ML20137B623
| ML20137B623 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/13/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137B593 | List: |
| References | |
| 50-309-96-14, NUDOCS 9703240014 | |
| Download: ML20137B623 (69) | |
See also: IR 05000309/1996014
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U. S. NUCLEAR REGULATORY COMMISSION
' REGION I
Docket No: 50-309
License No: DPR-36
Report No: 50-309/96-14 I
Licensee: Maine Yankee Atomic Power Company (MYAPC)
Facility: Maine Yankee Atomic Power Station
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Location: Bailey Point )
Wiscasset, Maine
December 8,1996 through January 25,1997
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. Dates:
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Inspectors: Jimi Yerokun, Senior Resident inspector I
i Division of Reactor Projects I
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l William Olsen, Resident inspector i
Division of Reactor Projects
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Richard Rasmussen, Resident inspector
Division of Reactor Projects
Edward King, Security inspector ,
Division of Reactor Safety l
Randolph Ragland, Radiation Specialist
Division of Reactor Safety
Larry Scholl, Reactor Engineer
Millstone Special Projects Office
R. Croteau, Project Manager
l Office of Nuclear Reactor Regulation
i Approved by: Richard Conte, Chief, Projects Branch 5
j Division of Reactor Projects
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9703240014 970313
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PDR ADOCK 05000309 j
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EXECUTIVE SUMMARY
Maine Yankee Atomic Power Company
NRC Inspection Report 50-309/96-14 ;
This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a 7-week period of resident inspection;
in addition, it includes the results of announced inspections by regional specialist inspectors
in the areas of radiological controls and security and the cable separation problems.
Operations
Maine Yankee properly maintained and operated safety related systems such as the Residual
Heat Removal (RHR) system in accordance with station procedures and as described in the
station Final Safety Analysis Report. Station operators were knowledgeable of plant
procedure and technical specification requirements for system operation.
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Operations department management were properly focusing on and tracking operator
workarounds in an attempt to keep the numbers low. However, a number of operator
workarounds still existed and continued management attention and resolution is necessary.
The decision to perform the fuelinspections during the current outage was an example of
conservative decision making by Maine Yankee. The reactor disassembly and fuel sipping
operations observed were performed in accordance with procedures and were well
controlled. There were good controls for lowering the RCS inventory. Operations personnel
tracked important plant parameters such as RCS temperature and level and RHR ,
temperature and flow. Good use was made of the plant computer in selecting appropriate
alarm setpoints for these parameters to give operators early indication of a potential i
problem. However, a weakness was noted by the inspector when operators did not select H
an appropriate setpoint for the RCS temperature when RCS temperature was at about 108 i
degrees. The computer alarm was still at 180 degrees and was changed to 120 degrees l
when this discrepancy was pointed out by the inspector.
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Poor performance was noted with the inadequate analysis that was performed in response
to the multiple failures of the spent fuel pool crane, CR-9. During the period from
January 7, through January 13,1997, the crane failed to function properly on four
occasions and each time fuel movement was resumed without formal documentation of the i
problem and a technical justification for continuing. In one case, irradiated fuel was used as
a load to troubleshoot the crane. This was found contrary to the requirements of Technical
Specification 3.13 and procedure 13-2. (VIO 50-309/96-14-01), Section 02.3.
Maintenance
Safety related equipment areas were maintained well. Material condition of systems, and
equipment was also maintained well.
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Enaineerina
Test personnel showed good technical capability and demonstrated good safety perspective
during their testing of the HPSI pumps and valves. Preliminary reviews showed that the
test was accomplished satisfactorily. Flow, pressure, temperature, and vibration data were
as expected. However, detailed review of test results is still ongoing by Maine Yankee and
the NRC and this item remains open. (URI 50-309/96-14-02)
Problems were identified in the areas of cable separation and logic testing surveillance
procedures. Maine Yankee was in the process of developing a plan to determine the extent
of the cable separation problems and was nearing completion of the logic testing reviews.
The initial scope of reviews directed by engineering was relatively narrow. However,
management intervened and directed a complete walkdown of the hydrogen analyzers.
(URI 50-309/96-14-03 and 96-14-04)
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Plant . Support
j in the area of plant support, we found that you continued to maintain adequate programs in
the areas of occupational radiation exposure and solid radwaste management and
transportation.
Notwithstanding, weaknesses in the contamination control program persists as evidenced
by the December 4,1996 discovery that personnel had been exposed to a discrete
radioactive particle in a chair used by Security /Firewatch personnel. Further, our review and
your own investigation of unscheduled gaseous releases appeared to have been hampered
by the lack of a clear designation of responsibility for, or ownership of, the overall
performance of the program and the summary of issues related to unscheduled gaseous
releases. In another instance, poor radiation work practice was observed when an
individual reached in and out of a contaminated area handling clean equipment outside of ,
the boundary. This was contrary to technical specification 5.11.1 and procedure 9-5-100,
Contamination Control / Decontamination Program, Section 7.5.1. Although the procedure
allowed for personnel to reach into contaminated areas for certain purposes, it required that i
personnel remove protective clothing prior to exiting contaminated areas, (Violation of )
TS 5,11.1) (VIO 50-309/96-14-05), Section R4.2. j
in the Security area, Maine Yankee maintained an adequate program. Management support
was ongoing as evidenced by the procurement of portable explosive detectors for personnel
processing, compic. tion of the vehicle barrier system, hiring of four additional security
officers, and the installation of permanent electrical outlets around the protected area to
support temporary lighting needs. Security training was being performed in accordance
with the NRC ap;. oved training and qualification plan and management controls for
identifying, resolvt 1, and preventing programmatic problems were effective,
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TABLE OF CONTENTS
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TA B LE O F C O NT E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
l. Operations..................................................... 2 ;
01 Conduct of Operatio ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
01.1 General Comments (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
01.2 Operator Workarounds ............................... 2
02 Operational Status of Facilities and Equipment ................... 4
02.1 Safety System Walkdown - Residual Heat Removal . . . . . . . . . . . 4 I
O 2.2 Containm ent Tours . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
O2.3 Operation of the Spent Fuel Crane (VIO 50-309/96-14-01) . . . . . . 5
4 O2.4 Fuel Sipping Operations .............................. 7
07 Quality Assurance in Operations ............................. 9
07.1 Plant Operations Review Committee ..................... 9
, 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 i
08.1 Review and Closeout of Licensee Event Reports . . . . . . . . . . . . . 10 I
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I I . M a i nt e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
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Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
M 1.1 G eneral Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 11
M2.1 Safety Systems Area Inspection . . . . . . . . . . . . . . . . . . . . . . . . . 11
M8 Miscellaneous Maintenance issues (92902) ..................... 12 4
M 8.1 Review and Closeout of Licensee Event Reports . . . . . . . . . . . . . 12 l
l Ill. Engineering ......................... ......................... 13
El Conduct of Engineering ................................... 13
E1.1 Testing of HPSI Pumps and Vaives (URI 50-309/96-14-02) ..... 13
. E2 Engineering Support of Facilities and Equipment .................. 14
E2.1 Loop Isolation Valve Torque Switch Failure . . . . . . . . . . . . . . . . 14 i
E8 Miscellaneous Engineering Issues ............................ 15 l
E8.1 Electrical Separation Issues (URI 50-309/96-14-03) . . . . . . . . . . . 15
E8.2 Safety-Related Logic Circuit Testing Update (URI 50-309/ j
96-14-04) ........................................ 17
E8.3 Review and Closecut of Licensee Event Reports . . . . . . . . . . . . . 18
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I V. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 19
R1 Radiological Protection and Chemistry Controls . . . . . . . . . . . . . . . . . . . 19
R1.1 ALARA .......................................... 19
R1.2 Stop Wor k Authority . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
R1.3 Video Monitoring of High Radiation Area Access . . . . . . . . . . . . 21
R1.4 Solid Radwaste Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
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R1.5 Radioactive Waste / Radioactive Material Shipping Program . . . . . . 23
R1.6 Implementation of the Revised DOT Shipping Regulations ...... 24
, R2 Status of RP&C Facilities and Equipment Tours . . . . . . . . ......... 25
R3 RP&C Procedures and Documentation .............. ....... .. 27
- R4 Staff Knowledge and Performance in RP&C . . . . . . . . . ........... 28
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R4.1 Knowledge of Radiation Work Permit, Radwaste Shipment and
Duratek System. . . . . . . . . . . ......................... 28 ,
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R4.2 Poor Contamination Control Practices (VIO 50-309/96-14-05) ... 29
R5 Staff Training and Performance in RP&C . . . . . . . . . . . . . . . . . . . . . . . 30
R6 RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 31
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R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . 33
R8 Miscellaneous RP&C lssues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
R8.1 Unanticipated extremity exposure occurred during Filter
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Replacement ...................................... 36
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R8.2 RIR 96-016 Discrete Particle Exposure in Chair Used by !
Security /Firewatch Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 )
R8.3 92904 " Followup Plant Support" . . . . . . . . . . . . . . . . . . . . . . . . 41
R8.4 Unscheduled Gaseous Releases . . . . . . . . . . . . . . . . . . . . . . . . 43 {
R8.5 Mis-characterization Contained in ERB-010 . . . . . . . . . . . . . . . . 46 *
R8.6 U FS A R Re vie w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 j
S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . . 48 j
S2 Status of Security Facilities and Equipment ..................... 49 l
S2.1 Protected Area Detection Aids . . .. . . . . . . . . . . . . . . . . . . . . . . . 49 !
S2.2 Alarm Stations and Communications ..................... 49 l
S2.3 Testing, Maintenance and Compensatory Measures . . . . . . . . . . . 50 '
SS Security and Safeguards Staff Training and Qualification . . . . . . . . . . . . 51 !
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S6 Security Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . 51
S7 Quality Assurance in Security and Safeguards Activities ............ 52 !
S7.1 Effectiveness of Management Controls . . . .. . . . . . . . , , . . . . . . . 52 j
S7.2 Audits........................................... 52 l
S8 Miscellaneous Security and Safeguards issues . . . . . . . . . . . . . . . . . . . 53 >
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S8.1 Updated Final Safety Analysis Report Review . . . . . . . . . . . . . . . 53 l
S8.2 Review and Closecut of Security Event Reports ............. 53 l
F8 Miscellaneous Fire Protection Issues .......................... 54 ;
F8.1 Review and Closeout of Licensee Event Reports . . . . . . . . . . . . . 54 l
V. M a n a g e m e nt M e eting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56
X1 Exit Me eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 ,
X 1.1 Cable Separation Debrief . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 )
X1.2 Security Exit Meeting ................................ 56
X1.3 Radiological Protection Exit Meting ...................... 56
PARTI AL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 )
INSPECTIO N PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59
ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60
LIST O F ACRO NYMS U S ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63
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Report Details
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Summarv of Plant Status i
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Maine Yankee began this inspection period in cold shutdown condition. Plant shutdown j
from 90% power was made on December 5,1996 to address inadequate cable separation !
issues and to investigate and address fuelleakage that had been detected during operation.
Confirmatorv Action Letter. CAL 96-015
On December 5,1996, Maine Yankee conducted a plant shutdown as required by plant .
Technical Specifications due to the manual reactor trip pushbuttons for the reactor i
protection system (RPS) being declared inoperable. This problem was identified as the !
result of on-going plant reviews involving NRC Generic Letter 96-01. Several cable
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separation problems were identified which affected four of the eight reactor trip breakers
(RTB's) and all four of the manual reactor trip pushbuttons. Following the identification of i
these deficiencies, a broader review was conducted and other cable separation deficiencies l
were identified. Maine Yankee notified the NRC that they were developing an expanded
review process for field verification to more fully identify the extent of the cable deficiencies
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Based on a telephone conversation between Mr. Graham Leitch, Maine Yankee Vice-
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President-Operations and Mr. Richard Cooper, NRC Region i Director of Reactor Projects,
the NRC issued a Confirmatory Action Letter (CAL 96-015) confirming Maine Yankee's
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commitment not to restart the plant until completion of the following items:
Initial review of the Generic Letter (GL) 96-01, Testing of Safety Related Logic by 1
December 31,1996, as committed to previously. '
Development of a plan and methodology for expanding the review to determine the !
extent of the cable separation problems at the facility, and to ensure that all safety ?
significant cable separation problems are appropriately dispositioned or resolved.
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Performance, concurrently with the two items above, of a root cause evaluation that
will address all hardware deficiencies identified and validate the comprehensiveness
of Maine Yankee's corrective actions.
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Meet with NRC representatives to discuss the results and conclusions, as they
pertain to the actions described above, and to gain the NRC Region 1 Regional
Administrator's agreement that Maine Yankee is sufficiently prepared to restart the
facility.
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Issue a final written report on the root cause evaluations and corrective actions at a
date to be discussed at the meeting.
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l. Operations
j 01 Conduct of Operations
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l 01.1 General Comments (71707)
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Using Inspection procedure 71707, the inspectors conducted reviews of ongoing plant '
i operations. Operators performed observed tasks in accordance with approved procedures. .
i Good management involvement with addressing the numerous " operator workarounds" was ;
evidsat. Control of evolutions were well coordinated such as with the fuel sipping 2
activities. However, activities associated with operation of the spent fuel crane were '
j initially conducted without adequate corrective actions being taken to address the problems
l with the crane.
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01.2 Operator Workarounds
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- a. Insoection Scoce (71707) I
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! The inspector reviewed Maine Yankee's processes and controls concerning operator
! "workarounds". The inspection was focused on those conditions that completed the ;
j normal operation of plant equipment and are compensated for by operator action.' l
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J b. Observations and findinas
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l The inspector reviewed Maine Yankee's operations department memorandum to the {
- Plant Operations Review Committee (PORC) concerning operator workarounds. In '
j addition, the inspector reviewed NRC guidance concerning the subject and conducted ;
a plant walkdown of the listed Maine Yankee workarounds to assess their effects, if i
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any, on plant safety,
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l Maine Yankee described the process to track and assess " Operator Workarounds" in i
- station procedure 1-2OO-10,." Conduct of Operations". The term operator ,
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workaround was defined as " Equipment deficiencies, including design deficiencies
l which can impact operator response during a transient or accident" (i.e. ;
j implementation of an abnormal or emergency procedure). '
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Station guidelines that have been used to determine operator workarounds include: I
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- Loss of or unreliable indication where comparable indicators are not readily l
j available and loss prevents monitoring a parameter needed for safe operation ;
or results in repeated monitoring from a remote location. ;
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! A nonfunctional annunciator or alarm.
Any deficiency that requires additional operator actions to be performed during l
i transient or emergency conditions which might degrade perforrnance by ;
delaying of inhibiting actions designated in AOP's or EOP's or annunciator
alarm response actions. ,
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l Any deficiency that requires a system normally operated in automatic to be +
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- Any plant design deficiencies which cause an unusual burden to operators
(system does not operate as originally designed due to design flaw or
oversight).
Maine Yankee developed a computer data base to track Operator Workarounds which
included the following items:
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All work orders designated as Operator Workarounds. :
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All design deficiencies that require compensatory actions and are not covered
by station work orders. i
All NRC Independent Safety Assessment Team identified Operator
Workarounds. ,
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The majority of the items on the operations department Day to Day Concerns
List. i
Known Procedural Operator Workarounds (Known). !
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The Maine Yankee Assistant Manager of Operations conducted a review of allitems !
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~ tracked in one fashion or another (approximately 100 items), and determined that
only 22 items met the station definition for operator workaround. Approximately i
50% of these items have procedural guidance initiated for the deficiency, most of .i
. these were in the form of EOP and AOP procedure changes. The EOP changes were
subjected to a validation and verification process to ensure that EOP was not 3
adversely affected by the change. :
The inspector reviewed the items and found that Maine Yankee was conservatively ;
tracking operator walkarounds. There was good effort being placed on identifying, !
tracking and fixing problems. For example, on a weekly basis, managers discuss the !
status of all open workarounds. A goal of not having' any older than 30 days was [
established.
c. Conclusions l
Maine Yankee operations department personnel were properly tracking operator
workarounds but a number of operator workarounds still existed. ;
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02 Operational Status of Facilities and Equipment
O2.1 Safety System Walkdown - Residual Heat Removal
a. Inspection Scope (71707)
The inspector conducted a walkdown of the Residual Heat Removal (RHR) system to
verify proper alignment for the current plant conditions and in accordance with the
applicable plant procedure, plant technical specifications and station final safety
analysis report.
b. Observations and Findinas
The inspector verified that the system was aligned in accordance with station
procedure 1.13-1, RHR Startup and Operation, by verifying the major flow path in the
control room and locally at accessible values and instruments. The system
components in the spray building including the pumps, heat exchangers, valves and
instruments, were verified to be properly lined up and in operating condition. Pumps,
motors, several valves and instruments were inspected for cleanliness and signs of
deterioration. No excessive boro 1 buildup was identified on any RHR system valve
during the inspection. The RHR system valves in the reactor containment were also
verified to be in the proper position and in good condition. Piping hangers and
supports were visually inspected for proper alignment and for any obvious
degradation. The inspected areas were also checked for ignition sources and were '
found to be free of these types of hazards. During the inspection, no scaffolding,
ladders or temporary equipment was found that would interfere with the system
components and prevent their proper operation. All major system components were
properly labeled, and properly lubricated. The ventilation system for the containment
spray building was in the process of undergoing a plant modification to install a new A
upgraded heating and ventilating unit (HV 7). A contingency plan was developed by
Maine Yankee to monitor the containment spray pump area temperature on an
increased frequency to ensure proper cooling to the RHR pump motors during this
period of operation. Freeze protection measures were verified to be in place. All
electrical circuit breakers were found in the required positions. No abnormal
conditions were noted during the period of inspection. The inspector did not identify
any deviations between the plant FSAR and current system ennfiguration or method
of operation.
c. Conclusions
Maine Yankee had the RHR system properly aligned in accordance with station
procedures and as described in the station Final Safety Analysis Report. The station
operators were found to be knowledgeable of plant procedures and technical
specification requirements for system operation.
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O2.2 Containment Tours
a. Insoection Scope (71707)
The inspector toured the reactor containment building on several occasions during
the period. These tours were to verify that station radiation controls were in place to ,
properly control outage work and to inspect the " aterial condition of the reactor - j
containment building. ,
b. Observations and Findinas !
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The area was maintained well with radiological controlled boundaries clearly defined i
and maintained. Postings were up to date. Access to the containment was being j
well controlled with personnel receiving the proper radiological controls briefing prior !
to entering the containment. The inspectors toured all three reactor coolant loops to !
verify that no abnormal conditions existed. All three loops were satisfactory. Visual :I
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inspection of major equipment around the containment did not reveal any unsafe !
condition. The primary component cooling (PCC) water pipes were inspected for any i
leakage indicative of through wall defects and none were found.
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c. Conclusions
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The primary containment was maintained well. Maine Yankee radiation controls ;
personnel properly controlled reactor containment access as required by station !
procedures. The radiation controls technicians were knowledgeable of existing plant j
conditions, work in progress and dose limits for all the accessible areas. *
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02.3 - Operation of the Soent Fuel Crane (VIO 50-309/96-14-01)
a. Insoection Scooe (71707)
During the period of January 7 through January 15,1997, Maine Yankee
experienced a series of problems with the spent fuel pool crane, CR-9. The inspector i
reviewed the background of these events and the actions taken by Maine Yankee to
address the problems,
b. Observations and findinas
Starting on January 7, operators were forced to stop moving fuel because CR-9
would no longer go in the up direction. After a non-working period, CR-9 started ,
working again and fuel movement continued. Operators assumed the brakes were l
dragging on the crane which caused heat and excessive resistance, j
On January 8, fuel movement was again stopped due to the failure of the crane and
fuel movement was resumed after a rest period. Shortly after fuel movement was
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restarted, it was secured due to unrelated problems with the load cell battery. At
this time, no trouble shooting, evaluation or formal documentation of the problem I
was performed. l
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On January 9, the oncoming plant shift superintendent (PSS) reviewed the problem
history with CR-9 and initiated a work order to troubleshoot the problem. No
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problems were noted with the crane or the brakes.
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On January 12, fuel movement was resumed and again, after several hours, the
crane failed to function. Again, fuel movement was resumed after a non-working
period. No formal evaluation was made as to why it was safe to lift fuel with the '
l crane in this condition. !
On January 13, the PSS again initiated actions to troubleshoot the crane. During-fuel '
movement, electrical maintenance measured the motor amperage to CR-9 and did not
' note any abnormalities. Later in the day, the crane again failed to function. Again,
fuel movement was resumed after a rest period. Still, no formal documentation or
evaluation of the problem was performed. j
On the evening of January 13, a decision was made to perform more troubleshooting
- prior to further fuel movement. Electrical maintenance determined that the motor for
CR-9, which was replaced in late 1996, had a duty cycle restriction of 20 minutes l
per hour, internal thermal overloads were heating up and openin0 to prevent the i
outward movement of the crane. Although the duty cycle restriction was identified !
in technical evaluation, T.E. 074-96, CR-9 Replacement Hoist and Trolley, dated
August 28,1996, the requirement to update the operations procedures was not !
identified.
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Fuel movement was stopped and a modification was performed to install a larger ;
motor without duty cycle restrictions. This motor performed adequately during i
subsequent fuel movements. )
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c. Conclusions i
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The inspector concluded that inadequate analysis was performed in response to the
multiple failures of the spent fuel pool crane, CR-9. The use of the degraded crane to
move irradiated fuel could have resulted in fuel being stuck in some position other
than properly seated in the pool. During the period from January 7, through
January 13,1997, the crane failed to function properly on four occasions and each
time fuel movement was resumed without formal documentation of the problem and
a technical justification for continuing. In one case, irradiated fuel was used as a
load to troubleshoot the crane.
Additionally, the engineering technical evaluation process failed to address the
requirement to revise the operations procedures to incorporate the duty cycle i
restrictions of the new motor. Once aware of the limitations, operations determined i
that the duty cycle restrictions were not acceptable for the crane to serve the {
intended function, i
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Technical specification 3.13, Refueling and Fuel Consolidation Operations, requires in I
Part, "A. Prior to each refueling, a complete checkout shall be conducted on fuel l
handling cranes that will be used to handle irradiated fuel assemblies." Operations l
orocedure 13-2, Fuel Handling in the Spent Fuel Pool, performs an operational check
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of the ability of the crane to raise and lower as a prerequisite to fuel movement.
j Contrary to the above, on four occasions from January 7, through January 13,1997,
i problems were identified that invalidated this checkout. In each of these cases, the
cause of the crane failure was not determined and the operability of the crane was 1
not evaluated prior to the movement of irradiated fuel. This is a violation of l
j Technical Specification 3.13. (VIO 50-309/96-14-01) i
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a i
j O2.4 Fuel Sionina Ooerations l
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a. Insoection Scope (71707)
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, Due to indications of fuel!eakage noted during the previous operating cycle, Maine i
j Yankee decided to take advantage of the current outage to investigate the source
{ and extent of the leakage. The inspectors observed plant activities to verify that
! they were conducted safely and in accordance with approved procedures.
l
1
- b. Observations and findinas
!
The inspector observed various portions of the activities involving the preparation for
i lowered inventory, and the reactor vessel head and upper internals removal, and fuel
j sipping operations in the reactor cavity and in the spent fuel pool areas.
i Reactor Coolant System Lowered Inventorv
i i
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, During the process of lowering the reactor vesselinventory in preparation for reactor j
l vessel head removal, the inspector conducted walkdowns of the control board, and i
j held discussions with plant operators to verify that plant parameters were being
i maintained within expected levels and that required safety systems were operational.
<
Some of the inc trumentation checked included the reactor vessel level and i
! temperature indications and computer alarm setpoints, and RHR flow and i
!
temperature indications and computer alarm setpoints. The evolution for lowering the
level was conducted in accordance with procedure 1-17-6, Lowering Inventory.
The RCS inventory was lowered to just below the reactor flange level at elevation
20 feet. The loops and core remained covered. The top of the loop pipes is at l
elevation 15 feet and the top of the core is at elevation 7 feet 9 inches. l
l
Two functional reactor vessel level and one cavity leve: bdic.ators were available on j
the control board. Annunciator panel RH3-10, Rx. Vessel Level, Narrow Range Hi- l
Lo, was functional to provide an early warning of the level being outside the I
expected range (19'-3" to 19' 9"). In addition, computer alarms for low and high
levels were available. Temperature indicators and recorders were available for RHR
inlet and outlet temperatures. Core exit thermocouple (CET-6), indicative of reactor i
temperature and the two narrow range levelindicators had been selected as fixed !
displays on the control room monitor. RHR temperature alarms were available on the l
plant computer. However, when the inspector checked the alarm values, he noted
that the setpoints selected (180 degrees) were not conservative relative to the
current temperature in the reactor of about 108 degrees. When this issue was
.
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8
brought to the attention of shift supervision, it was determined that it was
appropriate to reset the alarms to 120 degrees. The inspector verified the RHR flow
alarm on the computer which had been set at 1,000 gallons per minute.
Reactor Disassembiv
The inspector attended a pre-evolution briefing for the removal of the upper internals.
The briefing was thorough and included lessons learned at Maine Yankee and in the
industry. Roles and responsibilities, expected communications, radiological concerns
and foreign material exclusion considerations were covred in detail.
' The lift of the upper internals was accomplished without any significant problems.
The personnel involved appropriately utilized the procedure and good work practices
were observed during the evolution.
Fuel Sionina
Fuel sipping is a method used to detect fuel assemblies containing leaking fuel rods.
Fuel sipping consisted of lifting a fuel assembly from the core with the refueling
machine crane and sampling the water around the fuel assembly for fission product
gasses. After sampling, the fuel assembly is replaced in the core and the process
repeated for each of the 217 fuel assemblies. Sampling the water around the fuel
assembly was accomplished by a sample rig installed as a modification to the
refueling machine. The inspector reviewed the design of the rig and the 10CFR50.59
determination and identified no concerns.
During the sipping process, Maine Yankee identified nine fuel assemblies with leaking
fuel rods. Once the sipping evolution was completed, the leaking fuel assemblies
were moved to the spent fuel pool for further evaluation. Each fuel assembly
contains approximately 17C fuel rods. Ultrasonic testing was performed on the fuel
assemblies to identify the leaking fuel rods. The ultrasonic test was a simple go/no
go test that looked for water in the fuel rods. Fuel rods with water are identified as
the leakers. The ultrasonic testing identified 76 leaking fuel rods within the nine fuel
assemblies.
Preliminary reviews indicated that the leaking fuel rods may be due to fretting.
Fretting is (aused by movement of the fuel rods within the fuel grid structure due to
reactor coc'. ant flow. Fretting results in localized wear at contact points within the
structure. At the conclusion of the period Maine Yankee was still performing
analysis of the fuel to further determine the extent of the problems, the root causes
and corrective actions.
The inspector observed activities on the refueling bridge in the containment, in the
control room and at the spent fuel pool area. In the containment, FME controls were
in place and being weil controlled. However, it appeared that the licensee was not
well prepared to deal with the retrieval foreign materialin the cavity. When a
licensee personnel identified a piece of tape, no tool was available to be used in the
removal of the object so plant personnel had to fabricate a trash rake. At one point
the inspector observed operators having difficulties in lowering the spreader on one
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of the outer bundles. At that point, operators halted sipping activities and properly !
discussed the problem and the appropriate resolution. The resolution involved '
.
changing the sequence with which the bundles in the outer peripherals were moved l
to account for the expected lateral movement of those bundles. Sipping activities
'
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were resumed and accomplished satisfactorily. !
t
c. Conclusions ,
i
i
The decision to perform the fuelinspections during the current outage was an
example of conservative decision making by Maine Yankee. The reactor disassembly
j and fuel sipping operations observed were performed in accordance with procedures
2 and were well controlled.
l
j There were good controls for lowering the RCS inventory. Operations personnel
tracked important plant parameters such as RCS temperature and level and RHR
temperature and flow. Generally good use was made of the plant computer in )
selecting appropriate alarm setpoints for these parameters to give operators early I
indication of a potential problem. However, a weakness was noted by the inspectw j
when operators did not select an appropriate setpoint for the RCS temperature wher i
RCS temperature was at about 108 degrees. The computer alarm was still at 180 (
degrees and was changed to 120 degrees when this discrepancy was pointed out by
j the inspector. -
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07 Quality Assurance in Operations 1
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07.1 Plant Operations Review Committee 1
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a. Inspection Scone (71707)
i
i The inspectors attended Plant Operations Review Committee (PORC) meetings to
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verify that the meetings were safety focused and conducted in accordance with the
- requirements of the Technical Specifications.
5
b. Observations and Findinas
l On January 9,1997, the inspector attended a PORC meeting during which
i engineering and radiological issues were discussed. The engineering issues involved
,
an Engineering Change Notice (ECN) for the ongoing Containment Spray (CS) building
ventilation unit, HV-7, installation, and two Field Change Requests (FCR) involving
procedures for the Spent Fuel Pool Reracking. In addition, the Plant Root Cause
Evaluation (PRCE) for the containment electrical equipment environmental
qualification (EQ) issues was discussed.
- The meeting was conducted professionally and involved detailed discussion of the
"
technical issues. The committee members demonstrated good safety perspective i
i and good technical knowledge of the issues. Appropriate plant personnel were i
, present to provide detail information on the issues and to respond to questions from
the committee members.
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The inspectors attended a special PORC meeting which was convened to approve for
, use the latest revision of station temporary procedure " Emergency Safeguards
, System Disable and Enable." After the nature of the latest changes to the procedure
was discussed, the issue of the appropriate testing required after the systems were ;
unblocked discussed. It was determined that an appropriate test (loss of power) !
'could not be conducted because of plant condition and therefore a detailed functional !
] test would have to developed. The plant manager decided that this procedure would -
j not be performed and the emergency safeguards equipment would not be disabled. !
- To ensure that all operations personnel understood the ramifications of this decision '
- the plant manager requested that the operations manager provide additional guidance
- to operators of the possibility of inadvertent loss of RHR and the necessity of
J increased awareness and control of plant maintenance activities.
!
c. Conclusions
{ Prior to the start of each of the meetings, the PORC chairman verified that the i
technical specification required quorum was present for the meetings. The
]
j
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committee was professional, technically competent and maintained good safety l
focus as demonstrated by their review of the proposed temporary procedure to i
- disable emergency safeguards systems and subsequent disapproval of the process. !
) 08 Miscellaneous Operations issues j
! )
d
08.1 Review and Closeout of Licensee Event 9eports !
! a. Insoection Scope (907121
The inspector reviewc.i the following Licensee Event Reports (LERs)to determine if
i the requirements of 10 CFR 50.73 were complied with and to ensure that the
i licensee's corrective actions were appropriate and timely. The inspectors verified
i that each event description was consistent with their knowledge of the event, that
- the submittal of the report was timely and that the details of the event were clearly l
4
reported. The Mpector determined whether further information was required from
j the licensee, whether generic implications existed, and whether events warranted
further onsite followup.
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b. Observations, Findinas and Conclusions
. The events discussed in the following LER's wm inspected and documented in l
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various previous NRC inspection reports.
LER 95-15, Emergency Core Cooling system Valves Not in Compliance - NRC I
Inspection Report 50-309/95-24.
LER 95-16, Purge Valves Found in the On-Line Mode during Refueling - NRC
l Inspection Report 50-309/95-24.
4
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LER 96-03, Reactor Trip Due to High Steam Generator Level - NRC Inspection Report
50-309/96-02
1
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LER 96-15, Containment Hatch Interlock Conflict - NRC Inspection Report
50-309/96-08.
LER 96-30, incorrect Acceptance Criteria Specified for Containment Ventilation / Purge
System Filter Surveillance - NRC Inspection Report 50-309/96-12
LER 96-35, Requirement for Post Accident lodine Sampling inadvertently Removed
i From Procedure - NRC Inspection Report 50-309/96-12.
The inspectors determined that Maine Yankee's assessment of the events were
consistent with the inspectors findings during their inspection of the events. The
licensee implemented adequate corrective actions to address the events and the
reports met the requirements of 10 CFR 50.73.
fl. Maintenance
l
M1 Conduct of Maintenance ;
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M1.1 General Activities
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The inspectors observed portions of maintenance activities involving the reactor :
vessel head removal and surveillance testing of the High Pressure Safety injection l
System Pumps and Valves. Maintenance and Surveillance activities were conducted
I
safety and in accordance with approved docunmnts. l
M2 Maintenance and Material Condition of Facilities and Equipment
M2.1 Safety Systems Area inspection
a. Insoection Scope (71707,62707)
The inspectors conducted walkdowns of safety systems areas to asses equipment I
and area material condition and to ascertain that the licensee was properly identifying
and resolving problems. The areas inspected included the residual heat removal
system and service water system pump areas and the containment.
b. Observations and Findinas
The inspection included checks for equipment degradation, fluid leakages, non
functioning equipment, structural defects, uncontrolled / loose material, fire or safety
hazards and any abnormal situations. The inspectors also checked to see if all
discrepancies identified by the inspectors had been identified and tagged by the
licensee.
The areas were maintained well. No unsafe conditions were observed. Identified
leakages were minor and were previously identified by the licensee.
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c. Conclusions
The licensee was properly maintaining safety related equipment areas. Material
condition of systems, and equipment was also maintained well.
M8 Miscellaneous Maintenance issues (92902)
M8.1 Review and Closecut of Licensee Event Reports
a. Insoection Scope (90712)
The inspector reviewed the following Licensee Event Reports to determine if the t
requirements of 10 CFR 50.73 we e complied with and to ensure that the licensee's
corrective actions were appropriate and timely. The inspectors verified that each
event description was consistent with their knowledge of the event, that the l
submittal of the report was ".mely and that the details of the event were clearly .
reported. The inspector determined whether furtbr information was required from i
the licensee, whether generic implications existed, and whether events warranted
further onsite followup. ;
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b. Observations, Findinas and Conclusions
i
The events discussed in the following LER's were inspected by the inspectors and
documented in previous NRC inspection reports.
LER 95-13, Pressure Seal Bonnet Retention Screw Failures - NRC inspection Report
50-309/95-13. i
LER 96-12,ilater Intrusion of Turbine Driven Auxiliary Feedwater Pump Lube Oil i
Sump - NRC ispection Report 50-309/96-06. 2
LER 96-37, Leaking P-29C Gland Cooling Check Valve - NRC Inspection Report
50-309/96-13.
The inspectors determined that Maine Yankee's assessment of the events were
consistent with the inspectors findings during their inspections of the events. The
licensee implemented adequate corrective actions to address the evente and the
reports met the requirements of 10 CFR 50.73.
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Ill. Enaineerina
E1 Conduct of Engineering
E1.1 Testina of HPSI Pomos and Valves (URI 50-309/96-14-02)
a. Inspection Scoce (37551)
In response to questions raised by the NRC, Maine Yankee performed testing of the
HPSI system to verify adequate net positive suction head for the required modes of
pump operation. The inspector reviewed the procedure used,4-17-21, HPSI Pump
Testing at Minimum RWST Levels and HPSI Throttle Valve Sensitivity Testing, and
observed key portions of the performance of the test in the field.
I
b. Observations and findinas
, The NRC ISA team's questions involved the fact that the HPSI pumps were required
to operate near runout conditions with a minimum amount of net positive suction ;
head. Although previous testing had been conducted, the documentation of the l
previous tests was not adequate to fully demonstrate proper operation of the pumps.
Additionally, the effect of varying HPSI throttle valve positions on the total flowrate
from the HPSI pumps was not known. The ISA team was concerned that the i
allowed tolerances for setting the HPSI throttle valves could cause the pumps to 1
operate in a flow condition that would be detrimental to the pump because the
throttling characteristics of the HPSI valves were not known.
The test procedure established a flow condition of full flow to the reactor with two of
the HPSI flow control valves at their nominal limit switch openings, and the third
valve fully open to the limit stop. The reactor head was off which allowed injecting
water at high volume and low pressure into the reactor cavity. The pumps were run j
taking suction from the refueling water storage tank (RWST) until the RWST reached l
a level of about 98,000 gallons. This was about 10,000 gallons below the nominal
setpoint for the shift to sump recirculation. This procedure was performed for both
i the A and B pumps. The swing (S) pump was not tested because it had been
recently overhauled and therefore was considered bounded by the tests on the other
pumps. 4
l
The throttle valve sensitivity test measured the changes in flow as the HPSI throttle I
valves were opened. The throttle valves were moved in 0.020" increments to
determine the throttling characteristics of the valves.
During the test, flow, pressure and vibration data was recorded. The inspector noted
good coordination between operations and engineering during the performance of the
test. Preliminary reviews of the test results showed that the pumps would continue
to perform satisfactorily under the simulated conditions. No pump degradation was
noted. At the conclusion of the inspection period the detailed evaluation of the data
was still ongoing by Maine Yankee and the NRC.
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c. Conclusions ,
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Test personnel showed good technical capability and demonstrated good safety '
, perspective. Preliminary reviews showed that the test was accomplished
"
satisfactorily. Flow, pressure, temperature, and vibration data were as expected.
However, detailed review of test results is still ongoing by Maine Yankee and the
NRC. This item remains open pending completion of this review. (URI 50-309/
96-14-02)
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- E2 Engineering Support of Facilities and Equipment
i
E2.1 Looo Isolation Valve Toraue Switch Failure
a. jnspection Scope (37551)
On January 7,1997, the torque switch in the actuator for loop isolation valve, !
RC-M-21, failed while the valve was being shut in preparation for loop draining. The
inspector reviewed that actions taken to assure structural integrity of the valve
following this potential overtorque event.
.
b. Observations and findinas
On January 7,1997, the loop isolation valve, RC-M-21, was being shut by i
operations to facilitate loop draining for reactor coolant pump maintenance. The loop
isolation valves are normally de-energized open during plant operations and have no
safety function. During the closing evolution the breaker for the valves motor
operator tripped on thermal overload. Subsequent investigation by Maine Yankee-
determined that a shear pin in the torque switch had failed and the valve disc was
driven into the seat until the motor stalled.
Maine Yankee also discovered that due to the high current caused by the motor stall,
the wiring leads at the containment penetration were severely damaged. The
evaluation of the as found wiring was still ongoing at the end of the inspection
period.
The other concern was whether that the structural integrity of the valve was still
intact. Operations was able to confirm that the valve was fully shut and not leaking.
Engineering performed a calculation to determine the weak link related to the valve
and the maximum force which could have been developed by the motor. These
calculations showed that the valve motor would not develop enough force to damage
the valve.
To determine the actual force applied to the valve, motor -operated valve (MOV)
diagnostic equipment was used to measure the relaxation of the valve as the valve
was opened. This operation was performed and the actual stress on the valve was
measured to be less than the calculated stress. This was as expected based on the
conservative factors used in the calculations.
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i The failure of the torque switch was caused by a failure of a rollpin on the torque
switch shaft. Maine Yankee concluded that the failure resulted from the valve L
, sticking during the opening process when the valve was opened following the
previous outage. Sticking of the valve seat can cause rapid acceleration of the
torque switch shaft and damage to the roll pin.
c. Conclusions ,
The prompt evaluation of the structural integrity of the valve body demonstrated an ,
appropriate safety focus. This evaluation was completed prior to relying on the valve
as a pressure boundary for maintenance. The inspector concluded that the failure of
- the torque switch would have had no impact on the safety function of the valve as a
'
part of the RCS pressure boundary.
As of the end of the inspection period Maine Yankee was still developing corrective
actions for the roll pin failure and for the damaged penetration. l
E8 Miscellaneous Engineering issues
i
E8.1 Electrical Seoaration issues (URI 50-309/96-14-03) l
l
a. Inspection Scope (37551)
The inspectors reviewed the preliminary findings and actions taken by the licensee
regarding electrical separation issues that were identified in December 1996. The
inspection included interviews with plant personnel and reviews of plant records.
b. Observations and Findinas
On December 5,1996, the licensee was in the process of developing control system
drawings to facilitate the performance of logic testing reviews requested in NRC
Generic Letter No. 96-01, " TESTING OF SAFETY-RELATED LOGIC CIRCUITS."
During field walkdowns and inspections to support the drawing preparation process
and testing issue resolution, plant engineers identified three electrical wiring and
cable separation issues that affected the reactor protection system. The issues
included a lack of wiring separation between the four safety channels at the reactor
manual trip push buttons, cables from vital DC buses 3 and 4 were routed in the
same cable tray contrary to the FSAR criteria; and control power cables from vital DC
buses 3 and 4 were connected to the wrong reactor trip circuit breakers. Although
these deficiencies did not have an immediate effect on the ability to manually or
automatically scram the plant, the plant was shut down and the deficiencies were
corrected.
During the correction of these deficiencies, additional cable separation issues were
identified. The licensee initially believed that the separation problems may have been
limited to plant modifications installed after January 1995. However, when plant
management directed a full walkdown of the two containment hydrogen analyzer
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16
instruments, additional separation issues were identified. The licensee then
concluded that a more comprehensive review to identify the extent and root cause(s)
of the issues was necessary.
During the inspection conducted on December 18-20,1996, the licensee was in the
process of formulating a plan and developing procedures to address the electrical
separation issues. The licensee agreed to submit a written plan to the NRC when it
had been approved by plant management.
The inspectors noted that the licensee was including a review of other separation
issues that had been identified previously. These included an issue in 1991 where an
auxiliary shutdown panel purchased to support a plant modification was found to
have separation problems. These problems were determined to be isolated and the
result of inadequate purchase specifications, in 1992, a review was performed to
deterrnine if there were separation violations in the plant due to having inadequate air
gaps between cables where they transitioned from one raceway to another. No plant
deficiencies were identified at that time and the FSAR and the cable installation
specification were changed to clarify the air separation criteria. In 1994, separation
deficiencies similar to the recent findings were identified and the licensee determined
that they were the result of modification errors between 1979 to 1981. All
modifications performed during that period were reviewed.
Since the recent findings, the licensee identified that there were conflicting
statements in the Final Safety Analysis Report (FSAR) regarding cable separation
requirements. The licensee performed Technical Evaluation (TE) 226-96, " Cable and
Wire Separation Criteria," to provide a summary of all separation requirements; to
provide a basis for each requirement; and to provide the information necessary to
complete a 10 CFR 50.59 evaluation to support necessary changes to the FSAR.
]
The licensee developed procedure 17-52, " Cable Separation Assessment and j
Walkdown Procedure," to control the identification of raceways to be inspected
'
during plant walkdowns; to provide instructions for the performance of raceway
walkdowns and reviews; and to provide direction for configuration evaluation and
corrective actions. !
l
Utilizing these documents the licensee plans to screen all plant modificB
performed since initial construction to identify those that may have int. ed cable !
separation problems. In addition to inspecting the cables affected by the
modifications, the licensee plans to collect information on the origina!!y installed plant
cables in the inspected raceways. This information will be reviewed to assess the
adequacy of the original cables with respect to cable separation.
On December 18,1996, the NRC issued Confirmatory Action Letter (CAL) No.
1-96-015 to confirm in writing actions that the licensee planned to take to resolve j
the electrical separation and logic testing issues prior to plant restart in summary
'
these actions include: 1) the completion of the initial safety-related logic testing
review pursuant to GL 96-01,2) a determination of the extent of the cable
separation problems and resolution of all safety significant issues,3) the performance
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of root cause evaluations that will address deficiencies that were identified and to
validate the comprehensiveness of the corrective actions, and 4) a meeting with the
NRC to describe the results and conclusions as they pertain to the above actions.
c. Conclusions
At the time of this inspection the licensee was in the process of finalizing the plans
and procedures for resolving the cable separation issues. Until management
intervened and directed a complete walkdown of the hydrogen analyzers, the scope
of reviews directed by engineering was relatively narrow. The inspectors noted that
the licensee reviews for previously identified issues may also have been too narrow
in scope. Those reviews were, in one case, limited to modifications performed from
1979 to 1981, and in the other case, to modifications installed after 1995. The
adequacy of the previous actions can be better assessed when additional installation
information is available following the implementation of the current plan. This issue
remains unresolved pending NRC review of the licensee actions including the
following (URI 50-309/96 14-03):
e the finalization and implementatiors of a plan to determine the extent of the
cable separation problems;
- the resolution of identified deficiencies; and,
- the performance of a root cause determination and identification of necessary
corrective actions.
E8.2 Safetv-Related Loaic Circuit Testina Update (URI 50-309/96-14-04)
a .- Insoection Scope (37551)
The inspectors reviewed the actions being taken by the licensee in resporse to NRC
Generic Letter 96-01, " Testing of Safety-Related Logic Circuits."
b. Observations and Findinas
On January 10,1996, the NRC issued Generic Letter 96-01 which requested that
licensees review the adequacy of testing of reactor protection system, emergency
diesel generator load shedding and sequencing, and actuation logic for engineered
safeguards features system to ensure that the technical specification surveillance test
requirements were being met. The licensee initially planned to complete this review
in 1997. However, in August 1996, a special NRC team inspection identified logic
circuits associated with safety-related pumps that were not being periodically tested.
The licensee actions in response to this issue resulted in the identification of
additional test deficiencies and also a missing wire in the high pressure safety
injection (HPSI) start circuit. The licensee then expanded its review of testing to
include all of the circuits that were within the scope of GL 96-01. Several
deficiencies were identified and corrected prior to restart. (Discussed in Inspection
Report 50-309/96-11).
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The August 1996 review was intended to ensure that the testing was adequate to
support plant restart and the licensee planned to perform an additional test procedure
review to formally address GL 96-01. This review would include the preparation of
integrated control system drawings to facilitate the review and would include
independent verification of each review. The licensee was in the process of
performing the review at the time of this inspection. The licensee had identified a
number of questions regarding the testing adequacy and was in the process of
dispositioning the findings.
One finding involved the reactor trip circuit breakers. These circuit breakers can be
opened to initiate a reactor scram by either deenergizing the undervoltage coil or by
energizing the shunt trip coil. The licensee currently tests the breakers to ensure that
either one of the coils, when actuated separately, would open the circuit breakers.
However, the logic testing review revealed that during the reactor protection circuitry
testing, the coils receive simultaneous actuation signals. Therefore, the testing, does
not independently verify the proper operation of both sets of contacts on the logic j
relay. The licensee performed additional testing on the relay contacts and the results
were satisfactory.
Following the completion of this inspection, the licensee notified the NRC in
accordance with 10 CFR 50.72 of a finding that the logic circuitry that would start
the EDGs in the event of a safety injection signal coincident with a degraded power
grid condition had not been adequately tested. This finding resulted in the licensee
declaring both train of the emergency core cooling systems inoperable. As a result
the plant was placed in cold shutdown in accordance with technical specification
1
requirements.
The inspectors questioned whether the deficiencies that were identified in the current
review should have been identified during the August review. The licensee had not
yet performed this assessment.
c. Conclusions
This issue remtins unresolved pending licensee completion of the logic testing
reviews; the resolution of test deficiencies, and NRC review of these activities.
4 (URI 50-309/94-14-04)
E8.3 Review and Closeout of Licensee Event Reports
a. Insoection Scooe (90712)
The inspector reviewed Licensee Event Reports to determine if the requirements of
10 CFR 50.73 were complied with and to ensure that the licensee's corrective
actions were appropriate and timely. The inspectors verified that each event
description was consistent with their knowledge of the event, that the submittal of
the report was timely and that the details of the event were clearly reported. The
inspector determined whether further information was required from the licensee,
whether generic implications existed, and whether events warranted further onsite
followup.
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19
b. Observations. Findinas and Findinas
The events discussed in the following LER was inspected by the inspectors and
documented in NRC Inspection Report 50-309/96-08.
LER 96-11 Service Water Punip Cutiess Bearing cooling Water System Design.
The inspectors determined that Maine Yankee's assessment of the event was
consistent with the inspectors findings during their inspections of the event. The
licensee implemented adequate corrective actions to address the event and the report
met the requirements of 10 CFR 50.73.
IV. Plant Support
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R1 Radiological Protection and Chemistry Controls
Reviews were performed of occupational radiation exposure, solid radioactive waste
management and transportation of radioactive materials, and radiation protection
organization and management controls. Specific areas reviewed included maintaining
radiation exposures as low as is reasonably achievable (ALARA); use of health
physics stop work authority; video monitoring of high radiation area access; snlid
radwaste management; the radioactive materials and radioactive waste shipping
program; implementation of revised department of transportation regulations; status
of facilities and equipment; procedures and documentation; staff knowledge and
training; organization and administration; and quality assurance including program ,
audits and surveillances. Reviews were also performed of the details associate with I
an unanticipated extremity exposure and unplanned exposures to a radioactive
particle in a chair used by Security /Firewatch personnel. Finally, licensee responses !
to previous violations were reviewed, and a evaluation of the facility condition versus -l
the requirements in the UFSAR was performed.
R 1.1 ALARA
a. Insoection Scope (83750)
Reviews were performed of the radiation exposure goals used to rnaintain radiation
exposure as low as is reasonably achievable (ALARA). Information was gathered ;
through discussions with cognizant personnel, and by reviews or radiation dose l
surnmaries provided on January 17,1997, to the Station ALARA Council (SAC). !
b. Observations and Findinas
The 1996 radiation exposure goal had been set to maintain total radiation dose to
less than 40 person-rem. Actual dose received was determined to be 56.39 person-
rem, which exceeded the established goal by 16.39 person-rem. Licensee staff
stated that although opportunities for improvement were identified (e.g., deta;ied job
planning), the primary reason for exceeding the 1996 dose goal was due to emergent
outage work. This included approximately 12 person-rem from followup activities
associated with the 1995 Steam Generator Sleeving outage; installation of relief
.
.
20
valves on primary component cooling lines; and containment entries associated with
cable separation issues. The inspector reviewed graphs of accumulated dose versus
time for the year 1996, and noted that dose accumulation rates supported the
conclusion that Maine Yankee would have met the established 1996 yearly radiation
exposure goal, if not for the performance of emergent work.
The inspector was also informed that the 1997 normal operations dose goal had
been set at 50 person-rem. The 1997 outage dose goal had not been set because
outage scope had not been fully defined. However, the current goal for the fuel
sipping / cable separation outage was set at approximately 40 person-rem.
c. Conclusions
Maine Yankee exceeded the established 1996 ALARA dose goal due to the
performance of emergent outage work, and reasonable efforts were made to reduce
radiation exposures.
R1.2 Stoo Work Authority
a. Inspection Scoo (83750)
The inspector performed a review of current and historical use of health physics stop-
work authority. Information was gathered by a review of guidance provided in
procedure 9-302-2, " Job Coverage," Rev. 3, and through interviews with Health
Physics management, supervision, and technicians.
b. Observations and Findinas
Procedure 9-302-2, " Job Coverage," Rev. 3, included specific criteria for health
physics technicians to use to stop jobs for legitimate radiological reasons.
Examples included work not performed in accordance with radiation work permits
(RWPs); work not performed in accordance with good radiological controls; initiation
of electronic dosimetry alarms; and changes in radiological conditions or job scope. l
The inspector noted that Attachment A to the procedure also included a "RWP Pre- '
Job Briefing" sheet that included specific guidance to inform workers of conditions
on situations in which they should cease work due to radiological concerns.
4
During interviews, Health physics technicians were found to be familiar with
procedural guidance pertaining to stop work authority, and indicated that they would
not hesitate to exercise stop work authority for legitimate radiological reasons.
Supervisory and management personnel were also interviewed regarding past
examples of use of "stop work authority." Management personnel indicated that l
during the 1995 outage, some technicians raised concerns that Maine Yankee
management would not support HP technician use of stop-work authority. To
address this concern, Maine Yankee management put out a March 2,1995, Outage
Manager Directive that indicated that "RC technicians had the authority and !
responsibility to shut down jobs for radiological safety concerns, and the outage
organization would support technicians in these instances." During interviews,
_ _ _
.
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21
three examples of past use of HP stop work authority were discussed. Although, it
was apparent that some management discussion of the validity of the use of stop-
<
work authority had occurred (which had the potential to act as a chilling affect), in
each case, radiological control supervision ultimately supported the use of stop work
authority.
c. Conclusion
Based on this review, the inspector concluded that policies for use of HP "stop-work
authority" were clearly communicated to HP technicians through procedural
guidance, and management supported legitimate use of HP stop-work authority. No
specific examples were identified where Maine Yankee management prevented HP
technicians from exercising legitimate stop-work authority, and no examples of
retribution or punitive actions were identified when legitimate "stop-work authority"
was exercised. No violations of NRC requirements were identified.
R1.3 Video Monitorina of Hiah Radiation Area Access
a. Insoection Scope (83750)
A review was performed of the licensee's use of video monitoring equipment to
monitor access to high radiation areas. The inspector reviewed Technical
Specification 5.12; examined video monitoring equipment used at the health physics
check point and Vapor Containment (VC) access; and interviewed cognizant
personnel,
b. Observations and Findinas
Technical specification 5.12, paragraph 2, states that doors to locked high radiation
areas "shall remain locked except during periods of access by personnel under
administrative controls. Access administrative controls shallinclude an approved
RWP, measures to prevent unauthorized access, and measures for control of
exposure. Access administrative controls may include remote surveillance (such as
use of closed circuit TV cameras) by personnel appropriately qualified in radiation
protection procedures."
The inspector examined video monitors present at the health physics check point that
were used to remotely monitor personnel access to sveral frequently accessed
locked high radiation areas located in the balance of plant (BOP). These areas .
included the Let Down Pipe Tunnel on -11 foot elevation of the Primary Auxiliary
Building (PAB), the entrance to the RCA Building, and the Decon Pad and resin
storage tank (TK-85), both located within the RCA Building. The inspector noted that
video images were clear and video monitors were useful for monitoring authorized
and unauthorized personnel access.
The inspector also examined video monitors located at the Vapor Containment
access and observed HP Technician performance, with regard to monitoring video
cameras for Technical Specification 5.12 High Radiation Area doors. The inspector
noted that although Technical Specification 5.12 allows for use of closed circuit
-_ - - - - -- - . . - . _
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4
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22
television monitoring by individuals trained and qualified in radiation protection
'
i procedures, the use of cameras is only effective if they are adequately monitored.
4
The potential dm exist for HP technicians with multiple duties to become distracted, ,
and thereby noc adequately monitor high radiation area access. The inspector was
'
' informed that band on suggestions made in 1995 by Health Physics staff members, {
'
Maine Yankee revised their practices to use only " dedicated" technicians to monitor 1
personnel access to Tech Spec 5.12 doors located within the Vapor Containment l
attentive to the monitors; and that video monitoring equipment was used effectively ;
- to monitor personnel access to high radiation areas. No specific examples of .
- technicians failing to adequately monitor high radiation area doors via video cameras ]
i
- were identified by NRC inspectors. 1
!
l C. Conclusion
4
i
, Based on this review, the inspector concluded that the licensee effectively used '
video monitoring equipment to remotely monitor personnel access to specific high ,
radiation areas. Video cameras were operational, video images were clear, and HP
technicians were attentive to video monitors. No specific examples of technicians 1
failing to adequately monitor high radiation area doors via video cameras were '
identified by NRC inspectors. No violations of NRC requirements were identified.
R1.4 Solid Radwaste Proaram I
a. Insoection Scoce (86750) !
l
The inspector performed a review to evaluate if the licensee maintained current j
copies of applicable regulations, provided management approved procedures, and had )
-
an adequate basis for certifying that radioactive wastes intended for disposal were
properly classified, described, packaged, marked and labeled for transportation. In
addition, the inspector reviewed program initiatives to reduce dry active waste
(DAW) generation. Information was gathered by a review of records and procedures,
and interviews with cognizant personnel, i
b. Observations and Findinas
'
The licensee used a combination of direct isotopic sampling, scaling factors, and
dose-to-curie conversions to determine the isotopic and curie content of radioactive ;
waste containers. Waste streams were sampled and sent to an offsite laboratory on :
a periodic basis to determine the radioactive content. Hard to measure radionuclides I
(beta and alpha emitters) were related to the gamma emitting isotopes through i
scaling factors. These evaluations enabled licensee staff to properly
determine waste classification in accordance with 10 CFR 61.55. The inspector
examined several radwaste shipping records to evaluate the accuracy of waste
classification. No discrepancies with radioactive waste classification were identified.
i
s
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23
The inspector examined and verified that the licensee had up-to-date copies of
49 CFR Parts 100-179,10 CFR Part 20, and 10 CFR Part 71; applicable regulations
for the state of South Carolina; and licenses for facilities to which shipments of
, radwaste or radioactive materials were made. 1
The inspector also reviewed dry active waste (DAW) generation for 1995 and 1996. ,
in 1995, a DAW generation goal of 12,750 cubic feet was established for Maine l
-
Yankee. This amount was raised to approximately 16,150 cubic feet due to the l
performance of the steam generating sleeving outage. Ac+ual DAW generation !
exceeded this value by approximately 2,500 cubic feet (18,645 ft'). The 1996
compactible DAW goal was reduced to 7,200 ft'. The actual compactible DAW
generation in 1996 was 6,469 ft' which was approximately 700 ft' below the i
established goal. Although sorne reduced generation was attributed to training and j
awareness associated with the new program to prevent clean items from being mixed )
with contaminated items (i.e., Green-is-Clean program), and efforts to reduce the use I
of disposable coverings, licensee staff attributed reduced DAW generation in 1996 !
'
simply to a decrease in performance of activities that create DAW.
Finally, the 1997 DAW goal was established based only on past history and was set
at 15,000 ft'. Licensee staff indicated that the goal could be refined once total work
scope for 1997 was established.
c. Conclusion
l
Based on this review, the inspector made the following conclusions.
--
The licensee had a good basis for certifying that radioactive wastes intended
for disposal were properly classified, described, packaged, marked, and
labeled.
--
Although DAW generation goals were established and performance versus
goals was tracked, the use of goals to reduce DAW generation was
perfunctory.
R1.5 Radioactive Waste / Radioactive Material Shiocina Procram
a. insoection Scone (86750)
No outgoing shipments of radioactive waste materials were performed during the -
inspection. The inspector reviewed the radioactive waste and' radioactive material
shipping program through a review of various shipping records, and through
interviews with cognizant personnel.
b. Observations and Findinqs
The inspector randomly selected and examined records of shipments to evaluate
compliance with shipping regulations and the adequacy of shipping records.
Examples of shipping records reviewed included shipments of resin filled high
-. . .
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24
integrity containers (HICs), materials for volume reduction (incineration / metal melt),
dry active waste (DAW), and radioactive material (RAM) shipments such as
radioactive sources.
The inspector noted that the licensee takes pictures of shipping vehicles and
packages to make a record of vehicle placarding, and in some cases package labeling
and methods used for loading and storage, and blocking and bracing packages. No
,
discrepancies in vehicle placarding, package labeling, or vehicle loading were
identified. The inspector noted that each shipping record contained radiation and
contamination surveys of packages and vehicles, and appropriate shipping
documentation. The overall quality of shipping records was determined to be good,
and no discrepancies were identified with individual shipping records.
c. Conclusions
Based on this review, the inspector concluded that radioactive waste and radioactive
material shipments were prepared and made in accordance with applicable shipping
regulations, and shipping records were adequately developed and maintained.
R1.6 Imolementation of the Revised DOT Shiocina Reaulations
a. Inspection Scoce (Tl 2515/133)
The inspectors evaluated implementation of the revised 49 CFR Parts 100-179 and
10 CFR Part '71 regulations by a review of a computerized shipping program,
procedures, training rosters, and through interviews with cognizant personnel.
b. Observations and Findinas
A computerized radioactive waste shipping program named RADMAN was primarily
used to prepare radioactive waste shipments. The inspector verified that the
RADMAN program had been updated to incorporate the revised regulations. The
inspector noted that the licensee relies on use of the RADMAN system to classify
and prepare radioactive waste shipments, and in practice does not classify and
prepare shipments by hand using procedural guidance. The licensee had not yet
implemented the requirement for use of the International System (SI) of units;
however, this regu!ation is not a requirement until April of 1997. Finally, the I
'
inspector reviewed a training roster and noted that personnel responsible for
certifying the adequacy of radioactive materie' shipments had been trained in the
revised 49 CFR Parts 100-179 and 10 CFR Part 71 regulations. .
)
i
c. Conclusions
The implementation of the revised 49 CFR Parts 100-179 and 10 CFR Part 71
regulations was adequate,
l
1
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R2 Status of RP&C Facilities and Equipment Tours
a. Insoection Scope (86750)
,
l
The inspector performed tours to evaluate radiological control boundaries, postings,
contamination controls and monitoring, radioactive material control, housekeeping, ,
material condition, and the storage of radioactive materials / wastes. The inspector
,
also looked for signs of ground water intrusion into areas used for radwaste storage
(e.g., radwaste bunker, RCA Building). Areas toured included the Primary Auxiliary
Building (PAB), the Vapor Containment (VC), the hot shop, the entrance to the
'
- Radwaste Bunker, the RCA Storage Building, the LSA Building, and the Low Level
Waste and Equipment Temporary Storage Building (LLWETSB). ;
i b. Observations and Findinas
- r
,
PAB, VC, and Hot Shoo
l
Radiological boundaries in the Primary Auxiliary Building (PAB), Vapor Containment, I
and Hot Shop were clearly delineated and well maintained, and radiological postings
were informative. Access controls to high radiation areas were good including j
locked doors, door alarms, use of video monitors for high radiation area doors,
briefings, and RWP controls. Health physics technicians interviewed were ;
knowledgeable of ongoing work, radiological controls, procedural guidance, and
provided briefings that were commensurate with hazards. All contamination
monitoring equipment inspected were operational and within calibration. All
radioactive material containers were appropriately labeled with dose rate information,
date of survey, and initials of individual performing the survey. Work areas were well ,
illuminated and walkways and isles were clear and free of debris. No standing water
was observed.
Radwaste Bunker
The Radwaste Bunker is an underground radwaste/ material storage area located on
the northwest side of the site. The maximum curie (radioactivity) limit allowed for
the radwaste bunker was 1600 curies. Material inventory records indicated that
87.7 curies of radioactive material were stored in the bunker. That included
radwaste filters, a neutron source, and reactor coolant pump rotating element and
maintenance equipment. The inspector examined the entrance inside the bunker.
The inspector observed storage space, with no standing water and minor debris.
No discrepancies were identified.
RCA Storaae Buildina
The RCA Storage Building is a radioactive material /radwaste processing and storage
area located on the northwest side of the site. Equipment in the building includes the
Duratek water processing system, resin dewatering area, resin storage tank (TK-85),
waste holdup tank (TK-95), waste resin holdup tank (TK-109), and the decon pad
(radwaste container decontamination area). In addition, various equipment and
.
~
26
radioactive materials were stored in the RCA Building. Although many areas had
been decontaminated and house keeping had significantly improved, the RCA
Building remained cluttered with drums, hoses, and stored equipment.
LSA Buildina
The LSA Building is a radwaste processing / storage area that includes a scale,
compactor, high radiation DAW container, a high rad filter high integrity container,
biological waste, and a storage bunker. Although work areas were clear and free of
excess debris, the south end of the building was cluttered with stored equipment,
drums, and miscellaneous materials.
Low Level Waste and Eauipment Temocrary Storaae Buildina (LLWETSB)
The LLWETSB is a large radwaste/ material storage building located on the north side
of the site. The LLWETSB was designed to store up to 7000 curies of radioactivity,
and up to five years of waste generated at Maine Yankee, as well as contaminated
and non-contaminated reusable equipment. Radwaste packages were present in a
" ready to ship" condition (i.e., sealed, free of external contamination, and properly
labeled). Materialinventories were maintained current, and records indicated that
radiological surveys inside and on the perimeter of the building had been performed in
accordance with procedural requirements. The building was welllit, and walkways
and aisles were clear and free of clutter with only minor debris. No standing water
was identified.
Duratek Liauid Radwaste Processina System
,
i
The Duratek system is a liquid radwaste processing system used to filter and de-
ionize liquid radwaste, and transfer (sluice) media from Duratek demineralizer vessels
'
to high integrity containers (HICs). The system is located in RCA Building and is
temporarily connected to Maine Yankee's permanently installed piping system. The
inspector was informed that efforts were underway to change the Duratek liquid
radwaste water processing system from a temporary system to permanent plant
equipment. This effort involved the assignment of permanent plant valve numbers to
the Duratek system, procedure revisions, and changes to the UFSAR. The inspector
,
noted that since the station has committed to using the Duratek system for
resin / liquid radwaste processing, changing the status of the Duratek system to
permanent plant equipment was appropriate.
c. Conclusions
Based on this review, the inspector made the following conclusions.
- Radiological boundaries including radiation areas, high radiation areas, and
contaminated areas were well defined and well maintained.
-- Access controls to high radiation areas including use of RWPs, radiological
briefings, door alarms, radiological postings and remote video monitoring were
effective.
!
!
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--
Health physics technicians were knowledgeable of ongoing work, radiological
controls, procedural controls, and provided briefings commensurate with
hazards.
--
Some radioactive wastes awaiting processing and packaging, had been stored
for an extended time period (greater than several years). Examples included
vacuum filters in the spent fuel pool; filters and reactor coolant pump rotating
elements in the "High Radiation Bunker;" and vacuum sludge, filters, and in-
core-instruments (ICl) " hot-end" stored in the LSA building.
,
--
Housekeeping improvemats were generally noted in the PAB, containment
building, and RCA and LSA buildings.
'
--
The use of multiple radwaste processing and storage areas, that are not
physically connected (e.g., High Rad Bunker, LSA building, RCA building,
LLWETSB), presents the Maine Yankee staff with significant material '
handling / processing challenges. l
R3 RP&C Procedures and Documentation I
a. Inspection Scope (86750J
i
'
The inspector performed a review to determine if the licensee had provided
management approved, detailed instructions and operating procedures for all
personnel involved with radioactive material and radwaste transport. The inspector
gathered information by a review of procedures and discussions with cognizant
personnel.
b. Observations and Findinas
The inspector reviewed the following procedures.
9-13-100, Radioactive Waste / Radioactive Material Shipping and Handling Program, i
Rev.7 '
9-14-100, Radwaste Process Control Program," Rev. 5
9-15-100, Radwaste Minimization Program, Rev.1 ,
'
9-313-6, Shipment of Radioactive Waste For Burial, Rev. 5
9-313-25, Shipment of Radioactive Material, Rev. 3 i
9-313-19, Shipment of Radioactive Waste For Processing, Rev. 3
9-313-7, " Preparation of Radioactive Solid Waste For Disposal," Rev. 4
9-313-23, " Packaging Radioactive Material For Shipment," Rev. 2
9 313-17, " Surveillance of the LLWETSB, High-Rad Bunker, and LSA Building,"
Rev.4
9-313-16, " Transfer To And Storage of Radwaste And Material in The Low Level ,
Waste And Equipment Temporary Storage Building (LLWETSB)," Rev. 7
Procedural guidance was available for activities involved with the preparation and
shipment of radioactive wastes and radioactive materials. Some procedures were
general in nature and some minor discrepancies were identified. Example, 9-14-100, j
l
,
.
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28
Radwaste Process Control Program, Rev. 5, section 7.3 refers to " operations
personnel (including Duratek operators)." The inspector noted that the Duratek
operator actually reports to the Chemistry Department. Licensee staff informed the
inspector that technical and administrative procedure upgrade revisions were in
progress. A contracted radwaste technical expert had completed technical procedure
upgrades, and the radwaste staff was currently working with the radiological
controls department to standardize procedure format. The inspector verified this by
randomly reviewing proposed changes on draft procedures. Although no significant
deficiencies were identified in existing p.ocedural guidance, the inspector noted that
licensee efforts to make technical end administrative procedural improvements was
prudent.
R4 Staff Knowledge and Performance in RP&C
R4.1 Knowledae of Radiation Work Permit, Radwaste Shioment and Duratek System.
,
a. Insoection Scone (83750)
The inspector evaluated Health Physics technician knowledge of radiation work
permit (RWP) controls for work in the Vapor Containment; the Radwaste Shipper's
knowledge of methods for classifying and preparing radwaste shipments; and the
Duratek System Operator's knowledge of the Duratek liquid radwaste processing
system. Information was gathered through reviews of radiation work permits,
reviews of shipping records, and interviews.
b. Observations and Findinas:
Briefings provided by Health Physics technicians assigned to the Vapor Containment
included a discussion of planned work activities, reviews of radiological condition and
radiation work permit controls, descriptions of nccessary communications with the
Health Physics staff. The scope and depth of the briefings appeared to be
commensurate with hazards. HP Technicians were able to describe the scope of
ongoing work, and demonstrated good knowledge of procedural guidance for job
coverage activities and use of stop work authority.
The Radwaste Shipper demonstrated very good knowledge during the preparation of
radwaste and radioactive material shipping documentation using sample data,
procedural guidance, and computer software. The Radwaste Shipper was able to
describe, in detail, the preparation and process of radwaste and radioactive material
for shipping.
The Chemistry Duratek System Operator exhibited good knowledge of the Duratek
liquid radwaste processing system by describing Duratek system components, and
describing methods for sluicirg resins (media transfers), dewatering of high integrity
containers (HICs), and replac.ng filters. The operator also described proposed
procedural changes planned to convert the Duratek system to permanent plant
equipment.
1
1
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29
c. Conclusion
i
Based on this review, the inspector concluded the following.
--
Health physics technicians assigned to the Vapor Containment were
knowledgeable of ongoing work, radiation work permit controls, procedural l
.
requirements, and provided briefings that were commensurate with hazards. - ,
--
The Radwaste Shipper was knowledgeable of methods used to prepare and
process radwaste/ material shipments. l
--
The Chemistry Duratek System Operator exhibited good knowledge of the j
Duratek system and of procedural controls for operating the Duratek system.
1
R4.2 Poor Contamination Control Practices (VIO 50-309/96-14-05) l
!
a. Insoection Scope (71750)
!
l
During the observation of the high pressure safety injection (HPSI) pump testing,
described in section E1.1 of this report, the inspector observed poor radiation worker
practices which could have lead to the spread of contamination. The inspector
reviewed the corrective actions taken in response to this occurrence and the related
licensee procedures. I
b. Observations and findinas
f
During the HPSI test, a portable vibration instrument was used to measure HPSI
pump vibrations at various locations on the pump. This required a technician to i
'
place a probe-on the pump which was roped off as a contaminated area. The - _
inspector observed the technician using a gloved hand to reach across the 3
'
contamination area barrier to place and remove the probe to and from the pump.
However, after placing the probe into the contaminated area, the gloved hand was
used to manipulate the hand held meter outside of the contaminated area. This
practice was repeated for several points on the pump.
The inspector notified health physics who took action to stop the evolution and verify !
no contamination was spread to the meter or the individual. Testing was later ,
resumed using multiple probes fixed to the pump such that entry into the
contaminated area was not required to obtain the data,
t
Although the data was being taken for a special test, similar data is also obtained
monthly for routine testing of the pump. When questioned, the technician stated he
was using his normal technique to obtain the data. This indicates a lack of oversight
of radiation worker practices in the field. i
Procedure 9-5-100, Contamination Control / Decontamination Program, allows
personnel to reach into contaminated areas for certain purposes. However, the l
procedure also states that personnel shall remove protective clothing prior to exiting i
contaminated areas.
s
--. - - .a -
.
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I
in response to this issue, a poor work practice item (#97-004) was written. ;
However, from January 23, until January 31,1997, no action was taken by the
licensee to resolve this matter. On January 31, the inspector discussed the issue
with a radiation protection supervisor. The supervisor initiated actions to enter the
item into the learning bank program and initiate immediate corrective actions. One
action was to evaluate the need for more specific guidance regarding reaching into
contaminated areas to perform simple evolutions.
&
c. Conclusions !
i
Reaching in and out of a contaminated area and handling clean equipment outside of
'
the boundary is an example of poor radiation work practices and a violation of
licensee procedures. Additionally, the fact that this practice has been occurring for
some period of time raises concerns with the quality of health physics oversight of
field activities. Technical specification 5.11.1, requires in part that procedures for
personnel radiation protection shall be prepared. Procedure 9-5-100, Contamination !
Control / Decontamination Program, Section 7.5.1, allows personnel to reach into ;
contaminated areas for certain purposes and requires that personnel shall remove y'
protective clothing prior to exiting contaminated areas. Contrary to the above, on
January 23,1997, a radiation worker failed to remove protective clothing from his !
hand prior to exiting the contaminated area around the high pressure safety injection ,
pump, P-14A. (VIO 50-309/95-14-05)
The delay associated with entering this item into the learning bank process and the '
failure to promptly evaluate the need for immediate corrective actions is another >
example of deficiencies in the area of problem identification and resolution. .j
i
R5 Staff Training and Performance in RP&C !
a. Inspection Scoce (83750)
i
2
The inspector reviewed training provided to the Radwaste Coordinator, Radwaste
Shipper, and Radwaste Technicians. Information was gathered through interviews ;
'
and a review of training records.
b. Observations and Findinas ,
l
The inspector reviewed training records and found that the Radwaste Coordinator,-
Radwaste Shipper, and current Radwaste Technician had current training in ,
regulations pertaining to the transportation of radwaste including 49 CFR Parts j
100-179,10 CFR Part 20, and 10 CFR Part 71. In addition, these individuals also I
had current training on the use of computer software used to classify shipments of :
radioactive waste and materials.
However, training provided to Rad Material Handlers and Contract HP technicians l
relative to their responsibilities in Radwaste were not well defined (no established {
training matrix), and training records for contractor radwaste technicians were not I
well maintained. Example, of five contract HP Radwaste technicians selected, only
three sets of training records could be readily retrieved. The inspector also noted
i
.
_. _ _ . - _ . . _ _ - - _ _ ___ _ _ _ _ _ . _ _ _ - . . .__ __ _ _ _
,
.
5
31
that there appears to confusion between the training department and the Radwaste ,
Group in the ownership and maintenance of training records for individuals in the
! radwaste group. Licensee staff responded to this observation by stating that a i
j Radwaste Group training matrix would be established, and the. Training Department l
would assume responsibility for maintaining all training records. !
2
-
c. Conclusion
,
Based on this review, the inspector made the following conclusions.
a
Training provided on transportation regulations and use of computer software ;
) --
'
d
used to prepare radioactive waste / material shipments was very good and
current.
j -- Radwaste training weaknesses were evidenced by the lack of a training matrix !
for individuals assigned to radwaste, and discrepancies in contractor ;
Radwaste-HP Technician training records.
.
! R6 RP&C Organization and Administration
i a. Inspection Scope (83522_). ;
- The inspector performed a review of the organization and administration of the .
,-
radiological controls organization. Information was gathered by reviews of current
l and proposed organizational charts, through discussions with cognizant personnel,
and a review of training material for a new plant wide problem
d
identification / resolution system (Learning Process). l
. b. Observations and Findinas
'I
Oraanization and Staffina
'
Inspectors were informed that as a result of a desire to improve performance and
upgrade the radiological controls program, organizational and staffing studies were
. conducted. The goal of these studies was to identify necessary staffing
a augmentations that would allow for program upgrade without distracting personnel
- involved with program implementation, and to develop a proposed permanent
- organizational structure with defined staffing numbers.
.
During the immediate program improvement phase, the following staff augmentations
,
have occurred. A representative from the institute of Nuclear Power Operations had
- been assigned as the Assistant Radiation Protection Manager, responsible for
managing day-to-day program implementation. This addition was designed to allow
'
the Radiation Protection Manager to give fuller attention to managing program
improvement initiatives. Approximately seven professional technical personnel were
i
added to make program improvements such as procedure upgrades, implement
j automated access control, and perform program assessments. In addition, about 40
senior level Health Physics technicians were added to the organization to support
i program improvements and the current outage.
!
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4
w, -
, e
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32
Longer range plans involved increasing the permanent staff with two Health Physics 1
Supervisors, ten Health Physics technicians, four full time Decon Technicians, one ;
Radwaste Shipper, and two professional Health Physics Specialists. l
Proaram Performance Assessment l
i
l
The inspectors were informed that the Nuclear Safety Assessment Review I
Committee (NSARC), Radiation Protection Subcommittee was disbanded, and an i
Independent Radiological Oversight Committee (IROC) was established. IROC reports l
directly to the Vice President of Operations, and the Committee includes three
technical and one organizational expert. The former NSARC Radiation Protection ,
subcommittee chairman was also a member of IROC. Maine Yankee management I
stated that this change was designed to gain greater oversight of radiological l
controls program at Maine Yankee.
I
1
Cultural issues
The inspectors were informed that efforts were underway to address cultural issues
in the radiological controls department at Maine Yankee. These efforts involved
conducting various team building activities, performing integrated scenario training I
with other work groups such as maintenance, and providing specialized l
organizational efficiency (e.g., Covey training). The inspector noted that while this
initiative may support overall program improvement, clear evidence of substantial j
change in cultural attitude has yet to be demonstrated.
l
l
Work Coordination !
l
The inspectors were informed that concerted efforts had been taken to improve work
planning and coordination. Examples included assigning a member of the ALARA <
staff to the maintenance planning organization and the outage planning and l
integration team (OPIT), and developing generic radiation work permits with specific I
radiological controls incorporated into maintenance work packages / orders. The !
ALARA supervisor stated that these changes allow for radiological controls to be l
, integrated into maintenance planning packages during the early stages of )
development, and allow controlling documents with specific radiological controls to
be taken into the field.
Learnina Process
The inspectors were informed that a new problem identification / problem resolution
program, named "The Learning Process," had been implemented. Six current
problem identification / resolution programs were merged into the Learning Process
including Radiological incident Reports (RIRs) and Unusual Occurrence Reports
(UORs). The Learning Process is a computer based program that allows any plant
worker to identify issues and enter them into the learning process. Issues are
screened for risk significance, address interim corrective actions, and recommend
" Issue Owners." This program was designed to eliminate multiple systems, allow for
consistent evaluatic,n, lower the reporting threshold, and allow for better information
dissemination and trending.
l
. - - - . . . .- - .. . .. . ~ - - - . . --. .
-
4
- -
33
c. Conclusions
Based on this review the inspectors concluded the following.
--
Licensee management were supportive of, and driving, program
improvements.
--
The current radiological controls organization was configured to make
, substantive program improvements, without distracting personnel involved
with program implementation.
--
Although the new " Learning Process" (consolidated problem )
'
identification / resolution program) had many potential benefits, the
effectiveness of the program had not yet been demonstrated.
]
5
--
Licensee efforts to improve work coordination, address cultural issues, and
i
develop a comprehensive problem identification / resolution program
'
demonstrated a commitment to performance improvement. Not withstanding,
clear evidence of substantial performance improvement has yet to be
, demonstrated. i
4
- R7 Quality Assurance in RP&C Activities !
-
a. Insoection Scope (83750)
,
i
The inspector performed a review of quality assurance in the area of radiological
-
controls and radioactive waste transportation. Specific areas reviewed included
radiological incident reports, the annual quality assurance audit of the radiological-
1
controls and radwaste transportation programs, and miscellaneous surveillances."
! I
j b. Observations and Findinas
Radioloaical Incident Reports (RIRs) :
5
-
The inspector reviewed eight Radiological incident Reports (RIRs) generated from
July 16,1996 - December 4,1996. RlR documentation included an event
descriptions, immediate corrective actions, causes, and long term corrective actions.
Issues placed into the RIR system received a thorough review as evidenced by
detailed root cause analyses, Event Review Board evaluations, and independent third
party reviews. The inspector noted that corrective actions implemented, specifically
addressed identified weaknesses. Based on this review, the inspector concluded that
the RIR system appeared to be effectively used to identify and evaluate radiological
deficiencies. However, the effectiveness of corrective actions has not yet been
demonstrated.
I
- . . .-. - ~ . .- - . . -- -
.-
!*
34 l
QA Audit Reoort MY 96-03/09 " Combined Rad Controls - Radwaste Audit"
The inspector reviewed the results of Maine Yankee Audit MY-96-03/09. The audit i
- was a combined radiological controls - radioactive waste audit, and the purpose was l
to assess the Maine Yankee's effectiveness in the implementation of the radiation j
protection program. Special technical assistance was provided by Yankee Atomic in
the performance of the audit, and a QA Supervisor from Houston Light and Power ,
also assisted. Results of the audit revealed two deficiencies requiring formal j
responses, and two observations. Overall, the audit found that the Radiation
Protection Program was considered satisfactory to protect personnel from
radiological hazards; controls were in place to allow for operational implementation of l
the program; and implementation of the administrative program was adequate. l
Examples of identified deficiencies and observations / recommendations including I
,
licensee response appears as follows:
1
Some radiation protection QA records were not maintained in accordance with
'
--
procedures in that some radiation work permit (RWP) records were not stored ;
in fire proof cabinets and were not transmitted to the document control group l
in a timely manner. j
l \
In response to this deficiency, the Radiation Protection (RP) section placed the l
. files in fire-proof cabinets or arranged to have the records sent to the !
j document control center.
l
-
The radiation protection section's self assessment and corrective action l
program was not meeting requirements of Administrative Procedure D.65,
Self-Assessment in that facility inspections and observations of procedure
- implementation were not performed.
]~
In response to this deficiency, the Quality Programs Department committed to
addressing this deficiency in the next Self-Assessment / Corrective Action Audit
(Quality Programs).
--
Training records for radwaste workers were not being completely maintained
by training department as required by procedures.
In response to this observation, the RP section agreed to coordinate the
development of specialty training lesson plans with the training department,
and to turnover all training records to the training department.
-- Recommendations were made to increase administrative controls to prevent
personnel contaminations in.non-contaminated areas within the restricted area.
In response to this observation, licensee staff made procedural changes,
initiated training for procedural changes, increased survey frequency in
specific areas, provided additional survey guidance in required reading,
increased staffing of decontamination personnel, and increased
decontamination efforts in overhead areas and general walkways.
.
'
35
The inspector noted that the use of independent auditors was a good practice.
Based on this review the inspector concluded that the audit was overall broad in
scope, of good quality, and effective in the identification and evaluation of program
strengths and weaknesses. However, the overall effectiveness of the audit, in terms
of bringing about substantive improvements, had not yet been demonstrated.
Further, these licensee identified and to be corrected violations are being treated as a
Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.
i
ennual Radwaste Proaram Assessment
The inspector reviewed the Annual Radwaste Program Assessment for 1995. The
program assessment included evaluations of procedures, radwaste minimization,
training, process control program, staffing, documentation, corrective actions, and
recommendations for enhancement. The assessment was performed by the
Radwaste Coordinator and highlighted program strengths and weaknesses. The
inspector noted that the self assessment was good in that it identified areas for
improvement such as the la :k of a dedicated labor force for Radwaste Handlers, lack
of site wide sensitivity to issues of radwaste generation and minimization, and the
lack of a training matrix to specify required training for Radwaste Handlers and
contractor Health Physics technicians. However, acceptance criteria for the
assessment were generally broad. For example, the acceptance criteria for the self
assessment were as follows: ensure regulatory and procedural compliance, ensure
proper tracking of action items, and ensure proper training and qualification of
personnel. The inspector also noted that recommendations for improvement
appeared to be treated as goals rather than action items. Licensee staff responded to
this observation by stating that additional specific acceptance criteria would be
considered for incorporation into future self assessrt + , and recommendations for
improvement were planned to be incorporated into the Learning Process which would
allow for a more formal problem review, tracking, and response.
Quality Assurance Surveillances
The inspector reviewed nine QA surveillances performed in the area of radiological
controls performed from August 1995 to December 1996. The inspector noted that
no surveillance were performed in the area of radwaste management / transportation
during this period.
Surveillances identified program strengths, deficiencies, and numerous
observations / recommendations for improvement. Details of findings were described,
apparent causes were identified, and recommendations for improvement were made.
Information was gathered by reviews of procedures, training records, and numerous
in-plant field observations. Corrective Action Requests (CARS) were written for all
deficiencies. Examples of QA Surveillance Report deficiencies were as follows:
-- 96S-OO1, Contamination control practices were inadequate 1
-- 96S-018, Keys to the Technical Specification 5.12 High Radiation Doors were
not adequately controlled.
1
I
l
i
!
!
-. - - _ . - - . . - . . - -.-.--- ...-.--- --.
!-
4
i
'.
--
96S-026, Formal lesson plans for specialty training programs for contractor
training programs were not developed
The inspector noted that findings were insightful and based on numerous ~ hours of in-
field observations, and corrective actions were initiated to address each deficiency.
Although corrective actions appeared to effectively address the findings, clear
evidence that corrective actions have resulted in substantive program improvements
had not yet been demonstrated. For example, weaknesses persisted in the
contamination control program (See Section R8.2), and due to work prioritization,
actions had not been taken to substantially improve contractor training. Further,.
these licensee identified and to be corrected violations are being treated as a Non-
Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.
c. Conclusions
Based on this review, the inspector made the following conclusions.
-
Issues placed into the radiological incident reporting system were effectively
reviewed to identify and evaluate radiological deficiencies. However, the
effectiveness of corrective actions remains to be demonstrated.
--
The annual QA audit of radiological controls and radwaste transportation was
overall broad in scope and of good quality.
--
Acceptance criteria for the Annual Radwaste Program Assessment were
generalin nature and recommendations for corrective actions were treated as
goals rather than actions items.
- --
QA surveillances were effectively used to identify radiological control program
weakness. However, problem resolution efforts were not fully effective as
evidenced by continuing weaknesses in the area of contamination control and
contractor specialty training.
R8 Miscellaneous RP&C lasues
R8.1 Unanticioated extremity exoosure occurred durina Filter Replacement (Closed,
Unresolved item 50-309/96-09-01)
The inspector performed an onsite review to evaluate the details associated with an
unplanned extremity exposure of 6.954 rem that was reported to have occurred on
April 16,1996, during the replacement of a filter on a radwaste processmg system,
inspectors gathered information through a review of radiological incident report file
RIR 96-010, interviews with cognizant personnel, and discussions held during an
NRC Management meeting conducted in the NRC Region I office on August 9,1996.
The licensee's investigation included a review of records including dosimetry,
radiological surveys, and radiation work permits; a review of current work practices
regarding the use of extremity dosimetry; interviews with cognizant personnel
including the individuals who performed the filter replacement; a review of Maine
.
.
37
Yankee's filter handling experience; the performance of filter replacement mockups; a
review by the Nuclear Safety Assessment Review Committee (NSARC); and an
independent review by industry experts.
Based on the onsite review, it appears that the most likely event that caused the
unanticipated extremity exposure of 6.954 rem occurred during the April 16,1996
radwaste filter replacement. This appears to have resulted from the failure to survey
the filter housing seating surface, after the filter housing was removed and prior to
cleaning the seating surface. Although the extremity exposure of 6.954 person-rem
is a small fraction of the 50 rem dose limit to the extremities, the failure to survey
the filter housing seating surface, after the filter housing was removed and prior to ,
cleaning the seating surface is a violation of procedure 9-302-5, Radiological Survey
and Control of Hot Particle Areas," which requires a survey to be performed "upon
exposing unsurveyed surfaces (e.g., systems / component breaches)."
We note that the following corrective actions were taken in response to this event.
--
An comprehensive evaluation of procedures was performed with regard to l
planning and providing coverage to work associated with hot particles.
--
Instructions and controls for discrete radioactive particles were consolidated
into one major procedure: 9-305-106, Radiological Surveys and Controls for
Discrete Radioactive Particles, Rev. 0
-
Extremity dosimetry is now required to be used on all primary liquid system
filter replacements.
--
Extremity dosimetry used for filter replacements with dose rates greater than
1 R/h will now be sent for immediate processing. '
-
An evaluation of the use of long handled tools for filter replacements was
performed.
--
An in depth dose investigation was performed and dose assignments were
reviewed by independent technical experts.
--
The radiological incident report was included in Technician Required Reading
file, and the event was reviewed in continuing training. ;
The inspector noted that this event was self-identified by the licensee, could not
have reasonably been prevented by corrective actions implemented within the last
two years, corrective actions sufficient to prevent recurrence were implemented, and
the violation was not willful. Accordingly, this licensee identified and corrected
violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of
_ _ _ _ _ _ _ _ _ _ _ . _ . .____ . _. __ _ _ _ _ _ _ _ _ _ . _ _
- .
i
1* '
38
! R8.2 RIR 96-016 Discrete Particle Exposure in Chair Used by Securitv/Firewatch Personnel
a. Insoection Scope (83750) .
The inspector performed a review to evaluate the circumstances surrounding the
December 4,1996, discovery of a discrete radioactive particle found in a chair used
by Security /Firewatch personnel. Information was gathered by review of radiological
incident report number 96-016, interviews with cognizant personnel, and verification
of skin dose calculations using the VARSKIN computer software,
b. Observations and Findinas
l
On December 4,1996, the radiological controls staff conducted a non-routine
radiological survey of f abric chairs within the radiologically controlled area. During
that survey, a discrete radioactive particle _ was found in a chair located inside the I
Containment Access Guard Shack that was routinely used during power operations !
by Security /Firewatch personnel. Upon discovery, immediate corrective actions were !
taken to remove all fabric chairs from the restricted area, and to survey other !
commonly encountered items such as phones, door knobs, tables, etc. i
Detailed evaluations were performed to establish a dose model for estimating
personnel dose. This was based on qualitative and quantitative measurements of the
radioactive particle; evaluations of the location of the particle within the fabric chair; i'
evaluation of the shielding provided by the chair foam, chair fabric, and clothing;
estimates of personnel occupancy; evaluations of how personnel sit in chairs; and
skin dose calculations performed using the computer software code VARSKIN.
The particle was counted by gamma spectroscopy and determined to be a fission
fragment. The age of the particle (post irradiation) was determined by using the
decay law to calculate the difference between the observed ratio of certain isotopes
to the ratio expected for 65-day aged fuel'. The age of the particle was determined
to be 3.3 years post irradiation. Once the age of the particle (post irradiation) was
determined, the total particle activity on 1/1/96, including non-gamma emitting
isotopes such as Sr-90, Y-90, and Pm-147, and short lived isotopes, was calculated
using the decay law. The particle was modeled as a point source beneath 2.39 mm
of material (i.e., chair covering materials and uniform of workers) at 0.450 g/cc and
. the dose rate was. averaged over 1 square centimeter at a tissue depth of 7 mg/cm 2.
The software program VARSKIN MOD 2 was used to calculate the dose rate per pCi
of activity for each isotope, and the total dose rate was summed,
i
The exposure model assumed the following:
--
A discrete radioactive particle (fission fragment) became embedded in the I
fabric of a chair used in containment during the 1995 steam generator
sleeving outage.
NRC NUREG/CR-5873, "VARSKIN MOD 2 and sADDE MOD 2: Computer Codes for Assessing skin Dose from Skin
Contamination." December 1992 !
-. - _ . . - _-- - .~ - . _ - . . - - - . . - . .. .- .- ,
.
l
39
--
On 1/1/96, the fabric cha; was removed from containment and placed in the
Containment Access Guard Shack.
!
-
--
On 1/1/96 the discrete particle fission fragment had an age of 3.3 years post j
irradiation anc: a total activity of 0.218 curies.
!
3
--
The discrete particle was a point source shielded beneath 2.39 mm of material
(i.e., chair covering materials and uniform of worker) at 0.450 g/cc. ,
--
The particle dose rate was averaged over one square-centimeter (cm'), at a
tissue depth of 7 mg/cm'.
--
The particle activity remained constant during the entire year (i.e., the model
conservatively assumed that there were no activity losses due to radioactive j
decay). ;
i
--
The maximum dose rate to the skin was 0.109 rem /h. !
-
Each time an individual sat in the chair, they stayed perfectly still for one hour.
f
--
Due to the steep dose rate gradien, around discrete particles, and taking into !
consideration the way people sit in chairs, it was assumed that it was unlikely :
that the same square centimeter of tissue was irradiated more than once, even
after repeated exposures. l
Using this model, the maximum dose an individual would have received, even from
repeated exposures, would have been 0.109 rem to the skin.
The licensee then submitted the dose model and calculations to independent industry d
experts for review and evaluation. Both Maine Yankee and their consultants ,
concluded that the estimated activities of radionuclides in the particle and the !
calculated dose of 0.109 rem was a reasonable upper estimate of the dose received ;
by workers from exposure to the subject particle, and since 0.109 rem was a small 1
fraction of the 50 rem skin dose limit, further refinements in dose estimation were -!
not warranted.
l
)
The licensee then used security logs to identify each individual that may.have used J
the chair in 1996, and a skin dose of 0.109 rem was entered into each individual's i
personal exposure file. l
Based on the NRC review of the dose model and calculations, inspectors made the
following observations.
--
The methods Maine Yankee used to estimate and assign radiation dose were
reasonable, and followed accepted industry practices.
1
.
.
!
40
--'
Although the exposure model presented included inherent uncertainties, the ;
actual doses received by personnel were likely to have been a small fraction of ;
the 50 rem skin dose limit.
--
No observable biological effects would be expected for any of the individuals
in this exposure scenario.
Root Causes and Causal Factors
Maine Yankee determined that the root cause for this event was inadequate- j
contamination controls. Barriers for contamination controls including the
performance of proper surveys for personnel and materials leaving contaminated
areas failed to detect the presence of the discrete particle, in addition, the failure of
the routine surveillance program to detect the presence of the discrete particle was 3
also cited as a causal factor. )
i
Lona Term Corrective Actions '!
I
Maine Yankee has taken the following long term corrective actions.
--
Implemented a routine survey program to direct frisk commonly used plant l
equipment and security watch stations occupied by personnel in the Restricted !
Area (phones, chairs, pages, door knobs, etc.) )
--
Included lessons learned for this topic in Health Physics Technician I
recertification training.
--
Initiated the development of a tool control program to minimize the movement
of tools in and out of the restricted area. !
!
--
Incorporated this incident into Plant Root Cause Evaluation No. PRCE-211 to l
develop recommendations to improve contamination controls.
J
The inspector noted that Maine Yankee's corrective actions were adequate to ;
prevent recurrence of this specific event, and were generally good. However, the ;
inspector noted that weaknesses in the contamination control program persist. For l
example, procedural guidance presented in 9-5-11, " Contamination
Control / Decontamination Program" requires items being removed from a
contaminated area to be " wiped down to remove gross contamination," and to be
surveyed for contact exposure rates. Procedural guidance does not require items
being removed from a contaminated area to be " frisked" for fixed contamination,
unless the item is also being removed from the restricted area. Consequently, using
current procedural guidance, the potential exists for items with low levels of fixed
contamination, to be released from contaminated areas (within the restricted area).
Based on this review, it appears that the root cause of the unanticipated skin
exposures of 0.109 rem, resulted from the failure to perform adequate surveys at the
contaminated area boundary to the Vapor Containment, or the failure of the routine
surveillance program to identify the presence of the particle. Although 0.109 rem is
.
.
41
a small fraction of the 50 rem allowable skin dose, the failure to perform adequate
surveys is a violation of NRC requirements, specifically 10 CFR 20.1501, " Surveys
and Monitoring." However, we note that this event was self-identified by the
licensee, could not have reasonably been prevented by corrective actions
implemented within the last two years, corrective actions sufficient to prevent
recurrence were implemented, and the violation was not willful. Accordingly, this
licensee identified and corrected violation is being treated as a Non-Cited Violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy.
c. Conclusions
--
The methods Maine Yankee used to estimate and assign radiation dose were
reasonable, and followed accepted industry practices.
--
Although the exposure model presented included inherent uncertainties, the
actual doses received by personnel were likely to have been a small fraction of
the 50 rem skin dose limit.
--
No observable biological effects would be expected for any of the individuals
in this exposura scenario.
--
Corrective actions were adequate to prevent recurrence of this specific event,
and were generally good.
--
Although efforts were underway to improve contamination control and
contamination monitoring, the potential exists for items with low levels of
fixed contamination to be released from contaminated areas, using current
procedural guidance.
--
Contamination controls continue to be a program weakness at Maine Yankee.
R8.3 92904 "Followuo Plant Suncort"
(Closed) Violation 50-309/95-06-02
l
The inspector performed a review to evaluate licensee response to NRC Violation !
50-309/95-06-02, " Failure to Maintain Survey Records." The violation was cited j
because the licensee could not locate a copy of a radiation survey that was i
performed at about 4:50 p.m. on February 13,1995, when elevated radiation levels
were identified on a hydrolaser wand tip in the reactor cavity upender pit. The l
inspector reviewed the July 13,1995 Reply to Notice of Violation for NRC Inspection j
Report 50-309/95-06, reviewed procedures and training documents, and discussed
corrective actions taken with various members of the health physics staff.
The inspector verified that specific corrective actions described in the licensee's
response letter, dated June 13,1995, were complete. The following corrective 1
actions were taken. The licensee developed procedure 9-303-113, Rev. O, that l
provided enhanced instructions on survey preparation, review and routing; i
administered Radiological Controls Continuing Training Lesson Plant (RCCT 97-15), !
1
. i
,o 1
42
that included instruction on improving survey documentation; and issued a March 31,
1995 " read and sign" memorandum entitled " Performance of Routine and RWP
Radiological Surveys." The inspector noted that the corrective actions taken
specifically addressed the violation and identified weaknesses, and were adequate for !
closure. This item is closed. l
(Closed) Inspector Followuo item (IFI) 95-06-04
IFl 50-309/96-04-04 " Quality of Radiological Surveys and Documentation" was
opened because inspectors noted that the overall quality of radiological survey
documentation was degraded. Examples included inconsistency in use of units
(mR/h versus R/h), some survey data was illegible, and there was some omission of
pertinent conditions (e.g., water level, surfaces dry or wet, equipment in work area).
The inspector reviewed licensee corrective actions taken to address these concerne.
The inspector reviewed procedures and training documents, discussed corrective
actions taken with various members of the health physics staff, and reviewed
examples of radiological surveys.
The licensee took the following corrective actions. Developed procedure 9-303-113,
" Survey Documentation and Review," Rev. O, to include detaiied instructions on
survey preparation, review and routing; administered Radiological Controls Continuing
Training (RCCT 97-15) to health physics technicians that included instruction on
improving survey documentation; issued a March 31,1995 " read and sign"
memorandum entitled " Performance of Routine and RWP Radiological Surveys" that
provided specific management expectation for the documentation of radiological
surveys; and contracted with a vendor to improve the overall quality of radiological
survey maps. The inspector noted that these corrective actions specifically
addressed the identified weaknesses and appeared effective. In addition, the
inspector reviewed a sampling of radiological surveys from the Primary Auxiliary
Building (PAB) and the Vapor Containment (VC). Surveys were clear, legible,
complete and useful for ALARA planning purposes. Finally, the inspector reviewed a
sampling of radiological survey maps produced by a vendor and proposed for use
(awaiting final supervisory approval), and noted that these survey maps were of very
high quality. The inspector concluded that the corrective actions for this IFl were
very good, and sufficient for closure. This item is closed.
(Closed) Violation 95-06-03
The inspector performed a review to evaluate licensee response to NRC Violation
50-309/95-06-03, " Dosimetry Not Located in Highest Region of Exposure." The
violation was cited for the f ailure to relocate whole body dosimetry to the knees,
which was determined to be the whole body region of highest exposure, during work
with a reactor coolant pump rotating element. The inspector reviewed the July 13,
1995 Reply to Notice of Violation for NRC Inspection Report 50-309/95-06, reviewed
procedures and training documents, and discussed corrective actions taken with
various members of the health physics staff.
.
.
43
The inspector verified that specific corrective actions described in the licensee's
July 13,1995, Reply to Notice of Violation letter were complete. Procedure
9-7-100, " Dosimetry Program" was revised to include specific instructions on
requirements for relocating dosimetry, and this topic was addressed in continuing
training provided in 1995. The inspector noted that these corrective actions
specifically addressed the violations and identified weaknesses, and were adequate
for closure. This item is closed.
(Closed) Violation 96-04-03
The inspector performed a review to evaluate licensee response to NRC Violation
50-309/96-04-03, " Failure to Brief Workers to Minimize Exposure." This violation
was issued for the February 29,1996, failure to inform personnel of precautions and
procedures to minimize their occupational radiation exposure relative to maintenance
activities on the CH-138 valve (vent valve on the volume control tank). The
inspector reviewed the April 25,1996 Reply to Notice of Violation for NRC
Inspection Report 50-309/96-04, reviewed procedures and training documents, and
discussed corrective actions taken with the maintenance manager.
The inspector verified that specific corrective actions described in the licensee's
response letter, dated April 25,19965, were complete. Procedure 67-200-14,
" Pre-Job Brief Guidelines" Rev. O, had been developed that included directions for
preparing and conducting pre-job briefings; Specialty Training Lesson Plan MS-15-02,
"Using Pre-Job Brief Guidelines" was developed and based on a review of training
records, implemented; and Maintenance and Operations memoranda addressing pre-
job briefings and improved communications were issued. The inspector noted that
corrective actions appeared effective, and specifically addressed the violation and
identified weaknesses. This item is closed.
R8.4 Unscheduled Gaseous Releases
a. Insoection Scoce (84750)
The inspector performed a limited review of licensee policies for estimating effluent
releases, making notifications to the State of Maine when estimates exceed planned
releases, and licensee efforts used to investigate discrepancies between planned and
actual releases. Information was gathered by a review of correspondence between
the State of Maine's Radiation Control Program and Maine Yankee Atomic F: ier #
Company regarding required State notifications for unscheduled releases, reviews of
selected Unusual Orcunence Reports (UORs), and interviews with cognizant
personnel.
.- .- . - _
-
la
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44
b. Observations and Findinos
, State of Maine Notifications
{ Maine Yankee provides the State of Maine public notification of estimates of gaseous
and liquid effluent releases via a Public Information L.ine (1-800-762-7104). Maine
, Yankee also has an agreement with the State of Maine to notify the State if actual
releases exceed estimates by an agreed upon amount. For example, the following
table summarizes the policy for required notifications to the State of Maine for
! gaseous releases.
.
Estimated Gaseous Release Published on The State of Maine Must Be Notified if the
i
the Public Information Line (Pil)
'
Actual Release Exceeds the Estimate
, 800-762-7104 Published on the Pil by the following
'
(X, = Planned release in curies) (X, = Actual release in curies)
X, < 0.01 Ci X, A 0.02 Ci
-
0.01 Ci < X, < -10 Ci X,12 times original estimate
10 Cii X, < 50 Ci X,1 Original Estimate + 10 Ci
50 Cii X, < 100 Ci X,1 Original Estimate + 20 Ci
100 Cii X, < 250 Ci X,1 Original Estimate + 25 Ci
X,1250 Ci X,11.1 times original estimate
The inspector noted that based on Maine Yankee's effluent release notification
agreement with the State of Maine, some effluent release notifications to the State of
Maine were required even at very low effluent release levels. For example, if the
original gaseous release estimate published on the Public Information Line was 0.03
curies, Maine Yankee would be required to notify the State of Maine if the actual-
release was 0.00 Ci, or 0.03 curies above the original estimate. On the other hand,
if the original estimate was 10 curies, Maine Yankee would not have to notify the
State of Maine if the actual release was 19 curies, or 9 curies above the original
estimate.
Review of Gaseous Releases
The inspector reviewed examples of gaseous effluent releases that required
notification to the State of Maine. Information was gathered through reviews of
Unusual Occurrence Reports, and interviews with cognizant personnel. Examples of
gaseous releases reviewed appear in the following table.
.
.
45
Examples of Maine Yankee Gaseous Effluent Releases
That Required Notification of the State of Maine
Estimated Actual State dstimated Dose Reason / Source of
Date Release Release Notification at Site the Release
curies curies Required Boundary
(mrem)
Giireleased from
02/29/96 <0.01 0.16 Yes 3.OE-5 volume control tank
via CH-138 valve
10/14/96 <0.01 0.103 Yes 5.2 E-6 No known cause
10/30/96 <0.022 0.053 Yes 3.8 E-6 Chemistry sample
line left on
Maintenance
11/19/96 0.09 0.77 Yes 3.9E-5 activity on charging
pump vent
MI
The inspector noted that, in accordance with Maine Yankee's effluent release
notification agreement, the State of Maine (and consequently the NRC) was required
to be notified of the releases listed in the above table, even though offsite dose
consequences were negligible.
The inspector was informed that Maine Yankee had recently made a change to the
method used for estimating releases. In the past, Maine Yankee used a default value
of 0.01 curies for estimating gaseous releases when stack sample results were less
that the lower limit of detection. Maine Yankee now intends to use the actuallower
limit of detection (LLD) value for estimating stack releases, when stack samples are
below the detection limit. Since the LLD value is approximately a factor of 10 larger
that the historical default value, the use of the LLD value for estimating gaseous
releases below the detection limit will effectively reduce the number of State
notifications that have to be made. The inspector noted that this program change
only affects releases that would have negligible offsite dose consequences.
The inspector noted that information related to unscheduled gaseous releases was
somewhat difficult to review due to the fact that no one group (e.g., Chemistry,
Operations, or Health Physics) had overall responsibility for accumulating or
compiling information related to unscheduled releases. The Operations department
maintained Unusual Occurrence Reports, the Chemistry department performed
sampling and maintained a log book that documented releases, and the Health
Physics department had performed air sampling to help identify the source of some
releases, in response to this observation, the inspector was informed that Maine
.
,.
46
Yankee was attempting to better identify the source of releases through team work
with the Operations, Chemistry, and Health Physics department, and that details of
unscheduled gaseous releases were being entered into the new " Learning Process,"
which should allow for a more comprehensive or global review of issues related to
unscheduled offsite releases.
c. Conclusions
Based on this review, the inspector made the following conclusions.
--
Gareous releases that required notification of the State of Maine in 1996 had
minimal offsite dose consequences.
--
Maine Yankee's use of the Lower Limit of Detection value for estimating
gaseous releases below the stack sampling detection limit, will reduce the
number of required notifications made to the State of Maine for minor
releases.
--
Maine Yankee's methods for planning and estimating gaseous releases, and
investigating the causes of unscheduled releases were hampered by the fact
that no one group had overall responsibility for compiling an overall summary
of issues related to unscheduled releases.
RB.5 Mis-characterization Contained in ERB-010
a. Insoection Scope (83750)
The inspector reviewed licensee actions with regard to the discovery that
inaccuracies were identified in Maine Yankee's Event Review Board (ERB) Report No.
ERB-010, " Unplanned Worker Exposure In RCP Impeller Decontamination Container,"
April 27,1935, that was submitted to the NRC. Information was gathered by a
review of ERB-010; a January 31,1997 letter from Maine Yankee to NRC Region I
clarifying ERB-010; and interviews with cognizant personnel,
b. Observations and Findinas
j Maine Yankee management informed the NRC that as a result of inquiries made by a
member of the public and information brought forward by Maine Yankee staff
, members, additional review of Event Review Board Report No. ERB-010 and of the
March 24,1995 Unplanned Exposure Event, had been performed. Based on
additional review, Maine Yankee concluded that ERB-010 mis-characterized the role
of one particular person mentioned in the report, specifically, the " Balance of Plant"
(BOP) Bartlett Supervisor. Maine Yankee management stated that Event Review
Board Report No. ERB-010 indicated that the Bartlett BOP Supervisor had a direct
role in the event, when in f act, the Bartlett BOP Supervisor had only a peripheral role
in the event. This conclusion was supported by additional information gathered as a
result of an investigation subsequent to the ERB, relative to the actual involvement of
the Bartlett BOP Supervisor in this event. A member of the Event Review Board also
.
.
47
stated that the Bartlett BOP Supervisor's involvement in the March 24,1995 event
was mis-characterized because information obtained during interviews was not fully
verified due, in part, to imposed time constraints.
Maine Yankee management stated that in order to clarify the information provided to
the NRC in Event Review Board Report No. ERB-010, a letter clarifying the role of the
Bartlett BOP Supervisor' involvement in the March 24,1995, Unplanned Exposure
Event would be submitted to the NRC Region I office. (Note: This letter was
received in the NRC Region I office on February 5,1997.)
Based on interviews conducted with Maine Yankee Health Physics technicians and ;
supervisory personnel, the inspector also concluded that the Bartlett BOP Supervisor i
did not have a direct role in the March 24,1995 Unplanned Expr,5um Event.
1
ERB-010 identified numerous deficiencies in the Maine Yankee radiologi:al control J
program, and Maine Yankee accepted responsibility for the events leartng to the l
March 24,1995 administrative overexposure. Accordingly, the mipcharacterization
of the Bartlett BOP Supervisor's involvement in the March 24. W95 event as
documented in ERB-010, appears to have been inadvertent. The inspector
determined that it was proper for Maine Yankee to have clarified the NRC docket
with regard to the mis-characterization. Notwithstanding, the corrected information
had no material significance relative to NRC inspection findings or subsequent
enforcement action. Consequently, there is no regulatory action required in this
matter.
c. Conclusion
Based on this review, NRC made the following conclusions.
--
Maine Yankee submitted a letter to the NRC Region i Office dated January 31,
1997, that clarified information provided to the NRC in Event Review Report
No. ERB-010. This clarification indicated that a Bartlett Balance of Plant
supervisor had no direct role in the March 24,1995 event. l
-- The corrected information contained in the January 31,1997, correspondence
to the NRC Region I office, relative to Maine Yankee Event Review Board
ERB-010, had no material significance relative to NRC inspection findings or 4
subsequent enforcement action. Consequently, there is no regulatory action
required in this matter.
R8.6 UFSAR Review
A recent discovery of a licensee operating their facility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures and/or parameters
,
_ _. .__ __ __ -. . _ _ _ _m __ . . _ _ _ _
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4
to the UFSAR description. While performing the inspections discussed in this report, l
the inspectors reviewed the applicable portions of the UFSAR that related to the ;
areas inspected. I
The inspector reviewed selected sections of Chapters 11, " Radiation Protection" of
i
the Updated Final Safety Analysis Report (UFSAR), pertaining to radiological controls,
to evaluate the accuracy of the UFSAR regarding existing plant conditions and l
j practices.
,
c. Conclusions
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No UFSAR discrepancies were identified during this review. l,
l S1 Conduct of Security and Safeguards Activities
l a. Insoection Scoce (71750) I
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! The inspector reviewed the security program during the period of January 6-9,1997.
Areas inspected included: effectiveness of management control; management
support; protected area detection equipment; alarm stations and communication; l
testing, maintenance and compensatory measures; and training and qualification.
,
{
The purpose of this inspection was to determine whether the licensee's security '
program, as implemented, met the licensee's commitments in the NRC-approved i
security plan (the Plan) and NRC regulatory requirements.
'
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b. Observations and Findinas '
Management support was ongoing as evidenced by the procurement of portable.
explosive detectors for personnel processing, completion of the vehicle barrier
system, hiring of four additional security officers, and the installation of permanent
electrical outlets around the protected area to support temporary lighting needs.
Alarm station operators were knowledgeable of their duties and responsibilities and
security training was being performed in accordance with the NRC-approved training
and qualification (T&O) plan. Management controls for identifying, resolving, and
preventing programmatic problems were effective as demonstrated by a reduction in
security related events.
Protected area (PA) detection equipment satisfied the NRC-approved physical
security plan (the Plan) commitments and security equipment testing was being
performed as required by the Plan. Maintenance of security equipment was being
performed in a timely manner as evidenced by minimal compensatory posting
associated with non-functioning equipment. In a licensee requested meeting held in
Region I on June 5,1996, the licensee presented its actions to correct recurring
programmatic weaknesses involving the improper control of safeguards information,
security vital area keys, and designated vehicle keys. Additionally, the licensee
discussed an issue concerning the improper installation of the vehicle bomb threat
.. - . _ - - - - -- -- - -- . .- -
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1
barrier and a weapons handling event involving an accidental discharge. The
inspector reviewed the licensee's corrective actions during this inspection and -
>
- determined that the efforts to resolve the weaknesses appeared to be effective.
c. Conclusions l
l The inspector determined that the licensee was conducting its security and
.
safeguards activities in a manner that effectively protected public health and safety ,
l and that the program as implemented met the licensee's commitments and NRC
i requirements.
] S2 Status of Security Facilities and Equipment '
)
l S2.1 Protected Area Detection Aids
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! a. Insoection Scope (71750)
i
The inspector conducted a physical inspection of the PA intrusion detection systems
(IDSs) to verify that the systems were functional, effective, and met licensee
.
commitments.
b. Observations and Findinas
j On January 7,1997, the inspector determined by observation that the IDSs were
functional and effective, and were installed and maintained as described in the Plan.
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i c. Conclusion
!
- The PA intrusion detection aids met the licensee's Plan commitments.
!
! S2.2 Alarm Static.ns and Communications
'
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a. Insoection Scope (71750)
i Determination whether the Central Alarm Station (CAS) and Secondary Alarm Station
j (SAS) are: (1) equipped with appropriate alarm, surveillance and communication
j capability, (2) continuously manned by operators, and that (3) the systems are
i
independent and diverse so that no single act can remove the capability of detecting
a threat and calling for assistance, or otherwise responding to the threat, as required
by NRC regulations.
'
j b. Observations and Findinas
Observations of CAS and SAS operations verified that the alarm stations were
equipped with the appropriate alarm, surveillance, and communication capabilities.
Interviews with CAS and SAS operators found them knowledgeable of their duties
and responsibilities. The inspector also verifiea through observations and interviews
j that the CAS and SAS operators were not required to engage in activities that would
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interfere with the assessment and response functions, and that the licensee had
exercised communications methods with the locallaw enforcement agencies as
committed to in the Plan,
c. Conclusion
The alarm stations and communications met the licensee's Plan commitments and
NRC requirements.
S2.3 Testina, Maintenance and Compensatorv Measures
a. Insoection Scope (71750)
'
Determination whether programs were implemented that will ensure the reliability of
security related equipment, including proper installation, testing and maintenance to
replace defective or marginally effective equipment. Additionally, determination that
when security related equipment failed, the compensatory measures put in place
were comparable to the effectiveness of the security system that existed prior to the
failure,
b. Observations and Findinas
Review of testing and maintenance records for security-related equipment confirmed
that the records were on file, and that the licensee was testing and maintaining
systems and equipment as committed to in the Plan. A priority status was being i
assigned to each work request and repairs were normally being completed in a timely I
manner for all work, necessitating compensatory measures. However, the inspector :
questioned the testing rational adopted by the licensee for certain IDS zones that j
have been returned to service after maintenance and repairs. The licensee is carrying l
out the same test procedure used during the required weekly functional test after j
maintenance and repairs. The inspector discussed this with security management j
and explained that more stringent testing after maintenance and repairs would be
prudent to ensure that the equipment is capable of meeting allits performance
requirements. Based on this concern, the inspector requested that the licensee
conduct testing of some IDS zones to ascertain that the zones were functional and
operable. The tests were performed per the plan procedures and the inspector was
able top ascertain that the zones were operable. Although there wqere weaknesses ;
in the procedures, they were adequate at verifying IDS zones operability after.
maintenance and repairs,
c. Conclusions
Security equipment repairs were timely and the use of compensatory measures was
found to be appropriate and minimal. Weaknesses were identified in the Post
Maintenance Testing for IDS zones. The licensee agreed to revise the post
maintenance testing procedure for IDS zones as a program enha cement.
_. _
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S5 Security and Safeguards Staff Training and Qualification
a. Insoection Scoce (71750)
I
Determination whether members of the security organization were trained and J
qualified to perform each assigned security related job task or duty in accordance
with the NRC-approved T&O plan.
b. Observations and Findinas
On January 9,1997, the inspector met with the security training staff and discussed
training department enhancements and program initiatives implemented since the + ;
previous program inspection conducted in December 1996. One enhancement I
discussed involved including the security procedures with the training lesson plans,
the development and implementation of a tactical weapons course used during the
most recent requalification training session and the conduct of tactical response drills
as a means of reinforcing the importance of initial response deployment. j
Additionally, the inspector interviewed a number of security force members (SFMs) !
to determine if they possessed the requisite knowledge and ability to carry out their ;
assigned duties. ;
)
c. Conclusions
The inspector determined that training had been conducted in accordance with the
T&O plan, Based on the SFMs responses to the inspector's questions, the training I
provided by the security training staff was considered effective. l
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S6 Security Organization and Administration
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a. Insoection Scoce (71750) l
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A review of the level of management support for the licensee's physical security
program was conducted.
b. Observations and Findinas
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The inspector reviewed various program enhancements made since the last prc. gram
inspection, which was conducted in December 1996. . These enhancements included
the procurement of portable explosive detectors for personnel processing, completion
of the vehicle barrier system, hiring of four additional security officers, and the ;
installation of permanent electrical outlets around the protected area to support
temporary lighting needs.
c. Conclusions
Management support for the physical security program was determined to be
excellent.
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S7 Quality Assurance in Security and Safeguards Activities
S7.1 Effectiveness of Manaaement Controls
a. Insoection Scooe (71750)
A review was conducted to determine if the licensee had controls for identifying,
resolving and preventing programmatic problems. I
b. Observations and Findinas ,
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The inspectors determined that the licensee had controls for identifying, resolvmg,, !
and preventing security program problems. These controls included the performance i
of the required annual quality assurance (QA) audits, the licensee's self-assessment i
program under which approximately two hundred (200) assessments were conducted l
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in 1996, the security contractor's Performance Assessment Program, and the
performance of root cause analysis for each human performance error. .The licensee
also utilizes industry data, such as violations of regulatory requirements identified by
the NRC at other facilities, as a criterion for self-assessment. j
c. Conclusions
A review of documentation applicable to the licensee controls, including results,
indicated that performance errors were being minimized and that controls were
effectively implemented to identify and resolve potential weaknesses.
S7.2 Audits
a. Insoection Scope (71750)
The inspector reviewed the licensee's NRC-required audit of the security program to
determine if the licensee's commitments as contained in the NRC Plan were being
satisfied. j
b. Observations and Findinas
The inspector reviewed the 1996 QA audit of the security program conducted
September 9-13 and 17-18,1996, (Audit No. MY-96-04). The audit was found to
have been conducted in accordance with the Plan. To enhance the effectiveness of
the audit, the audit team included two independent security specialists. The audit
report identified no adverse findings and had six observations, which were not
indicative of programmatic weaknesses but, if corrected, would enhance program
effectiveness. The audit results had been disseminated to the appropriate levels of
management. The inspector determined, based on discussions with security
management and a review of the responses to the observations, that the corrective
actions were effective.
.
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c. Conclusions
The review concluded that the audit was comprehensive in scope and depth, that the
findings were appropriately distributed and that the audit program was being properly
administered.
S8 Miscellaneous Security and Safeguards issues
S8.1 Uodated Final Safety Analysis Report Review
A recent discovery of a licensee operating its facility in a manner contrary to the
UFSAR description highlighted the need for a special focused review that compares
plant practices, procedures, and parameters to the UFSAR description. Since the
UFSAR does not specifically include security program requirements, the inspector
compared licensee activities to the NRC-approved physical security plan, which is the
applicable document. While performing the inspection discussed in this report, the
inspector reviewed Section 9.5 of the Plan, Revision 23, dated May 14,1996, titled,
" Vehicles." Based on discussions with security management, reviews of procedures,
and observations, the inspector determined that vehicles were being maintained and
controlled as described in the Plan and applicable procedures.
S8.2 Review and Closeout of Security Event Reoorts
a. Insoection Scoce (90712)
The inspector reviewed Security Event Reports to determine if the requirements of 10
CFR 73.71 were complied with and to ensure that the licensee's corrective actions
were appropriate and timely. The inspectors verified that each event description was
consistent with their knowledge of the event, that the submittal of the report was -
timely and that the details of the event were clearly reported. The inspector
determined whether further information was required from the licensee, whether
generic implications existed, and whether events warranted further onsite followup.
b. Observations, Findinos and Conclusions
The events discussed in the following SER's were inspected by the inspectors and
documented in various previous reports or discussed in this report.
SER 95-S01 Failure to Maintain Compensatory Measures for a Degraded Protected
Area Barrier - NRC Inspection Report 50-309/95-02.
SER 96-S02 Safeguards Information Found Uncontrolled - NRC Inspection Report
50-309/96-13.
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SER 94-S02 Dearadation of a Security Access Controlled Barrier (Closed)
The inspector reviewed the SER and determined that Maine Yankee properly reported
the degraded security barrier and provided the required compensatory measures as
required by the station Security Plan. The information in the SER was determined to
be appropriate for the circumstances and this was not a generic issue. The root
cause determination and corrective actions were appropriate and timely. The
corrective actions were to immediately take the ventilation gate out of service, a j
modific.: tion to the gate was implemented and the alarm switch was adjusted to
ensure the alarm would activate at the proper time. In addition security supervisors a
received reirforced training on the need to properly assess conditions brought to their I
attention. The inspector inspected the barrier and determined that the licensee's l
corrective actions appear to properly resolve the problem. l
The inspectors determined that Maine Yankee's assessment of the events were
consistent with the inspectors findings during their inspections of the events. The
licensee implemented adequate corrective actions to address the events and the
reports met the requirements of 10 CFR 73.71. I
F8 Miscellaneous Fire Protection issues
F8.1 Review and Closeout of Licensee Event Reports
a. inspection Scoce (90712)
The inspector reviewed Licensee Event Reports to determine if the requirements of
10 CFR 50.73 were complied with and to ensure that the licensee's corrective
actions were appropriate and timely. The inspectors verified that each event i
description was consistent with their knowledge of the event, that the submittal of
the report was timely and that the details of the event were clearly reported. The
inspector determined whether further information was required from the licensee,
whether generic implications existed, and whether events warranted further onsite
followup. l
b. Observations. Findinas and Findinas
The events discussed in the following LERs were inspected by the inspectors and
documented in various NRC Inspection Report.
LER 95-14, Cardox Zone 1 Activation Wiring Defect NRC Inspection Report
50-309/95-24.
LER 96-17 and 96-17-01 Fire Barrier Penetration Seal Discrepancy - NRC Inspection
Report 50-309/96-08.
___ _. _. _ _ _ _ _ _ _ _ _ _ _ _
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The inspectors determined that Maine Yankee's assessment of the events were
consistent with the inspectors findings during their inspections of the events. The !
licensee implemented adequate corrective actions to address the events and the ;
reports met the requirements of 10 CFR 73.71. !
l
LER 95-05) Rust and Scafe Found in Technical Specification Reauired Fire y
Sorav/ Foam System
The inspector reviewed the LER reporting the rust and scale found in the fire i
spray / foam systems to determine if the requirements of 10 CFR 50.73 were ,
satisfied.-The event was reported in a timely manner with all the required !
information. The rust and scale were found during a refueling outage in portions of
the Turbine lube oil Reservoir Sprinkler and Seal Oil system. The rust and scale
flakes were of sufficient size and quantity that they may have plugged the flow
orifices or spray nozzles. The turbine lube oil reservoir sprinkler and seal oil system i
utilize Aqueous Film Forming Foam (AFFF) to more effectively extinguish oil fires and
is highly corrosive. The root cause of the event was due to improper flushing and
draining of the turbine lube oil reservoir sprinkler and seat oil system sprinkler piping
after inadvertent actuation. The corrective actions were to clean, replace or repitch
the affected piping. The three other non-technical specification systems were j
disassembled inspected and repaired as necessary. Station procedure 19-71, " Fire
Protection System Flush" was developed and issued on December 28,1995, to
provide guidance for flushing the Hydrogen seal Oil unit, Lube Oil Storage Reservoir, j
Emergency Diesel Generators 1 A & 1B and Turbine Driven Feed Pump (P-2C). The ;
procedure was well written with appropriate guidance to perform the required i
flushing activities, however there was no specific guidance as to what constituted an j
acceptable flush. This LER remains open pending further NRC review. l
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j
J V. Manaaement Meetinas
X1 Exit Meeting Summary
i
The inspectors presented the inspection results to members of the licensee on February 10, !
,
1997. The licensee acknowledged the findings presented. !
X 1.1 Cable Seoaration Debrief
f
On December 20,1996, the preliminary results of the cable separation issues -j
j, inspection conducted by regional specialists were discussed with Mr. S. Nichols and
l
- others of your staff. The inspectors did receive proprietary material during the
inspection.
I X1.2 Security Exit Meetina -,
e ,
i On January 9,1997, the inspector met with the licensee representatives at the
i conclusion of the Security inspection. At that time, the purpose and scope of the
,
inspection were reviewed, and the preliminary findings were presented. The licensee i
acknowledged the preliminary inspection findings.
X1.3 Radioloaical Protection Exit Metina ,
i On January 17,1997, the inspector met with the licensee representatives at the
"
conclusion of the radiological controls inspection. ,
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
- +- R. Blackmore, Plant Manager !
- -
J. Frothingham, Quality Program Department Manager !
i *
S. LeClerc, Quality Program Section Head :
- +
"
R. Hayward, Training Manager ;
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J. McCann, Licensing Section Head ;
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M. Veilleux, Maintenance Manger ;
- +- S. Smith, Operations Manger !
l + P. Metivier, Security Manager :
I
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E. Heath, Radiation Protection Manager
i +- G. Leitch, VP, Operations )
P. Radsky, Chemistry Section Head j
- -
J. Connel, Manger - Technical Support i
. *+#- J. Weast, Licensing Engineer !
-* E. Soule, Plant Engineering Department Manager l
1
! + J. Grant, Plant Support Manager j
?
+- W. Tracy, Quality Assurance Supervisor !
j + P. Cunningham, Operations Security Supervisor !
'
+ H. Torberg, Security Programs Coordinator j
i + C. Urquhart, Security Chief, American Protective Services (APS) '
- # W. Barry, Supervisor, Plant Engineering Department .
! # T. Gifford, Assistant Manager, Corporate Engineering Department l
- - S. Nichols, Manager, Corporate Engineering Department !
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D. Ash, Radiological Controls
l J. Blair, Radiological Controls !
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A, Capristo, Supervisor, Radiation Protection )
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S. Edgerly, Quality Programs Division
E. Fox, Radiation Protection Technician l
- J. Hurst, Assistance RPM
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R. Leddy, RP Consultant ;
y
- D. Morehouse, QA Engineer i
! -
M. Readinger, Radwaste Coordinator 1
- F. Smith, Chemistry Section Head
i R. Willis, RP Supervisor
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M. Wyman, Staff Training Supervisor
- C. Young, Radwaste/ Transportation specialist
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NRC i
- #- J. Yerokun, Senior Resident inspector i
+- W. Olsen, Resident inspector
J. White, Chief, Radiation Safety Branch, DRS
-
R. Ragland, Radiation Specialist ;
- +- R. Rasmussen, Resident inspector 1
- L. Scholl, Reactor Engineer I
- R. Croteau, Project Manager, NRR
Other
- + #- P. Dostie, Maine, State Nuclear Safety inspector
- U. Vanags, Nuclear Safety Advisor
Denotes those present at the February 10,1997, Exit Meeting.
+ Denotes those present at the January 9,1997, Security Exit Meeting.
- Denotes those present at the December 20,1996, Cable Separation Debriefing.
-
Denotes those present at the January 17,1997 Radiological Controls Exit Meeting
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INSPECTION PROCEDURES USED
i
IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing
Problems
IP 62707: Maintenance Observation i
IP 64704: Fire Protection Program
IP 71707: Plant Operations ;
IP 73051: Inservice inspection - Review of Program l
lP 73753: Inservice Inspection '
IP 83729: Occupational Exposure During Extended Outages ,
IP 83750: Occupational Exposure '
IP 86750: Solid Radwaste Management and Transportation ,
IP 90712: Followup Review of Non-Routine Events !
IP 92700: Onsite Followun of Written Reports of Nonroutine Events at Power Reactor i
Facilities
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lP 92901: Followup - Op1 ations ,
IP 92902: Followup - Mamtenance l
IP 92903: Followup - Engineering ,
'
IP 92904: Followup -Plant Support
IP 93702: Prompt Onsite Response to Events at Operating Power Reactors
Tl 2515/133 Implementation of Revised Part 49 CFR Parts 100-179 and 10 CFR Part 71 ;
Regulations
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ITEMS OPENED, CLOSED, AND DISCUSSED
ltems Opened:
50-309/96-14-01 VIO Inadequate use of the Spent Fuel Pool crane in January,1997
to move irradiated fuel contrary to Technical specification 3.13,
Refueling and Fuel Consolidation Operations and Procedure 13-
2, Fuel Handling in the Spent Fuel Pool. (Section O2.3)
4
20-309/96-14-02 URI NRC review of the HPSI pumps and valves test results. I
Preiiminary reviews showed that the test was satisfactory. l
However, detailed review of test results is still ongoing by
Maine Yankee and the NRC. (Section E1.1)
504,09/96-14-03 URI Electrical cable separation issues remain open pending NRC
review of licensee actions including the following: the i
I
finalization and implementation of a plan to determine the
extent of the cable separation problems; the resolution of
identified deficiencies; and, the performance of a root cause ;
determination and identification of necessary corrective actions. ;
(E8.1)
50-308/96-14-04 URI The adequacy of Safety Related Logic Testing (GL 96-01)
remains unresolved pending licensee completion of the logic
testing reviews; the resolution of test deficiencies, and NRC
review of these activities. (Section E8.2)
50-309/96-14-05 VIO Contrary to Technical specification 5.11.1 and Procedure 9-5-
100, Contamination Control / Decontamination Program, a
radiation worker failed to remove protective clothing from his
hand prior to exiting the contaminated area around the high
pressure safety injection pump. (Section R4.2)
l
Items Closed:
l 50-309/96-09-01 URI Unplanned extremity exposure including potential for the failure
to perform a radiation survey (h9.)
50-309/95-06 G2 VIO Failure to Maintain Survey Records (survey associated with
upender pit hydrolase want could not be located).
! (Section R8.3)
,
50-309/95-06-03 VIO Dosimetry NOT located in Highest Region of Exposure
(associated with RCP element handling can 3/24/95)
(Section R8.3)
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50-309/96-06-04 IFl Quality of Radiological Surveys & Documentation
(Section R8.3)
[ 50-309/96-04-03 VIO Failure to Brief Workers to Minimize Exposure (Section R8.3)
1
, LER 95-15 LER Emergency Core Coc".ig system Valves Not in Compliance -
NRC inspection Rer..srt 50-309/95-24. (Section 08.1)
- LER 95-16 LER Purge Valves Found in the On-Line Mode during Refueling -
}. NRC Inspection Report 50-309/95-24. (Section 08.1)
i LER 96-03 LER Reactor Trip Due to High Steam Generator Level- NRC
, inspection Report 50-309/96-02. (Section )8.1)
q LER 96-15 LER Containment Hatch Interlock Conflict - NRC Inspection Report !
50-309/96-08. (Section 08.1) '{
.
LER 96-30 LER Incorrect Acceptance Criteria Specified for Containment
- Ventilation / Purge System Filter Surveillance - NRC Inspection l
, Report 50-309/96-12. (Section 08.1)
.
LER 96-35 LER Requirement for Post Accident lodine Sampling inadvertently
Removed From Procedure - NRC Inspection Report 50-309/ I
q 96-12. (Section 08.1) !
LER 95-13 LER Pressure Seal Bonnet Retention Screw Failures - NRC
Inspection Report 50-309/95-13. (Section M8.1)
' '
LER 96-12 LER Water Intrusion of Turbine Driven Auxiliary Feedwater Pump !
4
Lube Oil Sump - NRC Inspection Report 50-309/96-06.
(Section M8.1)
LER 96-37 LER Leaking P-29C Gland Cooling Check Valve NRC Inspection i
Report 50-309/96-13. (Section M8.1)
I
LER 96-11 LER Service Water Pump Cutiess Bearing cooling Water System 1
,
Design - NRC Inspection Report 50-309/96-08. (Section E8.3)
4
SER 95-S01 SER Failure to Maintain Compensatory Measures for a Degraded
Protected Area Barrier - NRC Inspection Report 50-309/95-02. ,
(Section S8.2)
SER 96-S02 SER Safeguards Information Found Uncontrolled - NRC Inspection ,
,
Report 50-309/96-13. (Section S8.2)
.
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SER 94-S02' SER Degradation of a Security Access Controlled Barrier.
(Section S8.2)
LER 95-14 LER- Cardox Zone 1 Activation Wiring Defect NRC Inspection Report
50-309/95-24. (Section F8.1)
LER 96-17 (-01) LER Fire Barrier Penetration Seal Discrepancy - NRC Inspection i
Report 50-309/96-08. (Section F8.1) i
Items D:scussed:
LER 95-05 LER Rust and Scale Found in Technical Specification Required Fire l
Spray / Foam System. This LER remains open pending further
NRC review. (Section F8.1)
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LIST OF ACRONYMS USED
AEOD Office for Analysis and Evaluation of Operationa; Data
ALARA As Low As is Reasonably Achievable
ANSI American National Standards Institute
CFR Code of Federal Regulations
CVCS Chemical and Volume Control System
DAW Dry active waste
DCC Document control center
DOT Department of Transportation
EA Escalated Action
ERB Event Review Board
ESF Engineered Safety Feature
gpm Gallons per minute
GPO Government Printing Office
HIC High integrity container
ICI In-core instrument
IFl Inspector followup item
IFS Inspection Follow-Up System
IMC Inspection Manual Chapter
IPAP Integrated Performance Assessment Process
IROC Independent Radiological Oversight Committee
ISI in-Service inspection
LER Licensee Event Report
LLD Lower limit of detection
MD Management Directive
mR Milliroentgen
NCV Non-Cited Violation
NMSS Office of Nuclear Material Safety and Safeguards
NRC Nuclear Regulatory Cornmission
NRR Office of Nuclear Reactor Regulation
NSARC Nuclear Safety Assessment Review Committee
OE Office of Enforcement
01 Office of Investigations
OPIT Outage Planning and Integration Team
PAB Primary Auxiliary Building
PIPB inspection Program Branch
PPR Plant Performance Review
QA Quality Assurance
RA Regional Administrator
RA Restricted area
RCP Reactor coolant pump
RIR Radiological incident report
RP Radiation Protection
. . . . _ _ _ = _ . --
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RP&C Radiological Protection and Chemistry
RWP Radiation work permit
SALP Systematic Assessment of Licensee Performance
Si international System of Units I
Tl Temporary Instruction
TLD Thermoluminescent dosimeter
TS Technical Specification
UOR Unusual Occurrence Report
URI Unresolved i'am ,
VIO Violation l
WO Work Order !
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