ML20236M315
| ML20236M315 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/02/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236M310 | List: |
| References | |
| 50-333-98-02, 50-333-98-2, NUDOCS 9807140018 | |
| Download: ML20236M315 (36) | |
See also: IR 05000333/1998002
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Docket No.:
50-333
License No.:
DPR 59
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Report No.:
98-02
Licensee:
New York Power Authority
Facility:
James A. FitzPatrick Nuclear Power Plant
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Location:
Post Office Box 41
Scriba, New York 13093
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Dates:
April 20 - May 31,1998
Inspectors:
G. Hunegs, Senior Resident inspector
R. Fernandes, Resident inspector
J. McFadden, Radiation Specialist (5/18-22)
J. Noggle, Senior Radiation Specialist (5/18 22)
L. Peluso, Radiation Physicist (5/18-22)
C. Sisco, Operations Engineer (4/20-24)
- Approved by:
D. Lew, Chief, Projects Branch 2A -
Division of Reactor Projects
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9807140018 980702
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ADOCK 05000333
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EXECUTIVE SUMMARY
James A. FitzPatrick Nuclear Power Plant
NRC Inspection Report 50-333/98-02
This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a 6-week period of resident inspection
and announced inspections by an operations engineer, two regional radiation specialist
inspectors and a radiation physicist.
Operations
On May 1,1998, operators appropriately inserted a manual reactor scram due to
alarms indicating a control rod drift condition and operator actions to control reactor
pressure vessel water level were acceptable. However, the licensee identified an
operator performance issue concerning the use of emergency operating procedures
(EOPs). Specifically, operators used an incorrect procedure to verify that all control
rods were full-in. The failure to carry out the actions of the correct procedure
during the use of the EOPs was determined to be a violation. (VIO 50-333/98002-
01)
The licensee has had a previous failure to properly execute emergency operating
procedures. In a September 16,1996 transient, operators did not enter EOP-3
when entry conditions were present. Corrective actions for that error included
reinforcernent of management expectations and operator training. This additional
failure to properly execute EOPs indicated that difficulties in proper implementation
of EOPs have continued. (Section 01.1)
Two plant procedures were not adequate concerning assigned duties of the on-shift
senior nuclear operator and plant operations during degraded core flow conditions.
Specifically, an administrative procedure was not adoquate, because it allowed the
on shift operator to direct the licensed activities of I! censed operators. Also,
abnormal operating procedure AOP-8, Loss of Coolant Flow, was not adequate in
that a power to flow map was incorrect and inconsistent. The inadequate
procedures were determined to be two violations of NRC requirements. (VIO 50-
333/98002-02& 03) (Section O3.1)
Violations of requirements were identified with respect to the LORT program
implementation concerning the inadequate control of the annual operating
examination duplication (VIO 50 333/98002-05)and the sampling of required
content of the examination (VIO 50-333/98002 06& 07). The operators failure to
complete the required licensed operator requalification training (LORT) program and
the continuing poor attendance by some operators of the current LORT program is
an unresolved item. (URI 50-333/98002-04) The annual operating examinations
were conducted in a professional manner with thorough and objective
documentation and evaluations of each operator and crew performance. The
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Executive Summary (cont'd)
' inspector agreed with the evaluators that all observed individuals and crews passed
the examinations. (Section 05.1)
Maintenance
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On May 1,1998, the rod position information system power supply failed, resulting
in false control rod drift indications, which led operators to insert a manual reactor scram. Although the maintenance rule was applied to the rod position information
system, appropriate preventive maintenance had not been identified and, therefore,
not performed. The review' conducted in 1995 to determine what preventive
maintenance tasks were appropriate for the rod position information system power
supply was weak, and therefore system reliability was affected. (Section M1.3)
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The equipment failure evaluation, which was conducted for the. failed rod position
information system ' power supply, was thorough and inclu' ed an industry
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experience review and an extent of condition review to determine the appropriate
preventive maintenance task and frequency for similar systems. Additionally,
licensee troubleshooting efforts associated with the rod position information system
were wellimplemented. (Section M1.3)
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Turbine bypass valve degradation was determined to be caused by stem to disc
connection degradation. A weak maintenance procedure resulted in improper
reassembly of the turbine bypass valves by the licensee. The licensee corrective
actions were appropriate. (Section M1.4)
Enoineerina
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.The safety issues associated with the use of the reactor building crane to move
spent fuel were adequately addressed by the licensee's safety evaluation process.
However, the quality of the original safety evaluation was poor because reference to
other licensee documents was necessary to assure that heavy load requirements
were met. (Section E.8.1)
Plant Suooort
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The licensee effectively maintained and implemented the Radiological Environmental
Monitoring Program in accordance with regulatory requirements. The licensee
performed a comprehensive review of an anomalous indication of iodine-131 in an
environmental milk sample.- (Section R1.1) Overall, the licensee effectively
maintained system operability, and performed calibrations for the strip chart
recorders for the meteorologicalinstrumentation. (Section R1.2) The licensee met
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the Quality Assurance (QA) audit requirements. The audits were thorough and of
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sufficient depth to assess the strengths and weaknesses of the Radiological
Environmental Monitoring Program (REMP) and Meteorological Monitoring Program
(MMP). (Section R7.1) The environmentallaboratory continued to implement
effective Quality Assurance / Quality Control programs for the REMP, and continued
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Executive Summary (cont'd)
to provide effective validation of analytical results. The laboratory demonstrated the
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ability to accommodate and incorporate difficult media and geometries into the
program. The programs are capable of ensuring independent checks on the
precision and accuracy of the measurements of radioactive materialin
environmental media. (Section R7.2)
Based on plant tours during the inspection period, plant areas were in compliance
with established radiological posting procedures and regulatory requirements, and in
general, most of the principal radiation exposure sources that were accessible
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during plant operations, have been minimized through effective pipe flushing
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activities. (Section R1.3)
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. The licensee has implemented an effective as low as is reasonably achievable
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(ALARA) program based on declining station exposures and reductions in source
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term input. The ALARA program has been supported by management through
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several exposure reduction initiatives including: recirculation system chemical
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decontamination, zinc injection, and the systematic reduction of cobalt
contributors, such as the replacement of non-cobalt containing control rods and
valve components. (Section R1.4)
Calibration and use of portable radiation protection (RP) instruments, counting
instruments, whole body counting instruments, thermoluminescent dosimeters, and
the use and control of respiratory protection equipment were effectively
implemented as evidenced by current calibration documentation and tours of
applicable facilities. No discrepancies with regulatory requirements were observed.
(Section R2.1)
On April 9,1998, radiological monitoring for the control rod drive relief valve
replacement was weak. During the maintenance activity, an unexpected condition
occurred which was that steam emitted from the normally cold, radioactive waste
system. Mechanics and RP technicians did not demonstrate a questioning attitude
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when faced with this unexpected condition as the unexpected condition was not
fully evaluated nor were radiological surveys taken. There were no radiological
consequences as post job surveys did not show contamination or abnormal airborne
radioactivity levels. (Section R4.1)
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RP program oversight between September 1997 and April 1998 consisted of an
effective combination of QA surveillance, QA audit, RP self-assessments, and RP
program feedback through the station problem identification and resolution program.
(Section R7.3) -
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Executive Summary (cont'd)
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On March 17,1998, a shipment of radioactive material was made to another
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licensee and was improperly classified as excepted packages with package surface
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radiation levels exceeding 0.5 millirem per hour, which constituted a failure to
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comply with 10 CFR 71.5(a) with respect to 49 CFR 173.421(a)(2). The violation
is of concern because it was not self identified during the preparation and review of
the shipping paperwork package. (VIO 50-333/98002-11)(Section R8.2)
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TABLE OF CONTENTS
EX EC UTI VE S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TAB LE O F CO NT ENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi
Summary of Plant Status
............................................1
1. O p e ra t io n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01
Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 Manual Reactor Scram Due to Multiple Control Rod Drift Alarms
(VIO 5 0-3 3 3/9 8002-01 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
03
Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . 3
03.1 Operating Procedure Discrepancies (VIO 50-333/98002-02& 03) . 3
03.2 (Closed) Unresolved item (URI) 50-333/97009-01 . . . . . . . . . . . . 4
05
Operator Training and Qualification ........................... 5
05.1 Licensed Operator Requalification Training Program (Unresolved
item 50-333/98002-04,VIO 50-333/98002-05,06 & 07) ...... 5
ll. M ainte n a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
M1
Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
M 1.1 General Comments on Maintenance Activities (62707) ........ 7
M1.2 General Comments on Surveillance Activities (61726) . . . . . . . . . 8
M1.3 Rod Position Information System Power Supply Failure . . . . . . . . . 8
M1.4 Repair of Turbine Bypass Valve . . . . . . . . . . . . . . . . . . . . . . . . . 9
M8
Miscellaneous Maintenance issues (92902) .................... 10
M8.1 (Closed) Licensee Event Report (LER) 50-333/95004, Revision 1.
10
lll . Engine e ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1
E8
Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . .11
E8.1
(Closed) Unresolved item 50-333/97007-02 . . . . . . . . . . . . . . . 11
E8.2 (Closed) Unresolved item 50-333/96006-02 . . . . . . . . . . . . . . . 12
IV . Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4
R1
Radiological Protection and Chemistry (RP&C) Controls ............ 14
R1.1
Implementation of the Radiological Environmental Monitoring
Prog ra m ( R E M P) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
R1.2 Implementation of the Meteorological Monitoring Program . . . . . 15
R 1.3 Radiation Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
R1.4 As Low As is Reasonably Achievable (ALARA) . . . . . . . . . . . . . 16
R2
Status of RP&C Facilities and Equipment
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R2.1
Radiological Protection (RP) Instrument Calibrations and Use ... 18
R4
Staff Knowledge and Performance in RP&C ....................20
R4.1 Weak Evaluation of Unexpected Radiological Condition During
M aintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
R7
Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 21
R7.1
Quality Assurance Audit Program
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Table of Contents (cont'd)
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R7.2 Quality Assurance of Analytical Measurements . . . . . . . . . . . . . 21
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R7.3 Radiological Protection Program Oversight ................ 22
R8
Miscellaneous RP&C lssues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
R8.1 (Closed) Inspector Followup Item (lFI) 50-333/97008-04 ...... 23
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R8.2 Improperly Classified Shipment of Radioactive Materials
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(VIO 50-333/98002-1 1 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
F4
Fire Protection Staff Knowledge and Performance . . . . . . . . . . . . . . . . 24
F4.1
Response to Fire in Computer Room . . . . . . . . . . . . . . . . . . . . . 24
V. - Management Meetings
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Exit Meeting Sum m ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5
ATTACHMENT
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Attachment 1 - Partial List of Persons Contacted
- Inspection Procedures Used
-ltems Opened, Closed, and Discussed
- List of Acronyms Used
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Report Details
Summarv of Plant Status
. The unit began this inspection period at 90 percent reactor power, with operators in the
process of raising reactor power following single loop operations. Power was returned to
100 percent on April 20,1998.
On May 1,1998, operators inserted a manual reactor scram from 100 percent reactor
power due to indications of control rod drift. The plant was taken to the cold shutdown
condition and a short forced outage was conducted. On May 9,1998, the reactor was
taken critical and on May 11,1998, the plant was returned to power operation. On
May 12,1998, the plant was returned to 100 percent reactor power and remained there
for the duration of the inspection period.
1. Operations
01
Conduct of Operations'
01.1 Manual Reactor Scram Due to Multiple Control Rod Drift Alarms (VIO 50-
333/98002-01)
a.
inspection Scone (71707)
On May .1,1998, while the plant was at 100 percent reactor power, the control
room operators noted several rod drift alarms and manually inserted a reactor
scram. The inspectors observed the scram recovery from the control room and
portions of the subsequent shutdown. In addition, the inspectors reviewed the post
transient evaluation (PTE) and pertinent licensee procedures.
b.
Observations and Findinas
On May 1,1998, at 12:32 pm, annunciator 09-5-2-3 " Rod Drift" alarmed for
control rod 42-47. The reactor operator acknowledged the alarm and' selected the-
rod on the four-rod display for position location, however the~ position was blank.
- Several seconds later, the control room supervisor noted additional rod drift lights
on the full core display and directed that a manual reactor scram be inserted. it was
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subsequently determined that there was no actual rod movement but that the failure
of a rod position indication system (RPIS) power supply caused the erroneous
indications. The reactor was scrammed ten seconds after the receipt of the first
. alarm and the immediate operator actions for a reactor scram were performed.' The
. inspectors determined that the immediate operator response to the control rod drift -
alarm was appropriate.
'I'opical headings such as 01, M8, etc., are used in accordance with the NRC
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standardized reactor inspection report outline. Individual reports are not expected to
address all outline topics.
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Reactor water level dropped as a result of the scram, resulting in a Group 2
isolation, and continued to drop to the point at which the high pressure coolant
injection (HPCI) and reactor core isolation cooling (RCIC) systems initiated.
However, because of the short duration of the drop in the reactor vessel water
level, the HPCI and RCIC systems did not inject into the vessel because the signals
did not " seal in." Alternate rod insertion (ARI) initiated and both reactor water
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recirculation pt,mps tripped as a result of the water level transient. Reactor water
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level subsequently recovered due to both reactor feed pumps (RFPs) running in
automatic operation. An operator manually tripped the "A" RFP, as water level
' increased, however, before level could be controlled with the operating "B" RFP, the
high reactor water level trips for HPCI, RCIC, and the "B" RFP occurred. The "B"
RFP was subsequently started and used to control reactor water level. The
inspector determined that operator actions with respect to reactor pressure vessel
(RPV) level controls were acceptable.
The operators carried out the immediate actions for a reactor scram and an entry
into Emergency Operating Procedure (EOP)-2, RPV Control, was made due to the
water level transient. During the actions to verify completion of the scram, the
operators determined that six control rods did not indicate full-in and EOP-3, Failure
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To Scram, was entered. The PTE identified that the operators, at this point, entered
- the incorrect procedure for establishing rod position when presented with the EOP
entry condition. EOP-3, in part, directs the operators to enter and execute EOP
Support Procedure (EP)-3, Backup Control Rod insertion, to insert control rods when
the reactor protection system (RPS) and ARI have failed to fully insert control rods.
The operators, in an attempt to expedite the process to verify control rod position
indication, incorrectly utilized Abnormal Operating Procedure (AOP)-1' Reactor :
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Scram, to verify all rods in. After starting a second control rod drive (CRD) pump
and adjusting the CRD system to drive rods, three rods were determined to be full-
in; and, following an insert signal to the remaining three rods, all rods were
determined to be full-in. The inspector noted that, although AOP-1 and EP-3 allow
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the operators to utilize the CRD system to manually insert rods to establish rod
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position, these procedures employ different mitigation strategies. For example, EP-3. -
requires that ARI and RPS to be reset prior to driving control rods, which would
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allow operators to insert a scram a second time if conditions warranted such action.
The inspectors determined that the failure to enter the appropriate procedure
during the execution of the EOPs was a violation of NRC requirements.
Administration Procedure (AP)-12.03, Administration of Operations,' Revision 12,-
requires that operators shall observe a hierarchy for plant operating procedures and
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E hat EOPs and EPs have the highest priority. Additionally, EP-1, EOP Entry and Use,
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Revision 2, requires, in part, that actions taken per other plant procedures shall not
contradict or subvert actions specified by the EOPs. Operators incorrectly executed
EOPs by taking actions per other plant procedures which contradicted actions
specified by the EOPs. Specifically, operators had entered EOP-3, Failure to Scram,
which directed that operators execute EP-3, Backup Control Rod Insertion. Instead
of entering EP-3; AOP-1, Reactor Scram, was entered immediate corrective
actions for this error included briefing the shift operators on the correct procedure to
be used while carrying out the actions of EOP-3.
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The licensee has had a previous failure to properly imp!ement emergency operating
procedures. In a September 16,1996 transient, operators did not enter EOP-3
when entry conditions were present. Corrective actions included reinforcement of
management expectations and operator training. This additional failure to properly
execute EOPs indicated that difficulties in proper implementation of EOPs have
continued.
Equipment problems, in addition to the failure of the RPIS power supply, included
the "D" intermediate range neutron monitor spiking which caused a half scram
during the shutdown. The instrument was bypassed and the licensee subsequently
determined that the erratic indication was the result of a loose connection under the
reactor vessel. No other major equipment problems were noted during the
transient. Following the event, the licensee elected to take the plant to a cold
shutdown condition to repair the RPIS and perform additional maintenance.
c.
Conclusions
On May 1,1998, operators appropriately inserted a manual reactor scram due to
alarms indicating a control rod drift condition and operator actions to control reactor
pressure vessel water level were acceptable. However, the licensee identified an
operator performance issue concerning the use of emergency operating procedures
(EOPs). Specifically, operators used an incorrect procedure to verify that all control
rods were full-in. The failure to carry out the actions of the correct procedure
during the use of the EOPs was determined to be a violation. (VIO 50-333/98002-
01)
The licensee has had a previous failure to properly execute emergency operating
procedures. In a September 16,1996 transient, operators did not enter EOP-3,
Failure to Scram, when entry conditions were present. Corrective actions for that
error included reinforcement of management expectations and operator training.
This additional failure to properly execute EOPs indicated that difficulties in
implementation of EOPs have continued.
03
Operations Procedures and Documentation
03.1 Operating Procedure Discrepancies (VIO 50-333/98002-02& 03)
a.
insoection Scoce (71001)
During the review of unresolved item 50-333/97009-01,as documented in section
03.2, the inspector reviewed several procedures that were used in the conduct of
plant operations.
b.
Observations and Findinas
The inspector identified that AP-12.03, Administration of Operations, Revision 11,
was not adequate in that step 6.11.2 incorrectly provided instructions which would
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allow the on shift operator to direct the licensed activities of licensed operators,
instead of specifying that only a senior operator can direct licensed activities.
10 CFR 55.4 defines operator as any individuallicensed to manipulate a control of a
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facility and a senior operator as any individual licensed to manipulate the controls of
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a facility and to direct the licensed activities of licensed operators. The inadequate
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plant procedures concerning the duties and responsibilities of on shift operators is a
violation of NRC requirements. (VIO 50-333/98002 02)
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Further, the inspector identified that AOP 8, Loss of Coolant Flow, was not
. adequate in part. Specifically, areas of the power to flow map provided operator
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actions that differed from the power to flow maps of operator aid number 24 and
Reactor Analyst Procedure (RAP) 7.3.16, Plant Power Changes. The incorrect and
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inconsistent procedures concerning the power to flow maps are a second violation -
of NRC requirements. (VIO 50-333/98002-03)
c.
Conclusions .
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Two plant procedures were not adequate concerning assigned duties of the on-shift
senior nuclear operator and plant operations during degraded core flow conditions.
Specifically, an administrative procedure was not adequate, because it allowed the
on shift operator to direct the licensed activities of licensed operators. Also,
abnormal operating procedure AOP-8, Loss of Coolant Fiow, was not adequate in
that a power to flow map was incorrect and inconsistent. The inadequate
procedures were determined to be two violations of NRC requirements. (VIO 50-
333/98002-02& 03)
O3.2 (Closed) Unresolved item (URI) 50-333/97009-01: emergency operating procedure
terminology is not consistent with emergency procedure guidelines. - NRC inspection 1
Report 50-333/97009 documented that the EOPs and plant specific technical guide
(PSTG) stated " Verify initiation of" instead of " Initiate each of the following which
should have initiated but did not" which was not consistent with Emergency
Procedure Guidelines (EPGs) step " reactor control / level" (RC/L)-1. In addition, the
EOP change and the PSTG indicated that this was plant specific terminology and did
not specifically identify whether the wording change was meant to be identical in
interpretation. Also, the change from series to concurrent execution of initiate
isolations was not specifically addressed.
. The licensee revised their PSTG to clarify the wording of the applicable procedures.
The inspector reviewed the revised PSTG and the deviation taken for EPG step
RC/L-1. The revised PSTG clearly stated the change was a concurrent rather than a -
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serial execution step. Also, the revision explained the rational for the change to the
EOPs. The inspector noted that the facility evaluators sampled the operators
knowledge of step RC/L-1 by asking follow up questions following the dynamic
- simulator examination. Operators demonstrated a sound knowledge of step RC/L-1
indicating that training had been effective. The original finding did not reflect an
inadequate procedure. Based on a review of the documents and the demonstrated
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- knowledge of the licensed operators, the inspector concluded that this item is
closed.
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05
Operator Training and Qualification
05.1 Licensed Operator Requalification Training Program (Unresolved item 50-
333/98002-04,VIO 50-333/98002-05,06 & 07)
a.
Inspection Scope (71001)
The licensed operator requalification training (LORT) program was assessed using
selected portions of NRC Inspection Procedure 71001," Licensed Operator
Requalification Program Evaluation." The inspector observed performance and
evaluations of a shift and staff crew during the annual operating examination and
reviewed medical records.
b.
Observations and Findinas
- LORT Proaram Content and implementation
The inspector noted that the LORT program attendance records were not easily
auditable; As a result, the inspector performed a detailed review of LORT program
supporting attendance information and other data compiled which was provided by
the licensee. This data indicated that several individuals had not made up missed
training since the beginning of the training cycle in June 1997. Specifically, the
data indicated that 15 staff licenses had failed to attend cycle training June 2
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through July 17,1997, four staff licenses from July 21 through September 4,-
1997, one staff license from September 8 through October 23,1997 and 10 staff
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licenses from October 27 through December 11,1997. Training procedure (TP)-
5.05, LORT Program, Revision 1, Step 7.4.2 allowed the on-shift and staff licenses
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to make up this missed training within 14 weeks or be removed from licensed
duties. The inspector noted that, at the time of the inspection, licensed operators
composing the operating crews did not have to be removed from licensed duties for-
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failure to make up missed training. Also, staff licensed operators who missed
training were restricted from licensed duties.
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1The inspector noted that licensed operators, as required by 10 CFR 55.59(a)(1),
must complete the requalification program (also a standard condition of each.
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licensed operator's license per 10 CFR 55.53(h)). A restriction placed on licensed
duties;due to missed trained after a period of time is not a condition of th'ese
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regulations. The inspector determined that TP-5.05, step 7.4.2 is a root cause -
-(there may be other. root causes) for the attendance problem that has continued
among staff licenses. The adequacy of this procedure is being referred to the
Operator Licensing Branch, Office of Nuclear Reactor Regulation (NRR).
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Further, the inspector reviewed licensee letter [[::JAF-97-0184|JAF-97-0184]] dated May .16,1997,.
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- that stated, in part, that some licensed operators had not completed the required
LORT program from 1994 through 1996. The inspector reviewed licensee- provided
data and determined that 18 individuals (mostly staff licensed operators) had failed
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to complete the LORT program during this time period. Corrective actions as
indicated in this letter were apparently ineffective based on the 1997-1998 LORT
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]
.
l
6
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l
attendance data. In response to inspector questioning, the licensee provided a
!
letter, [[::JAF-98-0178|JAF-98-0178]], dated May 28,1998, giving additional details of individual
,
operators who missed training during the 1994-1997 period. The apparent
continuing poor attendance by some operators during the current LORT program is
j
unresolved pending further NRC staff review of program adequacy. (URI 50-
333/98002-04)
Exam Materials
The inspector reviewed a sample of the annual operating test materials administered
,
'
by the facility. The sample included dynamic simulator scenarios, and job
performance measures (JPMs). The materials reviewed had an appropriate level of
difficulty with expected operator performance standards clearly stated. The
inspector also determined that 7 scenarios and 19 JPMs were the test items
,
administered to approximately 60 operators. Over the approximate 6 weeks of
exam administration, the number of licensed operators exposed to the identical
exam materials was excessive or not sufficiently comprehensive to identify
)
retraining needs for the LORT program. As an example, dynamic scenario 98G was
administered to 42 individuals, JPM 98H was administered to 31 individuals. A
i
total of 68 individuals had signed security agreements concerning the annual
'
operating test. The inspector did not identify that any examination compromise had
occurred. The LORT program procedures provided no direction concerning the
j
duplication of examination materials, but operators did sign security agreements
indicating they would not divulge test material. The practice of duplicating exams
from week to week regardless of signed security agreements is a violation of NRC
requirements (50.54(1-1) and 55.59(c)(4)) because the duplicated examination
materials did not sufficiently allow for the determination of areas in which retraining
was needed to upgrade reactor and senior reactor operator knowledge. (VIO 50-
333/98002 05)
Based upon a review of training procedure TP-5.07," LOR" Examination
Development and Administration," Revision 3, the inspector determined that low
power or shutdown conditions are excluded from the sample of items to be
evaluated during the annual operating test. 10 CFR 55.45(a)(2) requires, in part,
licensed operators demonstrate the ability to operate the facility between shutdown
and designated power levels. The licensee's failure to include this important topic
during the ann'ual operating test is a violation of 10 CFR 55.59(a)(2)(ii),55.45(a)(2).
(VIO 50-333/98002-06) in addition, the inspector determined that on shift Control
Room Supervisors were excluded from being evaluated, again on a sampling basis,
their ability to execute the emergency plan. 10 CFR 55.45(a)(11) requires, in pan,
the licensed senior reactor operator to demonstrate the ability to execute the
'
emergency plan. The licensee's failure to include and sample this important topic
during the annual operating test is a violation of 10 CFR 55.59(a)(2)(ii),
55.45(a)(11). (VIO 50-333/98002 07)
l
_
. - _ _ - -
-
l
-
1
7
Exam Administration
l
The inspector observed the facility evaluators administer four dynamic simulator
examinations to one operating crew and one staff crew. The examinations were
conducted in a professional manner with thorough and objective documentation and
evaluations of each operator and crew performance. The inspector agreed with the
,
evaluators that all observed individuals and crews passed the examinations.
l
l
Medical Records Review
The inspector conducted a review of randomly selected licensed operator medical
records. The inspector determined the records were easily reviewed and were well
maintained and met the requirements of 10 CFR 55.53(l).
c.
Conclusions
Violations of requirements were identified with respect to the LORT program
implementation concerning the inadequate control of the annual operating
examination duplication (VIO 50-333/98002-05)and the sampling of required
content of the examination (VIO 50 333/98002-06& 07). The operators failure to
complete the required licensed operator requalification training (LORT) program and
the continuing poor attendance by some operators of the current LORT program is
an unresolved item. (URI 50 333/98002-04) The annual operating examinations
- were conducted in a professional manner with thorough and objective
documentation and evaluations of each operator and crew performance. The
inspector agreed with the evaluators that all observed individuals and crews passed
the examinations.
11. Maintenance
M1
Conduct of Maintenance
M1.1 General Comments on Maintenance Activities (62707)
The inspectors observed all or portions of the following work activities:
e.97-02265, control rod drive system check valve inspection
- 95-04547, control rod drive relicf valve replacement
- The inspectors observed that the work performed to the above work requests (WRs)
was conducted satisfactorily.
t
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_ _ _ _ - _ _ _ _ _ _ _ _ _ _ ______ -____
__
j;
I
c:
8
M1.2 General Comments on Surveillance Activities (61726)
l
The inspectors observed all or portions of the following surveillance activities:
o ST2H, Low Pressure Coolant injection initiation Logic System Functional Test
.
o ST3P, Core Spray Test
- ST9K, Emergency Diesel Generator Barring
- ST76E, Fire Hose Station Inspection
The licensee conducted the above surveillance activities appropriately and in
accordance with procedural and administrative requirements. As applicable, good
coordination and communication with the control room were observed during
performance of the surveillance.
l
M1.3 Rod Position Information System Power Supply Failure
o
a.
Insoection Scooe (62707)
On May 1,1998, the rod position information system (RPlS) power supply failed
. resulting in false control rod drift indications which led operators to insert a manual
,
I
reactor scram (see Section 01.1). The inspector reviewed the licensee's equipment
.j
l
failure evaluation (EFE), the maintenance rule basis document for the reactor manual
'j
control system (RMCS), and the licensee's corrective actions. The inspector also
observed troubleshooting efforts and replacement of the RPlS power supply.
'
b.
Observations and Findinas
On May 1,1998, RPIS power supply,- 03PSY5, failed. The failure caused
approximately 48 control rods to indicate a false control rod drift" which led
operators to insert a manual reactor scram. The RPIS power supply failure was -
attributed to a failed internal cooling fan which caused the RPlS power supply to
1
'
overheat and resulted in the failure of the RPIS power supply regulator circuit.
The inspector noted that the power supply was original plant equipment and had
been installed since' 1974. The power supply was not in the preventive
maintenance (PM)_ program and no PM tasks had been performed. Although the -
, _
licensee considered the system to be highly reliable, the inspector noted that, in
!
-1995, one of the RPIS power supplies had failed which had caused a loss of full-in
_
indication for control rods. The licensee review of the 1995 failure was limited and
'
was'a missed opportunity to identify PM tasks to be conducted.
.
The inspector concluded that previous reviews, conducted to determine what
preventive maintenance tasks were appropriate for the RPlS power supply, were
. weak. The maintenance rule basis document shows that the RMCS is in the scope
of the maintenance rule because information displays are used in emergency
-
<
operating procedures and failure of circuitry could result in a reactor scram. The
RMCS includes the RPIS and the rod drift annunciator was identified as critical to
ll'
-
.
.-
9
ensuring required RMCS functions can be met. The licensee appropriately classified
the RPIS power supply failure as a maintenance preventable functional failure.
The EFE was thorough and included a review of industry experience and corrective
actions. The EFE documented that another nuclear plant had implemented PM tasks
to inspect and clean a similar power supply on a two year frequency. The licensee
implemented corre,ctive actions to identify PM tasks and task frequency for the RPIS
power supply and reviewed other comparable systems to determine appropriate PM
tasks to be conducted.
The licensee troubleshooting efforts showed that the RPIS power supply failure
i
resulted in damage and degradation to other system components. To preclude
additional problems, the licensee replaced four power supplies and potentially
damaged components associated with the RPIS. The licensee troubleshooting
efforts were well implemented.
c.
Conclusions
On May 1,1998, the rod position information system power supply failed, resulting
in false control rod drift indications, which led operators to insert a manual reactor scram. Although the maintenance rule was applied to the rod position information
system, appropriate preventive maintenance had not been identified and therefore,
not performed. The review conducted in 1995 to determino what preventive
maintenance tasks were appropriate for the rod position information system power
supply was weak, and therefore system reliability was affected.
The equipment failure evaluation which was conducted for the failed rod position
information system power supply was thorough and included an industry experience
review and an extent of condition review to determine the appropriate preventive
maintenance task and frequency for similar systems. Additionally, licensee
troubleshooting efforts associated with the rod position information system were
well implemented.
M1.4 Repair of Turbine Bypass Valve
a.
Insoection Scone (62707)
l
l
The inspectors reviewed a licensee finding of poorly performed maintenance on the
'
number one turbine bypass valve. The inspection included a review of Maintenance
Procedure (MP) 94.09, " Main Turbine Bypass Valves," Rev. 3, maintenance rule
requirements, and a review of the licensee's actions to correct the problem.
b.
Observations and Findinas
On May 6,1998, post work testing on the number one turbine bypass valve
showed some anomalous conditions, and the licensee decided to investigate further.
- The licensee disassembled the valve and determined that the disc retaining nut had
!
_ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ - _ - - _ _ _
_
_ _ _ _ _ _ _ - _ _ _ - - _ _ _ _ _ _ - _ - - - _ - _ _ _
.
,
10
backed out approximately 1-1/8 inches, leaving only 2 3/8 revolutions of thread
engagement to prevent separation of the stem from the disc.
The bypass valve stem is retained to the valve disc by a bushing. The bushing is
threaded into the disc and locked in place by staking the disc retaining ring into four
slots on the bushing. The inspector noted that all four bypass valves were rebuilt
,
before the 1996 refueling outage by the licensee. The licensee's previous practice
had been for the vendor to assemble the valve stem and disc, where it was
apparently common knowledge that the disc material be extruded into the bushing.
The licensee concluded that maintenance personnel had inadequately staked the
disc to the bushing when assembling the valve in preparation for installation during
the 1996 refueling outage. The licensee determined that the procedural direction,
although adequate for the vendor craftsmen, needed additional direction for plant
maintenance staff.
1
The licensee's corrective actions included inspection of the other three bypass
valves, where similar conditions, although to a lesser extent, were identified.
Additionally, other valves in the plant were reviewed for similar configurations. The
licensee revised the maintenance procedure to include additional guidance and has
scheduled a review of other applications where prefabrication or preassembly
procedures may lack appropriate guidance for the plant staff.
The bypass valves are in the scope of the maintenance rule. The licensee
determined that the condition was not a functional failure as the valve was still
considered to be operable. The inspectors determined the maintenance rule review
conducted by the licensee to be appropriate and the actions to pursue the
inconsistent performance of the number one bypass valve during post work testing
to be good. The previous maintenance activities involving the assembly of the
bypass valves was poor.
c.
Conclusions
Turbine bypass valve degradation was determined to be caused by stem to disc
connection degradation. Weak maintenance performance resulted in improper
reassembly of the turbine bypass valves by the licensee. The licensee corrective
actions were appropriate.
.M8
Miscellaneous Maintenance issues (92902)
M8.1 (Closed) Licensee Event Report (LER) 50-333/95004, Revision 1: logic system
functional test procedure revision. The original LER, 50-333/95004,was closed in
NRC inspection report 50-333/95-08. That NRC inspection report documented that
no additional concerns or problems with the licensee's response were noted. The
licensee submitted Revision 1 to the LER on August 29,1995. The inspector
conducted an in-office review of Revision 1 to the LER and noted that the revision
provided an update of the status of corrective actions and changed the licensee
contact for the LER. The inspector noted that there was no other pertinent
inforn:ation provided in the revision that would affect the conclusions developed in
__ _---____ _ _ -
_ _ _ - _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ - _ - _ _ - _
_
1
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11
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NRC inspection report 50-333/95-08. Based on this in-office review, this LER is
closed.
Ill. Enaineerina
E8
IGiscellaneous Engineering Issues
E8.1
(Closed) Unresolved item 50-333/97007-02: Moving Spent Fuel With the Reactor
Building Crane
l
a.
Insoection Scoce (92903. 37551)
l
NRC inspection report 50-333/97007 documented issues concerning the use of the
reactor building overhead crane instead of the refuel bridge to move spent fuel to
l
peripheral cells in the spent fuel pool. The inspector reviewed the licensee's 10
i
Code of Federal Regulations (CFR) 50.59 nuclear safety evaluation (SE), James A.
FitzPatrick (JAF)-SE-97-003, Use of Reactor Building Crane to Move Irradiated Fuel
Assemblies in the Spent Fuel Pool, which was prepared to conduct the evolution
and discussed the conduct of the evolution with licensee personnel. Specific items
reviewed included reactor building crane travellimits, use of radio vice electrical
crane controls, use of an operator to disconnect crane power and heavy load
requirements. Additionally, on April 20,1998, a ccnference call was held with NRC
techniest staff to discuss the issue.
b.
Observations and Findinas
The proposed change in the fuel handling procedures to use the reactor building
overhead crane instead of the refuel bridge to move spent fuel assemblies
introduces some procedural and physical equipment changes to the fuel handling
operations that were not previously addressed in the updated final safety analysis
report (UFSAR). However, the NRC recognizes that the UFSAR is not very specific
about the use of the overhead crane and therefore the licensee's review of the
proposed use of the overhead crane is appropriately addressed by subjecting the
change to the 10 CFR 50.59 criteria.
The licensee's 10 CFR 50.59 nuclear safety evaluation provided reasonable basis
that the use of the overhead crane to move spent fuel was not an unreviewed
safety question (USO). The NRC determined that the travel limits of the reactor
building crane are adequately controlled and use of radio control for crane
mo..: ment was acceptable. Use of a dedicated person to disconnect the crane
power in the event of an overhead crane malfunction was determined to be
acceptable. Use of the overhead crane to move spent fuel assemblies creates the
j
possibility of a different way of initiating a fuel handling accident but is bounded by
the worst case accident and therefore has been evaluated previously in the UFSAR.
The inspector noted that the licensee's evaluation of postulated load drops
considered the combination of the load and associated lifting devices in the
- _ _ _ _ .
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_ _ - - _
--
-
_ _ - -
__.
- _ -_
. - _ - - - - - _ - - _
.9
.
12
analysis. However, the particular load handling situation results in the reactor
l'
building crane load block and hook traversing over the spent fuel which are required
!
to be considered a heavy load. The inspector also noted that the licensee's safety
evaluation, JAF-SE 97-OO3, Use of Reactor Building Crane to Move Irradiated Fuel
Assemblies in the Spent Fuel Pool, does not clearly address this specific issue.
Additionally, the reactor building crane mechanical load brake capabilities, with
respect to holding, were not addressed in the original approved safety evaluation,
and special requirements for interlocks and administrative controls for the overhead
crane control of heavy loads were not clearly addressed. However, the assurance
that heavy load requirements were ad&essed could be ascertained through referral
to other licensee documents and the safety evaluation was subsequently revised to
clarify the additional concerns noted.
c.
Conclusions
The safety issues associated with the use of the reactor building crane to move
spent fuel were adequately addressed by the licensee's safety evaluation process.
However, the quality of the original safety evaluation was poor because reference to
other licensee documents was necessary to assure that heavy load requirements
were met.-
E8.2 (Closed) Unresolved item 50-333/96006-02: Condenser Rupture Disc Actuation
a.
Insoection Scope (92903)
On September 16,1996, an automatic reactor scram occurred. A condenser
rupture disc and reactor feed pump rupture disc actuated, resulting in a steam
- release to the turbine building. [For brevity, condenser rupture disc and reactor feed -
_ pump rupture disc can be considered to be the A sme in this inspection report} The
condenser response related to this event was an unresolved item pending
determination of the design adequacy of the condenser and related controls. The -
inspector reviewed the licensee's engineering report, A42-00114, Design Adequacy
Evaluation of September 16,1996 Rupture Disc Actuation at the James A.
FitzPatrick Nuclear Power Plant, and the licensee's safety. evaluation, JAF-SE-96-
.060, Evaluation of Operator Action to Facilitate Main Steam isolation Valve (MSIV)
Isolation Under Specified Conditions. Sections of the JAF safety evaluation report,
UFSAR and the NRC Standard Review Plan pertaining to the condenser were also
L
reviewed.
b.
Observations and Find!nos
On September _16,1996, an automatic reactor scram was caused by personnel
error, inadvertent _ operation of the reverse power relay blocked the fast transfer of
plant buses to reserve power and resulted in the loss of the main circulating water
pumps which caused the loss of condenser heat removal capability. Condenser
" pressure increased until one of the main condenser rupture discs and one reactor
feed pump rupture disc actuated.
E
.i
,
.,
13
The licensee prepared an engineering report, A42-00114,to analyze the plant and
s
condenser response to the abnormal operating transient and determined that the
plant is in conformar ce with applicable design standards associated with the
condenser. The licensee's analysis determined that, given a loss of circulating
,
L
water flow from high power, a condenser rupture disc actuation could not be
.
precluded and that the September 16,1996 event was bounded by previous
!
analysis.e The inspector concluded that the licensee's engineering report appeared
to be tnorough based on the level of detail and design engineering involvement.
i
The inspector also noted that the licensee's safety evaluation report states that the
l
occurrence of an abnormal operating transient would not lead to significant olease
l
of fission products to the environment.
l
The licensee's safety evaluation, JAF-SE-96-060, Rev. O, Evalation of Operator
Action to Facilitate Main Steam isolation Valve isolation under Specified Conditions,
l
assessed the acceptability of manual action to close MSIVs during events similar to
i
p
that which occurred on September 16,1996. The licensee has completed several
j
l
modifications to the MSIV closure circuit. The current design of the MSIV low
i
bypassed when the reactor mode switch is not in "run" and the turbine'stop valves
are closed. By procedure, operators place the mode switch in the " shutdown"
position after a reactor scram to prevent automatic MSIV closure on low reactor
pressure and the consequent removal of the heat sink. This operator action
bypasses the MSIV LCV closure function which makes the condenser vulnerable to
rupture disc actuation on a loss of off-site power or a loss of power to station
_
auxiliaries. The licensee's safety evaluation report determined that manual MSIV
isolation is acceptable. The licensee also determined that, if circulating water were
lost to the condenser, MSIV closure would not preclude rupture disc actuation. The
,
'
NRC review determined that the conclusions stated in the licensee's safety
evaluation were reasonable.
The licensee determined that the UFSAR is inconsistent in discussion of the
l
potential for condenser rupture disc actuation and was incorrect in stating that
l-
closure of the MSIVs on LCV would preclude actuation of the condenser rupture
i
disc. UFSAR Section 7.1.1.2, Power Generation Design Bases, previously stated
l
' that the closure of MSIVs on LCV would preclude actuation of the c'ondenser-
I
rupture disc. The UFSAR did recognize that, in the unlikely event that the rupture
!;
disc should rupture due to the loss of circulating water to the main condensers, the
. resultant radiation dose would not exceed those resulting from the steam line break
inside the turbine building. The UFSAR does not discuss the potential for rupture
.
~
disc actuation as being a consequence of loss of offsite power or a loss-of-power to
the station auxiliaries. The UFSAR was also incorrect in documenting that the
MSIVs go shut following the loss of the resctor protection system (RPS) motor
generator. (MG) set. As the September 16,1996 event revealed, the MSIVs did not
go shut following the loss of the RPS MG set. The licensee has updated the UFSAR
- to reflect the operating experience obtained as a rese!t of the September 16,1996
- transient. The incorrect design bases information in the UFSAR was identified and -
corrected by the licensee as required by 10 CFR 50.71, Maintenance of Records,
Making of Reports,'and accordingly there was no violation of NRC regulations.
~
.
.
14
c.
Conclusions
During a September 16,1996 automatic reactor scram, a condenser rupture disc
and reactor feed pump rupture disc actuated, resulting in a steam release to the
turbine building. The licensee's analysis was sufficient to show that the condenser
response and design was acceptable. Modifications made to the main steam
isolation valve closure logic did not introduce the potential for rupture disc
actuation. The Updated Final Safety Analysis Report (UFSAR) was inconsistent in
the discussion of the potential for condenser rupture disc actuation and was
incorrect in stating that the closure of the main steam isolation valves on low
condenser vacuum would preclude actuation of the condenser rupture disc. The
incorrect design bases information in the UFSAR was properly corrected by the
licensee,
IV. Plant Support
R1
Radiological Protection and Chemistry (RP&C) Controls
R1.1 Implementation of the Radiciogical Environmental Monitoring Program (REMP)
a.
Inspection Scope (84750)
The following areas of the REMP were assessed and reviewed:
selected sampling and analysis procedures;
-
analytical data from 1998;
-
selected sampling techniques;
-
operability and calibration of air samplers;
-
1996 and 1997 Land Use Census results;
-
1996 and 1997 Annual Radiological Environmental Operating Reports; and
-
the licensee's investigation after identifying iodine-131 in milk during the
-
week of April 21,1997.
b.
Observations and Findinas
The sampling and analysis procedures provided appropriate guidance to perform
REMP tasks. Sampling techniques were appropriate to collect environmental sample
media. The air sampling equipment and water compositors were operable during
1997 to present, as evidenced by the sample logs and sample analysis results. The
air sampling equipment calibration results were within the established tolerances,
and calibrations were performed within the frequency specified in the procedure.
[
A Land Use Census was performed 1996 and 1997 during the growing season as
}-
required by the Technical Specifications (TS).
l
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.- _
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_ _ _ .
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3
15
The 1996 and 1997 Annual Radiological Environmental Operating Reports included
results of the environmental monitoring program, program changes, Land Use
Census, and interlaboratory comparison program, as required by TS. The reports
provided a comprehensive summary of the results of the REMP around the site and
I
met the TS reporting requirements,
lodine-131 (1-131) was detected in a routine indicator milk sample during the week
of April 21,1997 at a concentration of 0.5 picocuries per liter (pCi/L). The licensee
conducted an investigation and discussed this issue with the NRC in April 1997.
The licensee environmental laboratory investigated the analysis results by
reanalyzing the sample and confirming the results with another laboratory. The
licensee considered the investigation to be detailed. The licensee concluded that (1)
the' source of the I-131 could not be determined, (2) that it was unlikely the source
of the I-131 was from either the Niagara Mohawk Nine Mile Point (NMP) or from the
James. A. FitzPatrick (JAF) plant, (3) and the dose was insignificant compared to
the doses received from natural sources. The details and conclusions of the
investigation was documented in the 1997 Ar:nual Radiological Environmental
Operating Report as required by TS.
c.
Conclusions
1
The licensee effectively maintained and implemented the. Radiological Environmental
Monitoring Program in accordance with regulatory requirements. In .1997, the
licensee performed a comprehensive review of an anomalous indication of iodine-
131 in an environmental milk sample.
l
R1.2 Implementation of the Meteorological Monitoring Program
a.
Inspection Scone (84750)-
,
4
The status of the meteorological instrumentation including system operability,
channel calibrations, and the associated maintenance and calibration of the strip
,
chart recorders were reviewed for the period from July,1996 to May,1998.
!
b.
Observations and Findinos
.
t
The licensee's instrumentation and Control (l&C) department maintained and
,
calibrated the strip chart recorders located in the control room and the Technical
1
. Support Center (TSC). The strip chart recorders were calibrated annually and
according to the maintenance and strip chart recorder calibration procedure. The
i
results were within the licensee's acceptance criteria.
The NMP plant staff continued to perform semiannual channel calibrations for all the
j
- sensors (wind speed, wind direction, and temperature) at the primary, backup, and
I
!
inland meteorological towers for the NMP and JAF site. ' The JAF I&C department
supported the semiannual channel calibration by verifying the output readings in the
i
control room and TSC.
'
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.
16
. c.
Conclusion"
Overall, the licensee effectively maintained meteorological monitoring system
operability, and performed calibrations for the strip chart recorders for the
meteorological instrumentation.
R1.3 Radiation' Controls
,
a.
insoection Scope (8371Q)
Tours, that included independent radiation surveys of the reactor building, were
conducted to review the adequacy of radiological posting practices, and to
determine if plant radiation hazards were as low as is reasonably achievable.
,
Followup discussions with licensee personnel were held to determine the history of
and plans for addressing various radiation hazards.
l
l
b.
Observations and Findinos
. Areas reviewed were found to be posted and controlled in compliance with
. regulatory requirements. There were a number of radioactive piping runs in the
general access areas (e.g., fuel pool cooling, reactor water cleanup transfer, and
cavity drain line piping) that contribute to relatively high background radiation fields.
In addition, the chemistry sample sink area snd a recently installed alternate decay .
~
- heat removal system were chronic sources of radiation exposure to personnel during
j
plant operations. Except for the chemistry sample sink, the licensee has actively
pursued pipe flushing and hydrolysing activities on a periodic basis for the piping
' runs mentioned above. The licensee indicated that the alternate decay heat removal
'l
" system is planned to be permanently shielded during the Summer of 1998.
l
(IFl 50-333/98002-08).
!
c.
Conclusions
!
Based on plant tours during the inspection period, plant areas were in compliance
with established radiological posting procedures and regulatory requirements, and in
- general, most of the principal radiation exposure cources that were accessible
.
during plant operations, have been minimized through effective pipe flushing
..
' activities.
'
R1.4 -As Low As is Reasonably Achievable (ALARA)
- a.
Inspection Scope (83750)
)
'
-
i
A review of exposure performance indicators with respect to industry data was
4
performed utilizing the following licensee documents:
)
2
!
1
,
'..
'
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_ __
-
.
._
,
!
-
f
17
l
l
-1996 Annual summary of the Radiation Field Control and Cobalt Reduction
'
- Programs; and
1
-1997 Annual Summary of the Radiation Field Control and Cobalt Reduction
l
Programs.
b.
Observations and Findinas
,
!
. The licensee averaged 224 person-rem collective personnel exposure over the last
.
two years (representing one refueling cycle). This collective personnel exposure
,
"
compares closely to the boiling water reactor (BWR) industry average of 233
person-rem over the same 2-year time period (1996-1997). A comparison of the
licensee's recirculation piping system contact dose rates with the BWR industry
average of 1996 - 1997 data, indicates that the licensee is significantly lower than
j
the average BWR (119 millirem per hour (mR/hr) versus 218 mR/hr). The licensee's
.
. review of the significance of higher exposures than piping dose rates indicated that
l
additional opportunities to reduce exposure have been identified; these opportunities
include additional efforts to control steam affected area exposures during operating
conditions and improving the effectiveness of ALARA measures to lower outage
exposures.
-
Historically, the licensee has experienced high recirculation piping system dose
- rates. To address this condition, the licensee has conducted a chemical
.
l
decontamination of the recirculation system piping on 3 separate occasions, the last
of which occurred during the 1994 refueling outage. Other plant piping was not-
'
[
chemically decontaminated, although the licensen has been tracking and flushing
l:
,
some other piping systerns on a periodic basis.
,
b
L
' Cobalt-60 input to the plant has been monitored and reduced over recent years.
>
.
Cobalt-60 concentration in reactor water has trended down over the past several
. years and currently is below the vendor recommended target level (2E-5 microcuries
per cubic centimeter (uCi/cc).versus 5E-5 uCi/cc). This has been the result of
effective water chemistry cuntrol that has included the injection of depleted zinc
- oxide into the reactor water since March 1996. In addition, the licensee' has
replaced 62 percent of the control rods with non-cobalt contributors and
approximately 30 percent of valves which interface with reactor water have been
i
'
replaced with non-cobalt valve facing components. ~ A reactor water cleanup system
pipe replacement project is scheduled over the next two refueling outages primarily
as a dose reduction initiative. All of these actions have contributed to significant
radiological source term reductions and indicates a well managed program that has
'been well supported by licensee management.
c.
Conclusions -
The licensee has implemented an effective ALARA program based on declining
station exposures and reductions in source term input ' Ths ALARA program has
' been supported by management through several exposure reduction initiatives
-- including recirculation ' system chemical decontamination, zinc injection, and the
systematic reduction of cobalt contributors.
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R2
Status of RPAC Facilities and Equipment
R2.1 R:diological Protection (RP) Instrument Calibrations and Use
a,
. Inspecpon Scope (83750)
L
A sampling of portable RP instruments, source calibrator, gross sample counting,
gamma spectroscopy counting, thermoluminescent dosimeter, and whole body
l.
counting equipment calibration documentation was reviewed. Additionally, the
!
respiratory protection processing facility was reviewed. Instrument calibration and
!
counting geometries were observed and discussed with the plant staff. As a basis
for instrument calibration, recent radiochemistry analysis of waste streams and in- -
,
'
plant gamma spectral measurements were utilized. Other documents reviewed
included:
RTID 97-002, Nuclide Distribution and average beta energy found in contaminated
. smear sampias at the JAF NPP
j
RP-OPS-05.04, Revision O Radioactive Waste Data Base Control Program :
j
RTP-23, Revision 5 SAC-4 Scintillation Alpha Counter Operation and Calibration
RP-DOS-205 Stand-up Whole. Body Counter Operation and Calibration
RP-DOS-201, Revision 0 investigation and Evaluation of Bioassay Results
RP-DOS-01.OO, Revision O External Dosimetry Quality Assurance Program
!
- b'.
Observations and Findinas
y
L
. Based'on the 1996 outage recirculation pipe gamma spectral study, the average
gamma strength was 1.1 million electron volts (Mevs). The licensee utilized two -
400 curie (Ci) cesium-137 sources to calibrate portable RP instruments at a 0.662
Mev gamma energy. %e discrepanc,y between calibra'Jon energy and plant energy
was not accounted for in the instrument calibrations. The effect of this was
expected to result in an over response for Geiger-Mueller instruments, which is
conservative. The RP staff indicated this would be evaluated with appropriate
'
procedure changes as necessary. The licensee actions will be tracked as lH 50--
333/98002-09. Otherwise, the portable RP instruments reviewed were
appropriately calibrated traceable to National Institute of Standards and Technology
(NIST) to include proper daily response checks.
- Both.RP and chemistry laboratory air sample counting instruments utilized
appropriate NIST-traceable sources of plant radietion energies for calibration and
. daily response checks. In determining internal exposures from air sample gamma
soectral analysis, the. licensee does not account for non-gamma emittere, ' cwever,
this represented only 3.9 percent of the derived air concentration (DAC) frection
and was currently not significant as internal exposures at the licensee for the last
2 years have been below recording thresholds.
Whole body counter calibrations also utilized appropriate NIST-traceable sources
.
. that provided an appropriate gamma spectrum calibration range for the detection of
plant radionuclides. By reference to gamma peak resolution calibration data for the
'
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,
. .
19
sodium-iodide detectors, zinc-65 would most likely be undetected in the presence of
coba!t-60 due to the relatively poor peak resolution of the detectors. Based on
,
recent waste stream radiochemistry analysis, this may represent a loss of
(
approximately 2.9 percent of the internal exposure due to gamma-emitting
radionuclMes. Adding to this the non-gamma emitters of approximately 3.9 percent
'
(as mentioned above) would indicate that current bioassay measurements may read
approximately 6.3 parcent low. ,The licensee has not recorded any positive whole
body counts (greater than procedural threshold of 0.02 annual limit on intake (All))
.
for 1997 through May 1998.' The licensee indicated that an evaluation would be
j
conducted with respect to non-gamma emitting radio nuclides in air samples and
whole body counts as they affect the internal exposure program. (IFl 50-
333/98002-10)
A review of the respirator processing facility indicated only National Institute for
Occupational Safety and Health /Mine Safety and Health Administration
1
(NIOSH/MSHA) approved equipment is used, tested annually and inspected monthly
prior to use. Station air and the air compressor utilized for filling breathing air
l
bottles were tested quarterly to standards which are specified by the Compressed
l
Gas Association. Annual respirator user qualifications, including physical
examination, fit test, and respirator training are computerized and ensure only.
qualified personnel are issued respirators. All respiratory protection areas reviewed
I
met regulatory requirements and were adequate to limit exposures of individuals to
airborne radioactive materials.
'
Record external radiation exposures were determined by thermoluminescent
dosimeters (TLDs) at the licensee. TLD vendor processing by Duke Engineering and
Services Environmental Laboratory maintained National Voluntary Laboratory
Accreditation Program (NVLAP) accreditation in all radiation categories.
Interlaboratory proficiency testing'during the 4th quarter of 1997 demonstrated
excellent results with no deficiencies resulting from the latest NVLAP inspection.
To monitor vendor TLD processing, the licensee effectively implements a quality
control program by comparing spiked TLDs read results by the vendor on a quarterly
basis. No~ discrepancies with TLD processing were noted.
c.
Conclusions
Calibration and use of portable radiological protection instruments, counting
instruments, whole body counting instruments, thermoluminescent dosimeters, and
the use and control of respiratory protection equipment were effectively
implemented as evidenced by current calibration documentation and tours of
applicable facilities. No discrepancies with regulatory requirements were observed.
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R4-
Staff Knowledge and Performance in RP&C
[
R4.1 Weak Evaluation of Unexpected Radiological Condition During Maintenance
a.
Insoection Scone (71750)
l
On April 9,1998,the inspector observed a control rod drive pump relief valve being
reinstalled into the system. The inspector focused on the radiological practices
associated with the maintenance activity, discussed the radiological conditions with
j
_the radiation protection technician who'was' assigned radiological monitoring
]
responsibilities and reviewed the associated radiological survey forms. Radiation
j
protection procedures which provide instructions to radiation protection personnel in
J
the areas of radiological surveys and radiation work permits were reviewed. The
inspector discussed with senior licensee radiation protection management
expectations for acceptable radiological monitoring were.
b.
Observations and Findinas
i
' The relief valve was being reinstalled into the system located in the line between
the control rod drive system and the radiological waste system. Initial radiation and
contamination surveys showed that the contamination levels inside the pipe were
4000 disintegrations per minute per 100 centimeters square. When the job started,
there was no steam present. The inspector observed that, during reinstallation of
the reliet valve into the system, a small plume of steam emitted from the open end
of the pipe leading to the radiological waste system. This section of pipe is
normally cool and steam is not expected to be in the system. The RP technician
providing radiological monitoring decided that additional surveys were not required
and that it was acceptable to continue the work.- His decision was based on his
knowledge of plant chemistry which indicated that there were no fuel leaks, pre-job
survey results which showed that the contamination levels inside the pipe were
low, ahd the steam was'of a sufficiently low flow so the potential to transport
contaminated corrosion products was low. Additionally, the inspector noted that -
the mechanics did not question the source of the steam and continued to work
using rags to protect themselves from the hot pipe and backing away fro n the
steam as much as pcseble.
1
The inspector discussed his observations with senior licensee radiation protection
management.' t.icensee management expectations for radiological monitoring were
not met in this case in that the evaluation of an unexpected condition was weak.
"
Additional sur/eys, including airborne radiation surveys were expected. Licensee
corrective ac'. ions included additional coaching and counseling for radiation
technicians and mechanics.
!
Post job' surveys showed that there was no spread of contamination. The licensee
determined that there were several possible sources of steam into tha radioactive
waste system and generated problem identification reports to investigate the
source.
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21
c.
Conclusions
On April 9,1998, radiological monitoring for the control rod drive relief valve
replacement was weak. During the maintenance activity, an unexpected condition
occurred which was that steam emitted from a normally cold system. Mechanics
and radiation protection technicians did not demonstrate a questioning attitude
when faced with this unexpected condition as the unexpected condition was not
fully evaluated nor were radiological surveys taken. There were no radiological
consequences as post job surveys did not show contamination or abnormal airborne
radioactivity levels.
R7
Quality Assurance in RP&C Activities
R7.1 Quality Assurance Audit Program
'
a.
Inspection Scone (84750)
The inspector reviewed the following Quality Assurance (QA) audit reports:
A97-07J, Radiological Environmental Monitoring Program:
-
-
,
A97-18J, J. A. FitzPatrick Emergency Preparedness Programs.
-
b.
Observations and Findinas
The scope and technical depth of ths audits effectively assessed the programs for
strengths and weaknesses. Few findings and recommendations were identified as a
result of the' audits. The departments responsible to correct deficiencies responded
to these findings and recommendations in a timely manner. The above audits also
included a review of previous recommendations and deviation event reports (DERs).
The previous DERs had been closed.
c.
Conclusions
The QA audits reviewed were thorough and of sufficient depth to assess the
radiological Environmental Monitoring and Meteorological Monitoring Programs.
R7.2 Quality Assurance of Analytical Measurements
a.
Insnection Scooe '84750)
The following aspects of the Quality Assuance/ Quality Control (QA/QC) program
for the J. A. FitzPatrick Environmental Laboratory for the period of July 1996 to
May 1998 were reviewed:
the results of the internal QC program, including efficiency and resolution
-
checks, daily instrument energy checks, control charts of instrument
performance, and routine calibrations; and
l'
.
____________
..
22
-
the results of the QA program, including the Interlaboratory Comparison
(cross-check) Program.
b.
Observations and Findinas
' The QA/QC program for analyses of REMP samples is conducted by the licensee's
environmental laboratory. The laboratory implemented intrataboratory (QC) and
interlaboratory (QA) programs. The intralaboratory program included efficiency and
resolution checks, daily instrument energy checks, control charts of instrument
performance, and routine calibrations. The results for 1996 and 1997 were
compiled and documented in the Environmental Laboratory QA/QC Report. The
results from 1996 through 1998 were within the acceptance criteria. The
'-
- laboratory continued to participate in an Interlaboratory Comparison Program
provided b y a vendor (Analytics, Inc.). The laboratory's participation in this
program was effective.
In addition to the above required comparison programs, the laboratory participated-
in a cross check program with the Environmental Measurements Laboratory (EML),
Department of Energy. The analysis results of this program were generally in
agreement, with occasional disagreements in certain samples. The laboratory had
conducted an investigation and determined the cause of the disagreements. EML
provided sample media and geometries different from the usual sample rr.edia and
geometries provided by Analytics, Inc. and the licensee. The laboratory
.
accommodated and incorporated different and difficult media and geometries into
the program. The licensee published a disc' ssion and the results in the 1997 -
u
Annual Radiological Environmental Operating Report, as required by TS.
c.
Conclusion
The environmental laboratory continued to implement effective Quality
- Assurance / Quality Control programs for the Radiological Environmental Monitoring
' Program, and continued to provide effective validation of analytical results. The
laboratory demonstrated the ability to accommodate and incorporate difficult media
and geometries into the program. The programs are capable of ensuring
independent checks on the precision and accuracy of the measurements of
!
radioactive material in environmental media.
]
i
R7.3 - Radiological Protection Program Oversight
I
a.
Insoection Scooe (83750)
,
Quality assurance surveillance which were conducted between September 1997
'
and April.1998 and an October 1997 audit of the RP program were reviewed in
addition, an RP self assessment summary report covering RP self-assessments from
June through December 1997 was reviewed. Radiological problem identification
l
and resolution program DERs issued between January through May 13,1998 were
.
G
also reviewed,
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_ - - - _ - _ _ - - _ - - - _ - - _ _ _
- - _ _ _ -
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23
b.
Observations and Findinas
There were three QA surveillance pertaining to the RP program which were
conducted during the previous 8 month period. These surveillance were adequate.
An RP audit conducted ever a two week period consisted of six auditors and
'inclu-led two outside utility technical specialists. The audit was focused on relevant
h
RP areas and was of adequate depth. The RP self-assessment program covering the
last six months of 1997 consisted of two special subject studies (ALARA cost
benefit for hydrogen injection and posting of RCAs) and nine management -
observations. The RP self-assessment summary report provided good insights in
!
- perceived problem areas and provided good direction for the RP program.
'
A review of 31 radiological DERs which were generated between January and .
May 1998, indicated a high volume, low threshold problem reporting system. The
issues reviewed were generally of low safety significance and corrective actions
. were appropriate.
c.
Conclusions
Radiological protection program oversight between September 1997 and
April 1998, consisted of an effective combination of Quality Assurance (QA)
L
surveillance, QA audits, self-assessments, and feedback through the station
'
problem identification and resolution program.
R8
Miscellaneous RP&C lasues
R8.1 - (Closed) Inspector Followup Item (IFI) 50-333/97008-04: radioactive waste training
' program not well organized and documented. A matrix of required training fo-
,
' personnelinvolved with radioactive waste processing and radioactive material
l
l
- shipping was available,'and the related computerized training record database was
updated. ~ ~ Appropriate actions were completed to address this IFl.
'R8.2 Improperly Classified Shipment of Radioactive Materials (VIO 50-333/98002-11)
j
a.
Inspection Scooe (83750)
~
' The health physics related items'which were documented m 1 a licensee's problem
identification and resolution process, were selectively revievoed. Deviation Event
Report (DER) 98-00541 which pertained to an improperly classified radioactive
waste shipment was reviewed in detail.
~
b.
~ Observe; ions and Findinas
'
DER 98-00541 documented that a radioactive materials shipment was made on
-
March 17,1998 which was improperly classified.' On March 9,1998, three-
= packages (two high efficiency particulate (HEPA) unhs and a B-25 box containing -
HEPA hoses) were surveyed in preparation for shiprnent. Tha survey results
indicated that each' package exceeded 0.5 millirem /hr at contact (the Department of
a:
_
__ -_--_-
--___
.
.
24
Transportation (DOT) limit for excepted packages) on at least four sides and that the
radiation level at 30 centimeters for each package was less than 0.1 millirem /hr.
The original shipment plan intended the three packages to be placed inside a sealed
container, which would constitute the package surface, and the shipping paperwork
j
and classification was based on this planned configuration. However, the actual
shipment was made with the two HEPA units inside tne sealed container and the
B-25 box containing HEPA hoses was located outside the sealed container on the
flatbed. in this configuration, the sealed container's ex+ernal surfaces had radiation
levels equal to or less than 0.5 millirem per hour while the radiation levels on
contact with five of six sides of the B-25 box exceeded 0.5 millirem per hour with
the maximum contact reading being 1.8 millirem per hour.
The shipment was classified as Radioactive Material, Excepted Package-Limited
Quantity of Material,7, UN 2910. The recipient of the shipment conducted an
incoming package receipt survey. The survey showed a dose rate of approximately
1 millirem per hour on the box. 49 CFR 173.421(a)(2),for excepted packages of
limited quantities of Class 7 (radioactive) materials, requires that the radiation level
at any point on the external surface of the package does not exceed
0.5 millirem / hour. The receiving licensee did not accept the shipment and returned
the shipment to the shipping licensee. Based on this information, the shipment on
March 17,1998, constituted a failure to comply with 10 CFR 71.5 with respect to
49 CFR 173.421(a)(2). (VIO 50 333/98002-11) The inspector noted that the
violation was not self identified during the preparation and review of the shipping
paperwork pack 1ge.
c.
Conclusions
On March 17,1998, a shipment of radioactive material was made to another
licensee and was improperly classified as excepted packages with package surface
radiation levels exceeding 0.5 millirem per hour, which constituted a failure to
comply with 10 CFR 71.5(a) with respect to 49 CFR 173.421(a)(2). The inspector
noted that the violation was not self identified during the preparation and review of
the shipping paperwork package.
F4
Fire Protection Staff Knowledge and Performance
F4.1
Response to Fire in Computer Room
a.
insoection Scoce (71750)
On May 4,1998, at 7:35 a.m., a small fire started in a computer room located in
the Technical Support Center. The inspectors observed contail room operator and
fire brigade response to the fire. The licensee's corrective actions for the fire were
reviewed and discussions were held with fire protection personnel.
.
. _ _ - _ _
. - _ _ __
_.
_
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_ - - _ _ _ _
___ _ _-
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-
.
0
25
b.
Observations and Findinas
The first indication of a fire was an alarm in the control room indicating that the
Halon system in the computer room had initiated. Control room operators
appropriately announced the fire and directed that the fire brigade respond.
i
Communications were noted to be good and fire protection procedures were
l
referred to by the control room operators. The fire brigade responded in a timely
l
manner and with appropriate fire fighting equipment. The fire had been
extinguished by the Halon injection in less than 15 minutes. Licensee investigation
deterrrsined that the fire may have been caused by a degraded fluorescent light
ballast. There was no damage to the room or equipment.
c.
Conclusions
On May 4,1998, control room operators and fire brigade personnel responded
appropriately to a small fire in a computer room.
V. Manaaement Meetinas
X1
Exit Meeting Summary
The inspectors presented the inspections results to members of the licensee
management at the conclusion of the inspection on June 11,1998. Also, on
April 24,1998, Mr. C. Sisco presented the NRC's inspection findings in the area of
licensed operator requalification. In addition, on May 22,1998, Mr. J. Noggle and
Ms. L. Peluso presented the NRC's inspection findings concerning the radiological
control program and radiological environmental monitoring program, respectively.
The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined dur ng the
inspection should be considered proprietary. No proprietary information was
identified.
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O
ATTACHMENT 1
PARTIAL LIST OF PERSONS CONTACTED
Licensee
M. Colomb, Site Executive Officer
D. Lindsey, General Manager, Operations
J. Maurer, General Manager 4t:pport Services
A. Mckeen, Radiological and Environmental Services Manager
D. Ruddy, Director, Design Engineering
I
i
D. Topley, General Manager, Maintenance
A. Zaremba, Licensing Manager
J. Romanowski, Operations Training Supervisor
NRC
.
R. Conte, Chief, Operator Licensing - Human Performance Branch
q
J. Noggle, Senior Radiation Specialist
!
C. Sisco, Operations Engineer, Operator Licensing - Human Performance Branch
B. Thomas, Technical Reviewer, Plant Systems Branch
T. Walker, Senior Enforcement Specialist
J. Williams, Project Manager
H. Williams, Senior Operations Engineer
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 61726:
Surveillance Observations
IP 62707:
Maintenance Observation
IP 71CJ1:
Licensed Operator Requalification Program '.: valuation
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 83750:
Occupational Radiation Exposure
IP 84750:
Radioactive Waste Treatment, and Effluent and Environmental Monitoring
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
I
'1
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_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
-
.
Attschm:nt 1
2
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-333/98002-01
failure to carry out the actions of the correct procedure during
the use of the emergency operating procedures
50-333/98002-02
a plant procedure was not adequate concerning assigned
duties of the on-shift licensed operator
50-333/98002-03
an operator aide was not adequate concerning plant operations
during degraded core flow conditions
50-333/98002-04
continued poor attendance by some operators during the
current LORT program
50-333/98002-05
LORT program procedures regarding duplication of examination
materials
50-333/98002-06
exclusion of low power or shutdown conditions from the
annual operating test
50-333/98002-07
exclusion of emergency plan from the annual operating test
50-333/98002-08
IFl
shield alternate decay heat removal system by Fall,1998
50-333/98002-09
IFl
gamma calibration of radiation protection instruments one-half
of plant gamma enercy
50-333/98002-10
IFl
account for non-gammas in air samples and whole body counts
50-333/98002-11
improperly classified radioactive material shipment
Closed
50 333/95004-01
LER
logic system functional test procedure revision
50-333/96006-02
condenser rupture disc actuation
50-333/97007-02
moving spent fuel with the reactor building crane
50-333/97008-04
IFl '
training program was not well organized and documented
50-333/97009-01
emergency nperating procedure terminology is not consistent
with emergency procedure guidelines
Qiscussed
None
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e
Attachment 1
3
'
LIST OF ACRONYMS USED
As Low As Reasonably Achievable
All
Annual Limit on intake
Abnormal Operating Procedure
Administrative Procedure
Alternate Rod insertion
Bypass Valve
Boiling Water Reactor
Ci
Curie
Control Rod Drive
DAC'
Derived Air Concentration
DER
Deviation Event Report
Department of Transportation
Equipment Failure Evaluation
EML
Environmental Measurements Laboratory
Emergency Operating Procedure
EOP Support Procedure
Emergency Procedure Guidelines
Final Safety Analysis Report
High Efficiency Particulate
High Pressure Coolant injection
1-131
lodine-131
Instrumentation and Control
IFl
inspector Followup Item
JAF
J. A. FitzPatrick
Low Condenser Vacuum
LER
Licensee Event Report
Licensed Operator training Program
Mev
Million Electron Volts
Motor Generator
Maintenance Procedure
MrAr
Millirem per Hour
NIOSH/MSHA National Institute for Occupational Safety and Health /Mine Safety and Health
Administration
Nationallnstitute of Standards and Technology
Nine Mile Point
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
National Voluntary Laboratory Accreditation Program
pCi/L
Pico-curies per Liter
Public Document Roora
'
Preventive Maintenance
PSTG
Plant Specific Technical Guide
i
Post Transient Technical Guide
Quality Assurance
Quality Control
Reactor Analyst Procedure
i
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4
Attachmint 1
4
Radiological Controlled Are-
Reactor Core Isolation Cooling
RC/L
Reactor Control / Level
i
Radiological Environmental Monitoring Program
I
Reactor Feed Pumps
Reactor Manual Control System
Radiological Protection
>
RP&C
Radiological Protection and Chemistry
Rod Position Indication System
Systems Approach to Training
' Safety Evaluation
TLD'
Thermoluminescent Dosimeter
'
Training Procedure
TS
Technical Specification
i
uCi/cc
Micro-curies per cubic centimeter
Updated Final Safety Analysis Report
Unresolved item
Unreviewed Safety Question
Violation
L-