ML20199E471
| ML20199E471 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/20/1998 |
| From: | Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20199E413 | List: |
| References | |
| 50-333-97-08, 50-333-97-8, NUDOCS 9802020164 | |
| Download: ML20199E471 (31) | |
See also: IR 05000333/1997008
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U.S. NUCLEAR REGULATORY COMMISSION
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Region I
License No.: DPR-59
Report No.: 97-08
Docket No.: 50-333
Licensee: New York Power Authority
Post Office Box 41
Scriba, New York 13093
Facility Name: James A. FitzPatrick Nuclear Power Plant
Dates: October 27,1997 through December 21,1997
-' Inspectors: G. Hunegs, Senior Resident inspector -
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R. Fernandes, Resident inspector
J. McFadden, Radiation Specialist
Approved by: John F. Rogge, Chief, Projects Branch 2
Division of Reactor Projects.
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9802020164 980120
PDR ADOCK 05000333
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EXECUTIVE SUMMARY
James A. FitzPatrick Nuclear Power Plant
NRC Inspection Report 50-333/97 08 --
Operations
e The shutdown for the forced outage conducted on December 7 was safe and well
controlled. Good command and control, communication and procedure adherence
-were noted. Operator observations, involving a degraded residual heat removal
system pipe support and mislabeled containment isolation valve, demonstrated
good operation practices.- The reactor startup following the outage was performed
in a safe and prudent manner,
o An operator error was made while performing an electrical ground isolation
abnormal operating procedure. Specifically, breakers were operated out of 4
- sequence, resulting in the inadvertent automatic operation of high pressure coolant
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injection (HPCl) system valves. Although the valve operation had minor safety
consequences as the HPCI system was out of service for maintenance, the
improper performance of an abnormal operating procedure was determined to be a
--violation. Additionally,,the pre-evolutior, brief for the operations staff was weak in
that the assignment of personnel to conduct breaker manipulations was not made,
e The inspector observed portions of the Safety Review Committee meeting
- conducted on November 20-21,1997 and noted that the meeting demonstrated
good safety oversight of station activities.
Maintenance
e - During emergency diesel generator maintenance activities, extensive supervisor
involvement was noted. Additionally, pre-evolution briefs wers conducted for
activities where warranted and procedures were in use. Emergent issues including
a lost lube oil valve disc retaining nut and damaged piston assembly resulted in the - y
work activity taking longer than originally scheduled. These emergent issues were '
effectively addressed through good coordination between operations,
maintenance, quality assurance, technical services and supervisor oversight,
e The process to control work activities associated with troubleshooting to locate a
direct current ground was unsatisfactory and resulted in an invalid engineered
safeguards feature (ESF) actuation signal for the high pressure coolant injection
(HPCl) steam supply valves. The HPCI system was out of service for scheduled
maintenance. Operators did not recognize that the troubleshooting activities made
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the primary containment isolation system (PCIS) function inoperable and therefore
did not enter the appropriate Technical Specification Limiting Condition for
Operation (LCO) action statement. The licensee's immediate corrective actions
were appropriate and the root cause analysis was critical of the oaeration staff's
handling of the troubleshooting activities, but lacked in-depth review of the work
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Executive Summary (cont'd)
control process for the activity. Additionally, the licensee's use of junction boxes
for temporary storage of parts was considered to be a poor work practice. The
failure to enter the TS LCO was a violation.
e The work package to prepare for replacement of the low pressure coolant injection
(LPCI) battery was weak in that the impact of removing a portion of the battery
enclosure ,on LPCI battery operability was not considered prior to beginning the +
work. Although the work was stopped, the licensee subsequently determined that
- the work would not impact battery operability. Additionally, plant drawings for the'
structure were not reviewed prior to the work being performed which contributed
to confusion in performing the task.
Enaineerina
e Environmental qualification (EO) components for the high pressure coolant
injection (HPCI) system were erroneuasly removed from the EQ program in 1993,
and in f act, may not have originally met EQ criteria because of installed
unrecognized test jacks which affected the EQ of the system. The licensee
prepared a justification for continued operation (JCO) which provided reasonable
assurance that the equipment would perform its safety function. The licensee was
slow to pursue the JCO because the impact of this non-EQ component on HPCI
system operability was not initially recognized. Once the problem was recognized,-
.the licensee was aggressive in resolving the issue. The EQ issue was
appropriately resolved through removing the component connection to the test
-Jacks and inserting the previously removed components back into the scope of the
EQ prooram. The licensee's erroneous removal of HPCI components from the EQ -
program was a violation of 10 CFR 50.49.
e The licensee's program to monitor safety relief valve (SRV) leakage was effective.
Licensee management exercised good judgement in electing to shutdown the plant
to effect repairs to leaking SRVs.
Plant Sunoort
o Overall, the solid radioactive waste program and ectivities and the program for the
transportation of radioactive naterials and its related activities were well managed
and effective. The quality assurance audits and surveillance reports were
thorough, programmatic and well documented.
e Training for personnel involved with solid radioactive waste activities was
appropriate in scope and depth. However, the training program was not well
organized and documented and therefore the administration of the training program
was a weakness.
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Executive Summary (cont'd)-
e- Radiological controls in administrative areas relative to the elevated radiation levels-
-due to hydrogen injection were proper and adequate.
e- - On December 11,1997, an emergency plan joint drill was conducted with the
licensee and Nine Mile Point participating. The emergency preparedness (EP) drill
demonstrated solid performance of the EP staff and licensee organization. -
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TABLE OF CONTENTS
EXECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . II
TA BLE O F CO NT E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1. O PE R ATI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . 1
- 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
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'01.1 Operational Safety Verification . . . . . . . . . . . . . . . . . .-. . . . -. . . 1
01.2 Plant Shutdown due to Safety Relief Valve Leakage .......--...2
04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 3
04.1 Battery Ground Isolation Procedure Error (Violation 50-333/97007-01)
...............................................3
07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
07.1 Licensee Self Assessment Activities .....................5
. ll . M AI NT E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
-M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
M1.1 General Comments on Maintenance and Surveillance Activities . . . 5 -
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M1.2 "A"_ Emergency Diesel Generator Scheduled Maintenance . . . . . . 6
M4 Maintenance Staff Knowladge and Performance . . . . . . . . . . . . . . . . . . 7_ :
M4.1 Invalid Engineered Safeguards Feature (ESF) Actuation and Failure to.
Perform Technical Specification Required Actions While Performing
Troubleshooting (Violation 50-333/97008-02) . . . . . . . . . . . .. . . . .L7
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M4.2 Low Pressure Coolant injection (LPCI) Battery Replacement . . . . . 9
lil . E NGI N EERI NG . . . . . . . . . . . . . . - . . . . . - . . . . . . . . . . . - . . . . . . . . . . . . . . . . . . . . 10
E1- Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
E1.1 Environmental Qualification of components in the High Pressure
Coolant Injection System (Violation 50-333/97008-03)
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E2 Engineering Support of Facilities and Equipment ..................12
E2.1 ' Licensee Monitoring of Leaking Safety Relief Valves (SRVs) . . . . 12 -
E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 12
E8.1 = (Closed) Inspector Follow up Item (IFI) 50-333/96007-02...... 12 .i
E8.2 (Closed) Unresolved item (URI) 50 333/95006-03 ........... 13
IV. Plant Support ................................................14
R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . - 14
R1.1 Implementation of the Solid Radioactive Waste Program .......14
R1.2 Compliance with NRC and Department of Transportation (DOT)
Regulations for Shipping of Low Level Radioactive Waste (LLRW) for
Disposal and Transportation of Other Radioactive Materials
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Table of Contents (cont'd)
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R1.3 Elevated Radiation Exposure Levels Due Hydrogen injection . . . . 15
j R5 Staff Training and Qualification in RP&C (Inspector Follow-up Item (IFI) 50-
333/97-008-04) !
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R7 Quality Assurance in RP&C Activities . . ......................18
R8 Miscellaneous RP&C lssues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 ->
R8.1 (Closed) Violation 50 333/96007-08 . . . . . . . . . . . . . . . . . . . . 19
R8.2 (Closed) Violation 50-333/96007 09 . . . . . . . , . . . . . . . . . 19
P1 Conduct of EP Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
P1.1 Emergency Plan Drill . . . . . . . . . . . . . . . . . . ........... 19
P8 Miscellaneous EP lssues (EA 98-008)(NCV 97-008-G ,...........20
V. M AN AG EM ENT M EETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
ATTACHMENT
Attachment 1 - Partial List of Persons Contacts
- Inspection Procedures Used
- Items Opened, Closed, and Discussed
- List of Acronyms Used
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Report Details
lummarv of Plant Status
' The unit began this inspection period at 100 percent power. On December 7, the plant
was taken to cold shutdown to repair leaking safety relief valves (SRVs). The plant was i
taken critical on December 13 and returned to 100 percent power on December 17. The
- plant continued operation at 100 percent through the end of the inspection period,
l. OPERATIONS
01 Conduct of Operations'
01.1 - Operational Safety Verification
a. Insoection Scoce
The inspectors observed plant operation and verified that the facility was operated
safely and in accordance with procedures and regulatory requirements. Regular
tours were conducted of the plant with focus on safety related structures and
systems, operations, radiological controls and security. Additionally, the ,
operability of engineered safety features, other safety related systems and on-site
and off site power sources was verified. -The inspectors performed walk downs of-
accessible portions of several safety related systems.
The inspection activities 'during this report period included inspection during ,
normal, back shift and weekend hours. Regular tours were conducted of the -
following plant areas:
control room
secondary containment building
radiological control point
electrical switchgear rooms
emergency core cooling system pump rooms
security access point
protected area fence
intake structure
diesel generator rooms
Control room instruments and plant computer indications were observed for
correlation between channels and for conformance with technical specification
(TS) requirements. The inspectors observed various alarm conditions and
confirmed that operator response was in accordance with plant operating
procedures. Compliance with TSs and implementation of appropriate action
statements for equipment out of service was inspected. Plant radiation monitoring
' Topical headings such as 01, M8, etc., are used in accordance with the NRC
standardized reactor inspection report outline. Individual reports are not expected to
address all outline topics.
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system indications reviewed for unexpected changes. Logs and records were -
reviewed to determine if entries were accurate and identified equipment status or -
deficiencies. These records included operating logs, turnover sheets, system
safety tags, and temporary modifications. Control room and shift manning were
compared to regulatory requirements and portions of shift turnovers were
observed. Daily supervisor meetings were attended to assess personnel focus on
risk significant items and plant priorities,
b. Observations and Findinos
Overall, the licensee operated the plant safely. Plant activities were performed in
accordance with procedures and effective controls were implemented for safe
plant operation. Overall, equipment operability, material condition and
housekeeping conditions were good,
c. Conclusions
Overall, the licensee operated the plant safely and activities were performed in
conformance with requirements. Effective controls were implemented to achieve -
safe operation of the plant.
01.2 Plant Shutdown due to Safety Relief Valve Leakage
a. inspection Scoce
During the inspection period the licensee noted an increasing trend in safety relief
valve (SRV) leakage to the torus. On December 1, the licensee elected to enter a
forced outage in order to replace the leaking SRVs. The inspectors witnessed
various portions of the shutdown preparations, power reduction, and reactor
cooldown and depressurization activities on December 7. The inspectors'
objective was to determine the effectiveness of management controls in ensuring
a safe transition to shutdown, in addition, the inspectors observsd portions of the
reactor startup conducted on December 13,1997. Inspector attemion was
focused on reactivity control, operator procedure use and communications,
b. Observations and Findinas
The unit was shutdown per operating procedure (OP)-65, Start-up and Shutdown
Procedure. Power reduction was performed in accordance with reactor analyst
procedure (RAP)-7.3.16,I'lant Power Changes, and the main generator was
removed from service on December 7,in accordance with applicable ope:ating
procedures. The unit was in cold shutdown at 3:48 a.m. and the reactor mode
switch was taken to the refuel position at 4:47 a.m. on December 8.
The inspectors noted good command and control of unit shutdown activities.
Communications were professional and precise with three-point communications
used. Coordination of various shutdown activities by licensed operators was very
good. Appropriate oversight of personnel during manipulation of the reactor
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controls was noted. For example, a second checker for control rod motion and
selection was stationed. in addition, senior licensee management personnel were
assigned for shift coverage.
Prior to the startup of tha plant, the operations staff noted two plant deficiencies. -
One deficiency involved dislodged grouting material from behind a pipe support in
the residual heat removal system, and the second issue involved a mislabeled
containment isolation valve. Both issues were adequately addressed by the
licensee prior to startup, and reviewed by the inspectors. The latter event will be
1 further reviewed following the issuance of a licensee event report (LER).
The startup was characterized by clear operator communications and procedure
use, attentive management oversight, and effective control by shift supervision.
Shift tumover meetings were performed in a controlled manner and crew briefings
were good. Senior operations management personnel were designated to provide
continuous oversight,
c. Conclusions
The shutdown for the forced outage was safe and well controlled. Good
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command and control, communication and procedure adherence were noted. The
observations by the operators demonstrated good operational practices. The
reactor startup following the outage was performed in a safe and prudent manner.
04 Operator Knowledge and Performance
04.1 - Battery Ground Isolation Procedure Error (Violation 50-333/97007 01)
e. Insoection Scone
On October 23, the operators entered abnormal operatino procedure (AOP)-23,
Direct Current (DC) Power System Ground Isolation, in response to indications of a
ground on the B" battery. Testing involving the high pressure coolant injection
(HPCI) logic system had just been completed prior to the ground appearing on the
control room instrumentation, so the control room staff elected to proceed to the
portion of the AOP which isolates the HPCI logic circuitry. During performance of
the procedure, the operators failed to open the power supply breakers for 23MOV-
57 and 23MOV-58, the HPCI liooster pump suction from the suppression pool
downstream and upstream isoldion valves respectively. This resulted in the
valves automatically opening when the correct circuit breaker,71DCB2 Breaker 6,
HPCI Logic Power Supply, was opened in the improper sequence. The event
occurred during a HPCI maintenance limiting condition for operation (LCO) and
therefore the system was already considered inoperable. The inspector reviewed
procedures, plant logs and conducted interviews with station personnel involved in
the performance of the ground isolation procedure.
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b. Findinos and Ohscrvatiores
AOP-23, DC power System 0 Ground Isolation, provides steps which attempt to
locate the source of a ground in the DC power system. The procedure contains
general steps in the main body of the text and lists the specific breakers to be
utilized in the isolation of the grounds h on attachment to the procedure.- In b.lef,-
the procedure directs the operators to establish communications between the
control room and the operator at the specified breaker, perform any actions
required by the breaker attachment sheet, enter the any applicable LCOs, and open
the isolation breaker. The ground detector in the control room is then monitored
to see the effect, if any, of opening the isolation breaker. The process is repeated
until the ground is isolated. More specifically, in the attachment to the procedure,
tables identify the isolation breaker to be opened, its corresponding circuit or
component, and the actions required prior to opening the isolation breaker. In this
particular event, the operator be.:ame focused on selecting the proper isolation
breaker and omitted the requirements of the proceduro to open the power supply
breakers for 10MOV 57 and 10MOV 58.
In discussion with the plant staff, the inspector learned that the pre-evolution bdef
for the operations staff was not specific. The operators had been monitoring the
ground circuit prior to the alarming condition being reached, taken out the AOP, -
and discussed the most probable circuit to check based on recent HPCI system
testing. The inspector noted that all the control room staff had been included in
the discussions of current plant conditions, including the selection of an additional
operator to perform a peer check of the isolation breaker operation. However, the .
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assignment or discussion of who was going to open the breakers to 10MOV-57 -
, and 10MOV 58 was not discussed as part of the brief.
The impact of the procedure error was to cause the HPCI booster pump suction to
shift from the condensate storage tank (CST) to the torus. The torus suction
valves are designed to go open on low CST water level or high suppression pool
level to ensure that HPCI has a makeup water source. This action occurred
because the HPCI logic circuitry, following a power loss to the CST level
instrumentation when breaker six was opened, caused the suction valves to
automatically go open. As previously stated, the system was undergoing
maintenance and thus was already considered inoperable. The impact was limited
to unnecessarily challenging the HPCI logic circuitry and cycling valves.
Immediate correct!- n actions were to restore the power to the logic circuitry,
reposition the valves, and re-perform the procedure correctly. The electrical
ground was subsequently located and fixed.
c. Conclusions
The operator error in performing the actions of the AOP had minor safety
consequences, however, the proper performance of abnormal operating procedures
is of high importance and was determined to be a violation. (50 333/97008-01)
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Additionally, the pre evolution brief for the operations staff was weak in that the -
assignment of personnel to conduct breaker manipulations was not made..
07 Quality Assurance in Operations
07.1 Licensee Self-Assessment Activities
a. insoection Scooe
- During tlie inspection period, the inspectors reviewed multiple licensee self-
assessment activities, including portions of the Safety Review Committee (SRC)
meeting conducted on November 20 - 21,1997. Observations of the SRC
meeting are noted below,
b. Observations and Findinas
Recent plant history and issues and performance indicators, operational review and
human performance trends were discussed. Specific issues that were discussed in
depth included nuclear personnel turnover and engineering lack of rigor. SRC
members demonstrated a good questioning attitude and good interaction with the' -
Indian Point 3 representative were noted. Follow up items were developed where
appropriate,
c. Conclusions
. The SRC meeting demonstrated good safety overright of station activities. e
11. MAINTENANCE
M1 Conduct of Maintenance
M1.1 General Comments on Maintenance and Surveillance Activities
a. Insoection Scooe
The inspectors observed selected maintenance activities to verify that activities
were conducted in a manner sufficient to ensure reliable, safe operation of the
plant. The inspectors observed selected surveillance tests to determine whether
the tests were conducted in accordance with technical specification and other
requirements.
The inspectors observed all or portions of the following work activities:
WR 97-06988-06 Replace "B" control rod drive (CRD) pump and restore
temporary modification 97-095,
WR 97-08389-01 Investigate and repair source of water leakage in main stack
room.
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WR 97-08063-02 Troubleshoot / repair valve positioner, feedwater heater 33E 5A
l drain valve operator.
l WR 97 06476-00 Troubleshoot / repair pre-cooler drain lines.
The inspectors observed portions of the following surveillance activities:
ST-3P Core spray flow rate and valve inservice test.
ST SD Average Power Range Monitor Calibration.
ISP 94 Reactor Protection System Electrical Protection Assembly
Functional Test / Calibration.
b. Observations and Findinas
The inspectors found the work performed under these activities to be professional
and thorough. Technicians were experienced and knowledgeable of their assigned
task. Activities were conducted appropriately and in eccordance with procedural
and administrative requirements. Good coordination and communication were
observed during performance of the surveillance activities.
c. Conclusions
Overall, the above maintenance and surveillance activities were well conducted,
with good adherence to both administrative requirements and maintenance and -
surveillance procedures.
M1.2 "A" Emergency Diesel Generator Scheduled Maintenance
a. Insoection Scooe
The "A" emergency diesel generator (EDG) was scheduled for planned
maintenance from November 3 to 5,1997. Activities to be completed incit.ded
routine preventive and corrective maintenance, fuel oil replacement, and power
pack assembly (cylinder liner, pistons and associated components) replacement.
The inspector observed selected activities including procedure use, quality
assurance, and supervisor oversight. During the planned mainunance, emergent
work including a replaced power pack failure and the identification of a missing nut
on a lube oil check valve were also reviewed.
b. Observations and Findinos
During performance of maintenance procedure (MP) 93.11, the lube oil gallery
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supply check valve was inspected due to industry information which documented
a history of problems with the valve Mechanics identified that the valve disc
retaining nut was missing, and the disc was lying on the bottom of the valve. The
valve is a % inch swing check valve. The licensee initiated deficiency and event
report (DER) 97-1545 to investigate the problem and to analyze the impact of the
missing nut. The check valve is located between the lubo oil cooler and main lube
oil pump discharge in a line used for lube oil warm up when the engine is
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shutdown. On November 5, a licensee quality assurance (OA) inspector found a -
brass nut of similar proportions located outside the screenwell and the licensee
determined that the nut was the missing nut.
The licensee's basis for this determination was that the lube oil cooler had been
disassembled prior to the valve inspection. The fact that the nut was missing was '
not known at the time of the lube oil cooler inspection. Detection of tne nut i
during tube oiler cooler maintenance would be difficult, due to the size of the nut
and the amount of oil present. Since the nut was not found in the locations where ~
lt would be expected to be based on lube oil flow paths, the licensee concluded
that the nut was removed from the cooler without detection during cooler
maintenance. The inspector concluded that the licensee's analysis was -
reasonable. /
Another major activity completed was that the power pack assemblies for all
cylinders were replaced. During EDG post work testing, a high crankcase pressure
alarm was observed after about 15 minutes of EDG operation and the engine was
shutdown. It was determined that one piston and liner was damaged.
Specifically, the bottom end of the piston skirt was broken and other internal parts '
were damaged. The assembly was removed and replaced with a rebuilt
assembly, broken parts were retrieved from the lube oil sump and the lube oil
strainer and filter were changed. -The licensee is awaiting the results of an
equipment failure evaluation for the damaged power pack assembly.-
The EDG limiting condition for operation (LCO) was exited on November 8. The
- delay in completing the work activities was a result of emergent work.
c. Conclusions
Extensive supervisor involvement was noted. Additionally, pre-evolution briefs
were conducted for activities where warranted and procedures were in use.
Emergent issues including the lost lube oil valve disc retaining nut and the
damaged piston assembly resulted in the work activity taking longer than originally
scheduled. These emergent issues were effectively addressed through good
coordination Letween operations, maintenance, quality assurance, technical
services and supervisor oversight.
M4 Maintenance Staff Knewledge and Performance
M4.1 invalid Engineered Safeguards Feature (ESF) Actuation and Failure to Perform
Technical Specification Reauired Actions While Performing Troubleshooting
(Violation 50-333/97008-02)
a. Insoection Scooe
On October 24, while performing troubleshooting to locate a ground on the "B"
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DC power system, an inadvertent short across a pair of test jacks in an electrical
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panel caused a partial isolation signal for *.he high pressure coolant injection (HPCI)
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system. The HPCI system was not operating at the time and was in a scheduled
limiting condition for operation (LCO) for maintenance. The expected isolation of
the appropriate HPCI steam supply valves did not occur. The operations staff
subsequently determined that fuses removed from the HPCI logic circuitry during
troubleshooting, prevented the associated isolation valves from automatically
closing. The maintenance activity disabled the primary containment isolation .
function of the HPCI valves without entering the appropriate LCO. Following the
discovery of this, the licensee completed the actions required by Technical
Specifications (TS) by isolating the outboard HPCI steam isolation valve, 23MOV-
60. The inspector reviewed the licensee's root cause analysis for the event,
conducted interviews, attended management meetings on the event, and reviewed
station procedures to assess the event regarding safety significance and work
control processes.
b. Observations and Findinas
On October 24,1C37, maintenance activities to repair a ground problem were
conducted which rendered the primary containment isolation system (PCIS)
function of the outboard HPCI steam isolation valves inoperable, however, the
applicable LCO action statement was not entered. If one or more of the
containment isolation valves are inoperable. Technical Specifications require, in
part, that the affected penetration be isolated within four hours by use of at least
one deactivated automatic valve secured in the closed position. Operators did not
recognize that PCIS was disabled until after a maintenance error caused a short of
the logic circuitry which caused an invalid engineered safeguards feature (ESF)
actuation signal sixteen hours after disabling the logic.
The root cause analysis identified severai nppropriate actions. These included
the failure to recognize the impact on the PCJ function of the HPCI isolation
valves when removing logic fuses during surveillance test ST 2M, ECCS Trip
Systems Bus Power Monitors Functional Test, disabling the same PCIS function
during trouble shooting without entering the applicable LCO, and failing to enter
the correct LCO when the condition was recognized. Several causes were
identified by the licensee for the inappropriate actions identified above.
Surveillance test ST-2M was inadequate, in that it did not recognize disabling the
PCIS function of the HPCI valves, a less than adequate review of the short form
temporary operating procedure and protective tag out, and inadequate training on
a previous technical specification change which resulted in operators using the
incorrect section of the TS. The licensee developed twelve recommended
corrective actions, including revising procedures to capture the lessons learned,
training and review of the event with operators, and review of all surveillance test
and operating procedures to identify the impact of fuse removal on TS.
The inspector also reviewed the troubleshooting work request which led to the
event and determined that the impact on PCIS was also missed during the work
control process. The inspector noted that this issue was not addressed in the root
csuse analysis. The licensee subsequently reviewed the work control process and
determined that the troubleshooting process for this emergent work item relied on
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l the workers to determine the affects of their actions in the field during work
l execution. Had a detailed review of the logic by the workers and the operations
staff been conducted, the potentialimpact of the work could have been identified.
The mind set of the plant staff was that the system was in an existing LCO and
tagged out, therefore work would not impact the plant. This was an incorrect
assumption as identified when a technician inadvertently shorted two terminals in
a junction bux. The issues surrounding the performance of the troubleshooting
were discussed in Licensee Event Report 97-011 and will be addressed when the
LER is reviewed.
c. Conclusions
The process to control work activities associated with troubleshooting to locate a
DC ground was unsatisfactory and resulted in an invalid ESF actuation. Operators
did not recognize that the troubleshooting activities made the PCIS function
inoperable and therefore did not enter the appropriate LCO action statement. The
licensee's immediate corrective actions were appropriate and the root cause
analysis was critical of the operations staff's handling of the trouble shooting
activities but lacked in-depth review of the work control process for the activity.
The failure to enter the LCO was determined to be a violation (50-333/97008 02).
M4.2 Low Pressure Coolant Injection (LPCI) Battery Replacement
a. insoection Scone
The inspector observed preparations for the "B" LPCI battery replacement in the
reactor building. The mechanics were trying to remove several sections of metal
panels that make up one of the walls to "B" LPCI battery enclosure, in discussion -
with the maintenance personnel the inspector learned that the wall was much
more intricate than the maintenance crew had expected. The responsible engineer
was notified and after further discussion the licensee determined that the work
should not be continued. A horizontal top corner piece of the structure had been
removed to allow access to the vertical wall sections, but no other pieces were
removed. The inspector reviewed the licensee's work planning and discussed the
activity with the licensee personnel,
b. Observations and Findinas
Work request (WR) 96-05333-07,was written to remove panels from the west
wall of the "B" LPCI enclosure, to facilitate the installation of a temporary load
handling monorail. The monorail was to be used to replace the existing LPCI
battery cells with new cells during the upcoming scheduled LCO maintene -
period. In follow up interviews with the plant staff the inspector learned that the
original maintenance package did not consider potential fire protection and seismic
issues associated with the removal of various battery enclosure panels, in
discussion with the planning staff the inspector discovered that the enclosure
drawings were not reviewed as part of the work package planning which
contributed to a leck of detailin the work package. The licensee initiated a
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deficiency event report to investigate the issue and utilized another method to
exchange the battery nlis. The licensee's investigation concluded that the battery
enclosure was not necessary for LPCI battery operability. The battery replacement
work package was weak in that the impact of the work activity was not assessed.
Additionally, the work control procedure did not include a requirement to include
all structures, systems and components (SSCs) when reviewing work for seismic
concerns. The corrective actions were appropriate for the above findings,
c. Conclusions
The inspector concluded that the work package to prepare to replace the LPCI
battery was weak in that the impact of the work on the LPCI battery operability
was not considered prior to beginning the work. Although the work was stopped,
the licensee subsequently determined that the work would not impact battery
operability. Additionally, plant drawings for the structure were not reviewed prior
to the work being performed which contributed to confusion in performing the
tank.
Ill. ENGINEERING
E1 Conduct of Engineering
E1.1 . Environmental Qualification of components in the High Pressure Coolant Injection -
System (Violation 50-333/97008-03)
a. Insoection Scoce
On October 24,'1997, while performing troubleshooting for a DC ground, a nut
was dropped across test Jacks located in a junction box. The resulting short
- > caused a HPCl isolation signal. The identification of electrical test jacks on
junction boxes for HPCI and RCIC isolation circuits raised questions concerning the
operability and environmental qualification (EO) of the associated components.
The inspector reviewed the licensee's EQ program calculations, justification for
continued operation (JCO) and conducted a physical walkdown of the affected
areas,
b. Observations and Findinas
On October 24,1997, during troubleshooting on a pressure switch for the source
of a DC ground, a nut was dropped across two hot test points in a junction box,
located in the west crescent area, which initiated a HPCI isolation trip signal.
Fuses pulled for the troubleshocting prevented the actual system isolation. The
inspector noted that the junction box was marked as EQ, however, the pressure
switches locatrd in the junction box had been removed from the EQ program.
Test jacks were also located in the bottom of three additional junction boxes and
were not identified on plant drawings. The concern was that the test Jacks may
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potential existed to impact the HPCI steam line isolation function. Similar test
Jacks were located in junction boxes associated with RCIC,
At the time of the event, the licensee did not initially recognize the need to
determine the operability of the affected components. Subsequently, an
operability review for HPCI and RCIC was completed on November 4,1997, and
c licensee
a prepared a JCO, JAF EQ JCO-97-002, Plant Operation with Test
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Jacks installed in Junction Boxes JB-R2550D snd JB-R2550E for 23 PS-86A,B,C
ano D to justify continued operation with the test Jacks installed. The licensee's
operability review determined that the test jacks did not affect the operability of
the circuits.
The licensee reviewed the EQ status of the associated junction boxes. It was
determined that on March 3,1993, the licensee deleted approximately
15 c 'ponents from the EQ program for harsh environment plant electrical
equit w snt. The analysis was documented in JAF CALC-HPCI-00820 and was
prepared to show that HPCI electrical components would not be subject to a harsh
environment during a HELB. The licensee determined that a nonconservative
asrumption was made in thu calculation which resulted in removing the HPCI
components from the EQ program.
The licensee's corrective actions included walkdowns to identify any other similar,
test Jacks that posed EQ issues. The results of the wa!kdown determined the
extent of the condition was limited to HPCI and RCIC. Additionally, the licensee
removed the electrical connections to the HPCI and RCIC test lugs under a plant
modification. A longer term action review other components removed from the EQ
program was in progress.
. The inspector noted a station work practice where technicians occasionally used
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junction boxes to temporarily store various objects while working on components.
Typically, this practice was used when technicians ware working on grates where
there was not a readily available place to temporarily store small tools or
components, The licensee reviewed their practices in this area and determined
that this practice would no longer be used,
c. Conclusions
Following the initial event, the licensee was slow to pulsue the JCO because the
EQ aspects were not readily recognized. The EQ components were erroneously
removed from the program in 1993, and in fact, did not orienally meet EQ criteria
because of the unrecognized installed test jacks. A JCO was prepared which
provided reasonable assurance that the equipment would perform its safety
function. The EQ issue was appropriately resolved through removing the
connection to the test jacks and inserting the previously removed components into
the scope of the EQ program.10 CFR 50.49, Environmental Qualification of
Electric Equipment important to Safety for Nuclear Power Plants, describes EQ
program requirements. Contrary to these requirements, the licensee erroneously
,
removed HPCI components from the EQ program (VIO 50-333/97008-03).
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E2 Engineering Support of Facilities and Equipment
E2.1 Licensee Monitoring of Leaking Safety Relief Valves (SRVs)
a. Insoection Scoce
Because of a history of safety relief valve leakage and an industry event where an
SRV inadvertently opened, the licensee monitors SRV leakage. The inspectors
reviewed the licensee's SRV monitoring program and discussed the issue with
licensee personnel. Total SRV leakage recently increased to a point where the
licensee elected to shutdown and repair the leaking SRVs as described in Section
01.2.
b. Observations and Pndinos
The licensee use: 11 Target Rock 2 wage pilot actuated safety relief valves for
pressure relief of the reactor vessel. The licensee monitors SRV tailpipe
temperature and calculates leakage based on torus heat up rate. The licensee had
previously developed an action plan to schedule a plant shutdown at a torus .
heatup rate corresponding to an SRV leakage rate of 400 lbs/hr and to shutdown
the plant at a torus heatup rate corresponding to 600 lbs/hr. As of November 21, ,
the licensee was operating with indication of 3 leaking SRVs and a leak rate of
450 lbm/hr with most leakage attributed to "C" SRV main seat leakage.
The inspectors monitored the licensee's performance related to SRV leakage. The
licensee closely tracked SRV performance through daily torus heat up rate 4
calculations and observations of SRV tailpipe temperature. In addition, the
inspectors noted that SRV performance is routinely scheduled for discussion at the
department manager's meetings.
c. Conclusiqng
The licensee's program to monitor SRV leakage was effective. Licensee
management exercised good judgement in electing to shutdowr the plant on
December 7th to effect repairs to leaking SRVs.
E8 Miscellaneous Engineering issues
E8.1 (Closed) inspector Follow up item (IFI) 50-333/96007-02: Affect of reactor water
cleanup and contro! rod drive flow on alternt decay heat removal (ADHR) pre-
operational testing. During refueling outage 12, the inspectors noted that the
control rod drive (CRD) and reactor water clean-up (RWCU) sys.tems were in
service providing approximately 240 gallons per minute flow to the reactor vessel
and providing additional refueling cavity mixing during the pre-operational testing
of the ADHR system. This was of concern to the inspectors because the intent of
the pre-operational testing was to ensure that the alternate deca / heat removal
system was capable of removing the heat generated by the spent fuelin both the
reactor vessel and in the spent fuel pool. The test was to demonstrate that the
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ADHR system would remove heat from the reactor cavity and the spent fuel pool
using natural circulation. The inspector's concern was that the added circulation
and cooling water may result in non conservative results with respect to the
capabilities of the system. Additionally, the original calculations for the heat
removal capacity did not account for the additional circulation provided by the
CRD and RWCU systems, which were in service during the test. The licensee
subsequently revised the computer model used to independently verify the General
Electric design calculations on the ADHR system, JAF cal.C DHR 02380, ADHR
System Thermal Hydraulic Analysis, and incorporated the CRD and RWCU system
flows. The revised calculation demonctrated that there was no significant change
in the natural circulation flow characteristics of the reactor vessel and the surf ace
water temperature of the SFP differed from the original calculation by one degree
Fahrenheit, which was considered to not have a materialImpact.
E8.0 (Closed) Unresolved item (URI) 50 333/05006 03: Components Missing from
Control Room Ventilation Drawings. As part of a system walkdown, the inspector
identified that two motor operated dampers (MODS), MOD 113 and MOD 114, and
flow element (FE) 102 were omitted from as built drawing FB 35C, Rev.12,
Equipment Room Heating, Vent and Air Conditioning. Hcwever, the inspector
noted that the components were identified on the cont:ol rocin flow diagram FB-
45A and identified in the Final Safety Analysis Report (FSAR), The inspector was
concerned that despite a temporary modificaHon ar 1 a minor modification being
processed for two safety relat6d components, this deficiency in the as built
, drawing was not identified by the licensee. The licensee subsequently determined
that the error occurred during original construction and that the drawing was not
required to be updated because of its classification. The System Engineering
Standing Order, (SESOF2, classified the drawing as a type "C" drawing, and per
Design Change Manual, (DCM) 22, drawing changes are not required until five
changes have been posted to the drawing or with the department man %er's
approval. The classification of the drawing signifies that it is used to facilitate
design and maintenance that has a low freauency use. The current practice is for
users to verify the current status of changes to drawings in the drawing status log
prior to use. In the review of the event, however, the nicensee determined that the
modification process should have listed the drawing as an "affected drawing" so
that the appropriate revision process would have been implemented. The
modification removed the motors from the dampers and put them in a f ail safo
position and they currently do not provide any safety function. The inspector
reviewed the corrective actions including revision of the appropriate drawings,
updating the plant equipment database and training on the issue. The inspector
determined the corrective actions were appropriate and that a violation did not
exist because the licensee practices were in accordance with their procedural
requirements for drawing update and were appropriate to the circumstances.
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IV. Plant Suonort
R1 Radiological Protection and Chemistry (RP&C) Controls
R1.1 Implementation of the Solid Radioactive Waste Program
a. Insoection Scone
The inspector reviewed the licensee's solid radioactive waste program.
-Information was gathered through observation of activities, tours of the
radiologically controlled areas including the radioactive waste building, discussions
with cognizant personnel, and review and evaluation of procedures and
documents,
b. Observations and Findinas
'In reviewing the implementation of the solid radioactive waste program, the liquid
wasta processing and solidification methods were inspected, the dry active waste
operation was evaluated, and storage locations were inspected. Liquid radioactive
waste was processed through a vendor filter /demineralizer skid. There was a
" green is clean" program, and potentially contaminated materials were sorted and '
frisked to minimize the generation of radioactive waste. Offsite contracted
services were available for equipment / parts decontamination and for
supercompaction or incineration of dry active waste. The interim waste storage
building provided five high bay areas for storage of low level waste and a separate
area with shielded concrete cells for storage of higher level waste such as high
integrity containers (HICs) filled with spent powder or resin. A large portion of this'
stcrage capacity was still avsilable.
.The volume of solid radioactive waste and especially of dry active waste had
steadily and significantly decreased over the last several years. Numerous
initiatives to reduce waste were evident. This included the establishment of a low-
level radioactive waste reduction team. Reusable wrist and ankle straps in place
of masking tape for protective clothing purposes, reusable bags, the wearing of
i cotton glove liners into the whole body contamination monitors, and the
evaluation / implementation of good radioactive waste reduction practices from
other licensees and from a utility research group have contributed to the decrease
in radioactive waste volume.
Housekeeping was good; aisle ways were clear and clean; storage areas were
clean and orderly; contaminated areas were minimized; radioactive material was
clearly and properly labeled and stored in an orderly fashion.
c. Cgnplusions
The implementation of the solid radioactive waste program was well managed and
effective.
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R1.2 Compliance with NRC and Deptatment of Transportation (DOT) Regulations for
Shipping of Low Level Radioact!ve Waste (LLRW) for Disposal and Transportation
of Other Radioactive Materials
a. Insoection Scona
The inspector reviewed the licensee's transportation of radioactive materials.
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Information was gathered through observation of activities, discussions with
cognizant personnel and review and evaluation of procedures and documents.
Temporary Instruction (TI) 2515/133,lmplementation of Revised 49 CFR Parts
100179 and 10 CFR 71 was completed during this review,
b. Observations and Findings
The shipping records for several past radioactive waste shipments were reviewed.
There records were found to be appropriate and complete The Inspector
observed the transfer of a HIC containing spent resln into a shipping cask. The
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waste classification and Department of Transportation (DOT) shipment type
determination for this shipment were evaluated and met regulatory requirements.
After this waste shipment was on the public highway, the inspector tested the
emergency response information telephone process (10 CFR 49, Subpart G,
Emergency Response Information) by calling, in the evening, the emergency
response telephone number which was on the radioactive waste manifest. The
inspector's call was answered, and the emergency response information described
in 10 CFR 49.602 was made available by the recipient of the callin a timely
manner,
c. Conclusions
Good perfolmance was demonstrated in the area of packaging and transportation
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of solid radioactive waste.
RI .3 Elevated Radiation Exposure Lavels Due Hydrogen injection
a. inspection Scone
The inspector reviewed the licensee's radiological controls in administrative areas
relative to the elevated radiation levels due to hydrogen injection into the reactor
coolant system, ha.foimation was gathered through observation of a radiation
survey, conduct of a radiation survey, tours of the affected locations, discussions
with cognizant,pers.onnel, and review and evaluation of procedures and
documents,
b. Observations and Findinga
Elevated dose rates outside the radiologically controlled area (RCA) due to
hydrogen injection, and the radiological controls in those areas were reviewed.
This review focused on the second floor of the old administration building and on
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, the second floor of warehouse No.1 since these areas were representative of the
affected areas which were outside the RCA, within the protected area, and which
were occupied by mainly administrative personnel. These locations were included
in the licenses's routine radiation survey program.
Dose rates on elevation 290 of warehouse No.1.'arled from 10 to 70 microrem
per hour with reactor power level at 100% based on the licensee's routine
radiation survey results. Dose rates on elevation 286 of the old administrative
building varied from 10 to 150 microrem per hour with reactor power level at
100% based on the licensee's routine radiation survey results and on an
independent survey by the inspector using a calibrated licensee radiation survey
meter. At the time of these surveys, the hydrogen injection rate was
approximately 18.5 cubic feet per minute. The highest dose rates in the latter
area were along the outside window areas, and the dose rates gradually decreased
with distance away from these window areas. ,
Current occupancy factors were observed for each of the two areas and appeared
to approximate 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per wesk. Assuming a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> work week, continuous
occupancy during the work week, and 50 work v.'eeks per year,150 and
10 microrem per hour would equate to 300 and 20 millirem per year, respectively.
For perspective, the average background radiation level in the United States is
about 10 microrem per hour or approximately 100 millirem per year (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per
day and 365 days per year).
10 CFR 19.12, " Instruction to Workers," requires that all individuals who in the
course of their employment are likely to receive in a year an occupational dose in
excess of 100 millirem shall receive radiation protection instruction commensurate
, with potential radiological health protection problems present in the workplace.
Radiation safety training records for several white badged individuals (non-
radiation workers)in each of these areas were inspected. The individuals were
confirmed to have received general employee training, which includes basic
radiation protection training on an annual basis. The inspector confirmed that the
training included discussion of the guidance in Regulatory Guide 8.13, " Instruction
Concerning Prenatal Radiation Exposure."
10 CFR 20.1502 requires the use of individual radiation monitoring devices for
, adults likely to receive a dose in excess of a total effective dose equivalent of
500 millirem per year; and for the embryo / fetus of a declared pregnant women for
whom the embryo / fetus is likely to receive a dose in excess of 50 millirem during
the entire pregnancy. Inspector observations noted that a number of the
administrative (non-radiation worker) personnel had been provided individual
radiation monitoring devices even though this was not required by 10 CFR
20.1502. In some of these administrative areas, one would be likely to receive a -
dose in excess of 50 millirem within a nine month period, and, in such a case,
10 CFR 1502 would require the use of a individual radiation monitoring device for
the embryo / fetus of a declared pregnant woman. The inspector confirmed that a
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declared pregnant woman had been provided such dosimetry.
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The inspector also reviewed licensee documents titled * Radiological Technical
Information Document (RTID) No. 93 011, Basis for Individual Monitoring
Requirements Within the Restricted Area of the JAFNPP, December 13,1993,"
"First Quarter 1997 Restricted Area Dose Evaluation, April 23,1997," "Second
Quarter 1997 Restricted Area Dose Evaluation, July 18,1997," " Third Quarter
1997 Restricted Area Dose Evaluation, November 3,1997," and "RTID 96-005,
Radiological Assessment of Doses / Dose Rates in Non RCA Occupied Areas,
June 19,1996."
Based on monitoring and survey data by the licensee and the NRC, observations in
several affected areas, and review of training and documents describing the
licensee's evaluation of this issue, the inspector concluded that the radiation
badging requirements in 10 CFR 20.1502 and the training requirements in 10 CFR
19.12 were being followed,
c. Conclusions
Radiological controls in administrative areas relative to the elevated radiation levels
due to hydrogen injection were proper and adequate.
R5 Staff Trainin9 and Qualification in RP&C (Inspector Follow up ltem (IFI) 50 333/97-
008 04)
a. Insoection ScRD.t
The inspector reviewed the qualifications and training of selected radioactive
waste personnel. Information was gathered through discussions with cognizant- -
personnel, and review and evaluation of documents,
b. Observations and Findinas
Training department personnel stated that the training for radioactive waste-
processing, handling / transferring, packaging, and shipping was provided by
contractors and that the training courses were reviewed and approved by licensee
personnel before implementation. The inspector reviewed the course materials
used for the training of the radioactive waste handlers and shippers. The scope
and depth of the course materials was fully adequate. The inspector verified that
these individuals had been recently trained in the aforementioned topics and that
the two individuals who were responsible for classifying waste and determining
DOT chipment type had been retrained on the applicable computer program in mid-
1997. Additionally, it was confirmed that allindividuals authorized to sign
shipping paperwork had received recent training on the shipping reguistions.
However, a documented description of the required training for radioactive waste
processors, handlers /transferors, classifiers, and shippers was not available, and
the training record database was incomplete in that the latest training for the
computer program used for classifying and typing waste shipments had not been
entered. .These administrative deficiencies were considered a program weakness.
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The licensee stated that a matrix of required training and the frequency of same
for radioactive waste processors, handlers /transferors, classifiers, and shippers
would be developed and kept available for review and that the training record
database would be updated and kept updated. This issue will be reviewed during
a subsequent inspection (IFl 50 333/97008-04).
c. Conclusiong
The training and retraining for the radioactive waste handlers /transferors and
shippers was appropriate in scope and depth, and records of this training were
adequately maintained. However, a documented description of required training
was not available, and the training record database was incomplete. Therefore,
the training program was not well organized and documented.
R7 Quality Assurance in RP&C Activitles
a. Insoection Scoce
The inspector reviewed the licensee's quality assurance (OA) activities for solid
radioactive waste management and transportation of radioactive materials.
Information was gathered through discucslons with cognizant personnel and
review and evaluation of documents,
b. Qharvations and Findinas
Audit A9617J, conducted in the Fall of 1996, covered the implementation of the
revised DOT and NRC radioactive material shipping requirements. This audit
resulted in two Deviation and Event Reports (DEPis) and two Recommendations. -
The inspector's review of the audit checklist and audit report showed that the
audit was thorough and programmatic.
Audit A97 05J, conducted in February of 1997, covered the Process Control
Program (PCP) and Regulatory Guide 1.21. This audit resulted in the issuance of
two DERs and one Recommendationin the PCP area. The audit checklist and
audit report portions dealing with the PCP were reviewed and were found to be in-
depth efforts.
Six QA surveillance reports, performed from November 1996 to September 1997,'
were evaluated. These reports covered receipt inspections of radioactive waste
shipping casks and liners, shipment inspections of radioactive waste in casks,
review of documentation packages for several waste shipments, and the release of
material from the radiologically controlled area. There were no resultant DERs or
recommendations based on these reports. The surveillance reports showed that
the surveillance activities were detailed and well documented.
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c. Conclusions
The cuality Assurance audits and surveillance reports were thorough,
programmatic, and well documented.
R8 Miscellaneous RP&C lasues
R8.1 (Closed) Violation 50 333/96007 08: Failure to follow plant Technical
Specification for locked high radiation area entry (i.e., contractor in the drywell
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with his alarming dosimeter turned off). The inspector reviewed the corrective
actions described in the licensee's response letter dated February 21,1997. The
corrective actions were reasonable and comprehensive. No similar problems were
identified.
R8.2 (Closed) Violation 50 333/96007-09: Failure to follow a formal quality assurance
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program (i.e., failure to promptly identify and correct a deviation involving the lack
of current certification of technicians for use of a computer program code used to
classify shipments). The inspector reviewed the corrective actions described in
the licensee's response letter dated February 21,1997. The corrective actions
were appropriate and complete. No similar problems were identified.
P1 Conduct of EP Activities
P1.1 Emergency Plan Drill
a. Insoection Scoce
On December 11,1997, an emergency plan joint drill was conducted with the
licensee and Nine Mile Point participating. The purpose of the drill was to
demonstrate that various emergency preparedness functions could be performed
jointly from the emergency operations f acility (EOF). The drill was a partial scale
drill and had limited participation by Oswego County.
The inspector observed and evaluated the performance of licensee emergency
response personnelin the EOF including staffing and activation; facility
management and control; accident assessment and classification; offsite dose
assessment; protective action decision making and implementation; notifications
and communications; and interaction with the Oswego County personnel,
b. Observations and Findinos
The emergency was properly classified. The reactor condition and emergency was
continuously reassessed._ Environmental sampling teams were appropriately
deployed. Offsite dose assessment and protective action recommendations were
appropriate. Communications within the Emergency Operations Facility were
frequent with proper notifications and interaction with county personnel noted. A
particular strength noted was the good coordination between the emergency
directors from the Nine Mile Point and FitzPatrick f acilities in setting priorities.
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c. Conclusions
The emergency preparedness drill dernonstrated solid performance of the EP staff
and licensee organization.
P8 Miscellaneous EP lssues (EA 98-008)(NCV 97 008 05)
(Closed) IFl 50 333/97002 03(NCV 97 008 05): Adequacy of emergency
procedures governing evacuation from areas near the new fuel storage vault and
f ailure to meet requirements of 10 CFR 70.24 for new fuel criticality monitors.
This issue involved the f ailure to have in place either an adequate criticality
monitoring system for storage and handling of new (non irradiated) fuel or an NRC
approved exemption to this requirement contained in 10 CFR 70.24. The issue
was previously left as an inspector follow up item pending additionalinternal NRC
guidance regarding the adequacy of the existing monitoring system as well as the
.
emergency procedures governing evacuation from areas near the new fuel storage
vault.
10 CFR 70.24 requires that each licensee authorized to possess more than a small
amount of special nuclear material (SNM) maintain in each area in which such
materielis handled, used or stored a criticality monitorin0 system which will
energize clearly audible alarm signals of accidental criticality occurs. The purpose
of 10 CFR 70.24 is to ensure that, if a criticality were to occur during the handling
of SNM, personnel would be alerted to that f act and would take appropriate
action.
Most nuclear power plant licensees were granted exemptions from 10 CFR 70.24
during the ' construction of their plants as part of the Part 70 license issued to
permit the receipt of the initial core. Generally, these exemptions were not
explicitly renewed when the Part 50 operating license was issued, which
contained the combined Part 50 and Part 70 authority, in August 1981,the
Tennessee Valley Authority (TVA),in the course of reviewing the operating
licenses for its Browns Ferry facilities, noted that the exemption to 10CFR 70.24
that had been granted during the construction phase had not been explicitly
granted in the operating license. By letters dated August 11,1981,and
August 31,1987, TVA requested an exemption from 10 CFR 70.24. On May 11,
1988, NRC informed TVA that the previously issued exemptions are still in effect
even though the specific provisions of the Part 70 licenses were not incorporated
into the Part 50 license." Notwithstanding the correspondence with TVA, the
NRC has determined that, in cases where a licensee received the exception as part
of the Part 70 licenses issued during the construction phase, both the Part 70 and
Part 50 licenses would be examined to determine the status of the exemption.
The NRC view now is that unless a licensee's licensing b. sis specified otherwise,
an exemption expires with the expiration of the Part 70 license. The NRC intends
to amend 10 CFR 70.24 to provide for administrative contcols in lieu of criticality
monitors.
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The NRC has concluded that a violation of 10CFR 70.24 existed at FitzPatrick due
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to the inadequacy of the existing new fuel vault radiation monitor as a
comprehensive criticality monitoring system for new fuel handling and storage.
The NRC has also determined that numerous other licensees have slmilar
circumstances that were caused by confusion regarding the continuation of an
exemption to 10 CFR 70.24 originally issued prior to issuance of the Part 50
license. After considering all the factors that resulted in these violations, the NRC
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has concluded that while a violation did exist, it is appropriate to exercise
enforcement discretion of violations involving Special Circumstances in accordance
with Section Vil B.6 of the " General Statement of Polley and Procedures for NRC -
Entorcement Actions" (Enforcement Policy), NUREG 1600. Pending amendment to
10 CFR 70.24, further enforcement action will not be taken for f ailure to meet '
l 10 CFR 70.24 provided an exemption to this regulation is obtained by NYPA
before the next receipt of fresh fuel or before the next planned movement of fresh
fuel at FitzPatrick. This item is tracked as non cited violation (NCV 97 008-05).
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l V. MANAGEMENT MEETINGS
X1 Exit Meeting Summary
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The inspectors presented the inspections results to members of the licensee
management at the conclusion of the inspection on January 13,1998. The
licensee acknowledged the findings presented.
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The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary. No proprietary information was
identified.
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ATTACHMENT 1
PARTIAL LIST OF PERSONS CONTACTED
Licensee
G. Brownell, Licensing Engineer
M. Colomb, Site Executive Officer
D. Lindsey, General Manager, Operations
J. Maurer, General Manager, Support Services
A. McKeen, Radiologicel and Environmental Services Manager
T. Phelps, Radiological Supervisor
D. Ruddy, Director, Design Engineering
J. Solini, Sr. QA Engineer
D. Topley, General Manager, Maintenance
A. Zaremba, Licensing Manager
INSPECTION PROCEDURES USED 1
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37551 Onsite Engineering
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62707 Maintenance Observations
61726 Surveillance Observations
. 71707 Plant Operations
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71750 Plant Support
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83724 External Occupational Exposure Control and Personal Dosimetry
86750 Solid Radioactive Waste Management and Transportation of Radioactive Materials
92702 Follow up on Corrective Actions for Violations and Deviations
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Tl 2515/133 Implementation of Revised 49 CFR Parts 100179 and 10 CFR 71
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Attachment 1 2
ITEMS OPENED, CLOSED, AND DISCUSSED
Onened
50 333/97008-01 VIO Improper performance of DC ground abnormal operating
procedure resulted in HPCIlogic actuation.
50 333/97008 02 VIO Failure to enter Technical Specification Limiting Condition for
Operation while troubleshooting electrical grounds.
50 333/97008 03 VIO Erroneously removal of HPCI components from the
10CFR50.49 environmental qualification program.
50 333/97008-04 IF! Radioactive waste training pro 0 ram was not well organized and
documented
50 333/97008 05 NCV Failure to meet 10 CFR 70.24 requirements or to obtain a valid
exemption from this regulation, i
G91td
50 333/95006-03 URI Components missing from control room ventilation drawings
50 333/96007 02 IFl Affect of RWCU and CRD flow on alternate decay heat
removal preoperational testing
50 333/97002-03 IFl Adequacy of emergency procedures governing evacuation from
areas near the new fuel storage vault and f ailure to meet
requirements of 10 CFR 70.24 for new fuel criticality monitors
50-333/96007 08 VIO Failure to follow plant Technical Specification for locked high
!
radiation area entry
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50-333/97008 05 NCV Failure to meet 10 CFR 70.24 requirements or to obtain a valid
exemption from this regulation.
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l EA 98-008 NCV Adequacy of emeroency procedures governing evacuation
from areas near the new fuel storage vault and f ailure to meet
i requirements of 10 CFR 70.24 for new fuel criticality monitors.
l 50 333/96007-09 VIO Failure to follow a formal quality assurance program
Discussed
None
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Attachment 1 3
LIST OF ACRONYMS USED
ADHR Alternate Decay Heat Removal
AOP Abnormal Operating Procedure
CFR Code of Federal Regulations
CRD Control Rod Drive
CST Condensate Storage Tank
DC Direct Current
DCM Design Change Manual
DER Deficiency & Event Report
DOT Department of Transportation
EDG Emergency Diesel Generator
EOF Emergency Operations Facility
EQ Environmental Qualification
ESF Engineered Safety Feature
FE Flow Element
FR Federal Register
HIC High Integrity Container
HPCI High Pressure Coolant Injection
IFl Inspection Follow up item
IR inspection Report
LCO Limiting Condition for Operation
LER Licensee Event Report
LLRW Low Level Radioactive Waste
LPCI Low Pressure Coolant injection
MOV Motor Operated Valve
MP Maintenance Procedure
NCV Non-Cited Violation
NRC Nuclear Regulatory Commission
OP Operating Procedure
PCIS Primary Containtnent isolation System
QA Quality Assurance
QC Quality Control
RAP Reactor Analyst Procedure
RCA Radiological Controlled Area
RP&C Radiological Protection and Chemistry
RTID Radiological Technical Information Document
RWCU Reactor Water Clean-Up
SESO System Engineer Standing Order
SRC Safety Review Committee
SSC Structures, Systems & Components
Tl Temporary instruction
TS Technical Specification
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Attachment 1 4
-TVA Tennessen Valley Authority
UFSAR Updated Final Safety Analysis Report
VIO Violation
WR Work Request
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