingneartheboronrecoveryevaporatorroom.
>w 4.2.2 RIR 95-040 3IRIR 95-040 was written to address a temporary administrative loss of control 7 ?over access to a high radiation area. On December 10, 1995, at approximately<
my1,8:00 a.m., the key to a locked door for a high radiation area in which the radiation intensity to a major portion of the body could have exceeded 1000 W
mrem per hour was not maintained under the administrative control of the Plant.
' : Shift Superintendent on duty, when an auxiliary operator inadvertently left a. '
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key to a high radiation area door uncontrolled at the exit point to a
- contaminated area on the 21 foot elevation in the primary Auxiliary Building 4
near the letdown heat exchanger cubicle.
n 4.2.3 (VIO 50-309/95-26-02) RIR 95-043
.RIR 95-043 was written to address t temporary loss of access administrative-controls that occurred when a health physics technician that was assigned to prevent unauthorized access to a high radiation area was found to be I
inattentive to duties. On Docenbor 12.11995, at approximately 2:20 a.m.,
access administrative control to a high~ radiation area in which the radiation intensity to a major portion of the body could have exceeded 1000 mrem per hour was not maintained when a senior health physics technician assigned to maintain access control to a high radiation area located on the -2 foot
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. elevation of the Containment Building was found to be inattentive to duties, i-Based on followup investigations for each of these events, licensee staff concluded that no unauthorized entries to the high radiation areas occurred as 76 ;f a result of the temporary loss of t.dctnistrative control to these areas.
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LThis failure to maintain administrative control of keys and access to high
@$$radiationareasisconsideredanapparentviolationofHRCrequirements.
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.The inspector performed a selected review of licensee self-assessments
- -vPperformed in the area of radiological controls. The licensee's performance in this area was evaluated by a review of self-assessment performed since the
""last inspection of this program area, including quality assurance / quality control (QA/0C) surveillance, radiological incident reports, and poor work
. practices. The inspector noted that the QA/QC surveillar:ce and RIRs included excellent observations, identified program areas for improvement, and were being properly communicated to management. The inspector also noted that
' identified deficiencies were related to administrative controls and did not j
have significant health and safety consequences. Overall, the self-i assessments of the radiological controls program were very good.
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No NRC violations, other than the one previously identified in Section 4.2 above, were identified.
6.0 EXIT INTERY!EW
! The inspector met with licensee representatives denoted in Attachment 1.0 of
'"V* = -this report at the conclusion of the inspection on December 14, 1995. At that time, the purpose, scope, and findings of the inspection were reviewed, and preliminary findings were presented. The licensee acknowledged the preliminary findings.
On December 28, 1995, the inspector contacted the licensee representative denoted in below of this report to inform the licensee of additional findings-
. based on an NRC management review of preliminary inspection results. The licensee acknowledged the inspection findings.
KEY PERSONNEL CONTACTED i
LICENSEE PERSONNEL J. Bashaw, Technician - Radiological Controls R. Brickford, Outage Manager A. Capristo, Acting Section Head - Radiological Controls J. Connel, Manger - Technical Support M. Finn, Supervisor - Radiological Controls R. Hayward, Manager - Quality Programs E. Heath, Manager - Radiation Protection G. Leitch, Vice President - Operations C. Shaw, Plant Manager S. Shelanskey, Section Head - Radiological Controls T. Shippee, Supervisor - Radiological Controls S. Smith, Manager - Maintenance W. Tracy Acting Supervisor - QA J. Weast, Licensing Engineer
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3 ' ;tt J. Yerokun, Senior Resident inspector l.
-[ 3 W. Olsen, Resident Inspector
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- Indicates those in attendance at the exit meeting held on December 7, 1995.
A followup exit meeting was held with E. lienth by telephone on y
December 28, 1995, e
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' K DOCKET / REPORT NO:;; w' 50-309/05-26 Q?
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Maine Yankee Atomic Power Company so w YMY[p" FACILITY::
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Maine Yankee Atomic Power Station,
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Siscasset, Maine i
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M ?!NSPECTION DATES:t
,[ December 4-7, 1995 d
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- INSPECTOR:
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Edward B. King, Sec ty Inspector Date M t:
Emergency Prepared and i
Safeguards Branch L
46 L Division of Reactor Safety mo.
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I.tMphs" APPROVED BY:
Ar-WI.i RichardiR. Keimig, Chief
' Date Mif Emergency Preparedness and
- j - HQ Safeguards Branch i
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Division of Reactor Safety e
T['"AreasInspected: A ccre security inspection of the Maine Yankee Atomic Power 1
l MM Station's security program was conducted in the following areas: Previously j
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Identified Items; Effectiveness of Management Control; Management Support and Nh Audits; Protected Area Detection Equipment; Vital Area Access Control of -
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' Personnel; Testing.. Maintenance and Compensatory Heasures; and Training and O
Qualification.
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Results:
Inspected aspects of the licensee's physical security determinedtobeappropriatelydirectedtowardsassuringpublicbo$ramwere a th and i
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safety. Three previously identified items were closed based on inspector l
F4 & ? observations, document reviews and discussions with licensee management.
I f NA E J ' Management support is ongoing and the required audits were thorough, indepth
% % ~.and comprehensive in scope. Security training was being performed in j
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accordance with the NRC-approved T&Q plan and vital area access control was!
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, 4 'being limited to only authorized personnel who need access to perform their-j A
duties.
Protected Area (PA) detection equipment was installed and operated es-
- $ @6 M,~ committed to in the physical security plan (the Pl and timely repairs were i
- gQ being completed on security equipment. A minor defic Q
the control of safeguard information. However, no safety concerns or
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I ht N e 1.0 SECURITY PROGRAM INSPECTION -
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PWa The inspector reviewed the security program during the week of December 4-7, 1995. The purpose of this inspection was to determine whether r
- l the licensee's security program met the regulatory requirements of the 4(( I,. licensee's NRC-approved security plan (the Plan) and
.t i-l corrective actions for previously identified items were adequate to close the
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2.0 PREVIOUSLY IDENTIFIED ITEMS l
2.1 (Closed) VIO 50-309/95-02-02 t
The licensee failed to implement adequate compensatory measures which resulted j
in an uncompensated degradation of the protected area boundary.
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' h' A review of the licensee's corrective actions by the inspector verf fled that the actions were offective. The corrective actions included the revision of Security Procedure 15-4.5, titled, " Compensatory Security Measures," which i
provided guidance for security force members (SfMs) on the use of notification markers or placards on temporary barriers indicating that the security i
department must be notified prior to removal. Additionally, a change was made i
to maintenance work packages requiring notification to the security department prior to removing any temporary security barriers. No deficiencies were noted. This item is closed.
2.2 (Closed) VIO 50-309/95-03-02 The licensee failed to perform security lighting tests once within every consecutive seven-day period, as committed to in the Plan.
The inspector reviewed the licensee's corrective actions which included revisions to the Security Lighting Procedure 15-300, titled, " Security 4
i Lighting Surveys." The word " weekly" was replaced with the term "every seven 4
days." Additionally, the licensee provided a briefing memorandum, which was reviewed with the security force members during shift change, that provided a clarification of weekly testing. A review of the weekly security lighting surveys for a four month period by the inspector indicated that the testing was being performed as required by the Plan. No deficiencies were noted.
This item is closed.
1 2.3 (Closed) IFI 50-309/95-03-01
[
Several SFMs failed to perform required personnel searches in accordance with approved security procedures.
W h sThe inspector determined, based on observations of protected area (PA) access
-L processing of station personnel and visitors during peak and off-peak V A u-activity periods, that SFNs were performing their duties in a consistent j
manner and in accordance with approved security procedures.
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This item is closed.
hhk 3.0 EFFECTIVDIESS OF MANAGDIENT CONTROL.S
' The inspector determined that the licensee had controls in place for
"'q identifying, resolving and preventing security program problems. These
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controls included the licensee's self-assessment program, the security oy?
contractor's Performance Assessment Program, performance of root cause A
analysis for each human performance error, conduct of annual quality assurance "N
(QA) audits, and ongoing shift supervision oversight of program
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implementation. A review of documentation applicable to the program indicated that initiatives to minimize security performance errors and identify and j
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resolve potential weaknesses were being carried out. Additionally, the d
inspector reviewed the quarterly safeguards event logs for the first three J
quarters of 1995 and determined that security personnel errors had decreased..
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'llowever, on December 6,1995, at approximately 2:34 a.m., an SFM found a Q r.x document in the security training department office, located inside the PA in an area controlled by security, that was stamped safeguards information (SGI).
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.The document was in a stack of papers on the desk of one of the security s
training supervisors.
It was determined that the SGI document was 2
y' uncontrolled and the licensee made a one-hour non-emergency event report to E
the NRC in accordance with 10 CFR 73.21. The licensee later determined that:
i yV the document, even though stamped as SGI, did not contain any information that should be so classified. The inspector discussed the event with licensee management, and informed the licensee that only documents that actually contain SGI, should be stamped SGI, and if it is not SGI but is a part of an e
SGI document, then it must be controlled as SGI as required by 10 CFR 73.21.
}
To correct the problem, the licensee stated that all documents stamped SGI m
will be reviewed and documents determined not to contain SGI will be removed 9
from that category. The licensee determined that such an initiative would 1
greatly reduce the number of documents which would have to be controlled.
i
. Additionally, the security force was instructed during shif t change about'the i
need to control SGI properly. The Itcensee subsequently rescinded the event notification to the NRC.
};
4.0 NAMAGEMENT SUPPORT and AUDITS I
4.1
' Management Support i
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Managesent ' support for the Itcensee's physical security program was determined to be generally adequate. This determination was based upon inspector's 4
review of various program activities during this inspection, as documented in L
this report. Security program enhancements made since the last physical f j(
security inspection conducted in February 1995, were as follows:
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The procurement and installation of two new portal metal detectors and 1 NT two new hand held metal detectors which have enhanced personnel hjL; processing at the main access control point; m
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enhance comunication capability with the local law enforcement
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agencies; and j
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security force members to be retrained and requalified prior to implementation.
4.2 Audits i
The inspector reviewed the 1995 QA audits of the fitness-for-duty (FFD) i program, conducted April 10-14, and April 19, 1995, (Audit No. MY-95-048), and
~
the security program, conducted September 11-15, 1995, (Audit No. 95-04) and determined that the audits were conducted in accordance with NRC requirements.
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To enhance the effectiveness of the audits, both audit teams included three i
independent technical specialists.
l Hef ther audits found any deficiencies, but the FfD audit had one observation and the security audit had five observations. None required a written
".wm response and were not indicative of programatic weaknesses. The inspector determined, based on discussions with the ffD Administrator and security management, that the audits were very comprehensive in scope, the findings j
were reported to the appropriate level of management, and that the programs were being properly administered in accordance with their respective NRC requirements.
5.0 PROTECTED AREA (PA) DETECTION EQUIPRENT The inspector conducted a physical inspection of the PA detection aids on December 5, 1995. lhe inspector determined by observation that the barrier and detection aids were installed and maintained as described in the Plan.
6.0 VITAL AREA (VA) ACCESS CONTROL OF PERSOM EL A 2 PACKAGES The inspector determined, based on a review of the licensee's VA revalidation process, that individuals are granted access to specific VAs on an as needed basis. A review of applicable documentation and discussions with security supervision by the inspector verified that the access lists for each VA are updated and approved by the cognizant licensee manager or supervisor at least once every 31 days. The review ensures that only individuals whose specific duties require access to VAs during non-emergency conditions are included on the access lists.
l 7.0 TESTING, MAINTEXANCE Am COMPENSATORY MEASURES l
7.1 Testing and Maintenance lhe inspector's review of testing and maintenance records for security-related j
~.e.-equipment confirmed that the records comitted to in the Plan were on file and that the licensee was testing security equipment as committed to in the Plan.
' A review of these records indicated that a priority is assigned to each work request and, based on the level of prioritization, timely repairs were being completed. The inspector assessed the priority system as being adequate.
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8.0 SECURITY TRAINING AM) QUALIFICATION f
- The inspector selected at random and reviewed the training, physical, and l
- firearms qualification / requalification records for seven SFMs. The inspector A determined that the training, physical, and firearms qualification /
s requalification had been conducted in accordance with the security training
{t and qualification (T&Q) plan and that it was properly documented.
Several SFNs were interviewed to determine if they possessed the requisite knowledge to carry out their assigned duties. The results indicated that the individuals were knowledgeable of their job requirements. Additionally.
tthroughout the inspection, the inspector observed SFNs performing their duties J in'a professional manner and in accordance with applicable security procedures I
and post orders.
9.0 EXIT MEETING The inspector mot with the licensee representatives denoted in Attachment I.0 of this report at the conclusion of the inspection on December 7,1995. - At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were presented. The licensee acknowledged the preliminary inspection findings.
Key Persons Contacted LICENSEE AND CONTRACTOR PERSONNEL G. M. Leitch, Vice President, Operations C. Shaw, Plant Manager J. Connell, Manager, Technical Support Department P. Motivier, Security Director P. Cunningham, Security Supervisor, Operations H. Torberg, Security Programs Coordinator R. Hayward, Manager, Quality Programs Department J. Weast Licensing Engineer C. Urquhart, Security Chief. American Protective Services (APS)
G. Everett, Deputy Chief, APS G. Nichols Training Coordinator APS P. Dostie, State Nuclear Safety Inspector, State of Maine
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J. Yerokun, Senior Resident Inspector r
W. Olsen, Resident Inspector i
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Indicates those in attendance at the exit meeting held on December 7,1995, 4,
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Maine Yankee Nuclear Generating Station Information a
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Management meeting to discuss the recent performance of the radiation protection program.
MNEETING AT:
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9:30 a.m., November 21, 1995 iBs e.:.,
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. INSPECTOR:
R. Raglahd, Radi on Sped list Date L
Radiation afe y Branc s
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.m S A. Capristo',' Supervisor'- Radiological Controls T
J. Connc11, Manager - Technical Support i
E. Heath, Manager - Radiation Protection
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- R. Mellor, Chairman - Huclear Safety Audit and Review Committee
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G. Leitch, Vice President - Operations S. Shelansky, Section Head - Radiological Controls l
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- C. Shaw, Plant Manager - Maine Yankee
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w QS 1.2 HRC Personnel pe M 1,r J. Rogge, Chief, Branch 8, Division of Reactor Projects (DRP) i f y J. White, Chief, Radiation Safety Branch (RSB), Division of Reactor Safety
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.R. Ragland, Radiation Specialist, RSB, DRS E"
' 2.0 MEETING
SUMMARY
An open management meeting was held at the Maine Yankee Nuclear Generating
,s Station Information Center in Wiscasset, Maine, on November 21, 1995. The purpose of the meeting was to discuss the recent performance of the radiation
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protection program in light of the licensee's staff's identification of continuing deficiencies that were similar in nature to issues that were the subject of previous NRC enforcement action (identified in NRC Inspection b
Report No. 50-309/95-06), and reviewed in NRC Inspection Report 50-309/95-06, dated April 21, 1995. The NRC indicated that the licensee's corrective
.m actions were not effective or not being effectively implemented, and less than
- <m adequate performance was continuing to be demonstrated. The licensee made a c
" y.P -. formal presentation and provided a handout. A copy of this handout is y
attached. Recent improvement initiatives such as team building and re-engineering of the corrective action process were encouraged, and efforts to
~v strengthen self-assessments and plans to conduct industry bench marking were 1
commended. While the NRC is convinced that the licensee sincerely intends to improve performance NRC concerns will remain until actual performance l
improvement is evident.
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NSARC Comments R. Mellor
,5 j. Concluding Remarks G. M. Leitch m
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w w g[ Report / License No.r, '50-309/96-04; 0PR-36 M e, 7
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"'d Facility Name: 7' d n Maine Yankee Atomic Power Plant L<
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March 1-8, 1996 nspection At:
C ' r Wiseassett, Maine '
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R. Nimitz, Senior Ridiation Specialist Date
-t Radiation Safety Branch A'
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- s, Division of Reactor Safety
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CYito ri,h/ir.rH m/A1}%. l. l t
L. Peluso,' Radiation 5secialist
' Dat's
$@%M Radiation Safety Branc1
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t Olvision of Reactor Safety W'
L/T4 a
Y:
ph 1
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JWulliams,. Senior Operations Engineer'... I Datre -.e eratorLicensing-Human ~ farformance Branch h",,
(
vision of React r afety W
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_ /]/
y Approved By:
M 8rN J.Jhite, Chief V
/
Date l Ja'diation Safety Branch
.}
/ Division of Reactor Safety 3
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..fy;fAreasInspected: The inspection was a reactive health and lafety inspection Jof the circumstances, licensee evaluations and corrective a'ctions associated N[dUplanningandmaintenanceactivitiesassociatedwithvalv
~Z with the unplanned release on February 29, 1996, of high specific activity
? V radioactive noble gas from valve CH-138. Areas reviewed included work d
$ 7 tailpipe, the applied radiological controls provided for the work activity Mwith radioactive effluent release limits.a % including occupational exposures of affected i
[1
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O / Results: No significant internal or ex:ternal occupational exposure occurred to the affected individuals. The short term, unplanned release of noble gas E
{Was well within NRC radioactive effluent limits. -The individuals' efforts to I
Msecure the leak were effective thereby resulting in minimal personnel
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j 7st w hygf n Lh~ alm w& a v @4 p 9A,vfz yw;9 %k > #MMf&bMWWW;.;' %;r.i:.1.0
- INDIVIDUALS C0KTACTED '
M ,l~+ J:m e: %n w m nfW y n yh. i L7 f N U.V % R..W W.. QQ & Qf!MV+yQ Q1;f yf% w@& w;l:a .s MWhg Up l-1.;4 X. Principal Licensee Employees ( A Q,ijf,.,,pxx\\ j! de,. 6.9~ n y W %Mgdw m t Oa sw p !%MAfk,q . M d g& &*F i f*J. Connell, Man &f&f&sQ, ? Y Qi &Q Qhffff j Qf ss
- R. Hayward, Acting Manager-Quality Programs y, g g
/q $ gbQ;Q ager,tTechnical Support X y g y@g min g-f*E. Heath, Manager - Radiation Protection Q 4., T 1 S W R A l b y @% L yngp d $~; h h g h@ M
- B.'
Hartin, Dosimetry Expert M % W E/ f*G. Leitch, Vice Preside 4
- R Heixell, Hechanical maintenance Section Head.1 ~, fMWZ QG b
r.y i' y sMw$ df FAN q g y g ~.1 f*P. Radsky, Chemistry Department Head y @3 @Q M ' g %y. O F (y V #*C; Shaw. Plant Manager e.. t. a;
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- S. Shelanskey,' Supervisor - Radiological Controls
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%#.K m N*T. Shippee, Section Head - Radiological Controls " ' M (# %"'C'*S. Smith,OperationsManager. Qj[d@6,'*W. Tracy, Supervisor - Quality Assurance m o- / d L4.,
- L 2, N, c d dg ;' f*H. Veilleux, Maintenance Manager M' R~ " L * -
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- J. Weast, Engineer - Licensing
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%gg, - 1.2 NRC Employees t 4 j 7 g .y l$ @ , f*J. Yerokun, Senior Resident Inspector c 4c A?
- B. Olsen, Resident Inspector
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- J. Williams,. Senior Operations Engineer
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- ,"; '!.3' Others n
T m WM,'-
- P. Destie, State of Maine - Nuclear Safety inspector
,.? e A.w.n. y . e~ 4 ,~ .e MTM # Denotes those present at the preliminary exit interview on March 3, 1996. J +g4 m.s m.p, $yQD s
- Denotes those present at the exit interview on March 8, 1996.
C q$4 - d' Wi M@M@ The inspectors also contacted other licensee personnel during the c the inspection.- -a ww-3._ jhp2.0 PURPOSEANDSCOPEL? A - '%y ,y/{ gns x Ww r The purpose of the reactive inspection was to determine the circumstance. g""n f D&'m, leading to, the consequence of, and the causal factors for, of leak of highly y z ?j %yy* > radioactive noble gas to the Primary Auxiliary Building as a result of a work activity involving a radioactive system component, CH-138, w 4.)h' ?y M,M
- '(
4 s. a 4-A'- f %'.? The inspectors reviewed and evaluated the maintenance activities'on the valve,t * ~ A A 5 g$t@@@M occupational exposures of the involved in bp Mthe radiological controls provided for the work activity including the @.Cyph l
- N radioactive effluent release limits.rIn order.to ascertain the circumstances :,.
MG%hlicenseeevaluations,andcorrectiveactionsassociatedwiththeevent.:the M} W M M inspectors met with cognizant licensee personnel to discuss the E t m
- %)@Rilinterviewed the involved workers and health physics personnel, and reviewed
@$ d @@% affected valve CH-138.0 applicable documentation associated with the event, including the- ,-documentation involved:in the work process controls associated with the, h' 1 1 - r gM ' " pa c v4 k)h$ fg Nb.h%sM$M@a m fy sM3ih>1 Od5'Y M,_k h. s ww mn m Mhgln, Wy> @W Q.%,k+. sn ~ ~ }s p& W ,a N lC ' ,:y %p ;%hMd$y%. %"p.%, ", UlMQ,v O M n% C 'L 'O 19;;
- 4 MpN(Y W.M W #
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i-S' < W@::&r,qqggg f m n; % ;; g W p y g%g. w.4 w 9 p %w ;M g mpfg y R;m jyf f, qyn 49 f m M A,y f: g w y# Q & Q &n m $ a g QQ3 w n. w $ &j g y WWs Q s gyg p ny f b %w? O R M %p @ flL pf9 Q f%fb p W Miff 2 i, t a/: maus .ww j $p id p EVENT.DESCRIPT!0 0 y [g @m y y w g 2 '(3.0 d D~t TU 5 % #- N(Ci%1MNNbMdW 0n the morning of Febr&M8MG% jd UMy%(c hp 4:mnpmg 4mec:hanic (Mechanic A)pa permanent Haine Yankee uary 29,?!996, at~about~7:00 a.m., 440 l /was'assignedWMk Rev.000, which" directed the installation of,a pip ? 70rderg(WO)'No,195-622-00,f MC \\ fcap on the vent tailpipe of. valve CH-138 located at about the 36' elevation ofMM7 the PAB.qThe valveris a vent' valve of.the level
- column of the VCT;and i M
M}%[AWenvironment:and(under normal operating c e 2 'only isolation' WMW s 4 the VCT gas space. 4The VCT is a 4080 gallon enclosed vessel $hp [jf%q/k@? with' reactor coolant ~. which;actsias a let-down voluine.for reactori, coolant'and.is partially filled ip iQ l g Ms fmaintained'underabout:a24poundpersquareinchgauge(psig) hydrogen 3A . M $@g noble gases transported by the let c N blanketmThe gas! space normally contains high concentrations of radioactiveM i M g@@~ qQ ww'forrthe'gqngw g2y w;ug;in February 1995 w i yn yumgjpNkg 49 6 g O pTheW0 task was originally initiated [h ggwasiin" shutdown conditionias'a conservative a i WW m any potential leakage of. noble gas past CH-138 (a ball valve).nThe task was t i % jMto be performed while'the' station was shutdown.,% However, cat-some unknown timewji& MQgg the WO was changed to' allow work on the. system with the reactor at power.% TheMM'~ ~ & &Qindividualfwho changed the work activity to allow the work with"the reactor at@ ' QM W power was'not known h pMp$The' work activity!(b@#10gdEhmA ased'&p pmdogpgqWP%ggppM@Wu H e%1%@@ d Wm Mc on technical instructions contained in 10MM i p g involved tappinggusing a metal die,ithe 1/2 inch diameter (approximate 4 mQpginch long); valve vent tailpipe protruding from the valve, and placing a pipe U 4 $mA cap on'the newly threaded end.y The work location was in the overhead of thel, P Mg%PAB and was to be accessed via a ladder. A contractor mechanic (Mechanic B) gbf was also assigned to swart the work activity.rIn preparation for the task, y v a l j kHw d the two mechanics discussed the WO on the morning of February 29, 1996, 6 N3 en 1: M8M@obtained permission from the plant shift supervisor to aerform the work, asce i j i YMbproceeded to the'HP control point to obtain the necessary radiological. kW%icontrols briefing.L While at the HP control point, the mechanics were informed L l 4 i y 3%qmby HP technician A,:that their work activity was_ identified on the work
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) d N,plannim schedule,1and that radiological surveys were to be )erformed by a' gDM1ocation of work in. order for HP technician B to access;the are )e placed at ther 9 s i T % p% #4 et wm_g;$mdygm c % ;qM ; W + w i At~ about 9:30 *a". muon February 29,~ 1996,',HPitechnician B performed theN, i [W4 radiological; surveys and briefed the two mechantes as to the radiologicall 4 m TM$ conditions in.the work area (i.e.1 maximum contamination levels $ % Was 1,000 disintegrations per minute per 100 centimeters' squared (dpm/100 Q 8h.cm")). aThe mechanics were, informed that no significant1 radiation levels or8KM s VJcontamination: levels was*present in'their work location and thatithey were toke % Wavoid,thel duct'workilocatod at the rear and sides of;their work location.vNoN I 7;hMairborne radiosctivity samples were collected or prescribed inithe belief that[ s WWnW airborne radioactivity was present or, anticipated.t The mechanics'werewG I ~ Mprnided the: briefing' outlined in Attachment A of Procedura No.19-302-2{which? e >J Qrequired!that!workershleave;the work area and report:to health physicsWW% n 'y personnelsupon:any uhange'in work scope (cM nge:in.positionHlocation heethod p y ' @gW tofe work,Vsystem barrier / breach,stime to'compir.e work :that could-affcet Q W {d$ , radiological; conditions.S The mechanics'were iubsequen)tly given fhs%MwQwTqQ; valve;ta11 pip @e:by HP technician,B. a g work on the QA y, @%Gv %_ 99WM %f W ?Qf 3p 46 kh h h h
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Edg ' u~ e n my ,a m, v - - *y a <c 43 -. lj f 7.q.V .~ p*. i f y;, af g y *. 'm.,s;z:S; ,1 n u!.~ e 3. 3. s m. w hh 5Notwlths ndi g th ne igible radiological conditions present et the V gl;g+]p@tfor the task as specified by their W0 and applicable radiation pr g 1 g WDprocedures.~RWPNo.- 96-00062, Revision 0, was to be used for the task. % % 6 ? -i $g Neither mechanic' signed-in on the RWP for the task nor were informed t1st'an'pfN M j]5.hj $% RWP was available for.the task., Consequent I/ g / w, p, g prescribed rubber l gloves for working on a contaminated system as prest.rfbed q the RWP.y p g g g p M y % y &.g4 g i y 4. f{4 , Q,{: p o,u.p.xw % : e .v my m m;%g m ~ (f At the start of the work' Mechanic B, while we;aring street clothes ~. belt,andcottonlinertypegloves,reachedoverapostedcontaminationareasA1 g$ aj !,i tape and proceeded to cut the threads on the vent tailpipe of valve CH-138. WG M J i h pMechanic A remained at the bottom of the ladder to support Mechanic A 1 time that the valve was a single isolation valve between the PAB environment W ^ 7 W $wband the VCT gas space. Mechanic A assumed that the valve was a vent va I 4 Qt from the charging pumps. Mechanic 8, who performed the work, had no MfM D K conception as to what the system was or that the component that he was working) P =. 6Mp % on was directly associated with a high concentration of radioactive L [34 tpg&1:w p: s s w l. tw o% (Noter Th'e inspectors 'noted ~that the mechanics' WO indicated, as the-f p S..- % N(.M(;M description of problem, that " valve tailpipe needs to be M9V gA(; M possible VCT gas leak from top of level tree".' - Neither worker explored or UNW W' My) ME personnel protective devices., Further, the WO d Mquestioned the nature of the possible gas leak or the need for any additionalM b QW g;MW 04 physics personnel for RWP generation purposes, did not include the 4 JN)WMl/2 inch _ pipe and installing a pipecap.)
- W possible VCT gas leakJ It only indicated that the work involved threading a Qgs e
7 g' i. N -h Wm m M '.3 er M wgk.. s.%sy .. a n' MM S - ,<e4 ' MVSM During initial cutting of the threads on the vent tailpipe of the valve (atLOrs W %QM approximately 9:45 a.m.), the cutting die fell from the inside of the ep .*W 6 R approximately 30 inch (single handle) ratchet being use P /W 4@A[+M@to obtain wrenches so that the leaking val D Y wrist. Mechanic B subsequently searched for the leak with his exposed wrist,Wd c K U iocated a leak on the valve (CH-138), and informed his co-worker (Mechanic ^ Mechanic B W, noted that the bolts for the valve were loose and could be turned by hand. . d @0 " Mechanic B remained on the ladder at the valve location and noted that the d. d 'l valve appeared to leak only when he applied torque to the ta11 pipe with the f g)e y " thread cutting tool.The mechanic was unaware that the gas was from the VCT Mg r which contained hydrogen and high concentrations of radioactive noble gases, h Neither mechanic (Mechanic A or 8) informed radiation control personnel that
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the work scope had changed or that a system breach was encountered. g W. i MQ echanic A' subsequently left the area, proceeded dowr ohe stairs to the Hot w N$yg;X1(machine shop located outside the PAB, obtained the necessary wrenches from his WMk. tool box, frisked himself and the tools for contamination upon exiting of the 6s shop did not identify any contamination, and then subsequently returned to $/M@# Mechanic B who tightened the leaking valve using both hands at the f the work location. At the work location, Mechanic A passed the wrenches to tf k % $ 2 d Mechanic B noted some gas escaping when tightening, took approximately onef kminutetotightenthevalve,andsubsequentlyannouncedthattheleakwas: M@ @h Q s XM ,l a; o MMMC 1 b ?W P xv4 '. y J .v ge m mm f h+w?hlll-Q' SNhkWK w. <s xlj W Y sl~ ^ e m:s 3"Q. g3" "T '"" s, i -ew; c jp%n W@,@: x.x n..^~ ; a W9a M?<N&- CpWM
hW.W $ W &g d %g$ $ $ y? ?? ?$ h ? $ & S k 5 & ?'y? $ V / G hWW .T v T M & % W % 8 M A & D
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<,yMbMVLW %N 49g g h j M-h y.9 p [ m u g% w. d ' g g, xy%ggWMMP g o L. M@m;n y t e,% W y - e., ~%.ogv.M.4l M ~ M W" % : .m L. g qM M6.kN.y;Uy$ QG %g:.%. W }.M > ; .M N p,o 9~ + g,r ws;mwar , a. g eny
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Q@pryp;!worklvation,MechanicAnotedacontinuousairmonito (other individuals to the work location of valve CH-138.. Upon return to the -
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@ ' M ieffrom th( work area. to be alarmingt Subsequently, an HP technician AM technic :n B)' initiated collection of airborne radioactivity air samp(HP, les. The-W ? mechanics were directed to leave the area.
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R ' they were identified to be contaminated. Subsequently, fourteen other 4}Q $ &p3nh workers were subsequently sent for whole bo v*
individuals were also later identified as contaminated. All contaminated 4/ * :later characterized as short-lived daughter products associated with the decay l%k[- of radioactive noble gases.
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h . e h;MW(4.0 m PLANNING AND CONTROL OF WORK ON C ' ct, u. -A- %ng4 '11 Genera 1(Scope'of Review)@Q ' "' c 5"',*;,('
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y y g$ . 3; c, m.- r-3 4,.. $MThe" inspectors' reviewed the planning and control of the werk perfonned on M y 'N M February 29, 1996, on valve CH-138 and its tailpipe. The inspectors also W L %hreviewed the implementation of applicable work process and ta $ q%@ M White Tags"). :The inspectors also reviewed t d d " White Tagging Procedure";"and Procedure No. 0-14-1.1, Rev 0,;" Work Without 4 A Mok.W% 00622-00, Revision 0) associated with the task of capping the vent J
- @ W N 'on valve CH-138..The reviews in this area were against criteria contained in' $.
a '( %ppithe applicable procedures and Technical Specification 5.81" Programs and: x. T M e Et 9 % W ;mProcedures," and 10 CFR 50, Appendix B. 4 ' 4v $$ys,,; k 3: 4, s q,, x, %p pg ,v v n m c a n 4m . _,. ~y ,f h f ~ 'y &.,e 'y* MlMMMl %lMNNNMAb%ddMMiM#I" Y' i
k $&Nkh hi Ak i ~on' review'Nb
- ThiYevaluation ofithenicensee's performance;iMthis; area was base 1
E(cognizant'personne1rfromtheilicensee's of, applicable documentation,ttours of the work area'and discussions withAM %Q j 4 j Mand health. shysics departments'.%The inspectors;also. discussed,the work, order gg r j (WO);with,tle. mechanics;who' performed the valve tailpipe capping.g ~ - E ggg% 4 9%%#@Gn@%@WAWqhopg;3Md g i Mh * %dd 'y 4.2%NRCObservationslandFindingsAbpp$jb . Q$iSq.p A f $$ %nuMMf48sk &&i&&W W 9 ( The' inspectors'irevie$fav % % d % d k W h W W n j w determined that the applicable war order;for;the,tas (WO 1995.sThe description of 75b ytheNo.t95-00622-00)'wastinitiatedonFebruary24, discussed'(seeSecti )roblemontheWO'at?that: time,$aspreviously M sof_tiis; report)'was!that the tailpipe for valve.CH-138 needed to be: threaded b iforf a cap in; order,to. prevent a possible VCT gastleak_from theLtop.cfathe; tievelitree N The; inspectors'idiscussions with cognizant' maintenance personne . $V indicated,+ that no' leak was' evident at the time the work: order.was~ written k only,that the cap.was3tobe;installedasaprecautionaryl measure;- kg gp % @ 3 y# 0n. Feb'WWN73aQM$rMWWQq%g@p(PSS)ygwf ut W ruary 24',51995/,the PlantlShift Supervisori assigned wor pr or $g AD"2" and the' plant ~ conditions offextended/ refueling shutdown;fo6the;Wo j > Splant waszin an extended outage at<that; time.gThe~ planner / scheduler pr&T i epared } N Between_the time' period February $the WO onlFebruaryi24 A1995,inf acco m 24,'1995, andJuly!!!,11995rthe:WOwasW $pW { y@ g k(o bsequently" reviewed and approved,t in accordance with: step:6I4.o su N 16-3,;bytheoperationshengineering,'radiationcontrols.4and'matrMnance S l 4 Jdepartments7%TheWOwas'thensubsequentlyprintedandsissuedforworkingon.' l2it1995 However,(the work was not, started at'thatitime n g q M .TJulyj%bMM i Fag .The, inspectors re@ vie $WhM&MtMMWA~MNSMWN>M @ted i w determined that sometime after;the Wo.was:pr n MND i issued the3 riorityiand. plant 7 conditions were changedi(inspectors we y 4 ' normal /plar.ctoperating' conditions tand priority.6@!that' allowed the task The; 3 'able,to determine'who changedithe plant" conditions ~ W. j p Mperformed under; normal plantroperating conditions or the bases fo j p h - %@h; determined that:no; safety re j priority 6.RNone of4this,information was required by the procedure. As a resulth and basedjon#discussionswithcognizant; personnel,theinspectorsd jA@ a 43 Mef Ng 10@%o11owing the changes!toithe WO@M@u&M"4@& WWHMWO WE Extended /Refue%:%@ $ Ne% % % W b g @ rig % @ M Mperformance) was approved by the maintenance W P s vertheless,iths o inal: WO '(specifying ling ~S r wor 'N H ided to the V! % maintenance. mechanics,for use and approval by t1e plantioperations shift % b T j supervisorrfon: February.29, r1996.bThe WO theoriginalpermit)wasM Mapproved and; released.for work;by the opera (under, department;on the:sa tions dp/ > i
- The! permit wascapproved by;the* shift operationsisupervisoryl(SOS)feven
- though M y he*permittindicated,to be worked under extended / refueling,shutdownmThe SOSWM i wasVquestioned by thelinspectors; relative to thista)provalf and the SOS
~SCl indicated he did.n;otiseejan apparent; problem;pthp11
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- February 29 il996,3the'plantN$p$ihM!Md@90@% reactor,Mip
'gOn$$GWR ,' n kuM at %1 jpressurized;withia hydrogen gasi;was;ketrand contained highly radioactiv blan Vgasesf60nithe morning'of February 29,1996sthe work was performed lvolve?and :valveitailpipeLresultinglin')elease' ofa high; radioactive'gaon;the % an%y;&qg]ygmaphhkb bghk g4,$ m qqpp#W r se hydrogen to;the AB.WqE @gp W ww gp? s ygggg
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,L f QQ A kab h k. pare i 1 ti ns NNork Planning, Control ann; mh1Ntbeb'MN,hm nway wW Q 'gQW w h.qqq' y4MmMa m Ms M The: inspectors review of the work planning,, control' an~d imple w w g W m% that Technical Specification 5.8.2 requires, ~ 1n part, that written pr M M S pe @$s dated February 1978.y Regulatory Guide 1 shall be established,' implemented, and maintained covering the applicable 4 g N 0f y 4 ifM procedures recommended in Appendix "A" of Regulatory Guide 1.33,.(Revision'2),$f y p W@hQ February 1978, recommends in Section 9, that maintenance that can" affect $pa j;p$i-j t' safety-related equipment should be properly pre-planned and performed inW;Q4 ?y $ppS s Wp &j[f accordance with written procedures,; documented instructio
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- + E. W d Maine, Yankee Procedure No; p W w c0-16-3, Revision 10. " Work' Order Process," cont ~ .b 7 4g u Q$$"the licensee's maintenance'that can affect safety-relatad equipment. @W @e l gykyw%~ inspectors identifled the following examples ~where' a v a-d,. J ', ' %~ DQh' The ~ M:y m~ , 4:. - .va ~ w w y) a qQW h49 recommended by Regulatory Guide 1.33 were not implemented in accordance with .T,,Wj M y S g vTechnical. Specification 5.8.2.(NOV 50-309/96-04-01). >6, 7' M g, oa v Q M @H, s @' Haine Yankee' Procedure 0-16-3, Revision 10, requir ( t C 3 the PSS assign a priority to the job and specify plant conditions under Wi4b which the work will be conducted. Procedure 0-16-3, Attachment C, Q$$f1, " Definitions,". h/d> (2 condition assign (page 33) defines a plant condition as a specific plant i ed to a WO, during the planning process, to indicate M Q S
- when the WO can be performed.
dMp ;, . WO N k 95-00622-00,'Rev 000, dated July 12,~1995, was used to initiate
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4 KeWA ..u, .o g.W/M-0 t and conduct the work of capping the valve tailpipe on February 29, 1996.f MK g?S ' The plant condition specified on the work order was " Extended / Refueling $ M V F ~S/D". The inspectors noted that contrsry to the WO, the actual plant ' g G W ' V condition when the capping activity was authorized by maintenance and i -,, ' vbg 4 operations supervisors, and performed or February 29, 1996, was 90%- Mn n W@o rt. actor power. This is an example of failure u adhere to Technical' ~ u . _G,.". '0, $ man %g..~; Specifications.8.2.M wk t mm s x ...eu., +. .w -r cW ' ' Haine Yankee: Procedure 0-16-3," Revision 10,: Attachment G, requires, in7 w s.. g g (y;A part, in Step 1.6.1, that before beginning corrective action, the-. 1 hM ~, ?; A responsible worker shall comply with the safety procedures and SfQQ:w@. Jconsiderations noted on the WO form. WO No. 95-00622-00, Rev 000, dated,3-QW 1uly 12, 1995, was used to initiate and conduct the work of capping the L x b d 9 % l valve tailpipe on february 29, 1996.. The WO form indicated that a xM y ?.e @w @n@ M m t .2 ~ radiation work permit was required.. .d7p My. @yt Q; obtained or signed-in on a ra mw3 nn m-w "%3 yN M $ P T work activity described on WO No. + hnMRTThe' inspectors n~ oted that' neither of the mechanics who performed the af. 95-00622-00,- Rev. 000, on ' a
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$ % h f. February 29,'1996, hQ$pM ? This is a second example of failure to a fQ 4Qf 5.8.2.;
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8 9l g ~p p>h 0g3mW$$fQ, .g. ppfgy i h.] k 'Q.i. m ,l > gy,i, lf; s v w .,w . ?l i Q' y b j ohm hMg Maine 1 Yank}e$biu'rocedure'm,g . my nh u. .w & m 0-16-3, Revision 10,' Attachment"G,Fr & w a m ..-4, emp ~< eP eq'ufrei,5fn W W QQ h p $ g g part Ein Step 1.1.1,(that maintenance and test 3 p/Sthis procedure shall'not exceed the approved work scope ,gMhX Kz g q W W work. Further, Step'l.6.3, requires that any changes in work scope must N m % $ W M Q f r be brought to the responsible department supervisor's attention, #. ?~M " The,'in;toectors noted that'during performance of tap y%@Q %. 3; & a A Qggy - 37.. - t M. ga .y 3 M. mechanics work h e 9 M,., . vent tailpipe on valve CH-138 on February 29, 1996, MM? of the scope of their approved WO Skg6, g1 fineither obtained additional approva(No. 95-00622-00, Rev. 000) and T g l to perform the out-of-scope work or,h ,. 1 m. g$ h@AT i brought the out-of-scope work to their supervisors attention prior to. f 'm @Q4. [n,.' Technical Specification 5.8.2.. performing the work. This is a third example of failure to adhere to ;; 0
- 1. ; ' '
' 'i @" S f M - w M ' m, n g, n.u 3 Main'e Yankee Procedure 0-16-3,' " Work Order Process," Revision 10,' ? jiR gi requires in Step 6.1 that the WO originator notify the Plant Shift r o l y;; M4 '. Superintendent (PSS) and hang a WO tag if conditions allow.' @Wgf; . The inspectors noted that when the WO originator initiated WO 95-00622, - a p ~ h "y@. ySQf the fourth example of failure to adhere to Technical Specification 5.8.2, .M the originator did not hang a WO tag and conditions allowed. This is 4 . kMp .a-
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/; w ;n. s> g 'hm g%y Order Process," Revision 10, relative to 10 CFR 50 Appendix 8,' Criterion V 7 ;; E. The inspectors reviewed the adequacy of Haine Yankee Procedure 0-16-3, " Work y 34TThe inspectors _ identified the following apparent example of inadequate. 4M9. procedures for performing maintenance on safety related systems. (VIO y /' j M e i 50-309 , W & n /96-04-02).., < n .,,m , m~ OJ y i gWh J y, '" > Wy#%W 10 CFR 50, Appendix B, Criterion V, requires that activities affecting'M I [NM.;% : ' quality be prescribed by written procedures appropriate to the J" ~, Nj
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,, circumstances. m ~.. W w;.n; ;a .n 3MW.e Maine Yankee Procedure 0-16-3, Revision 10,' controls the work order W 1
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5 kb E( 2; N, ?? WO No. 95-00622-00, which was written to cap the vent valve,tallpipe, 4 '?.. gg n, was originated on February 24, 1995, andreviewedbypersonnelfromtheb T ~1 operations, maintenance, engineering, and radiation control departments R. Jr - ,' W.D, yft - before being printed and issued on July 12, 1995.~ In July.1995 the 6 ? 9, . u k %j plant was in an extended / refueling outage.1
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. s y4u g<:; m m % w W.. .ru, t..y, Y G W c d.e 'Sometim,e after the # work order was printed,; the plant conditions for the ;6 m%Ap 'Q v e : a $.W. % % 9,:in hazards (hydrogeng radioactive gases, and a high e 446 " # considerations that could _ impact the job or personnel safety.u How) eve % WJ ril 7mW M M N ' these changes were not reviewed by the associated respons l W, N M V departments.n Consequently, the work order, as revised,'was' inadequate Sp P 1" l e.g.y did notfspecify spark resistant tools, did not provide for: O M "^% b j,df @ 7a(dequate instruction of workers, and did not provide for safety $f f k MMf.% tagging)(InsummaryFtheWOdidnotaddressorevenrecognizethenewf]ii %$u $ $M hazards and the consequent' potential on safety.,As a result, the- , fN 1,' inspectors identified that Procedure 0-16-3, for the generation of W0s . ?(. M g, #].;wl%;$f. - +m;p y3u 3,: s> 4 j Q;S 1 a s m @% m.; v e, & a w+ M ~ :;< f, n e ~ yXp ^ ~ ;p J e, 44 T : h f wq ; ,4 'e 11 / h,;. l' .p.aty J %.a t ,, 'c sf M h hS h b.' h b $ h ?g b $ $ Y k W ~
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WMMf~ @d@@4@$@ D[#gpa .4 y am; dis 7% $ c ;? %@4 w y ,w g m ddd$dMF' eM@K[4; Wh3Nh$g Mh$$@mL 4< WM,p .: Q W C %; %v m.v @,4p % hl Nv e?+ ppgk u f.m A Q 0 a.'Wm:m N M4 Q W;)M M' Lk . m,JQ Y S y g.m & "' t.91 Q Q &og p M. f W W yf &., l' Q W N f.Y ,,,f f ,a .w dy5 J A w...-. w.xa. ~. s gN s'4 t 'w, ? ! NA +p? ; M s 5::% Vf.m& f 4
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W - l:&plpg t g was inadequate in;v 4 V A ;a." c OpwfMn W Q4 thatLit ' allowed changes to~W0s without proper analyses p E . sg.2 review of the changes:and the potential affect on safetyw This Rg%s@of fallure to adhere to 10 CFR 50,LAppendix B,, Criterion V.W2M% &, example .The inspectors m$t@the %dWap+ditional negative"obse#N f mi x MMunn X ade following ad rvation. 4 . g, D W pThe' inspectors'ereview $ w m: [ WwpWMAWg#Mi % S4WB cts Ma#W9MWMM: Nfh noted that, Technical Specification 5.8.2 requires :in p -M of the work planning; control and implementation g $ 6 M .p@M the applicable procedures recommended in Appendix "A" of R }' procedures!shall be established, implemented,:and maintained covering n W h l M d Q M Guide 1.33.-(Revision 2),* dated February 1978. t* Regulatory Guide. l.33,y?p4q yg8(Revision 2),'datedFebruary1978,recommendsinAppendix.."A",wAf 'NdSSection1.thatproceduresfor.equipmentcontrol;(e.g.,;1ockin % Q tagging) be covered by written proceduresttMaine Yankee Procedure 9%d O-14-1, Revision 17,1" White Tagging Procedure", and Maine Yankee 4 WW M h M M Procedure No.'0-14-1.1, Revision 0, " Work Without White Tags,", provides $ f W l $% control of. white tags (i.e., tags that identify industrial hazards).1 Ai $g ( g$ h@ M R O . M Ww e r: Au am t a 3.2 A % M Maine Yankee Procedure No. 0-14-1, Revision' %"3t. W e + M g y b > 2.1,-that tagouts are provided to protect workers an m 17, states,Jin'partYin M i %q p NI$nknX e establishing boundaries which isolate energy. source 3%A M W M. Procedure 0-14-1 defines an energy source as any source over 15 psig 6 T , g MpiO ;(mechanical).0 Further,tStep 6.2.6.1 requires special consideration be%$Wd MyW MWh4 given,to; systems l&p%;mmm::%nw:containing. hazardou %gp g l M pg a wy vn.nkh mm 6 1 & $ ' q%p ;The inspectors noted that although valve C %p ? p S M iM % Wsource (the pressure behind the valve was 24 psig) and the valve served @q< bW f has an isolation between high activity noble gases as well as hydrogen 6f b v$ &@T % W jb'@q ~ the valve was"not." white tagged".TM% & MP MWTh' ^ inspec@ tors"discu%WfegiVI$ahh&&h*&'s. va 6B 'l & y; pnfu Neognizant licensee personneluFrom the Q %@y hWM& Wh $&Q W e ssed the;need for white tagging of th MMi b@%ffW@Wthere was#a. lack of consensusTamong lic @% S ft MO4 white tagging procedure was such that the cogni idkMid for,a~whiteltag&The inspectors were concerned that the clarity of the j W d<* W'?W $a were not able to clearly understand what conditio l E ' N W M 9for a white; safety tag to isolate workers from energy sources..As a f [,t gsp $[sDW result,' the inspectors considered the licensee's proced ' g' 4 tagging to be weak. *
- M
,p ^ N ? $m eThe'insp'itors nond that the licensee wrote ~a second WO (WO No. e a i %Rev 00) on March 1,1996,: to verify that the valve was leak tight after the e -9 event.LThis WO specified tightening the valve bolts (step 5) aid checking for f. t., .leakss(step 6); cThe WO appeared to address the hazards involve] e. g.,1 i ll 'specified spark free tools. and RWP).- However, operations release (d the worki iwith:no white tags'and a handwritten note that stated that the work was w %;,FM ' 1 dinspection only plus: snooping".x!t appeared, as discussed above,;that it was V l i tappropriatetowhitetagthevalve7php$ e g w r W[ imp $ddk N T@@@MMM m%m M g pd %wg@pw @b;M[%g g 4.M R f 3%DONMhMMKM MdE E i Ai q e sT m e d fme&spMmyngyWx V b:4v 74w E y31 we Mn%m&~MA%~N ' 'S in " "t: W Z,n' 7 % W I = - d'+Mygwj-m,DW @n.wwwn m, n x
- e xn
, UwRia ?N .1 w t ~ . QM" ' b i,' ?. y 3,, w
- g g y w
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_,m.= ~ ~ ,x ,i.- s n -,y.+ .-:, r / n~ 3 3 = i Y Ytr j' gg 1 4 e NRCConcusons(Nantenance)sk ' ~' l p y 4 Th' ,nspe of i o g 84 3 W/&#MiWgsipid8SjWWfdE#R.heslicensee s; work order pro pyQp WINK cumbersome and did not.provided adequate guidance to~ prompt effective $ 3p,maintenanceandoperationspersonnelauthorizedworkona d a #ereview of changes made to work ordersw Furtheriscognizantilicen U k ishutdowrl condition'/ y p g W W W$ operations despite w gymgggM@gymM 9 MwepON Q M d@ WRMMMWJ )Thework~orderpro% ce'ss~continuesto'% L ) that;several; examples of not following established work order . i Mwereipreviously;1dentified during NRCsinspections (Reference NRCWpM y i % Inspection; Report No.950-309/94-03 cdated Aprili7,%2994;iNRC Inspection' M C;ReportNo.150-309/94-21,,datedDecember:7,g1994;IandNRCInspectionO: W ' F1 WReport:NoP 50-309/95-22,$ W M TM A& M # % V $?' dated Oct M %MtQtWMWfnk iWA%5d -kW _Further.,the.11censee'Pcorrective actions for the previousievents did VM 'not.appearsto be' effective.) Specifica11yga Notice of Violation! = enclosed with NRC; Inspection Report No.350-309/95-22. indicated tha(N0V)$ t the licensce's mechanical maintenance personnel worked-out-ofsthe: scope;ofs hi work.on"aisafety-related valve.without approval causing an ortffcesto W umistakenly3 removed;from a high pressure safety!!njection.pumpM9 M ' recirculation.11neNfThe licensee's response to the NOV,1datedbu p Novemberf17;t199531ndicated mechanical, maintenance" workers'and$1 h t g; supervisors were briefed.on the; event:and on the importance#cfAfo11$im,g "19t$4 necessary, to' guidelines :and ~ obtaining proper;; app ng W fwork control, r '3M - d @no : appear,q M p p,qi q h g g ;i h q a mto b deviateifrom guidanceiin the ork 'T A g ; @t pgs $ Workers!falled toiexhibit a. questioning attitude:and ~ exhibited _ poor.. y@p, f fpdjudgementlwhendduring work' on valve CH-138 on February 2931996,ithe. MM izGworkers" continued to' work.onia non-emergency leak ofsthe; valve despite. ? i Nycompletellack of knowledge as to the radiologicalfor industrialssafety' @say 1 4Mdramificationsiassociatedwith:their:workP ~ ^ gfThetitcensee s>wh teitagging procedu%e p/d f % gpA g g p M 494 W p g64 7 b ^.. r rovid l'imi jspecift vg M guidanceiforteffectivel1taggingiof,high; energy; systems. 't fd individual s kinvolved jinitagging were unable t to: reach ta$ Forlexamp consensustasito. I W the!need for whiteitag'on:a valve 1(CH-138)sthat isolated:a high" nergy, 7 source
- '?
2 3A WiEM y 50 RADIOLOGICAL CONTROLS M 5 - se e < ev ew 44 h. Thet nspec a equacy an INness o gc D y . controls 7provided:duringithiPplanning'and performance:of the work on the,ventMph. tailpipe]ofsvalve"CH-138?on; February 129,t1996 M Thesinspectors also reviewed theiadequacy?oflthe311censee's! response to.the;1dentification of ai ~ g radioactivity 7inithe PA8 during performance cf the work.f The: Inspectors Q M y Q 3 isolectively! reviewed thelfollowing-areas:qgMRyggppggfhNh% / - ' ' ^ g \\
.+w, ., ge w m, g n - - m : ,,e a nh -a itial ann ng for;tfe: work activity; m M ,A @7/E radiological, controls 11nstructions for workers MM; f/4 : 4 l circumstances surrounding,the'eventiincluding wo'rke 'an radi fon h g g protection personnel response to;thel event; h;t M yL-@4 applied' radiological l controls ~provided fo
- E66 if e-including ra ation jg pwork:permitaradiologicalisurveys) initial, ion-goint dosfeetry; y
%g Joccupationaliradiation exposure ass (essments.(in g e affected workers;tand : p %'W ~" ""NgMy t-N eventireporting~ w4p ? ' g ?q f WrMege l .m.1 m aw a The:inspeNors performe ifn ependent testingif ventilation conditions at th e l 4 g ! valve location;(using smoke tubes):to ascertain'the. transport <and diffusion of ' (the. radioactive noble;gasesc,and the apparent exposure of woders:to the gases g 6
- The@linspectorsgp%dW~ MP,p4RMwNwMipwnc$@mp%dqtnygsgf j
m review was'against' criteria'containedlinl10 CFR 20,i,' Standard a . for' Protection gainstf Radiation,"-applicabloolicensee procedures 1andtpuW A s PTechnical; Specification 5.1r+" Radiation Protection Program "(and.10 CFR 50.72,~ if + I w "! mediate.Nottfication* Requirements.for Operating' Nuclear, Power' Reactors." Wg 4 0 h kkg h hhf n W 5.2;4 W ya @ g eg%%pg, % fh y R h gy%g M igg g1 h.W M 1 Planning for the Work Activity M g g y g %g#pWwg%g@ dggg g ptMMWfWGMip WSMWav + I;gThe inspectors noted"that t@eilicensee's healt 'p ys es' personnel ha ~ ~er p h ~ dinformed of the proposed work on the tailpipe cf va gN M AFebruary 29,i1996;VHowever, although the actual: work order tN% h Pgroup was/POSSIBLE;VCT GAS LEAK FROM TOP.0F, LEVE Lindicated Mof3the. potential < exposure to high concentrations of noble gases.Y >th physics did perform surveys of the loca gf. R1eak been present at the start of the work,ilt could 1 ave been readily&@g y% radiation work permit:(RWP) the workers we detected by the. potable survey meter.2 The inspectors ~also noted that-the CNH b96-00062. Rev.0) did not;1dentify their specific worker order (95-00622-00,k j Rev.10. P 3 p W a gNg%)gdhC6@inoted1that the: original'WO~is@ su pM Qi S E A i g e @d y @h d y s % h%Wg!Qp WM e WN s s% LastlyRthe>1nspectors pecified!%A Athattthe work was to'be performed during shutdow n, y# $g performed withithe reactor.at about 90 reviewed byv tradiation' protection-supervisor.y.The" work activity was'actuallyd t a Qindicated'a need for, improved, planning @, communication tand atte %when; planning work @WMhgg@gg "ggpMM;jpgW [ AQ"ZL py a W W W MMa gd 5.32.2 Instruction @s; W S S 4 d n g@ gagg% W 6 pf y m p " $ yA toWorkersg# w W MMWt T nspectorsPreviewhasidiscussedIin*SectionT4.2?a ove,lidentifiw dMw if W F. 16
- t t th physicsrpersonn;el,thatithe component?they would
~ ."workerst ', informed (MechaniciAtand[B)[ assigned to perform,the work; activity A pressurizedisystem thatrcontained both hydrogen;be wor WM j and high concentrations. D pradioactiv " noble gasesMFurtherhthe' workers ejot(informed:thattthe ^
- h l
Mb w M h ' %u m#k 14 SMN hN $MD Q W Mw&ay? - A* Yk Q a.f l,
- A s
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.i< Wa.ams W p M @Ny%& eh r6Muh W & & 1 hy Qg w &w:&m d g$w$m$w w% h@hy{M +7, associated with the vent tailpipe they were working'onkw k* ? $m o&m 3 1 @m@ythesingle; iso)lationvalve:isolatingth eW %s i i valve- (Cll-138 d l / j ' environmenthThe: mechanics were; informed to eave the area.i A iwork'scopgoccurrede Eg%gggggqUskp%gg$g j s tailpipe for valve CH-138,t Mechanic e t gggg g g DMMMWM% en NMf d 46 During the work on; h'Jn lk i i ' leak from the valve.gThe mechanics l'did not see a' concern with tight untered a j \\ hthe front;of the valvet facing him-AvalveheThe mechanicisubsequently p sortion of his wrist'around M$ j i Mmechanic; unknowingly exp(osed his pe)rson to high: specific activity' rad to searchifor, t1e leakM Asia'resultpthe % i j $noblegasts;(0.45,microcurieper<cubiccentimeter)sleaking from the VCT gas; i k h g @& The[ inspectors noted,that!!0 CFR 19.12 requiresilicensee'sito! instruct i j in,#among other matters,: precautions and procedures to minimize exposure to g M y -@MMprotectionprogramprocedures(ProcedureNo.9-302-2,Re rkdiation or radioactive materials.t g NM WM i % guidance for< instructing workers on precautions and procedures to mini lyf li f Procedure:9-302-2,;Rev.' 4 be used for this purpose.p Attachment A of theM %s i i 7 procedure specifies.thatLworkers are to beiinformed'to l W Many change,in work scope ^(change 11n~ sositionAlocation,t eave the 1
- method of workVsystemy f$ conditions) Ult!also9 barrier breach,r time to complete wor
() occurred that~ could affect 1radiologicalg i t M required that the workers;1 eave?the work'areatif the workk ) idoesn't go as plannedy }y%y,ggg@bw%ghg$MLadggfgM - W$ $M 'ThetinspeclorsLnoted the:tw$ksodad . AnMMW%bMW@th k t 9 %f i o maintenance mechanics were nottadequate as i j . informed ofgthe*precautioris{and procedures to" minimize their' exposure as; wMg . evidenced by;one mechanicisearching for ailocalized leak of; valve CHil38, W b using his; bare wristhon February 29,11996;mThe mechanic was unaware thatit e$e$ 5; d valve was a' vent? valve from the VCT and unaware >that the:gasileak encounteredq8W s y 6d?(i.escontacting his bare wrist)the two mecianics were? W emanating from'the VCT g further, was high s)ecific'activityinoble gasgMWh } , 9mechanicf assistin;g wastalso unaware of the content;of i g M 19.12((VIOj50-309/96-04-03) g g % Thistis(antappa Jpresence of. highly' radioactive gas' j k f W M M fh % M W % G M W n & @% DQNpg W Jgg g h; yP ) T %mf p(Th'MWDM@l Response *and . Mtwo y %f y5!2:3jorker Personne ! Response to t e Event ,~ 3 } i gsM yW $W there was' generally e i radioactivity was detected in AM j ? 'sthe. Primary Auxiliary Bu11 ding PAB) elevation.36MThe building was evacuated ' by,, health) physics!personnelland(workers were restricted,from further i g/g In: addition;tthefentrancesito;the building were posted as^ airborne @iWWT D, h, radioactivity.areasWFurther;theli C i affected workers M !nTaddition.1the licensee; initiated a radiological m f? B 1 Mechanic!B) sho* passed his; hand through;the gas Np d@The:11censeaf also reportedithe event to the NRC@; emanating from v i Further :the.licenseeiissued h i 'a memorandum (to;al.1; Maine, Yankee personnel /on March IL19966regardingMWh i expectations 1when! working lwithin the restr cted. area and; issued aLstallar g# hhh m _m. y.... m o - -.,@_NL n,,N,,D,* N E
_ -, ~. l .emn, ny$h.h % cd;,m Ng m.y s ewh l up,ym#q-u n m wpW:hm;p & g<.g,ggwP g*1 memora 4 l q 2; S6%M Md m%&,.w 4 My G M Ah el"on Harch'1;i1996.D " M g$ @ M R X A-poor MM h]$ work practice" was also issued for the two workers who did not k N@$g%RWP,to work on' valve CH-138 on February so issued a d W M D g }e M[Notwithstanding these actions,'and as' discu z@.m4segsmyJ/,e. " mf yn CQ $@p tion kh A[ good response to the identified leak as follows.Ma ,..du H; ? p;%@.;.gWwy:ker (Mechanic A), as discussed in Section 3 above %e sgh @; m.h'.e wrpm MW yg W JN?4 t 1 W 0ne wor . % : ~ to be extensively contaminated (set the frisker off-sca ?
- 1. g gb N'
d 8 Mvw pJN9F nevertheless re-entered the PAB to go back to the are indicated he believed that the elevated readings on the frisker may be-7 MWk l 4 coming from the die set near the frisker, however he did not appear to V T T-lhMMj&% W; display good judgement re-entering the PAB when he believed himself to ' W - $p $ be contaminated. Further, it appears that the worker did not adhere to? (g.h@$p%F l f <donning of gloves and shoe covers when found contaminate ~ J $g
- o s v f
? JW One worker (Mechanic B), as discussed in Section % -3 above, displayed poor;'q :x .e' % Q h p W judgement in passing his hand through an unknown gas emanating from a d A,
- pgqlpi 4 breached system and failing to inform health physics i
fv?p h h N i that he was verbally provided. 4, 4 g-gggygg]QQ?( s myo .g + n. J4ma n e 3~ k k, wp%&The inspectors had.the following additional negative observation. P MMM@ m, .g 'wgypx ,e e &c 4 4y 4 : + 1 QMp$M& y se.. r r., u. r. Q;WE'i[cTheinspectorsnotedthatonceitwasdeterminedthat'onemechanicmayc4 ~. ? W s n m DS' 'have received, on February 29,1996,* an unplanned exposure to his persori@%%' QJkk ~ from high specific activity radioactive noble gas, the individual was sg A- $ p & M W sent for whole body counting on that day.. However.1the individu dsk, MA W M7 h. access to the radiological controlled area. was not restricted in a W M'
- timely manner pending completion of his dose assessment.;
- Procedure'NoF %
i SM:h 9-307-3,: Revision 3, requires in Section 5.6.5, that for potential over . exposures or unplanned exposures,. the affected int' MNG ' M9M placed on the restricted list in accordance with Procedure 9-307-4.x The v' + D M f $ K individual was not placed on the restricted list'(i.et,'no'radiologica1 0 VJ
- E W C f X controlled area entry) and re-entered the radiological controlled area ?
y4 kM 3 % g y g: +L : shortly after the whole body count on February MQMN _. individual's access was subsequently restr 29, 1996.1 The % V g ptM 3 'following issuance of a " poor work practice"g on this individual. G $ $ Mg h @M @p h i,r: , mmy . v. dThis observations:g~ gy ,, indicates inattention to exposu& y n ' s 'y2 W ' MQ :. yi B; re' restriction guidanceWe; y Q%%@fhng&.xp m M M mnMwen? during dose assessments'an qq# OMM s % V%WQ%; QF hh&p%2np$a$$;g&;W M:&h6Md p&MWMW?% 1mf D m$ f f $ h k h ?;$ W h h h N N-$0 $k ym~,,,,h N0Y ap n w, e wwwg ~ m pm wa ws + g,c c. ~ L, t nggw w ; s., t n n c 4 gs. e, - ) o, Q e t
- 2 s
y .g 4 4 ~ } E^ ? wos. m.. w'w w w J:% #,,wy'3 m m p?3 ? a m e,- hk, ) h ~ Q'%srw a%, m;q qp4;M(w M,. c A. >m2 kk,h i y' .[ A ' [i = n xq - n 1, - h ~ g~ ny5:.7 e ss ,n n .M [* N tM*) N i t.fK $i*h Y; *,- ? \\ U[ .N [ / y fiQM:@&kUs% &l O R; id: ,q ._ 1 ,;n ,hi ,t a .y n .} 3 h N Y r YO Y a.~
2 aliRadiathn Exposure Assessmentsi(internal / externa " exposure S 5.2.4~0ccupat i mk uaa;no?!6findiv!" duals.were identiffeb x ,Thet nspecto review:indicatekthat k ? i contaminated byithe'radioactivejnoble gashincluding the two' maintenance Mi r mechanics who.weresinvolvedtin working on valve CH-138 7The individuals:were .MM j pidentified:as? contaminated during-their;attempttto exitcthe.radiologicalii%3 ~ ph deontrolledtarear ' g subsequently;rema(RCA) Vat the health physics control-point.t' The-individ j ined!in thefarea while the short-lived radioactive noble' gas a. y Q i g' particulate daughter l products decayed away and alliindividuals we M exit:the RCA;in.less;than,about,two hours without; alarming personnel i contamination' monitors. p. d 4 ;d m e& $ @ M b LM m N hM 3 s ? M WQM %;M &dMW W W f M(Subsequent wh51e body; counts.cf all individuals did not identify j [ k radioactive noble gas with the exception of Mechanic BJAThis' individual @MN $1ndicated some. apparent; uptake of radioactive noble gas.OSince regulatoryQBMM A11mits for, personnel exposure to noble gas are. based on external expo.sure tom *N i 1 ~i $3 the' gas %the licensee'used the whole body count, data to estimate l potential % or
- l l
dexternal. exposure to.the body.dThe licensee compared this: data with exposure @yt c : Qindicated by personne1 dosimetry (i.e.,;TLD a 4 PW WMN 3 dp;W MyMw yRegarding/ external; exposure 'of th l %ppocket dosimetry; indicated no external-exposure to the whole body., TheP 4@M 1icensee also processed the whole body TLD badge of. Mechanic;B and concluSfed% ENd d %that no'significant external, exposure occurred MW N %#:id$04/MNWupsu6M@ discus #dd.pp p@eg@$g - % ' MM I !h l5N bh qMN i ? ,The inspectors noted th R based on y jyy i y% performance,of smokeitesting'at the valve locationh radioactive noble rJas?by, Mechanic B was very;11mited and non-uniformA $g6 1 { Specifically the tsposure appeared to be. isolated to the;right;wristVand th' t tips of4the left fingers $The: inspector questioned the dose assessment;for 'go 1
- this individual based on these'observationsW ^%fM q 2;The licensee' subsequently performed calculations and g aff d;5 e th c.WG@
1 f i N bE(i.e.,'31ess than a maximum off about 200' millirad to.the? extremity ~ y i w dnd f glicensee believed that;the,TLD recorded ~dosensincelanuary!!Gl996M(IM!W@s kf g tf Wl$jWapproximatelyil46 millirem,toithe wholeibody r of the! eye)tadequately; represented.the' exposure lof the' extremity;and wholdi A 5 1g$n . g t 50 rem'and that;thelindividual's= maximum, potential: exposure;w sbodyntThe;. inspectors noted that the NRC exposure: limit:for theLoxtremitylis b i 4f l h i thisilimit%The' inspectors'Jfurther noted that the maximum potential ex g to the.~ extremity;was well within the 5 rem limitjrequi d $ ' N monitor [p ]e h_ p M,$lgK i P ng ff.thejcxtremity' sy;gggpDp p ' Mthw D Qg )4 q, w j %g.adiologicaNControl,(CoiE1usions);y[& q 3 R p gge ignifican "externalf@or.' interna "occupationa ' g gm pf Thelin~spectors4conciIidid no , expo sure: occurred !to;the}the Wal vel CH-138.Wgg indi idualstaffected byithe noble' gases l including r individual lwhgtibhtened . p* i u' W @ g& Q m % Ql5 d d @ M $ +@n@ M-@ m g W PC~m&M n ~ dWmwn gaMu M C+ u n w ?$WU eMM M M b A& & Ng a n w w@A ,A QQQCf' & %Mib@id% Dsd MM&iM
b f .hy &Q &nu mi I Q &,& h$$5by % f @ Q@L k M @, & ls p$ & & Q h Q Q ' W Q H .~ Qj N Q Aw a e* Ap %,p &M dg y4 .w Jio/ d d M f M N j $g dn & y W}4 3 4 G,A4Q 4 h ge n ~ wever weaknesses.in. planning M $ M M T M h N Mg g y w s a 1M ' MN 2MfM that p5h adequateinstructions5regradingprecautionsorproceduresto!minimizethe the workers did not; sign;the' applicable radiation work permit:or receive $ N 1 I ec Ma 4 ioccupationaltexposure M Thisilatterqissue wash dentified as an; apparent T @ Jh M $ Myiolation of 10 CFR 19l12gqg -@Q((wfdM.N@@MMMMK$f"M%ggg%gg g Mg?/5ggg%$ g/h d g y h M, W A h W E D Q W.l W i W y b W M. N W W M W h W.Nf [r/A
- @YMN R@Em $M%$yQ$M b
MC g['&ph.wmse%werph.D@C 4 6.0 % EFFLUENT MONITORING AND CONTROLO r D9 fn ZW 7 U W W % "W" @@f W 6.1 T General * (Scope of Review)MWW@9 WSWPDWPMM bbf L' v%wwwwww MMMGeMW94 M 1ThhinspectorsreviewedtheFebruary 29,1996/ unplanned release ofe4W;M N 4 jradioactive noble' gases' from valve CH-138,tte the PAB,tand-subsequently y s ? tenvironment via-the monitored Primary Vent Stack (PVS)MThe; calculated Mg g )9 {yl" Standards for Protection!Against-Radiatio i, environmental: doses;resulting from the release were. compared to applicable flimits and criteria' contained}in Technical Specifications;410 CFR Part(20, t $ l Q x c 1" Numerical: Guides for Design Objectivesiand Limiting Conditions for.0perationf NS 9 4fMaterial:in Light-Water-Cooled Nuclear Power Reactor: Eff $7 < M 9@?Ql2WM%%Q%%%CW%$&j&)q!gMNh%$ghh:q&pMt;)hWF s and the'0ff M % % aSite Dose Calculation Manuale(ODCM MW P M $gindependent; smoke;testitotascertain; performan i M NQ % psThe: inspectors'evaluatedilicensce's conducting an Q
- the direction and dispersion of,the~ gases
@p Mto the' PAB;$ independent calculationsito' confirm the 4tiantity;of; radioactive 4 M a Ml gases (released,*and the} dose;totthe'public and environment;' revie G N% h t applicable procedures'and documentatlon;: discussions:with personnel;land?$kW%M %dobservations;made during tours;of,thelPAB.g w@%%!@; 2 li 0bservations ' alnd ' Findin gs % WM W MCt% M hM% Q Q Q $df f Mf Q s6 h 3:C$gM ,, N b Q: s u g si pR mmmyspg se WA MMThelinspectorsVsmoke test of PAB general l area ventilation' flow path ff Mthat the radioactive'gaCimmediately flowed out of the VCT valve."away fromM M d gWand-to the:right'of. theiorker:FAs the gas moved toward the righte f hy approximately twelve. feet labove'the-floor nthe air flows on the 36 foot VkVW M ed e(levation of the PAB: dispersed the gases an)d mixing occurred resulting in fW WRM~ t MMsignificant dilution'of the noble gas.nBecause of several inputs of air:(at%@ F yhleast fcur), the noble gases were mixed into the PAB almost immediately andfW $ 6 ~ 9 g$wMradiation monitor. reading)s esubsequentlyivented to the.PVS.: As.the noble gases vented to the 9 f. Were recorded on the plant computer and.the strip MVchartM%g%QWW'been' terminated 1 chemistry pe 7 &%dm Wd%vgWtum:QM W. R $lWPW Ni: $'ESCM}the"releasehad 6r
- dMMD6,
'After samp1 R from the<VCTiand analyzed the: sample via gam y gas m p 3 pos
- theiconcentrationlof the dionuclidessin;the noble gas mixture' released to R Q
.tolthe e& The;licens;e wa.ra 9 the1PAB J sitheniable to" calculate the total; activity releasedhpd b e nvironment based.on;the! volume,of gases;releas;ed _from the' VCTsto;theQgg y o; PABrand the; total 1 concentration ofathe gases from an thiidatabthe"licensearcalculated the totalfrelease;ilsotopic?analysisMUsing M M e h $ w @ M M $ n % n% e %m @ m% % torbe0.16 curies (C1) h-a fd Tf h MN ww m n % g r e%ggg;W,pgqq% g@g">ggN.W lpf(pan c r m ~u@ w w m. p wW s ~N Q,VwMan 1 g.. Q 4Q p p_g W f 4Mm k WejS WG M 6 b4 h a n e ^ nkyp;ws Aww dy ggy- - g3 l lhdh[hk hh!k 1[ y y[ y aygtpcvs[.g g/n y(gg $ 3 (4 th k $ MNMM M -%h $$[ ~ s; l Y h
--n . j kfurther refine,sthis$$dhh&e"kn5dW$$8WNWk&& N y? li Phrate;at the's, radiation ~ monitor.%The licensee d T icalculation,Jih licensee used the information obtaine %gfromthePVS N d 6 PVS radiation monitor: strip chart, the sensitivityfof. the radiation m ite boundary by calculating the:totalicounts per minute from the % hbased on the primary calibration,<and the calculated flow rate.t From theseMg [k 4 & data,;the tota 1' activity released was 0.19 C1.x This:is in agreement with_the p M e d%gcalculated quantities lof noble gas radioact ( initial calculation of 0.16 C1.x The' inspectors attributed the difference inMM h@@dTheiinspector; reviewed th r M$ di f ferent : cal cul ati on ~ methods ~and. di f feren t e sources #o f data 3 The di f fe k$ in1these:results was-not*si nificant,iconsidering the magnitude of release.MNM i Af$q\\ Bas $ed on(the$$$#knviMMwdQiM& l n-l 7;.- inde 3If& n@ pendently verified the results.W@qWWhtota Mu l %2Q beMc%fM N ( se Q d 'k %effluentito the Whole Body and Skin of an individual at or beyond i id 1) he jarea boundary using the computer model GS2.gThe results were com applicable, regulatory requirements and limitsNand are listed in Table IngEbgj j i gh During:a',$ s M @in@spection P(See NRC Inspection; Report No.150 MWMt hM W gh M&MWN%%A M y a previous-12p % M h independent verification of the g!! dated July _9,11993);lan, d doses.to the public resulting, licen e for calculating projecte from discharges;of A g g 64 radioactive liquid and gases:to the environment, included-an assessment of thed vW i %$11censee's GS2 computer mode 1WThe res hWM Winspectors', PCDose computer; mode %.W@n@Wyl% W9E l f hp 9 ]& $ Y y @ y &MQ5 Q $$($Acalculated Dose' Ra W MygM@@ @g?W y M@WW N sq n $[4 $4gA"kg#0.37!mre A D MPercent f Li dth C ggmsw>e wmm w WP9Bkpt T ' ',9 skin $W0.66'mre&m/yr$wmp3000 mrem 7 WB W R 500 mrem /yr3 f 0 i 4 \\ $r amqww myw gg ' y m M & p m /yrip p igg f f 9 M p m p w wg e s p 0.022% g p ii Q Q U gyp m yr ? ; MNygnMMMN4M WMIMEM @34 Percen%hy Q QMJp vfMCalculated" Air Dose AirDoseLimit"? 5 ttof Linttgg@b$MMMM + b ' R. W@ % d pd W q ip ga e!i Be &mma % 2.lE-5' mrad /yr W % I0 mrad /yr g E . LGa p$&ME)e$Nh5W Ni i WW 6 ghii$$MERiM$$$MhgM$fMk@MMN ta 1;68E-5 20 mrad /yr o g 0.00008% A W M Wm, b$f 2 f(10-CFR'50iiAppendix<l e " h E M W M Q h M @ P dgY ? % 4 p y; &p F p k % p g & % & f % Q M f %jdhole Body g g g )y~ 3 3 p E QMM MW ha &QkfQQ nM y T 6.'4 4 Conclusions ~ % rap %k%C ove rev ew 'the inspectors @ determined D M 4o VmW 4PMMMM dMMthat the'J $s l aBasedon.theIMdWyd4R. assessment w l i Kvalid?and correct;i~no'regu,latory limits had been exceeded,?and that.theeMP W i dM11censee'siresults demonstrate that offsite. doses,and _ dose l rates were well Mg l 1 $withinLallowable limits;ggtgyQp M +h ms@w+ p p Mrp wM W WMV@W@W@m@remw, M dW w+,M,. M) #y&1LfW yy %gfg, p w'Qhb;X,.WNDhMMW u, mae m n.a. w ~ c a ,.w w l Aj yQWf Qy' "%. i, 7p M i &MQ ^ MQ~l 5.;Q ' 2 gMMM pM WMo l J@M%h.MM F d!$vsr% # e WW@WMW @NN[MWV%T m l e, f kh "b ~
1 {y. ,f- , + =, EU ' e, v 3,s n p . Ly = e y y h I 3 [. j[ , t %a ,be,f-s l6 e wy 17.0
- EXIT MEETINGS
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- f#a s
The' inspectors rnet on Marc 3"and 8,$ 1996,~ with licensee representatives; e u + E Mdenoted in Sectionil.0 at the conclusion of each portion of.the inspection, s 1 L The. inspectors summarized the purpose, scope,-and findings of the inspection p dThe; licensee acknowledged the inspection findings and generally agreed with, t _3 ,) 'the conclusions.% The licensee indicated that an independent Event Review t 7 Board (ERB) had been established to review this occurrence, to establish d y as? 9 causal factors, and to recommend corrective actions directed to the speciffc;, " ] > A and generic weaknesses ~and deficiencies. The licensee agreed to make the q 4f" ~ findings of the ERB available for NRC review upon completion of the effort W e w u h b ks Na ,; _y pap _y _wwg _w pym a x$ . ~ pa c m t ,~ n n s ~ suw e~ w na 3 7 m(W - L 7/;U40t4 g. 7Q ST49 J m + n - g[.GR cf~y ? WMN 54j. 4. N f " M i@U A;veN, , & a @" W U$ Mjyp gsg;gh? t ji : U ?% 2&j 4 "q@w? "D&& MV - bt ikf %QQ% %, >4 %. Q4f; s fgW 'y Q a bfYQnq w&e g%; g WD AWhi aw gm@/k@ # ? ' 7, x %yh,, tygNg: m; r -
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