ML20196J669
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| ML20196J669 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/30/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20196J610 | List: |
| References | |
| 50-443-97-03, 50-443-97-3, NUDOCS 9708050026 | |
| Download: ML20196J669 (57) | |
See also: IR 05000443/1997003
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Enclosure 2
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION I ;
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Docket No.: 50-443
License No.: NPF-86 j
Report No.. 50-443/97-03
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Licensee: North Atlantic Energy Service Corporation
Facility: Seabrook Generating Station, Unit 1
Location: Post Office Box 300
l Seabrook, New Hampshire 03874 ;
Dates: April 15,1997 - June 15,1997 l
, inspectors: William T. Olsen, Senior Resident inspector (Acting) l Cavid M. Silk, Resident Npector ! Alfred Lohmeier, Reactor Engineer, DRS
John R. McFadden, Health Physicist, DRS j
Accompanied by: Javier Brand, Resident inspector intern
Approved by: Richard Conte, Chief, Reactor Projects Branch No. 8
Division of Reactor Projects
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9708050026 970730
PDR. ADOCK 05000443
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EXECUTIVE SUMMARY
NRC Inspection Report No. 50-443/97-03 j
This integrated inspection included aspects of licensee operations, engineering, ;
maintenance, and plant support. The report covers an 8-week period of resident inspection l
j from April 15 through June 15,1997. l
Ooerations:
A preventable reactor trip occurred during the manual shutdown of the unit to enter the
fifth refueling outage. Plant indications available to the operators were overlooked.
Although the safety consequences of the event were low, it did subject the plant to a
unnecessary transient and challenge to the operators. The root cause analysis repcrt for
this event was very thorough, however, it indicated an overall lack of appreciation by the
staff for the impact of abnormal conditions on plant response. The inspector concluded
that staff performance leading to the trip indicated a weak assessment of the impact of the
IR detector current anomaly on plant operations (Section O2.1).
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Steam Generators (SG) "B" and "C" were inadvertently drained down during feedwater l
isolation valve stroke testing due to drain valves being improperly left in the open position. l
In addition to the system configuration control, operations personnel initially only assessed l
that "C" SG had drained down and failed to investigate the "B" SG low level indication
even after being questioned by the inspectors (Section 02.2).
i Fuel handling activities were conducted according to procedure in a well controlled manner l as the entire core was off loaded and subsequently reloaded. Reactor engineering l personnel charted assembly positions during fuel movement and monitored for an approach l to criticality by developing a 1/M plot during core reloading (Section 2.3).
Midloop activities were performed well and appropriately supervised. Safety considerations
l were manifested as activities which could have impacted the reactor coolant system or
midloop operations were restricted (Section 2.4).
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The licensee properly identified that the WRGM was inoperable and properly entered the
Technical Specification action statement. The increased sampling frequency was very
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conservative with an excellent safety perspective in evidence and the addition of the
portable sampling equipment provided excellent backup information. The failed power
supply was promptly repaired and returned to service. The inspector noted strong
i management involvement in the repair process to ensure that the instrument was promptly
returned to service (Section O2.5).
Maintenance:
l The Ten-Year In-service Inspection (ISI) Program status is within the targeted schedule and I consistent with the performance requirements for non-destructive examinations of Code '
Class 1,2, and 3 components and their supports, and is consistent with the augmented ISI
program indicated in the Updated Final Safety Analysis Report, Section 6.6.8. The
program plan was well prepared, documented, and implemented (Section M1.1).
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Executive Summary
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ISI examir5tions reviewed were performed in accordance with Code and NRC regulations
(Section M1.2).
l l The steam generator tube eddy current inspection program was conservatively planned,
and consistent with ASME Section XI, Reg. Guide 1.83, Plant Technical Specifica*;ons,
! EPRI Examination Guidelines, and licensee responses to Generic Letter 95-03 (Section ! M1.3). ,
Preparation for examination of the pressurizer head-to-shell weld showed good utilization of
technical resources. Weakness was shown in allowing the utilization of nominal drawing
dimensions, instead of as-built dimensions, to fabricate weld inspection qualification
models. This error was self-discovered und steps were taken to provide a satisfactory
l alternate inspection procedure (Section M1.4). l
A poor work practice was identified during work on the encapsulation vessel for the -i
recirculation sump. The inspector observed workers tossing safety-related flange bolts into ;
a bucket which could have damaged the bolts (Section M1.5).
- 'The foreign material exclusion program at Seabrook is excellent. Considerable manpower
and resources have been dedicated for this effort (Section M2.1). l
On May 27, tubing installed for a temporary pressure transmitter on the charging system
failed resulting in a radioactive spill of about 30 gallons because of under-rated tubing.
The inspectors determined within the past 18 months four tubing ruptures had occurred
because of under-rated tubing. This recent failure demonstrates inadequate corrective. )
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action and management oversight to prevent recurrence and is a violation of NRC )
requirements (Section M2.3).
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The licensee properly performed a significant plant modification by replacing the primary
l component cooling water heat exchangers. The modification was made to improve system '
reliability by increasing heat exchanger capacity resistance to corrosion. The 10 CFR
50.59 evaluation was thorough and properly documented. The work was well controlled
and system engineers provided excellent guidance to the workers (Section E1.1).
I' While reviewing activities and documentation related to operability verification testing for j
the main steam safety valves, the inspector determined that no 10 CFR 50.59 evaluation '
was performed for the new equipment used for the test. Although there were no safety
consequences resulting from the use of the new equipment as determined by bounding
UFSAR accident analysis, the inspector concluded that a potential weakness may exist in
the licensee's 50.59 process as no clear direction exists as to when an evaluation should
be performed (Section E2.1),
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Executive Summary
Evaluation of the failed leak rate test for containment spray valve CBS-V18 and subsequent ,
corrective actions to machine the valve disc and disc hanger to prevent wedging of the
valve disc, were adequate (Section E2.2).
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The licensee response to and assessment of the failed fuel rods was very good. Once the
location of the failed rods was identified, the licensee evaluated the commonalities,
developed probable causes, and implemented corrective actions to prevent recurrence
(Section E2.3).
Plant Suooort: j
The licensee maintained and implemented strong radioactive liquid and gaseous effluent
control programs, with capabilities to protect the public health and safety and the
environment. The management's commitments and support to the programs were noted. j
The licensee demonstrated very good effluent ALARA practices for the programs. The i
chemistry steff responded to QA audit findings and observations in a timely manner and
with sound technical bases (Section R1.1).
l '-The licensee continued to implement en overall effective Radiological Environmental -
Monitoring Program (REMP) including management controls, quality assurance audits,
- radiological environmental monitoring, and meteorological monitoring program. The Offsite
- Dose Calculation Manual (ODCM) was properly implemented. The 1995 and 1996 audit .
l reports effectively assessed program strengths and weaknesses. No deficiencies in the
Updated Final Safety Analysis Report commitments were identified (Section R1.2).
In the area of radiological controls, the licensee exhibited several positive performance
attributes in exposure monitoring and control and ALARA at the start of the outage. There
were two non-cited violations involving licensee-identified and corrected violations
pertaining to adherence to posted instructions and RWP requirements; and adherence to
access control procedures for radiologically controlled areas, including high radiation areas
(Section R1). j
The audits were of sufficient technical depth to effectively identify and assess program
strengths and weaknesses. The audits evaluated the technical adequacy of implementing l
procedures, TS requirements, and practices. Performance of the audits by the audit teams
was thorough, objective and of high quality as evidenced by the report documentation
(Section R7.1).
On May 22, the inspector identified a violation on NRC physical security requirements, in
that a licensee designated vehicle, inside the protected area, was left unattended with the
keys in the ignition and the engine running (Section S1.2).
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TABLE OF CONTENTS
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EX E C UTIV E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TA B LE O F C O N TE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v ;
l. Operations .................................................... 1 1
01 Conduct of Oper ations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 '
01.1 General Comments ................................. 1
02 Operational Status of Facilities and Equipment . . . . . . . . . . . . . . . . . . . 1 ;
O 2.1 Unexpected Reactor Trip ............................. 1 j
02.2 Inadvertent Steam Generator Drainings . . . . . . . . . . . . . . . . . . . 3
, O2.3 Re f ueling O perations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 !
O2.4 Mid-loop O perations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
O2.5 Failure of Wide Range Gaseous Monitor . . . . . . . . . . . . . . . . . . . 5 -
II. Maintenance .................................................. 6
- M1 Conduct of Maintenance .................................. 6
l - M 1.1 Review of the Ten-Year ISI Program Status . . . . . . . . . . . . . . . . 6
M1.2 Observation of inservice inspection (ISI) Activity ............ 7
M1.3 Steam Generator Tube Eddy Current inspection Preparation .... 9
M 1.4 Pressurizer Head-to-Shell Ultrasonic Test . . . . . . . . . . . . . . . . . 10
M1.5 Encapsulation Vessel for Recirculation Sump Isolation Valves
(CBS-V8 and V-14) ................................ 11
M2 Maintenance Support of Facilities and Equipment . . . . . . . . . . . . . . . . 12
M2.1 Foreign Material Exclusion (FME) Controls . . . . . . . . . . . . . . . . 12
l M2.2 -Diesel Generator-1 A Monthly Operability Surveillance Run . . . . . 13
M2.3 Pressure Tube Failure (Radiological Spill)in Primary Auxiliary
Building ........................................ 14
Ill . Enginee rin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
El Conduct of Engineering .................................. 15
E1.1 Modification of Primary Component Cooling Water Heat
Exch an g e rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l
E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . 16
E2.1 Main Steam Safety Valves Testing ..................... 16
E2.2 Leak Rate Test Failure of Containment Isolation Check Valve
CBS-V18 ....................................... 18
E2.3 Operating Cycle 5 Failed Fuel Rods . . . . . . . . . . . . . . . . . . . . . 19
E2.4 Failure of Unitized Starters During Surveillance . . . . . . . . . . . . . 22 i
l IV. Plant Support ................................................ 23 ! R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 23 .
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- R1.1 Implementation of the Radioactive Liquid and Gaseous Effluent j
- Control Programs ................................. 23 ;
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R1.2 Implementation of the Radiological Environmental Monitoring
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R1.3 Meteorological Monitoring Program (MMP) . . . . . . . . . . . . . . . . 27
R1.4 Refueling Outage Radiological Controls-External Exposure . . . . . 28
R1.5 Refueling Outage Radiological Controls-Internal Exposure ..... 30
R1.6 Refueling Outage Radiological Controls-Radioactive Materials,
Contamination, Surveys, and Monitoring . . . . . . . . . . . . . . . . . 30 .
- R1.7 Refueling Outage Radiological Controls-As Low As Reasonably '
l Achievable (ALARA) ............................... 32 ,
R1.8 Other Changes to the RP Program . . . . . . . . . . . . . . . . . . . . . . 32 j
i R2 Status of RP&C Facilities and Equipment ..................... 33 i
R2.1 Calibration of Effluent / Process Radiation Monitoring Systems
l (RMS)................... ..................... 33
R2.2 Calibration of Area Radiation Monitoring Systems (ARMS) . . . . . 34
R 2.3 Air Cle aning Syste m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 )
R3 RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 35 i
R5 Staff Training and Qualification in RP&C . . . .................. 36
l R6 RP&C Organization and Administration ....................... 37 -
R6.2 Management Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
R7 Quality Assurance in RP&C Activities ............. .......... 38
R7.1 Quality Assurance Audit Program . . . . . . . . . . . . . . . . . . . . . . 39
R7.2 Quality Assurance of Analytical Measurements ............ 40 l
R8 Miscellaneous RP&C issues ............................... 41 !
R8.1 Review of Updated final Safety Analysis Report (UFSAR) I
l C o m mit m e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 l R8.2 (Closed) LER 50-443/96-009-00: Missed surveillance PCCV/ l rate of change monitor alarm. . . . . . . . . . . . . . . . . . . . . . . . . . 41 , l R8.3 (Closed) LER 50-443/97-005-00: Misposition of Main Steam l
Line Radiation Monitors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 ;
S1 Conduct of Security and Safeguards Activities . . . ..... ........ 42
S1.1 General Comments (71707, 71750) .................... 42
S1.2 Uncontrolled Vehicle in Protected Area . . . . . . . . . . . . . . . . . . 42
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V. Management Meetings . . . . . . . . . . . ....... ...................... 43 l
X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . ........ .. 43
X3 Other NRC Activities . ............... .................. 44
PARTIAL LIST OF PERSONS CONTACTED .. ..... . ................... 45
i INSPECTION PROCEDURES USED . . . . . . ... .......................... 47 !
LIST OF ACRONYMS USED ............... ......................... 48
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Report Details
Summarv of Plant Status
At the beginning of this inspection period, the facility operated at 100% rated thermal
power for 62 days, with routine minor power reductions performed to support instrument
calibrations and turbine valve testing. Operators shutdown the unit on May 5 to begin the
fifth refueling outage (OR05).
1. Operations
01 Conduct of Operations i
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01.1 General Comments (71707)
The inspectors routinely conducted independent plant tours and walkdowns of selected
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portions of safety-related systems during the inspection period. These activities consisted of the verification that system configurations, power supplies, process parameters, support system availability, and current system operational status were consistent with Technical Specifications (TS) requirements and Updated Final Safety Analysis Report (UFSAR) descriptions. Additionally, system, component, and general area material conditions and housekeeping status were noted. l
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In general, routine operations were performed in accordance with station procedures and l plant evolutions, and were completed in a deliberate manner with clear communications and effective oversight by shift supervision. Control room logs accurately reflected piant
- activities. Shift turnovers were comprehensive and thoroughly addressed questions posed
l by the oncoming crew. Control room operators generally displayed good questioning l attitude prior to releasing work activities for field implementation. The inspectors found. l that operators were knowledgeable of plant and system status.
O2 Operational Status of Facilities and Equipment
l l 02.1 Unexpected Reactor Trip
a. Insoection Scop _e
On May 10, while performing a normal shutdown to enter the fifth refueling outage, the
reactor tripped from approximately 8% power due to an intermediate range (IR) high flux
trip actuation when permissive P-10 reset. Control room operators appropriately responded
to the reactor trip and implementd the applicable procedures. The inspector reviewed the
j root cause analysis (RCA), the circumstances, and activities related to the reactor trip to l assess licensee response to this transient. ,
b. Observations and Findinas
The licensee determined that the reactor trip occurred due to the high-neutron-flux trip
signal from detector IR N35 not being cleared prior to P-10 reinstating the IR high neutron
flux trip. Permissive P-10 blocks the IR high-neutron-flux trip when operating at power
levels greater than 10% and reinstates the trip when i ower is below 8%. The licensee
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determined that the flux trip signal had not cleared because of changes in IR N35 detector
current and a drift in its power supply. The net effect was that indicated power on the IR
detector was higher than actual power. Based upon the effects of the detector current
changes and power supply drift, the trip signal from IR N35 would have reset at
approximately 7% power. IR N36 was affected in a similar manner, however, its trip
signal had cleared prior to the resetting of P-10.
The licensee determined the root causes to be inadequate monitoring and trending of IR
channels, a lack of knowledge of the effect of detector current on reset and trip setpoints,
and a lack of detail in the shutdown proceduro that created an over-reliance on operators
to diagnose an imminent reactor trip. Several staff personnel were aware of the unusual IR
detector currents, but did not initiate actions to assess or address the condition. The
change in the detector currents was caused by an outward radial shift of neutron flux. The
licensee had become aware of detector power supply drifts that had affected the power
range and source range detectors but did not consider the effect on the IR detectors. The
shutdown procedure cautioned the operators about the significance of P-10 and operators
did have indication on the Reactor Protection System Bistable Panel that a trip condition
existed for IR high flux. The operator, however, failed to stop the power decrease and ;
clear the trip condition before going below 8% power.
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The RCA indicated that severallicensee personnel were aware of precursors and therefore !
had an opportunity to prevent the trip. The training staff, who maintain the training
simulator, compare plant parameters with simulated parameters. The staff noted the
change in the IR detector current and had incorporated it on the simulator's IR detectors.
During subsequent licensed operator training for low power scenarios, crews experienced
several effects (IR flux trip and IR rod stop at lower than expected values) from the change
in the IR detector currents. The licenseo failed to appreciate the significance of these
effects in the simulator scenario to assess the potential impact upon the plant. No ACR.
was generated as a result of occurrences in the simulator.
To prevent recurrence of this event, Operations management initiated the following
corrective actions: j
e Revise the shutdown procedure to verify the IR high-neutron-flux trip bistable lights l
have de-energized prior to decreasing power below 10%.
e Evaluate the effect of core flux changes on nuclear instrumentation, adjust the l
monitoring frequency as necessary, and develop a strategy based on the evaluation.
e Increase the frequency of monitoring IR for power supply degradation.
e Incorporate the lessons learned from this event into operator training.
- The inspector determined that the corrective actions appeared to address the root causes
l of this event. Further, the inspector determined that the RCA report was very good. l System conditions and licensee staff performances related to this event were thoroughly l evaluated.
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c. Conclusions
A preventable reactor trip occurred during the manual shutdown of the unit to enter the
fif th refueling outage. Hant indications available to the operators were overlooked.
Although the safety consequences of the event were low, it did subject the plant to a
unnecessary transient and challenge to the operators. The RCA report for this event was
very thorough, however, it indicated an overall lack of appreciation by the staff for the
impact of abnormal conditions on plant response. The inspector concluded that staff
performance leading to the trip indicated a weak assessment of the impact of the IR
detector current anomaly on plant operations.
02.2 Inadvertent Steam Generator Drainings
a. Insoection Scope (71707)
On June 13, with the unit in Mode 6 (refueling), a Lo Lo Stea>n Generator (SG) level (14%
narrow range) condition occurred on the 'C' SG due to a feedwater system drain valve
being inadvertently left open. The inspector reviewed the licensee's response and
circumstances related to the inadvertent draining of the 'B' and 'C' steam generators. -.
b. Observations and Findinos
While perforrning a stroke test on the 'C' SG Feedwater Isolation Valve, FW-V-48, SG level
dropped from approximately 40% to 14% causing an emergency safeguards feature
actuation. A reactor trip and emergency feedwater actuation occurs at 14% SG level. The
loss of SG inventory was due to drain valve FW-V-116 which was inadvertently left in the
open position. -The water drained from the SG discharged to the circulating water system
and ultimately to the ocean. Drain valve FW-V-116 should have been closed to maintain
the system boundary. (The feedwater check valve, FW-V-332, upstream of FW-V-116,
was the intended system boundary.) The licensee is investigating this event to determine
the root cause. (IFl 50-443/97-03-01)
Operators were not cognizant of the loss of SG inventory due to the SG level deviation
alarms already being present at 40% level. The deviation alarms occur at i 5% from
50% NR level. The inspector noted about two hours after this event, that the 'B' SG level
was approximately 20% while the levels in 'A' and 'D' SG were approximately 40%. The
inspector questioned the operators about the discrepancy. The operators stated that
nothing was unusual about the situation and attributed the lowering of 'B' SG level to
chemistry sampling. However, on June 14, the inspector was informed that 'B' SG level
had decreased from 40% due to the same reason as the 'C' SG The drain valve (FW-V-
115) between the feedwater isolation valve and check valve had been open and was a
release path for 'B' SG inventory when that isolation valve was stroke tested.
c. Conclusions
The inspector concluded that insufficient system configuration control contributed to the
event occurrence. Operations personnel initially only assessed that SG 'C' had drained
down and failed to investigate the SG "B" low level indication even after being questioned
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by the inspector. More importantly, the ineffective operato, awareness of SG level / trend
contributed to the slowness in detecting the inadvertent draindown of both SGs. Due to
the condition of the plant, the inspector determined that the safety significance to the
event was low because the SGs were not required to function as a heat sink.
02.3 Refueling Operations
a. Insoection Scone
The inspector reviewed licensee procedure RS0721, Refueling Administrative Control, and
observed licensee refueling related activities in the control room and in containment to
assess compliance with regulations and the licensee's performance in this area.
b. Observations and Findinas
The licensee adhered to the procedural guidance and cautions in the procedure during fuel
movement. Activities were conducted in an orderly and professional manner. Fuel
assembly movements and positions were monitored and tracked by reactor engineering
personnel in the control room and 1/M plots were performed appropriately to monitor
approaches to criticality.
While in containment, the inspector did not notice the audible source range detector
indication. According to the licensee, the detector scale was set at x1000 and was
providing an audible signal about every 60 to 90 seconds. The licensee is to suspend fuel
handling operations if the system becomes inoperable. However, at that scale setting and
frequency, the signal was unnoticed by the inspector in containment. (Only four
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assemblies remained to be loaded into the core at that time). The inspector was concerned '
that it would be difficult to recognize if the signal became inoperable. Therefore, the
inspector considered the high scale setting to not be a good work practice as the purpose
of the audible signal was diminished. l
The inspector verified that the licensee was in compliance with the TS regarding criticality
concerns in the spent fuel pool (SFP). T.S. 3.9.13 specifies the identification and storage j
location of three types of assemblies within the SFP. The inspector verified that the
licensee had identified the assemblies according to type and that the restrictions regarding ;
their relative storage location in the SFP were met.
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c. Conclusions
Overall, the inspector determined that the licensee's refueling activities were performed l
well and in accordance with procedure. The licensee agreed to review the setting of the
audible source range detectors.
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02.4 Mid-loop Operations
a. Insoection Scoce
On May 16 and 17, the inspector observed portions of the reactor coolant system (RCS)
drain down to mid-loop (RCS level below the top of the hot leg nozzles, which is between
minus 71 to minus 73.5 inches), to install the SG nozzle dams for eddy current testing, in
accordance with operations procedure OS1000.12, Operation with RCS at Reduced
inventory /Midloop Conditions. This activity was started after the fuel was fully offloaded
from the core into the SFP.
The inspector reviewed the procedure, attended pre-evolution briefings, walked down test
equipment and verified installation of levelindicators inside containment, and verified
proper indication in control room. In addition, the inspector verified proper residual heat
removal (RHR) system operation.
b. Observations and findinas
The inspector noted good coordination, communication, and management oversight during
the activity. Personnel were knowledgeable of the procedure, plant conditions and
required termination criteria of the evolutions to address abnormal conditions if required.
Adequate RCS levelindication and redundancy was provided by four separate level
indicators. Operators maintained excellent focus regarding RCS inventory and decay heat
removal throughout mid-loop operations.
c. Conclusion
The inspector determined that Seabrook Station personnel properly performed the critical
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evolution of drain down to allow removal of the SG nozzle dams, in a safe and controlled
! manner. Adequate measures, briefings, and activities were implemented to prevent ,
previous industry problems which have occurred while performing similar evolutions. No
l deficiencies were identified by the inspector.
02.5 Failure of Wide Range Gaseous Monitor
l a. Insoection Scone l l On April 30, Seabrook Station declared the station exhaust Wide Range Radioactive
Gaseous Radiation Monitor (WRGM) inoperable due a failed power supply. The inspector
verified the licensee's actions to comply with station TS and attended a Station Operations
l Review Committee (SORC) meeting.
b. Observations and Findinas
The Seabrook Station operations and radiation controls departments properly evaluated the
problem with the WRGM and entered the remedial actions of TS 3.3.3.9 as required. The
TS action statement requires grab samples every 12 hours and sampling every 24 hours
when the equipment is out of service. The station initiated sampling every 6 hours and
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analysis every 12 hours. In addition, a station temporary modification was initiated to
provide a method er continuous gas sampling of the primary vent stack in the event of a
failure of the WRGM low range monitor. A temporary portable gas radiation monitor was
installed at tha WRGM grab sample lines in the primary auxiliary building elevation 53 foot.
The temporary portable gas sampling monitor provided continuous monitoring for '
indication / trending only.
l A SORC meeting was convened to approve a temporary modification to install a portable {
continuous radioactive gas monitor to provide additional backup monitoring of station -
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exhaust during the period that the WRGM was out of service. During the SORC meeting
l station management asked very detailed questions regarding the effects of the temporary
equipment on other permanently installed equipment and the duration of the temporary
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installation. It was determined that no safety hazards existed from the installation of the
! temporary equipment. On May 6, a new WRGM power supply was received from the l vendor, was installed and tested satisfactorily, and the unit was retumed to service. The l inspector verified the proper installation and operation of the temporary gas sampling I
equipment in the primary auxiliary building. The equipment was properly labeled as being a
temporary modification.
The inspector noted that this radiation monitor power supply was previously placed in
l maintenance rule category A.1 due to repeated failures of the power supply. The station is l continuing to monitor the equipment performance.
I
c. Conclusions l
l The inspector determined that the licensee properly identified that the WRGM was i inoperable and properly entered the action statement. The increased sampling frequency l ' ~ was very conservative with an excellent safety perspective in evidence and the addition of
the portable sampling equipment provided excellent backup information. The failed power
l supply was promptly repaired and returned to service. The inspector noted strong
management involvement in the repair process to ensure that the instrument was promptly
'
returned to service.
II. Maintenance
M1 Conduct of Maintenance
M1.1 Review of the Ten-Year ISI Program Status
l l a. Insoection Scope ,
The inspector reviewed the status of Seabrook Station first ten-year American Society of ; Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PVC) Section XI
- inservice inspection (ISI) program plan at the beginning of the third period of the first ten-
4
year interval. The inspector reviewed the quality of preparation for monitoring the program
i ! L __
. _ . - . - -. . _ - - -. - - .-
l ! l- !
7
and the status of the inspection activities targeted for program completion at the end of
the second period of the first ten-year interval.
b. Observations and Findinas
The inspector found that the program plan in effect at Seabrook Station meets the
l '
regulatory requirements of 10 CFR 50.55a(g). The first ten-year interval ISI program has !
'
been in effect since the start of commercial operations, 8/19/90. The schedule was found
to be in accordance with Program B of IWA-2400, ASME B&PVC Section XI,1983 Edition
and Addenda through Summer 1983.
l
The inspector found that the program plan covered the first ten-year interval performance
requirements for non-destructive examinations (NDE) of Code Class 1,2, and 3 i
components and their supports under ASME B&PVC Section XI Subsections IWA., IWB, !
IWC, IWD, and IWF. Also, in effect at Seabrook Station is the augmented ISI program
indicated in the Updated Final Safety Analysis Report (UFSAR), Section 6.6.8. to protect
against failure of high energy lines penetrating the containment structure.
In the Seabrook ISI program plan, the inspector found a comprehensive listing of applicable
codes, programs, manuals, and administrative controls (personnel qualification, procedures,
records, and reports). Exceptions to ASME B&PV requirements were indicated for Class 1, l
2, and 3 system components and their supports. The programs, manuals, administrative i
I
plans were found to clearly document inspection expectations.
The inspector found the implementation of all elements of the program to be on schedule,
and in accordance with rules of ASME B&PVC Section XI Program B of IWA-2400.
Licensee expectations for completion of the program elements at the conclusion of the first
10 year interval is good,
c. Conclusions
The first ten-year ASME Section XI ISI program plan status at the beginning of the 3rd i
period of the first interval is within the targeted schedule and consistent with the
requirements of Section XI of the ASME B&PVC, and 10 CFR 50.55a(g). Licensee
expectations for completion of the program elements at the conclusion of the first 10 year
interval is good. The program plan was well prepared, documen+ed, and implemented.
M1.2 Observation of Inservice inspection (ISI) Activity
a. insoection Scooe
The inspector observed and/or reviewed the results of ISI at the plant planned during
i OROS. This included the following: l l l l
1
__ _
,
.
8
b. Observations and Findinos
ISI of Reheat System Pioe-to-T Weld
The inspector reviewed the penetrant testing (PT) and ultrasonic testing (UT) of the Reheat
System (RHS) pipe to T weld, Component RH0180-01-01, Class 2, Drawing 1-NHY-
800180lSI Revision 5, Joint #1. The PT included cleaning, lighting used, surface
condition, surface temperature, gage identification, and post examination cleaning. The
examination used Spotchek SKC-S cleaner / remover, SKL-HF penetrant, and SKD-S2
developer. The weld and .5 inches on each side were examined. The examination was
properly performed. Qualification and review signatures were verified. No reportable
indications were noted.
The inspector reviewed the UT process and examined the calibration data for the Stavely
Sonic 13b instrument and the KRA (.5 inch 2.25 mhz shear) and Megasonics (.14x.30 4
mhz longitudinal) search unit using Sonotech Ultragel 11 couplant. The examination was ;
properly performed. Qualification and review signatures were verified. No reportable
'
indications were noted.
ISI of Main Steam Line Elbow-to-Pioe Weld
The inspector observed and reviewed the main steam line (MSL) elbow-to-pipe weld MS
4000-02 09 magnetic particle test (MT) and UT results on weld MS 4000-02 09, class 2,
and code category augmented C5.51, as required in the UFSAR, Section 6.6.8.
The inspector observed the MT examination performed on the MSL elbow-to-pipe weld and
found the examination to be correctly performed in conformance to procedure No.
ES1807.003 using a Parker Yoke and dry red particles under a flashlight and drop light.-
The joint was examined in as-welded condition. No reportable indications were noted.
The inspector reviewed the qualifications of the NDE examiners and found them to be
acceptable.
The inspector observed the UT examination performed on the MSL elbow-to pipe weld.
The UT calibration record for the Stavely Sonic 136 instrument, and the KBA 2.25 mhz
shear wave search unit, were reviewed and found to be acceptable. No reportable
indications were noted. The inspector reviewed the qualifications of the NDE examiners
and found them to be acceptable.
ISI of Reactor Coolina System "C" Pumo Fivwheel
The inspector reviewed the results of a UT inspection on the reactor cooling system (RCS)
"C" pump flywheel. The ultrasonic calibration record shown on data sheet 97-25-053 for
! the Stavely Sonic 136 instrument, and KBA 2.25 mhz longitudinal pickup were found to be l acceptable. The qualification and review signatures were verifiM. No reportable
indications were noted. ,
! l l l l
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.
9
Snubber Functional Test Failure
The inspector examined the results of a failed MSL class NNS-1 snubber 5929-RM-1
functional high drag test. The failure was due to internal corrosion, and the snubber will be
replaced with a Phoenix Hydraulic Snubber. Functional tests were performed of 10% of
the same size and type of snubber. This expanded sample all passed their functional tests.
c. Conclusions
ISI examinations found no reportable defects under PT, MT, and UT. Reheat System Pipe-
to-T Weld, Main Steam Line Elbow-to-Pipe Weld, and Main Steam Line Elbow-to-Pipe Weld.
The snubber ISI found only one case of failing the functional test, and the snubber was
replaced. The tests were performed in accordance with ASME B&PVC Section XI rules and
NRC regulations. No defective welds were found.
M1.3 Steam Generator Tube Eddy Current inspection Preparation
a. Insoection ScoR2
The inspector reviewed the history of plugged SG tubes since the beginning of commercial
operation including review of the plugging history of eddy current (E/C) inspection of SG
tubes, and the basis for the planned E/C inspection during OR05.
b. Observations and Findinas
The inspection was directed at an assessment of the tube plugging history to-date, and the
plans for the E/C inspection of SG tubes which was scheduled to start shortly.
On review of the plugging history of SG tubes, the inspector found the following tubes
have been plugged:
Steam Generator "A" 12 tubes
Steam Generator "B" 6 tubes
Steam Generator "C" 13 tubes
Steam Generator "D" 5 tubes
Although the number of tubes requiring plugs is low, the inspector recognized that the
operating life is less than 7 years. Most steam generator degradation problems have been
found only after longer periods of operation. The E/C results to date indicate wall thinning
attributable to flow induced vibratory relative motion between the tube and its intended
support.
The SG E/C inspection during OROS will consist of full length bobbin inspection of 100%
of the tubes in SGs "B" and "C". All these tubes had been previously inspected, and all
tubes with identified anti-vibration bar (AVB) wear from previous inspections were
reinspected in May, 1994. 24 AVB flaws were detected in SG "B" and 44 AVB flaws
were detected in SG "C". On the basis of assessments of flaw growth rates, the licensee
estimated the remaining wall after two cycles of operation and concluded that four tubes in
- _____________ _
.. .. . . . .
3
. _
.
10
SG "B" and 6 tubes in SG "C" will require plugging during this outage in addition to 3 more
tubes in each of SGs "B" and "C" having more than one AVB flaw of at least 20%.
Additionally, Seabrook plans to perform rotating plus point probe inspection at the top of
the tubesheet on 50% of the hot leg tubes of SGs "B" and "C", where circumferential
cracking can occur, 50% rotating probe inspection of the row #1 U-bends, and a small
sample of rntating probe inspection of dents, dings, tangential flag signals, and free-span
signals. Two plugs will be changed in SG "B", and 3 plugs will be changed in SG "C",
both in the cold side due to material susceptibility to cracking.
Most of the plugged tubes are a result of AVB wear or a result of manufacturing defects.
AVB degradation can occur when the AVB clearances are excessive, and allow the
supposedly restrained tubes to move in a vibratory manner when acted upon by the flow
stream. This was a phenomenon carefully investigated by Westinghouse at the Research
and Development Laboratories in tests of U-tubes in flow environments simulating that in a
SG.
c. Conclusions
The inspection program was conservatively planned, and consistent with ASME Section XI,
Reg. Guide 1.83, Plant Technical Specifications, Electric Power Research Institute (EPRI)
Examination Guidelines, and licensee responses to Generic Letter 95-03. Seabrook
engineering has prepared an E/C inspection based on an engineering evaluation of wall
thinning rates estimated by past E/C data and an acceptance criteria established for
remaining wall required for continued operation.
1
M1.4 Pressurizer Head-to-Shell Ultrasonic Test - I
a. Insoection Scoce
The inspector reviewed preparation for pressurizer upper head-to-shell weld UT
examination using Time of Flight Diffraction (TOFD) technique qualified at the EPRI NDE
Center in Charlotte N.C. expressly for this application. (This inspection was documented in
NRC Inspection Report No. 50-443/97-02, Section Ill.E. 2.1.)
NRC inspection of the UT data evaluation for the pressurizer shell to upper head weld was
done on June 11-12, at the Seabrook site, after the UT data acquisition was complete and
the data evaluation was in progress,
b. Observations and Findinas
The inspector reviewed the results of the qualification tests conducted at the EPRI in order
to form a basis of comparison of the qualification process at the Seabrook site. A
l
calibration block was constructed for the Charlotte qualification process that duplicated the
nominal dimensions and materials of the pressurizer. Because of the sloping geometry at
the head to shell taper, it was necessary to use the UT technique called TOFD to have
repeatable capability for near-full volumetric flaw detection.
-- . . _ . . ..
.
11
The inspector walked-down the TOFD set-up within the containment after shut down of
the plant for ORO5. The equipment used was identica! to that used in Charlotte, as were
the three NDE technicians who were qualified using the same equipment. The inspector
asked the technicians to reproduce the image of a weld flaw found in Charlotte, and found
the resulting picture was identical to that taken during the initial qualification process.
During the requalification process within the containment, the licensee found that the
mockup used in the process qualification was built to nominal drawing dimensions of the
pressurizer, and not to the "as-built" dimensions of the pressurizer at the head-to-shell
juncture weld. As a result, the licensee used an alternate procedure to provide for an
acceptable weld examination. During the inspection, the inspector reviewed the UT
transducer positions used during the weld examination, sampled the UT findings and
compared the UT responses from pre-existing reflectors in the weld area for the techniques
used. The data from the TOFD,0,60 and 70 degree examinations were appropriately
confirmatory. No service induced indications were identified in the weld area and the pre-
existing indications were within construction acceptance limits.
l
c. Conclusions I
Preparation for examination of the pressurizer head-to-shell weld showed good capability in
utilization of technical resources. Weakness was shown in allowing the utilization of
nominal drawing dimensions, instead of as-built dimensions, for weld inspection
qualification models. This error was self-discovered, and steps were taken to provide a
satisfactory alternate inspection procedure for the pressurizer head-to-shell weld ISI
requirements.
1
M1.5 Encapsulation Vessel for Recirculation Sump Isolation Valves (CBS-V8 an't V-14) I
a. Insoection Scone (62707)
On June 11, the licensee made a 4 hour Non-Emergency Report to the NRC under 10 CFR
50.72 (b)(2)(i), to document the failed leak rate test for the Encapsulation Vessel for i
Recirculation. Sump Isolation Valves CBS-V8 and V14. On June 16, the inspector
inspected the work being performed on these valves. The Local Leak Rate Test (LLRT)
requires that the valves to maintain a test pressure of 52 psig. The leakage on both
vessels were from joints that were not disassembled during this outage. The encapsulation
vessels consist of three flanged sections, joined at two places, each with a groove and an
i O-ring to prevent leakage.
b. Observations and Findinas
The inspector identified in the -26'O elevation mechanical penetration area of the Primary
Auxiliary Building (PAB), that maintenance workers were tossing the bottom flange bolts of
the encapsulation vessel for valve CBS-V14 into a bucket. These safety-related bolts were
landing on top of each other and some were missing the bucket and hitting the wall. The
inspector questioned the workers about this practice. They responded that they were
awara that these bolts and the encapsulation vessel were safety-related components,
, however, they did not feel that tossing the bolts had the potential for causing damage to ! t
.
.
12
,
the bolts, and increasing the possibility for flange leakage. The inspector then expressed
his concern and informed the workers that this was not an acceptable practice. The
,
workers then proceeded with their work in an acceptable manner.
4
Management directed that the bolts be inspected for damage and initiated an adverse
condition report (ACR) to evaluate the poor maintenance practice. All 48 bolts were
inspected and tested. The licensee determined that none of the bolts exhibited damage
from the poor work practice that was observed by the inspectors. However, a total of 25
bolts were replaced due to corrosion. After the first repairs were made to prevent flange
leakage, the vessel for valve CBS-V8 failed the LLRT test once more while the vessel for
valve CBS-V14 failed two more times after repairs were made. The licensee issued Minor
Maintenance (MMOD) 97-0577 to redssign the gasket to address these repeated failures.
The inspector found this document to be adequate. After implementation of this )
<
modification, the encapsulation vessels passed the required leak test. '
J
Per UFSAR, Section 6.2.4.1, and Technical Clarification TS-009, dated 3/2/94, the subject
, encapsulations are not required to be in place to ensure containment integrity, because
they are not in direct contact with containment atmosphere, but rather, they are designed
to prevent release of any leakage from the recirculation sump isolation valves CBS-V8 and
V14, or related piping to the environment during design basis accident conditions. The-
valves are, however, designed to withstand the maximum calculated containment internal j
pressure (50 psig) for the design basis loss of coolant accident, as described in the UFSAR.
l
l
c. Conclusion
The inspector concluded that the tossing of the safety-related encapsulation vessel flange
bolts was a very poor maintenance practice and could have contributed to further flange
leakage. Furthermore, the fact that the workers did not think that this was a bad practice
when questioned by the NRC staff, is of concern. However, since this incident appears to
be an isolated case at this time, the corrective actions performed by the licensee to inspect
all 48 flange bolts, to counsel the individuals and the entire mechanical maintenance crew,
were found to be adequate. )
M2 Maintenance Support of Facilities and Equipment
M2.1 Foreign Material Exclusion (FME) Controls
a. Insoection Scooe (71707, 62707)
l
l
The inspector evaluated Seabrook Station administrative controls to prevent foreign l
material intrusion into plant structures, systems, and components. These observations
took place prior to and during OR05. The objective was to follow-up and evaluate the
adequacy of Seabrook's corrective actions and enhancements implemented to address a
previous NRC violation identified in inspection Report 96-02. This violation was closed by
NRC Inspection Report 97-01 in April 1997.
- - - - . - - - - - - - - -
,
.
13
The inspector reviewed the foreign material exclusion (FME) procedure (MA 3.4),
interviewed personnel, and performed several plant walkdowns in all plant areas, including
the spent fuel pool and the containment. Major evolutions observed included; core offload
and reload, emergency diesel generator (EDG) overhauls, main turbine and generator work,
primary cooling water heat exchanger replacement, emergency feedwater turbine driven j
pump governor valve replacement, reactor coolant pumps motor maintenance, and main
steam safety valve testing.
1
b. Observations and Findinas
The inspector determined that Seabrook personnel, including contractors and vendors, had
embraced a very aggressive FME controls program, and that extensive manpower and
resources were dedicated for this effort. The inspector observed that boundaries which
included FME tape, signs, and rope were established to monitor and control every object,
,
tool, or equipment used to perform the tasks. A designated and well trained individual was
'
dedicated and posted at each job site for this purpose. In addition, every small tool was
secured with a lanyard to the worker to prevent it from being inadvertently dropped into
the system or component. The inspector also observed that prompt notification and
corrective were taken to document and address issues of deficient FME controls. For
example, on May 9, the inspector noted that two out of three replacement safety valves
stored in the east steam pipe chase had lost their FME covers. The licensee promptly
evaluated the situation, replaced the FME covers and ensured that a visual inspection was
performed prior to installing these valves. Also, when an operator inadvertently dropped a
pen into the reactor cavity, fuel movement was suspended until the item was recovered.
An ACR was initiated for each issue identified.
c. Conclusion
The inspector concluded that the current FME program at Seabrook is excellent and that
considerable manpower and resources have been dedicated for this effort.
M2.2 Diesel Generator-1 A Monthly Operability Surveillance Run
a. Insoection scooe
On April 30, the inspector observed a portion of the monthly diesel generator operability
run for emergency diesel generator (DG-1 A) as directed by station procedure OX1426.01,
DG-1 A Monthly Operability Surveillance, in the control room and in the diesel generator
room. The inspector observed the performance of the control room operator during the
surveillance testing and verified the proper operation of the diesel generator support
equipment during the surveillance run in the diesel generator room.
b. Observations and findinas
The inspector observed that the station operator conducting the testing in the control room
was knowledgeable of the procedure requirements and was properly monitoring the
equipment parameters on the main control board. All required diesel parameters were
taken during the test run with no abnormal readWs identified during the test. The
_ _ _ _ _
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ,
.
14
inspection tour in the diesel generator identified no discrepancies and it was observed to
operate properly. The room was found clean with no fire hazards present and no visible oil
leaks. The testing concluded satisfactorily with no deficiencies identified by the inspector.
c. Conclusions
The inspector determined that Seabrook Station operations personnel properly tested an
important safety-related system as directed by station procedures in a safe and deliberate
manner. All equipment was observed to operate properly and the testing was
accomplished satisfactory.
M2.3 Pressure Tube Failure (Radiological Spill)in Primary Auxiliary Building
a. Josoection Scone
The inspectors reviewed the licensee's actions to recover from a spill of radioactive water
in the PAB that occurred on May 27, due a failure of an under-rated instrument test tygon
tube. The failure occurred in the tubing for Flow Transmitter CS-FT-121 in the reactor
coolant charging system after a temporary pressure gauge with a much lower pressure
rating was installed for Emerg icy Core Cooling System (ECCS) valve testing. The
inspector also reviewed the licensee's actions to past tubing failures to determine if the
previous corrective actions were appropriate for the event to prevent reoccurrence. The
inspector also reviewed the ACR documenting the event and interviewed the instrument i
and controit (l&C) personnel involved in the installation of the temporary tubing. I
b. O_bservations and Findinas
The inspector determined that the licensee's initial response to the event was good. A
station engineering supervisor was egressing through the area and identified the leak. He
promptly informed the control room and remained in the area to prevent anyone from
walking through the contaminated area. The radiation controls personnel promptly
responded to the event and limited personnel access to the contaminated area. Cleanuo
efforts were promptly initiated and radiation boundaries were installed to limit access to
the contaminated area. The highest contamination reading was 200,000 counts per
minute above background. Approximately 30 gallons of water were spilled. The inspector
verified that the area was properly monitored with appropriate surveys and radiation
postings.
The licensee determined that the person involved did not install the appropriate tubing for
the pressure application, in spite of departmental supervisory guidance and coaching. The
installed tubing was rated for 250 pounds per square inch (psi) instead of the high pressure
tubing (3000 psi) required for the task. The high pressure tubing has a very distinctive
color from the one that was used and the department supervisor could give no apparent
reason for the error. The l&C Supervisor did determine that the person did not willfully
install the incorrect tubing. In addition, it was identified that several similar events had
previously occurred.
- _ _ _ _ _ _ _ - _ _
. - - - - - - - _._.-- - . . - _
- . - . . . . . . .-
t
; .-
!
i
15
The inspector determined that the following events have previously occurred involving
j temporary tubing failures:
* December 6,1995: Tubing burst for RH-FE-610 due to tubing with an inadequate
} rating - Installed 250 psi tubing vice 3000 psi tubing
j * June 20,1995: Tubing burst for the filtration unit for Main Steam isolation Valve -
l high pressure hydraulic fluid -Installed 100 psi tubing vice 3000 psi tubing
; * June 1,1995: Tubing burst for 345 KV circuit breaker air trouble alarm due to
'
improper valve lineup - Installed 250 psi tubing
4 4 * April 15,1994: Tubing burst for Local Leak Rate Testing test panel due to opening
!
the valve on the nitrogen bottle too fast
The inspector reviewed the root cause evaluation for the December 6,1995, spill event
? - that was attributed to improper test tubing. Human error was the root cause determined to i be the major causal factor. This analysis failed to determine if a programmatic issue
existed or if other procedural modifications were necessary.
- c. Conclusions
.
The inspector determined that the radiation protection department personnel properly
j responded to the event and provided the proper controls to recover the area that was j contaminated. The I&C department installation of improper tubing in spite of severa' past
i problems temporary tubing appears to have serious implications concerning inadequate
l corrective actions and management oversight to prevent reoccurrence. The potential for l serious personnel injury is a major concern. This situation is considered a significant
- weakness in the licensee corrective action program, and is considered a violation of 10CFR
l 50, Appendix B, Cnterion XVI, Requirements. (Violation 50-443/97-03-02)
.
Ill. Enaineerina
l E1 Conduc of Engineering ] '
E1.1 Modification of Primary Component Cooling Water Heat Exchangers
- l
a. Insoection Scoce
On May 21, the inspector observed a portion of the work activities in the PAB,in
, preparation for replacement of the "A" primary component cooling water (PCCW) heat
! exchanger, as directed :,tation design change request (DCR 96-016). The heat exchangers l' are being replaced due to repetitive degradation of the heat exchanger tube material which
l required retubing. The original tube material was a 90/10 copper nickel composition. The
j new heat exchanger tube materialis titanium, which has better corrosion resistance
i
characteristics. The inspector also reviewed the DCR package for completeness and
! reviewed the accompanying safety evaluation screening documentation. The work site ,
!
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. ._ . __ _ _. _ _. . __ __ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ __
,
.
16 I
was inspected during preparations for removal of the 'B' train heat exchanger and also
during the installation of the 'A' train heat exchanger. The inspector also verified FME
controls and control of contractors during the inspection. Interviews were also conducted
with the task construction supervisor on several occasions in the field on work progress,
b. Observations and Findinas
!
The inspector noted a very strong management oversight of contractor personnel and of
work activities performed during the installation phase of the heat exchanger replacement.
The proper FME controls were in place for the system piping and new heat exchanger. A l
fire watch was stationed in the area during welding / cutting operations and security I
l '
personnel were controlling access when the building roof hatch was opened. The control
of the crane for lifting the old and new heat exchangers was well controlled, and when a
question regarding the lifting capacity of the crane was raised, the lifting work was
,
stopped until all questions were resolved. The licensee demonstrated excellent job ,
l planning when it was determined that a portion of the heat exchanger support plate would ' f
interfere with permanently installed piping during removal of the old heat exchanger. The
- new heat exchanger was modified to allow proper installation without in field modification.
l The work area was kept in a very clean condition during the period of removal and l installation of the heat exchangers. The inspector verified that the opposite train of PCCW
was maintained in an operable status during the replacement process.
I c. Conclusions j The inspector determined that the licensee properly performed a very significant
modification to the stations permanently installed equipment. The work was well
controlled and the safety evaluation was thorough and properly documented. The system
engineers were directly involved in the task and provided excellent guidance to the
personnel that were installing the equipment. The inspector did not identify any
deficiencies during inspection of this activity.
E2 Engineering Support of Facilities and Equipment
E2.1 Main Steam Safety Valves Testing
a. Insoection Scooe (71707)
On May 9, the inspector witnessed portions of the operability verification test for the Main
Steam Safety Valves (MSSVs). This test was performed during Mode 1 with the plant at
approximately 70% of rated thermal power, using operating procedure EX1804.041, Main
Steam Safety Valve In-place Setpoint Verification, and work request RTS 97RE00119001.
The procedure requires as found set point verification within i 3% of the lift setting, and
provides instruction for subsequent lift setpoint adjustments if necessary to meet the
values described in TS Table 3.7-2. The test also verifies that the as-left lift setpoint is
within i 1 % of the lift setting. This test is performed using the Furmanite Trevitest
System.
1 !
.
.
17
b. Observations and Findincs
The inspector attended the pre-avolution briefing, held discussions with involved licensee
and the applicable contractor (Farmanite) personnel, reviewed the procedure, and
observed testing of four safety valves. The inspector reviewed the test results and found
that as-found and as-left acceptance criteria were met. The inspector noted excellent
planning and control by the test coordinator, good communication with the control room,
and management oversight. In addition, adequate personnel safety measures were l
implemented to prevent injury to personnel in case of equipment malfunction. During pre-
briefing, the licensee demonstrated strong safety measures by including identified industry
experience / problems in their discussions and taking applicable corrective actions to prevent
similar events.
During this review, the inspector noted that the Furmanite test equipment had been
upgraded and was different from the original test equipment. The licensee had performed ,
a safety evaluation (SE) per 10 CFR 50.59 requirements for the original test equipment, but I
did not perform a new SE for the upgraded test equipment. Upon review, the inspector i
concluded the new equipment is much lighter and permits the valve being tested to fully
lift, if required, unlike the original equipment. With the original equipment, the valve had to
be declared inoperable for the test duration, because the valve lift was limited to 50% of
full travel. The original evaluation takes credit for the limited valve lift of 50% so that in
the unlikely event that a safety valve were to stick in the open position, the total steam
released would be less than the maximum release of 970,000 lbs/hr as analyzed in UFSAR,
Section 15.1.4. It also states that the lifting of a maximum capacity safety valve is
bounded by the analysis.
' Followup inspection revealed that station procedure PA3.5, Administisi.on of Procedures
and Forms, controls the preparation and revision of station procedures. The existing
procedure that was used to test the main steam safety valves was revised prior to this
outage. The document review and approval cover sheet on the procedure lists four 10
CFR 50.59 applicability questions which may be answered under specified conditions.
Answering no to these questions eliminates the need for further 50.59 screening. These
conditions require that the procedure is prepared using the safety evaluation of another
docrnent such as Design Coordination Report (DCR), Minor Modification Temporary
Modification (MMTM), Temporary Setpoint Change on prior revision of the procedure. The
safety evaluation must have been reviewed and remain valid for the procedure and also the
procedures references the DCR/MMOD, temporary modifications, temporary setpoint
change, prior revision of the procedure or other documents used to prepare the procedure.
The inspection determined that using this guidance , licensee personnel were not required
to document a current review if it was decided that a prior revision of the document
contained an appropriate SE review.
! c. Conclusions l
The inspector determined that although no SE was performed for the new test equipment,
there were no safety consequences, because the inadvertent lift of a Safety Valve is
bounded by the accident analysis presented in the UFSAR, Section 15.1.4.2. However,
the in.spector concluded that the reliance on an or:ginal SE presents a potential weakness in
,_
__
.
18
the licensee's SE process. The inspector determined that a programmatic issue may exist
in the station procedure revision process and this item will be inspected in a subsequent
inspection. (IFl 50-443/97-03-03)
E2.2 Leak Rate Test Failure of Containment isolation Check Valve CBS-V18
a. Insoection Scoce
The inspector evaluated Seabrook's actions to address the Local Leak Rate Test (LLRT)
f ailure of the containment building spray (CBS) check valve CBS-V18, the subsequent
valve internal visual inspection, the corrective actions performed to repair the valve, and
the correctness of Seabrook's reportability of this issue. CBS-V18 is the 'B' CBS train
inside containment isolation valve for penetration (X-15). This penetration is also provided
with an outside containment isolation motor operated valve (CBS-V17). A similar
arrangement exists for the 'A' CBS train penetration (X-14), where motor operated valve
CBS-V11 and check valve CBS-V12 provide the containment isolation boundary.
b. Observations and Findinos
On May 17, with the Unit in Mode 6 (refueling) valve CBS-V18 failed the LLRT. Procedure
EX1803.003, " Reactor Containment Type B and C Leakage Rate Tests", requires that nc ,
individual penetration leak rate exceeds 37 scfh. A visual inspection performed by tne l
licensee, on June 5, determined the valve disc was stuck in the "Open" position.
Inspection of similar valve in the opposite train (valve CBS-V12), revealed that although the
disc was in the proper " Closed" position, and the valve passed the LLRT test, it was also
susceptible to binding when the disc was manually activated to the open position. On
June 10, the licensee made a 1 hour non-emergency report to the NRC under 10 CFR , l
50.72 (b)(2)(i). l
Both valve CBS-V12 and V18, are 8"-forged body, swing type, check valves manufactured ,
by Velan Corporation. There are seven simi;ar valves installed at Seabrook, with only !
these two valves being used in a safety-related system (Containment Building Spray), and !
the remaining five being used in the Condensate and Main feedwater Systems. The
licensee considers valves CBS-V12 and V18, unique because they were modified during
the construction phase, under Engineering Change Authorization (ECA) # ECA-19802647B,
dated May 19,1986, to replace the valve's disc, hanger, and hanger ring. The inspector
verified that both valves have satisfactorily passed all six (6) LLRT tests performed
previously. The licensee's inspection performed on June 5, per ACR 97-1448, determined l
that the disc in both valves, were impacting the valve body at several points in the upper
throat. Preliminary evaluations performed by the licensee, determined that an incorrect
disc hanger may have been installed on both valves during the construction phase under
ECA-19802647B. To correct this problem, the licensee machined both the incorrect
hangers and the back side of the valves discs to eliminate any contact points with the
valve body, as recommended by the valve manufacturer. Both valves passed the post
modification LLRT on June 10. CBS-V18 was last disassembled and inspected in 1995
during the fourth refueling outage, and the penetration passed the LLRT test. The licensee
concluded that CBS-V18 failed during one of the quarterly stroke testing of the up-stream
(outside containment isolation valve) MOV-CBS-V17, performed during the last operating
-- - - - - _ _ - . _ - . - _ - - - - - . _ - - . - .. - _ - _ - . - -
f
.
19
cycle (cycle 5), since the pressure head of the Reactor Wcter Storage Tank (RWST) is
enough to open the check valve. The licensee's root cause evaluation will be reviewed by
the inspector at a later time, and will be tracked as a follow-up item. (IFl 50-443/97-03-04)
The inspectors initial concern was the timing of the reportability made by the licensee on
6/10/97, for the failed valve, which was first identified on 5/17/97 via the LLRT. The
licensee had initially taken the position that no reporting requirements exist for the failed
LLRT test on valve CBS-V18, based on NRC's Ruling No. RIN 3150-AF 18, which states in
part that reporting "would be required when the total containment as- found, minimum ,
pathway leak rate exceeds the limiting condition for operation (LCO)in the facilities
'
l Technical Specification.", and incorporated this criteria in their LLRT test program, which is l implemented under procedure EX1803.003, and the Station Leakage Test Reference
Manual (SLRT). TS Section 6.15, " Containment Leak Rate Testing Program", states that
l
for type B and C tests (which valves CBS-V12 and 16 are) the acceptance criteria for total
i leakage of all combined penetrations shall be less than 0.60 La (443.6 scfh). To meet this - l
requirement, the licensee's program has stabilized a limit to each individual penetration at ;
<37 scfh (.05La). Since the local leak rate testing demonstrated that MOV met the
'
,
leakage criteria for this penetration ( < 37 scfh), the licensee determined that no
l reportability under the LLRT program was required. The licensee did determine that this l
condition was reportable pursuant to 10 CFR 50.72 utilizing NUREG-1022 Draft Rev.2,
which identifies a similar situation being reportable under 10 CFR 50.72 (b)(2)(i), as an ;
event, found while shutdown, that had it occurred while the reactor was in operation,
! ' would have resulted in the nuclear power plant, including its principal safety barriers, being .
'
seriously degraded or being in an unanalyzed condition that significantly compromises plant
safety". This guidance lists the loss of containment isolation valve function during plant ;
l operation as a condition that results in a loss containment function or integrity. The failure
of the check valve in the open position is significant because during a design basis loss of
coolant accident (LOCA), coincident with a loss of power and a single failure of the "B"
EDG a leakage path through penetration X-15 (lost of containment boundary) would have i
been possible via the failed check valve (CBS-V18).
c. Conclusion ,
The inspector concluded that Seabrook's evaluation of the failed LLRT for containment
spray valve CBS-V18, the subsequent corrective actions to machine the valve disc and
disc hunger, to prevent wedging of the valve disc with the body were adequate. Also the
implementation of the applicable corrective actions for similar identified valve CBS-V12
was prudent. However, the inspector concluded that the licensee's failure to promptly.
report this self-identified situation as required by 10 CFR 50.72 is a minor violation and is
treated as a non-cited violation due to meeting the requirements of Section IV of the NRC
Enforcement Policy. (NCV 50-443/97-03-05) '
I
E2.3 Operating Cycle 5 Failed Fuel Rods
a. Insoection Scoce
l The inspectors reviewed the licensee's response to the identification of leaking fuel ; i assemblies during Cycle 5 operations. The findings and recommendations of the root 1
- .
*
!
l
.
* l
20
cause investigation team were reviewed to determine the appropriateness of the findings
and to determine if a generic failure was involved due to this type of failure mechanism.
b. Observations and Findinos
On December 10,1996, the licensee identified that reactor coolant system noble gas and
iodine activity had increased by a significant factor from the previous steady state levels.
These increases were well below the TS limit of 1 microcurie per gram dose equivalent l-
131. However, this was a very clear indication of a failed fuel rods. From the date of the
original observation until the end of the cycle, the reactor coolant system (RCS) fission
product activity continued to slowly increase. in addition, there were indications that
additional fuel pins had also f ailed. At this time it was believed that the failed fuel pins
were located in the second burn fuel batch.
Subsequently, during core offload special fuel sipping equipment revealed that four fuel l
assemblies from a first burn batch of Westinghouse Vantage 5H Zirlo clad fuel ~ assemblies '
contained f ailed rods. After the identified leaking fuel assemblies were transported to the
SFP, followup ultrasonic testing of the assembly fuel pins determined that five failed fuel
,
pins existed among the four assemblies. (One failed pin in three assemblies and two failed
l pins in one assembly). The leaking fuel rods were identified to be first burn fuel
assemblies, Westinghouse Vantage 5H with Zirlo Cladding, 4.8% enrichment, assemblies
l with 128 IFB A (boron burnable poison coated fuel pellets) fuel rods, next to thimble guide l tubes and fuel rods exposed to above average core power. When the licensee's personnel
- attempted to remove the leaking fuel pins, three were found broken in the mid point of the
rod, approximately at the same area, between the assembly fourth and fifth grid. The
other two leaking rods were successfully removed from the fuel assembly and stored in the
spent fuel pool. The two intact fuel pins were also found to have breaches of the fuel
cladding, but not in the same area as the broken fuel rods.
The safety consequences of a breach in the fuel cladding is that fission product gases in
the fuel rods are released into the RCS and this increased radioa ity represents a
potential impact on the plant safety analysis. The inspector rev. .ed several daily plant
l
radiochemistry logs for the time period that the leakage was occurring and determined that
! station TS limits were not exceeded as the highest reading was approximately 6% of the
limit.
The inspectors requested that NRC headquarters review the licensee's root cause
l evaluation to determine the appropriateness of the conclusions and corrective actions and l to determine if a generic issue existed. The regional request was for the determination to l be made prior to plant restart. !
On June 17, an NRC headquarters Division of Reactor Regulation fuels specialist met with
Seabrook Station, Yankee Atomic Energy Company, and Westinghouse personnel at the
station to discuss the findings and licensee's conclusions of the Fuel Failure Root Cause
Evaluation Report. The specialist determined that although this was the first time this
many failures occurred in the first burn of the Vantage SH fuel, the failures were well
enveloped by the plant safety analysis and TS. The licensee root cause evaluation
determined that a probable cause of the fuel failures was the combined effects of power
_ _ -
,
.
21
history, core design, and an operational strategy that resulted in interaction between the
fuel pellets and the fuel cladding. The affected fuel assemblies apparently carried a very
large load (produced high power) for all of the last cycle. Another potential cause which
could not be completely eliminated is crud induced corrosion. The most probable cause
will be determined by future eddy current testing and/or possible hot cell testing.
The licensee uses incore nuclear instrumentation to assess nuclear peaking factors and to
perform core flux mapping. Tha inspectors inquired if the licensee had observed excessive
nuclear peaking factors in the core during Cycle 5 and was informed that no limits had
been exceeded during any of the regular surveillance. The inspectors had determined that
the licensee had not extensively reviewed incore nuclear instrumentation data to
investigate the cause of the failed fuel because no limits had been exceeded during the
surveillance. The inspectors then inquired where the incore nuclear instrumentation was
located relative to the fuel assemblies which contained the failed fuel rods. The inspectors
were informed that assembly G69 (which contained a failed fuel rod) was an instrumented
assembly. The inspectors reviewed axial flux maps generated from data collected by the ;
l detector in assembly G69 for December 1995, July 1996, December 1996, and March
1997.
l
Although no limits were exceeded, the inspector observed some unusual fluctuetions in the !
, December 1996 and March 1997 maps between the fifth and sixth grid where some clad l damage and unusual crud buildup had occurred. .The reactor engineering staff did not
know the reason for the unusual flux maps at that location. This observation was
forwarded to the Yankee Nuclear services Division (YNSD) for evaluation. The licensee's
preliminary assessment attributed the fluctuations to characteristics of the computer code
used to generate the flux maps.
To prevent recurrence, the licensee implemented several corrective actions. The remaining
128 IFBA assemblies were located in lower power regions for Cycle 6. Prior to restart, an
extended Mode 4 RCS cleanup was completed to minimize crud deposition. The pre-
conditioning power ramp rate was reduced from 3% per hour to 2% above 20% power.
To further minimize crud deposition, RCS inventory pH will be increased from 6.9 to
approximately 7.1 at the beginning of core life. The Vantage SH fuel assemblies that
exhibited problems during the last cycle (128 IFBA) were removed from the high power
region of the core to the periphery. Furthermore, no additional vantage SH fuel assemblies
were included in the new cycle 6 core design.
c. Conclusions l
The inspectors determined that the licensee appropriately assembled a very significant root
l cause evaluation team and conducted a very thorough evaluation in an excellent manner. t
All findings and conclusions were reasonable with very prompt corrective actions. The
new cycle 6 core design was modified based on the above findings and the fuel assemblies
that were found to have leak:ng fuel pins were removed from service.
.
The inspectors had no further questions of the licensee regarding this issue at this time,
however, core performance during Cycle 6 will be monitored by the inspectors.
4
e
'.
22
E2.4 Failure of Unitized Starters During Surveillance
a. Insoection Scope
The inspector reviewed the licensee's current act!ons to resolve another identified problem
with unitized starters at the station. A unitized starter failed to properly actuate during
routine surveillance testing. This problem has been previously identified in NRC inspection
report (50-443/97-80). Unitized electrical starters are used to e:ectrically start and stop
motor operated valves and energize other electrical equipment such as fans and heaters.
b. Observations and Findinas
On May 8, Seabrook Station technical support personnel determined that the unitized
starter for motor operated valve service water (SW) V-140 had failed to properly actuate
during routine surveillance testing. After station maintenance electricians inspected the
internals of the unitized starter, a circular clip (E-clip) was found lodged in the contractor
assembly. The source of the clip was identified as being part of the unitized starter pawl I
assembly. The pawl assembly pulls and locks the unitized starter onto the bus electrical
connections. Seabrook station technical support engineering personnel discussed the i
problem of the with the unitized starter vendor and a modification was developed to l
replace the E-clip with a jam nut and different spring. Also the existing E-clip and spring )
were to be removed to eliminate the potential of E-clip falling into the unitized starter l
contractor assembly. The potential of loose parts falling into the unitized starter could
affect safety equipment operability.
Station Modification MMOD 97-0553, " Modification to MCC Bucket Latching device", was l
incorporated into Seabrook Station electrical maintenance procedure LS0557.09, Revision
'
3, "480 Volt Motor control Center inspection, Testing and PM" to perform the field
installation. Station work orders were initiated to inspect and install the modification on all
safety-related unitized starters as a first priority. The modification package had a 10 CFR
l
50.59 evaluation that adequately described the change and its effect on the unitized starter. The inspector reviewed the modification package and determined that the proper reviews were conducted and that the proposed change scope was adequately described for implementation. The non safety-related starters will be modified as the surveillance comes l due. In view of the recent identified concerns with unitized starters and subsequent I category A.1 maintenance rule classification, these actions are appropriate. The licensee is currently considenng other options for these starters including replacement, if necessary to resolve this issue, and restore the equipment from the A.1 Maintenance Rule category. i
l The inspectors will continue to review this issue in future inspections. l
c. Conclusions
l The inspector determined that Seabrook station personnel properly identified a serious i
safety concern regarding the unitized starters and conducted prompt corrective actions to prevent recurrence.
i
,
.
23
IV. Plant Sucoort
R1 Radiological Protection and Chemistry (RPf4C) Controls
R1.1 Implementation of the Radioactive Liquid and Gaseous Effluent Control Programs
a. Insoection Scooe (84750-01)
The inspection consisted of: (1) tour of radioactive liquid and gaseous effluent process
facilities, and control room; (2) review of radioactive liquid and gaseous effluent release
permits; (3) review of unplanned or unmonitored release pathways; (4) review of
quantification technique for airborne tritium release; and (5) review of the effluent ALARA
program.
b. Observations and Findinas
l The inspector toured control room and selected radioactive liquid and gas processing
facilities and equipment including effluent radiation monitors and air cleaning systems. All equipment was operable at the time of the tour. Effluent and process radiation monitoring terminals in the control room and at the HP checkpoint were also operable. During review of selected radioactive liquid and gaseous effluent discharge permits, the inspector determined that discharge permits were complete and met the Technicd Specification /Offsite Dose Calculation Manual (TS/ODCM) requiremen'a for sampling and analyses at the frequencies and lower limits of detection established in the TS/ODCM. The in.spector noted that there were no unplanned /unmonitored radioactive liquid and gas releases since the previous inspection conducted in January 1996. The inspector also noted that the licensee had reviewed the effluent control programs continuously to
l implement the IE Bulletin No. 80-10, " Contamination of Nonradioactive System and
Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to Environment." The ins actor reviewed the chemistry study and Technical information Document No. 97- 006," ide Range Noble Gas Monitor (WRGM) Response to Noble Gas Release." The
l Document 97-006 was prepared by tha chemistry staff for response to the ACR 97-0514, ! which dealt with exceeding the alert for the low range of the plant vent WRGM while
performing routine monthly calibration of the waste gas oxygen monitor. Noble gas was purged from the waste gas system to the plant vent prior to calibration of the oxygen
I monitor, as required. The chemistry staff calculated an expected WRGM reading using l actual noble gas analytical measurement results and the conversion factor of the WRGM. l The calculated WRGM reading was 293 Ci/sec and the actual WRGM reading was 360
Ci/sec (about 19% over responded). The inspector concluded that the comparison
between expected and actual readings were very good and was acceptable within the
normal characteristics of a beta scintillation detector.
,
The inspector requested the licensee to demonstrate its capability for monitoring and
quantifying the airborne tritium. The licensee calculated the total amount of water loss
' trom the spent fuel pool (SFP) using the makeup water inventory. The licensee assumed
.
.
24
that water loss was due to an evaporation from the SFP release to the environment via the
plant vent. The licensee calculated the airborne tritium release using SFP tritium
measurement results. Calculated airborne tritium released through the plant vent during
1996 was to be about 23.01 curies. The licensee reported in the 1996 Annual Effluent
Report that 23.5 curies of airborne tritium was released. The inspector stated that the
licensee's assumptions and calculation methodologies were good and had an excellent
airborne tritium monitoring program.
The licensee recognized a design deficiency of the waste gas system, which would prevent
expeditious processing of large volumes of waste gas. Therefore, the licensee developed a
plan to enhance the waste gas system to allow larger volumes of processing in the fall of
1996. In December 1996, the licensee identified a small fuel defect, which requires
additional processing capacities for waste gas. On January 10,1997, a task force
(representatives from chemistry, HP, operations, engineering, outage management, and
incensing) was formed to review areas where effluent releases could be minimized to as
low as reasonable achievable (ALARA). This task force met weekly to discuss all waste
gas systems to increase waste gas hold up time (developed in fall 1996). The inspector
reviewed the following data to determine the soundness of the effluent ALARA program for
the refueling outage in May 1997:
* iodine dose equivalent trending data and iodine inventory in the reactor coolant
(iodine source term),
* entrained noble gas activities in the reactor coolant and noble gas inventory in the
reactor coolant and waste gas systems (noble gas source term),
* licensee's projected dose calculation results to the public, and l
* management support to this project.
Although the projected airborne iodine and noble gas source terms were small and could
release directly to the environment (it would result in a fraction of TS dose limits to the
public), the licensee management decided to implement the effluent ALARA program. For
example, the licensee will use temporary decay tanks and slow purge of the pressurizer, I
which would reduce the release rate to the environment, minimizing doses to the public. l
The inspector stated that the licensee's effort was an excellent example for the '
implementation of the effluent ALARA.
c. Conclusions
Based on the above reviews, the inspector determined that the licensee maintained and
implemented excellent radioactive liquid and gaseous effluent control programs with
capabilities to protect the public health and safety and the environment. Furthermore, the
licensee management support to implement the effluent ALARA program was noteworthy.
l l l i l ,
_ _ _ ._ __
,
.
25
R1.2 Implementation of the Radiological Environmental Monitoring Program
a. Insoection Scone (84750-02)
The Radiological Environmental Monitoring Program (REMP) was inspected against Sections
3/4.12.1 and 3/4.12.2 of the Technical Specifications (TS) and Regulatory Guide 4.1,
" Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants." The
following activities were conducted to assess the licensee's capability to implement the
program.
-
Review of REMP procedures and ODCM changes which pertain to REMP;
-
Review of the land use census results;
,
-
Review of sample results to confirm sample frequency and impact of the plant on
the environment:
-
Assessment of the method for evaluating the results of the samples
-
Observation of personnel collecting samples from selected sampling locations;
-
Examination of air sampling equipment relative to function, operability, and
calibration;
-
Review of the calibration process for the dry gas test meter, including the
secondary standards used to calibrate the dry gas test meter; and,
-
Review of results of prevailing wind determination for the last ten years to assess
any significant changes since pre-operation to the present.
b. Observations and Findinos
The licensee's sampling procedures, including the contractor's (Normandeau Associates)
sampling procedures, contained appropriate information and methods comparable to
industry standards and good practices. The inspector observed the licensee and contractor
personnel collect milk samples and exchange air particulate filters and charcoal canisters
from selected air samplers, and discussed certain sample techniques not observed.
Sampling procedures and practices were intended to minimize the chances of cross
contamination. Samples were collected from the locations and at the frequencies required
by the TS and OOCM. The analytical results demonstrated that the types and frequencies
of analyses were performed as required. The inspector noted that radiological dose to the
public was in conformance with the technical specifications. To enhance the data sourco
j of the environmental monitoring program, the licensee continued to collect and analyze i supplemental samples in addition to those required by the regulatory requirements. ! l l
e
.
26
The 1996 land use census was performed by September 1996, according to the procedure
and the TS requirement. Performance of the land use census was thorough and complete.
No program changes (e.g., changes in sample locations) were required as a result of the
census.
The inspector reviewed the wind direction assessments (wind roses) from the past ten
years and compared them to the pre-operational wind roses to detect changes, if any, in
- the prevailing wind directions. No significant changes were determined. The
, environmental monitoring control stations are still valid.
The inspector noted that the licensee established an audit program to assess results and maintain oversight of the contractor laboratory. The Laboratory Quality Control Audit Committee (LOCAC) is a combined effort of technical specialists from the five sponsor
, utilities (Seabrook, Pilgrim, Vermont Yankee, Maine Yankee, and Yankee Rowe). The
LOCAC audit was comprehensive, covering topics such as equipment calibration and
analytical procedures. Details of this audit are documented in Section R7.2 of this report.
This audit program appears to effectively assess the quality of the YAEL performance.
The air samplers were in operation and good physical condition. The licensee maintained a
maintenance program to minimize the amount of sample loss due to mechanical failure.
Each unit was inspected for general function every week and carbon vanes, poly-tubing,
and power cords were replaced every 12 months. The licensee had a calibration program
to ensure validity of samples collected. Once per year, both types of air samplers (Kurz
and dry gas meters) were calibrated with a dry gas test meter, which is in turn calibrated
every 3 years. The results of the calibrations were within the established acceptance
criteria.
l
Calibration of the Jry qas test meter (DGTM) was typically contracted to a vendor.
However, when the calibration came due, the licensee learned that the vendor would no
longer perform calibration services; and as a result, that due date of December 1996 may
be exceeded. The licensee decided to perform the calibration in-house by the Measuring )
and Test Equipment (M&TE) laboratory. M&TE created a procedure utilizing the vendor i
manual and calibrated the DGTM by April 1997, within the 25% grace period allowed to
calibrate the instrument. The licensee did not calibrate any air samplers until the DGTM
had been calibrated. No air samplers exceeded the calibration due dates. The M&TE
laboratory esta'olished knowledge in this area through review the calibration process.
Tracofft, to National Institute of Standards and Technology (NIST) was demonstrated as
evidenced by in,. calibration certificates of the primary standards (manometer and pressure
gauge) used to calib, we the MKS Califlow System, which is used to calibrate the DGTM.
Technicians' knowledge aad understanding of the above areas was very good. The
inspector reviewed training 'nd qualification records of personnel responsible for certain
REMP duties. The techniciart successfey completed the qualification and training
programs provided by the trairmy, der'stment. The training program was comprehensive
and contained sufficient detail to perform REMP duties effectively.
&
l l* 1
27
l c. Conclusiq.nji
Based on the above review, observation, and discussions, the inspector determined the
l licensee's performance in implementing the REMP continued to be excellent.
R1.3 Meteorological Monitoring Program (MMP)
a. Insoection Scooe (84750-02)
The Meteorological Monitoring Program (MMP) was inspected against TS
Section 3/4.3.3.4, UFSAR Section 2.3.3.3 and Regulatory Guide 1.23 commitments. The
l
following activities were conducted to assess the licensee's ability to implement the
program.
-
Review of calibration procedures, calibration results,$nd channel check logs;
-
Review of calibration results of individual sensors:
! -
Discussion of data acquisition and availability of data: 1
l -
Observation of the material condition of meteorological equipment; and,
-
Review of the normal and backup power supply to the primary tower and test
verification.
b. Observations and Findinas
Calibration results from July 1995 through December 1996 were reviewed. Calibration of
the meteorological system (control room to sensors) was performed, in accordance with
,
applicable procedures, at a quarterly frequency (i.e., more often than semiannual, as
l
required by TS). Equipment tolerances were verified as described in the procedure and the
l UFSAR. Individual wind speed sensors were periodically tested and performance-verified
by a vendor semiannually. The results were verified to be within the required tolerances.
l
Channel checks were performed every shift by operations personnel, more frequently than I
daily frequency required by TS. The inspector selected and reviewed the shift log from
April 1-24,1997, and noted that the channel checks were performed every shift during
that time period.
!
The inspector noted that the instrumentation used by Instrument & Controls (l&C) to
calibrate the meteorological system were properly verified and validated. The system was
checked by l&C for operability and system performance, daily. The inspector also noted
that the strip chart recorders were in good physical condition.
The normal and backup power supplies to the primary meteorological tower was verified by
the inspector. The inspector reviewed schematics 1-NHY-310103 and 1-NHY-310104 to
independently determine the electrical power sources for the meteorological tower
instrumentation. The inspector determined the normal and backup power supply to the
e
'. I 28
instrumentation is from the emergency distribution system through the motor control
center (MCC) E-523, and the gatehouse essential lighting panel, ED-MM-212C, EL32.
Normal power is supplied by bus E5, a vital bus, supplied by the Unit Auxiliary Transformer
(UAT). Backup power is supplied by the Reserve Auxiliary Transformer (RAT) or, if the
RAT fails, the Train "A" diesel generator. The inspector determir:ed that the load shed was
verified immediately prior to a refuel outage (every 18 months) as part of the surveillance
test procedure, EX 1804.001, " Diesel Generator 1 A 18 Month Operability and Engineered
Safeguards Pumps and Valve Response Time Testing Surveillance", Rev. 7, dated, May 07,
1997. The inspector reviewed the procedure and noted that the load shed tests were
performed immediately prior to the cast two outages and the results were satisfactorily
completed and verified, as per Section 4.8.1.1.2.f.4.a of the TS.
c. Conclusion
Based on the direct observations, discussions with personnel, and examination of
procedures and records for calibration of equipment, the inspector determined that overall,
the licensee's performance of maintaining and calibrating the meteorological monitoring !
instrumentation was very good. The data were available as required and were easily l
accessed from several locations, including the control room and the EOF as specified in the
UFSAR. The l&C performance in this area was demonstrated through a good
understanding of the meteorological instrumentation based on qualification, training, and
l
experiences. The backup power is supplied from the station's Train "A" diesel generator to ;
the equipment at the meteorological tower as described in Section 2.3.3 of the UFSAR. j
R1.4 Refueling Outage Radiological Controls-External Exposure l
a. Insoection Scone (83750-02)
The inspector reviewed the licensee's control of external exposures. Information was
gathered through observation of activities, tours of the radiologically controlled area (RCA) j
including the containment building (CTB), primary auxiliary building (PAB), fuel storage I
building (FSB), and waste processing building (WPB), discussions with cognizant personnel, I
and review and evaluation of procedures and documents.
b. Observations and Findinas
A review of the radiation work permits (RWPs) and the observed radiological work
activities indicated that a large amount of advanced planning for the outage by health
physics (HP) had taken place. RWPs and a schedule of outage activities requiring HP
coverage were posted at each control point. The inspector noted visible and active HP
coverage within the RCA. A review of the personnel exposure status report, which was
current as of May 21, indicated that the highest cumulative individual exposure for 1997
l was 373 millirem (mrem). No administrative dose limits had been exceeded. The badge ! records for six selected individuals were exarnined and contained the documentation
required in 10 CFR 20.2104 and 20.2106. There was one declared pregnant worker so far
for 1997 whose radiation exposure was being managed in accordance with 10 CFR
20.1208.
1
,
l
29
1
The effects of the increased leakage of airborne radioactive gases due to fuel cladding '
defects continued after the scheduled shutdown on approximately May 10. There were 56
personnel contamination reports (PCRs) due to noble gas from January through April 1997. i
Through May 20, there had been 241 PCRs due to noble gas for the outage. Due to the l
presence of this increased potential for immersion dose from noble gas, the licensee l
investigated the relative merits of various dosimeters for their noble gas monitoring I
capabilities. This investigation involved the use of test phantoms with various dosimeters j
and thicknesses of protective clothing (PC) and the generation of a Health Physics
Study / Technical information Document (HPSTID-97-007, Technical Evaluation of Noble
Gas Monitoring Capabilities, May 14,1997). Presently, the beta dose to the skin is l
monitored using air sampling analysis and stay-time calculations while the photon dose to i
the skin, the lens dose equivalent, and the deep dose equivalent are tracked on a daily l
basis by electronic dosimeters and monitored on a periodic basis by thermoluminescent '
dosimeters (TLDs). Seabrook Procedure HD0958.05, " Dose Assessment for Noble Gas
Environments," establishes a lower limit of 25 mrem per hour for the tracking and
assignment of net-beta shallow dose equivalent. A licensee evaluation of containment stay
times indicated that workers spend, at most, 33 hours per year in noble gas environments.
HPSTID-97-008, " Nots Gas Skin Dose Assignment for OROS," addressed the individual
doses due to the elevated noble Oas concentrations on May 11 and 12,1997, when the
dose rate peaked at about 28.2 mrem per hour (7.87 derived air concentrations (DACs)).
Of the 287 workers who were in the CTB during the elevated gas conditions,11
individuals received greater than 100 mrem beta skin dose (the maximum individual beta
skin dose was 187 mrem). Accordingly, this beta skin dose will be assigned to the
workers in accordance with the licensee's procedures.
The licensee identified one incident, ACR No. 97-1169, in which contracted workers
inadvertently performed backseating on reactor coolant pumps (RCPs) A, B, and C without
HP coverage as required by the RWP. Previously, the workers had performed the same
task on RCP D with HP coverage. A review indicated that the HP coverage requirement
was misinterpreted by the workers, that there were no adverse radiological consequences,
and that clear communication of HP coverage requirements during RWP briefings needed to
be emphasized. An HP supervisor met with both the HP technician who provided the initial
coverage on RCP D and the contracted supervisor to review the incident and to discuss
RWP compliance in the future. The failure of the contracted workers to follow licensee
procedure and to adhere to the RWP and to notify HP of the need for coverage for the
additional work is a violation of NRC requirements. Accordingly, this licensee-identified
and corrected violation is being treated as a Non-Cited Violation, consistent with Section
Vll.B.1 of the NRC Enforcement Policy.
c. Conclusions
Positive performance was evident in the well-managed handling of the anticipated elevated
airborne radioactivity levels due to operation with fuel cladding defects. The licensee
effectively investigated the relative merits of various dosimeters for their noble gas
monitoring capabilities. A licensee-identified and corrected violation involving adherence to
a specific RWP requirement was noted and determined to be effectively resolved.
_ .__ __ _. _ _. ___ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
,
.
30
R1.5 Refueling Outage Radiological Controls-Internal Exposure
a. insoection Scooe (83750-02)
The inspector reviewed the licensee's control of internal exposures. Information was
gathered through observation of activities, tours of the RCA including the CTB, PAB, FSB,
and WPB, discussions with cognizant personnel, and review and evaluation of procedures
and documents.
b. Observations and Findinas
The inspector determined that the licensee does not track DAC-hours. Justification for
discontinuing DAC-hour tracking was provided in HPSTID-93-017, " Evaluation of the Need
For Internal Monitoring at Seabrook Station." The justification was based on historical
bioassay data and measurements of airborne radionuclides, which showed that annual -
intakes in excess of 10% of the applicable annual limits on intake were unlikely to be
received. No dose had to be assigned as a result of internal exposure per regulatory
requirements. The inspector rated that numerous air samples were being taken before and
-
during radiological work activiues, were being used for decisions on personnel access to
airborne radioactivity areas, and were being used for area posting purposes. Whole body
counts (WBCs) are performed for individuah before initial RCA access and before
termination of site access, after contamination events, and on a random basis (4 or 5 -
radiation workers per month). Additionr.4ly, the licensee's IPMs (Installed Personnel
Monitors)/whole body frisking booths are used as passive whole body counters. HPSTID-
93-014, " LPM Frisking Booth Sensitivity to inhaled Radioactivity," documented the IPM's
sensitivity (DAC-hours / alarm) for an acute exposure to typically expected mixtures of
radionuclides as being 15 to 25 DAC-hours [37 to 63 mrem, committe effective dose ,
equivalent (CEDE)). The licensee's procedure, HD0961.29, " Internal Dosimetry
Assessment," required that internal dose be assigned to individuals if the CEDE is greater
than or equal to 10 mrem. The licensee reported that there had been no recordable (10
millirem) internal exposures thus far in 1997. Based on this information, the licensee was
in compliance with 10 CFR 20.1204, 20.1501, and 20.1502 as regards surveys,
monitoring, and determination of internal exposure.
c. Conclusions
The licensee's control of internal exposures was managed in a satisfactory and competent
fashion.
R1.6 Refueling Outage Radiological Controls-Radioactive Materials, Contamination,
Surveys, and Monitoring
a. Insoection Scone (83750-02)
l l The inspector reviewed the licensee's control of radioactive materials, contamination, l
surveys, and monitoring. Information was gathered through observation of activities, tours
of the RCA including the CTB, PAB, FSB, and WPB, discussions with cognizant personnel,
( and review and evaluation of procedures and documents, r i
~
.
.
31
b. Observations and Findinos
Source term, external to the reactor vessel, and radiation levels were higher this outage
than for previous outages. The licensee reported that some of the reasons for this were
the length of the operating period (463 days), the additional crud burst due to a reactor trip
at 8% power, and fuel cladding defects. The dose rates in the reactor coolant inner loop
areas were twice the levels encountered in past outages, and dose rates in the steam
generator bowls were expected to be higher than in the past. Good communication and
cooperation between HP and operations after shutdown led to successful efforts to remove
crud and hot spots in piping and system components where it had not been seen in the
past. These efforts led to lower dose rates in the areas affected, including the residual 1
heat removal (RHR) was. l
1
!
A modification was performed prior to reactor head removal, which allowed venting of the l
air space beneath the reactor head, and this alleviated a potential for significant airborne .
radioactivity upon lifting of the reactor head. Elevated noble gas (up to 8 DACs) and iodine 1
(a few % up to 20% of a DAC) airborne levels were initially encountered in containment
after shutdown due to leaks in valving around the pressurizer relief tank. Containment l
remained posted as a Caution-Airborne Radioactivity Area until May 21. As of May 23,-
the airborne lewla of noble gas and iodine on the refueling deck in containment had
decreased to less than 2% of a DAC, each. The licensee stated that containment air was
being exhausted through a filtered system [11,000 cubic feet per minute (CFM)] rather
than through the unfiltered (40,000 CFM) system until the lodine-131 in containment
decreased to the point at which the projected dose in 31 days for continuous unfiltered
release would be less than 0.03 mrem per month to any organ of a member of the public
(10% of a Technical Specification requirement). The licensee was performing numerous air
sampling operations throughout each day and demonstrated vigilant oversight of this
situation.
On May 17, there were two incidents involving contractors and their failure to maintain
high radiation area (HRA) barriers (ACR Nos.97-1091 and 97-1139) in accordance with
licensee technical specifications and procedure, and one incident involving contractors and
their failure to follow licensee procedure and to adhere to a posted requirement to notify
HP prict to entry into a truck containing packaged radioactive material (ACR No. 97-1099).
These incidents were identified by the licensee, resulted in no significant adverse
i radiological consequences, and resulted in immediate and comprehensive corrective actions
(including temporary denial of RCA access), review and evaluation of the incidents,
meetings with the individuals involved and their management, a site-wide news article, and
discip!inary actions. These failures of the contracted workers to adhere to a posted
requirement to notify HP prior to entry into the truck and to maintain HRA barriers are
violations of NRC requirements. These licensee-identified and corrected violations are
being treated as Non-Cited Violations, consistent with Section Vll.B.1 of the NRC
Enforcement Policy.
.
.
32
c. Conclusion _g
Efforts to remove radioactive crud and hot spots from locations not required to be flushed
in the past were successful due to good cooperation between HP and operations. The
licensee also had been pro-active in their efforts to minimize the potential for high airborne
radioactivity due to the reactor head lift. Licensee-identified and corrected violations
involving adherence to HRA and RCA access and control procedures were noted and
determined to be effectively resolved.
R1.7 Refueling Outage Radiological Controls-As Low As Reasonably Achievable (ALARA)
l
a. Insoection Scone (83750-02) !
The inspector reviewed the licensee's ALARA activities. Information was gathered through
observation of activities, tours of the RCA including the CTB, PAB, FSB, and WPB,
discussions with cognizant personnel, and review and evaluation of procedures and
documents.
l
b. Observations and Findinas
During tours of the RCA, temporary shielding (lead blankets) was evident in numerous
areas, especially where workers would be required to spend a significant portion of their
time and where dose rates were significantly greater than the average general area dose -
rates. The inspector witnessed an ALARA pre-job briefing for removal of a stuck stud on
the reactor vessel flange (5.3 person-rem goal). The inspector noted that there was good l
planning and preparation for this evolution, requiring significant interaction between l
contractors and site personnel from several different departments. The cutage person-rem
goal appeared to a challenging one, especially considering the effects of the additional crud
burst and fuel cladding defects on dose rates. As of May 23, the actual and projected
outage doses were 34.8 and 33.9 person-rem, respectively.
c. Conclusior.s l
A number of positive attributes were noted in this area including the beneficial use of
temporary shielding in containment, the extent of planning and preparation invested in the
stuck stud task, and the setting of a challenging outage person-rem goal. The ALARA
program appeared to be effective in maintaining occupational radiation exposures as low as
reasonably achievable.
R1.8 Other Changes to the RP Program
a. Insoection Scoce (83750-02)
The inspector reviewed the licensee's RP program for changes since the last inspection.
Information was gathered through observation of activities, tours of the RCA including the
CTB, PAB, FSB, and WPB, discussions with cognizant personnel, and review and
evaluation of procedures and documents.
- .=- - -- - - - . - - - . - - . - - - - - - - - . -
.
33
b. Observations and Findinas
Approximately 5 weeks prior to this inspection, the HP organization lost a health physicist j
through resignation of employment. However, at the time of this inspection, there were
three contracted health physicists and two individuals from site training providing full-time
support to the HP group. There appeared to be sufficient health physicist staffing and
resources in HP during this inspection. To be prepared for a potential significant increase
in outage scope (tube inspection work in 4 rather than just 2 steam generators), the
licensee added twelve health physics technicians to the staff.
The scrub program (alternate type of protective clothing for prescribed types of work) had
been implemented since the last inspection. Also, TLD bar code readers had been installed
at the HP access control points. This change provided greater control of RCA access.
Other program improvements included the purchase of two Merlin Gerin AM-16 area
monitors and of four AMS-4 Eberline portable continuous air monitors, the use of cameras
in the RHR vaults to reduce exposure, turbine building HP controls to verify the radioactive l
contamination status of secondary systems, automated incremental neutron dose tracking
in the computerized access control system, and generation of a user / technical manual for
the e.ccess control system.
c. Conclusions
The HP organization appears well supported by licensee management and is continuing to
improve the radiation protection program.
R2 Status of RP&C Facilities and Equipment
R2.1 Calibration of Effluent / Process Radiation Monitoring Systems (RMS)
a. Insoection Scooe (84750-01)
The inspector reviewed the most recent calibration results for the following selected
effluent / process RMS and its system flow rates. The inspector also reviewed the
licensee's RMS self-assessment and RMS work orders.
* Waste Liquid Test Tanks Radiation Monitor (R-6509)
* Waste Liquid Test Tanks Flow Rate
* Steam Generator Blowdown Flash Tank Radiation Monitor (R-6519)
* Turbine Building Sump Pump Radiation Monitor (R-6521)
i
a Main Steam Line Radiation Monitors (R-6481 1&3, R-6482 2&4)
! * Primary Component Cooling Water Radiation Monitors (R-6515 & 6516)
* Containment Purge Radiation Monitors (R-6527 A & B) ,
o Containment Purge Line Flow Rate l
( * Plant Vent Wide Range Noble Gas Monitor (R-6528) l .
* Plant Vent Exhaust Flow Rate
-
* Condenser Air Evacuators Discharge Monitor (R-6505)
,
* Waste Gas Compressor inlet Radiation Monitor (R-6503) ,
,
* Waste Gas Compressor Discharge Radiation Monitor (R-6504) !
.- -
-.
. _ _ - - _ __ . _ - - _ __ _ . _ _ _ _ _ _ . __ _ _ - ._
.
34
b. Observations and Findinos
The l&C department had the responsibility to perform electronic and radiological
calibrations for the above radiation monitors. The system engineer had the responsibility to
trend and track the above RMS. All reviewed calibration results were within the licensee's
acceptance criteria.
During the review of the above RMS calibration results, the inspector independently
calculated and compared several calibration results including linearity tests and conversion .
'
factors. The comparisons were very good.
c. Conclusions
Based on the above reviews, the inspector determined that the licensee maintained and
implemented a very good calibration program for the effluent / process RMS.
R2.2 Calibration of Area Radiation Monitoring Systems (ARMS)
a. Insoection Scoce (83750)
The inspector reviewed the most recent calibration results for the following selected ARMS
described in Sections 12.3.4 and 13.5.2.3.(b).3 of the UFSAR:
o PAB - High Range Area Monitors (65081 & 2)
e RHR - High Range Area Monitors (65171 & 2)
e FSR - High Range Area Monitor (6518)
e Personnel Hatch (Post-LOCA) Area Monitors (65361 & 2)
The instrument calibrations were reviewed with respect to:
(1) selection criteria for energy dependence, accuracy, and reproducibility; (2)
calibration method and testability; and (3) alarm set point methodology.
The inspectors also utilized the following documents as a basis to determine
whether the calibration procedures contained sufficient calibration steps to verify
the alarm set points. l
* ANSI /ANS.HPSSC-6.8.1-1981, " Location and Design Criteria for Area Radiation
Monitoring Systems for Light Water Nuclear Reactors,"
e ANSI N42.3-1969, IEEE No. 309, "American National Standard and IEEE Standard
Test Procedure for Guiger-Muller Counters," and
e ANSI N323-1978, "American National Standard Radiation Protection
- instrurnentation Test and Calibration."
i I 1 i L -
- . _ - _ . _ _ . .. -- .-. . . . - . - - .. . . . - - - . - .
.
35
b. Observations and Findinas
The licensee applied very good calibration methodologies for the above ARMS, including
l radiological and electronic calibrations. Alert and alarm setpoints calculation
methodologies were good. The licensee also emphasized maintaining minimum background
l
radiation level during the calibration which was excellent. Calibration procedures were
detailed and easy to follow for all necessary steps.
*
c. Conclusions
Based on the above reviews, the inspector determined that the licensee implemented a
very good ARMS calibration program.
R2.3 Air Cleaning Systems ;
i
l a. Insoection Scoce (84750-01)
The inspector reviewed the licensee's most recent surveillance test results (visual
inspection, in-place HEPA and charcoal filter leak tests, air capacity, pressure drop tcot;,
and laboratory tests for the iodine collection efficiencies) for the following systems:
o Containment Purge Exhaust System
o Primary Auxiliary Building Exhaust System
o Fuel Storage Building Exhaust System (Trains A & B)
b. Observations and Findinas
All reviewed test results were within the licensee's TS acceptance criteria. During the
previous inspection conducted in January 1996, the inspector recognized that the
responsible individual had very good knowledge not only for TS requirements, but also for
standard industry practices. The inspector determined that the licensee maintained and
implemented a very good air cleaning system surveillance program. !
c. Conclusions
Based on the above reviews, the inspector determined that the licensee maintained and
implemented a very good air cleaning system surveillance program.
R3 RP&C Procedures and Documentation
a. Insoection Scope (84570-01)
!
l The inspection consisted of: (1) review of selected chemistry procedures to determine
whether the licensee could implement the routine radioactive liquid and gaseous effluent
control programs and the emergency operations; (2) review of 1995 and 1996 Annual i
i Radioactive Effluent Reports to verify the implementation of TS requirements; and (3) l
review of the contents of the ODCM for performing the effluent control programs, including
i projected dose calculation methodologies to the public. I
-. - ., __
_- - ... _ ._. . - - - - . - - - _ - - _ - - - - - - -
!
.
l 36 , I b. Observations and Findinas i ! !
The inspector noted that reviewed effluent control procedures were detailed, easy to ,
follow, and ODCM requirements were incorporated into the appropriate procedures. The l
l licensee had good procedures to satisfy the TS/ODCM requirements for routine and l
emergency operations.
~
l l
The inspectors reviewed the 1995 and 1996 Annual Radioactive Effluent Release Reports.
l These reports provided data indicating total released radioactivity for liquid and gaseous
effluents. The annual reports also summarized the assessment of the projected maximum i
j individual and population doses resulting from routine radioactive airborne and liquid s
effluents. Projected doses to the public were well below the Technical Specification (TS) -{
limits. The inspectors determined that there were no anomalous measurements, omissions '
or adverse trends in the reports. ;
!
The ODCM provided descriptions of the sampling and analysis programs, which were ;
established for quantifying radioactive liquid and gaseous effluent concentrations, and for
calculating projected doses to the public. All necessary parameters, such as effluent
- radiation monitor setpoint calculation methodologies, site-specific dilution factors, and dose
factors, were listed in the ODCM. The licensee adopted other necessary parameters from
l Regulatory Guide 1.109.
c. Conclusions
l
Based on the above reviews, the inspector made the following determinations:
* effluent control procedures were suffic;3ntly detailed to facilitate performance of all
necessary steps for routine and emergency operations,
e the licensee effectively implemented the TSiODCM requirements for reporting
effluent releases and projected doses to the public, and
* the licensee's ODCM contained sufficier,t specification, information, and instruction
to acceptably implement and maintain tie radioactive liquid and gaseous effluent
control programs.
l R5 Staff Training and Qualification in RP&C
a. Insoection Scooe (83750-02_1
The inspector reviewed the qualifications and training of selected contracted HP
technicians. Information was gathered through observation of activities, discussions with
l cognizant personnel, and review and evaluation of documents. i l l . !
_ _ __ _
.
.
37
b. Observations and Findinos
The documented qualifications of the senior HP contracted technicians were reviewed and
were found to meet the technical specification requirements. The training of these
technicians at the site was in compliance with site procedures.
c. Conclusions
The qualifications and site training of the HP technicians contracted for the outage were
satisfactory.
R6 RP&C Organization and Administration -
a. Insoection Scope (84570-01)
The inspector reviewed the organization and administration of the radioactive liquid and
gaseous effluent control programs and discussed with the licensee changes made since the
last inspection, conducted in January 1996. The inspector also reviewed the management
-
support for the effluent control program through interviews and implementation of the ;
'
effluent ALARA program.
b. Observations and Findinos i
)
There were no changes since the last inspection of the effluent control programs. The l
chemistry department had the major responsibility to conduct the effluent control l
programs. Other groups (i.e., HP, operations, I&C, and system engineers) had supporting
responsibilities to the program. Staffing levels appeared to be appropriate for the conduct i
of routine and emergency operations.
The inspector interviewed the chemistry staff and management for the importance of the
effluent control programs and implementation of the effluent ALARA program. All
chemistry staff demonstrated very good knowledge of: (1) TS/ODCM requirements and its
bases, (2) projected dose calculation, and (3) implementation of the TS/ODCM during
normal and emergency operations. Management also demonstrated knowledge in these
areas and supported the radioactive liquid and gaseous effluent control programs (see
Section R1 of this inspection report).
c. Conclusions
Staffing levels appeared to be appropriate to conduct routine and emergency operations.
As aforementioned, management support of the effluent ALARA program was noteworthy.
I
!
~. - .- -. - - .- - - _ .. _ - -. - ..-- . . .... - . . - - .
3..-~-.-~
:
,
l. ;
;
- '
l
'
38
R6.2 Management Controls !
!
a. Insoection Scooe (84570-02) -
.
The inspector reviewed organization changes and the responsibilities relative to oversight
of the REMP and MMP, and the annual radiological environmental operating report to verify
the implementation of the TS. l
b. Observations and Findinas
'
The organization changes did not appear to have a negative impact on the oversight of the
REMP. The responsibilities relative to oversight of the REMP and MMP have essentially
.
remained the same.
l The annual radiological environmental monitoring reports for 1995 and 1996 (draft) )
provided a comprehensive summary of the results of the REMP around Seabrook and met i
the TS reporting requirements. No omissions, mistakes, or obvious anomalous results and
trends were noted.
l , ,
c. Conclusion
Based on the above review, the inspector determined that the licensee implemented good
management control and oversight of the REMP and MMP and effectively implemented the
TS requirements.
L R7 Quality Assurance in RP&C Activities j l l l a. Inspection Scone (84750-01) - t ! l l l The inspection consists of: (1) review of the 1996 audit and its responses, if any; (2) QA !
policy of the measurement laboratory; and (3) implementation of the measurement
laboratory QC program for radioactive liquid and gaseous effluent samples.
l b. Observations and Findinas l
The inspector reviewed QA Audit Report No. 96-A10-02. The inspector noted that the
audit team also included other technical personnel. The 1996 auait team identified one
finding and five observations / recommendations. The finding or observations were not !
safety-related, rather to enhance effluent control programs. The responses to these ,
findings and observations were completed in a timely manner. The inspector noted that l
the scope and technical depth of the audit was very good in assessing the radioactive
liquid and gaseous effluent control programs.
.
The licensee maintained very good QA policy and implemented the policy throughout the
! chemistry department, including the analytical measurement laboratory. The QA/QC ; ! program for analyses of effluent samples is conducted by Yankee Atomic Environmental , i Laboratory (YAEL). The YAEL has interlaboratory and intralaboratory QC programs. The l l QC program consists of measurements of blind duplicate, spike, and split samples. The ;
I
w - + - > p- e y-t 7 -P +- 4 --7-
._ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ . . _ _ . _ . _ . _ . . _ _ _ _
7 __ .. . _ . ___
_ -_
4
39
YAEL published a QC report quarterly, semiannually, and annually. The inspector reviewed
l the QC data of the Seabrook station for intra /interlaboratory comparisons listed in the
quarterly OC reports and noted that the OC data were within the YAEL's acceptance
criteria. The inspector also noted that the YAEL evaluated analytical results for accuracy
and precision, which was an excellent effort for validating analytical results. When
discrepancies were found, the responsible licensee consulted the YAEL personnel and
reasons for the discrepancies were investigated and resolved.
c. _Csnclusions
Based on the above reviews, the inspector determined that the licensee continued to
conduct excellent QA and OC programs for the radioactive liquid and gaseous effluent
control programs,
a. Inspection Scooe (83750-02) i
The inspector reviewed the licensee's quality assurance (QA) activities for their TLD
vendor's ability to accurately report radiation dose. Information was gathered through
discussions with cognizant personnel and review and evaluation of documents.
b. Observations and Findinos
The vendor's National Voluntary Laboratory Accreditation Program (NVLAP) certification -
was available and stated that, in accordance with American National Standards Institute
l (ANSI) Health Physics Society (HPS) N13.11-1993, the UD808 and UD814 badges were ! each qualified for Categories 1,11, Illa, IllB, IV, VA, VB, VC, VI, and Vil, and that the
combination of the two badges was qualified for Category Vill. The licensee performed
blind spiking of the TLDs in the first quarter of 1997 using Cesium-137. The average
reported dose was within approximately 11% of the delivered dose. This result was
acceptable for the radiation type tested.
c. Conclusions
The licensee's OA activities for their TLD vendor's ability to accurately report radiation
dose was acceptable for the radiation type tested.
R7.1 Quality Assurance Audit Program
' !
a. Insoection Scoce (84750-02)
[ The following Quality Assurance (OA) audit reports were reviewed against Section 6.4.3.8 l of Technical Specifications. l l -
Nuclear Safety and Assessment (NSA) Audit Report No. 95-A09-01, " Radiological
Effluent Technical Specification (RETS), Radiological Environmental Monitoring
Program (REMP), Offsite Dose Calculation Manual (ODCM)."
,
. - - - .- -
I*
- .
i
40
-
Nuclear Safety and Assessment Audit Report No. 96-A10-02, "RETS, REMP,
ODCM."
The 1996 Laboratory Quality Control Audit Committee (LOCAC), Audit Report was
reviewed to evaluate the licensee's oversight and assessment of the contractor laboratory,
b. Observation and Findinas
The inspector noted that the licensee conducted the NSA audits according to TS. The
licensee used the audit procedure, OP AE 3.0, as guidance to conduct the audits. The
scope of the audits reflected the environmental monitoring program requirements. All
aspects of the scope were completed. The inspector noted that the audit team leader
utilized a technical specialist to assess the REMP. The technical specialist (auditor)
reviewed and understood the TS, ODCM, and the pertinent program procedures. The
auditc was familiar with sampling and analytical practices and observed collection of
certain samples and reviewed the results obtained by the analytical laboratory. The
inspector determined that the auditor's recommendations, were appropriate and were
suggested to refine the REMP. Followup of the 1995 findings was complete for the QA
Assessment audits and was documented in the 1996 audit report.
The LOCAC audit was a joint effort audit conducted by members from the five sponsor
utilities to evaluate the programs of YAEL. During this audit, each member of the audit
i team selected an area of interest and performed a probing investigation in the area. The
areas of interest were analytical procedures, laboratory practices, quality assurance and
quality control program. The findings were appropriate and reasonable. The responses,
when required, were timely and appropriate. Followup for the LOCAC audit is not yet
complete.
c. Conclusions
Based on the review of the audits and discussions with the auditors, the inspector
concluded that the audits were of sufficient technical depth to effectively identify and
assess program strengths and weaknesses. The audits evaluated the technical adequacy
of implementing procedures, TS requirements, and practices. Performance of the audits by
the audit teams was thorough, objective and of high quality as evidenced by the report
documentation.
R7.2 Quality Assurance of Analytical Measurements
a. insoection Scooe (84750-02)
l The inspector reviewed the Quality Assurance (OA) and Quality Control (OC) programs i against Section 3/4.12.3 of the TS and recommendations of Regulatory Guide 4.15, i
" Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" to determine whether the licensee had adequate control witn respect to sampling, analyzing, and evaluating data for the implementation of the REMP.
i
- _ _ _ _ _ _ - - _ _
.
41
b. Observations and Findinas
The performance of the contractor laboratory, Yankee Atomic Environmental Laboratory
(YAEL), continued to be excellent. During an inspection at Millstone, the inspector visited
the laboratory and assessed the quality assurance program. See Section R7.2 of the
Combined Inspection Report Nos. 50-245/96-09, 50-336/96-09, and 50-423/96-09 for
details.
The YAEL implemented an interlaboratory comparison program, required by the TS,
through continued participation with Environmental Protection Agency (EPA) drinking water
program and the program provided by Analytics, Incorporated. The inspector reviewed the
analytical results of this program and noted the results were within the established
acceptance criteria.
c. Conclusion
Based on the above observations, the inspector determined that the performance of the
contract laboratory was excellent and the interlaboratory program was effective.
R8 Miscellaneous RP&C lssues
R8.1 Review of Updated final Safety Analysis Report (UFSAR) Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the Updated
Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused
review that compares plant practices, procedures, and/or parameters to the UFSAR
descriptions.
While performing the inspections discussed in this report, the inspector reviewed the
applicable portions of the UFSAR that related to the areas inspected. The inspector
verified that the UFSAR wording was consistent with observed plant practices, pacedures,
and/or parameters.
R8.2 (Closed) LER 50-443/96-009-00: Missed surveillance PCCW rate of change monitor
alarm.
Channel calibration of the Primary Component Cooling Water (PCCW) head tank rate of
change monitor is required on a refueling interval by the TS. A complete calibration of the
PCCW head tank rate of change monitor was not performed on the required frequency.
This LER is not considered safety significant because the radiation monitor had been
calibrated and alarms have been functional. The licensee promptly corrected this event.
The inspector determined that, at the time of the event, missed surveillance PCCW rate of
change monitor alarm, constituted a violation of TS Section 4.3.3.9, Table 4.3-5, Item 4a.
This licensee-identified and corrected violation is being treated as a non-cited violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-443/97-03-06)
l
l
_ -- ----
e
42
R8.3 (Closed) (URI 50-443/97-002-01) LER 50-443/97-005-00: Misposition of Main
Steam Line Radiation Monitors.
On March 14,1997, the licensee discovered that the four main steam line radiation
monitors were positioned downstream of the atmospheric steam dump valves (ASDV)
contrary to the descript!on in UFSAR Section 11.5.2.1. On the sarne day, the radiation
monitors were declared inoperable and appropriate actions were initiated. On March 20,
the licensee repositioned radiation monitors (consistent with the UFSAR) and declared them
operable. The licensee promptly corrected this event.
The inspector reviewed the most recent calibration results for four main steam line
radiation monitoring systems (see R2.1 of this inspection report). All calibration results
were within the licensee acceptance criteria.
The inspector also toured the main steam line radiation monitors and noted that the new
location of the radiation monitors was about 2 feet upstream of the ASDV. It should be
noted that the radiation monitors had capabilities to monitor at the old location
(downstream of the ASDV). This licensee-identified and corrected violation is being treated
as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev.
This item is closed. (NCV 50-433/97-03-07)
S1 Conduct of Security and Safeguards Activities
S1.1 General Comments (71707, 71750)
The inspectors observed security force performance during inspection activities. Protected
area access controls were found to be properly implemented during random observations.
Proper escort control of visitors was observed. Security officers were alert and attentive
to their duties.
S1.2 Uncontrolled Vehicle in Protected Area
a. Indoection Scope
The inspector conducted a routine inspection of the station physical security program by
verifying the all vehicles in the Protected Area were locked with the keys removed when
unattended.
b. Insoection Findinas
On May 22, the inspector identified an uncontrolled designated vehicle in the Protected
Area during a routine inspection tour. The unattended vehicle had the keys in the ignition
and the engine running. The inspector notified a security officer of the condition and the
individual took possession of the vshicle keys after stopping the engine. The security
officer found the operator of the vehicle out of sight in the vicinity of the vehicle while the
vehicle was being off loaded.
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Station management directed that an ACR be initiated to document the finding and to
provide recommendations for corrective actions. After the completion of the immediate ;
corrective action, which was to take possession of the vehicle keys by a security person,
several other corrective actions were implemented. These included attaching a
" cautionary tag" on the vehicle key ring by security personnel to aid the vehicle operators :
concerning the duties and responsibilities of the while being an authorized operator. Also ,
l the vehicle operator was coached and counselled by a security department supervisor ; '
concerning the event.
The Seabrook Station Physical Security Plan, requires that procedures be developed and l
implemented to control vehicles inside the station Protected Area. Security Department !
Instruction, SDl002.00, " Control of Vehicles", requires in part, that all Licensee Designated ;
Vehicles (LDV"s) when unattended must have the ignition locked, keys removed from the !
ignition and controlled by an authorized person. Contrary to the above, on May 22, at
approximately 10:30 am, the inspector found LDV 16-02 unattended, with the keys in the
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ignition, and the engine running and not controlled by an authorized person. This is a j
l violation of NRC requirements and the Station Security Plan. (VIO 50-443/97-03-08) ! l l c. Conclusions } l
!
The inspector determined that the licensee promptly took control of the LDV when notified !
by the inspector. The licensee's response was prompt and provided adequate immediate l
l corrective actions. However, in view of this being the second NRC identified violation ;
regarding the control of LDV's in the past year. !
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V. Mananement Meetinas
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X1 Exit Meeting Summary l
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- The radiation physicist presented the inspection results to members of the licensee j
management at the conclusion of the inspection on April 25. The licensee acknowledged
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the findings presented.
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The radiation specialist presented the inspection findings to members of the licensee
management at the end of the inspection on May 23, and during a telephone discussion on ,
May 30. The licensee acknowledged the findings. j
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The inspectors covering the inservice inspection presented the results to members of '
licensee management at the conclusion of the inspection on May 23. The licensee ;
acknowledged the findings presented. ,
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The inspectors presented the inspection results to members of licensee management, j
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following the conclusion of the inspection period, on June 25. The licensee acknowledged j
l 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.
X3 Other NRC Activities ;
Conference calls between NRC managers and technical staff specialists and licensee
managers and technical staff leads were performed on the following dates.
,
5/31/97 Fuel Failure Conference Call
,
6/3-5/97 Site Review by Reactor Projects Chief, Branch No. 8, NRC Region I
'
6/16/97 Site Review by Deputy Regional Administrator, NRC Region I
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PARTIAL LIST OF PERSONS CONTACTED
Licensee '
B. Beuchel, Engineering Performance Manager ,
**
W. Cash, HP Dept. Supervisor '
*
R. Cliche, Design Engineering Supervisor
W. Creb, HP Department Manager
M. DeBay, A.O.M.
*
W. DiProfio, Station Director
*
R. Donald, Senior Auditor l
*
B. Drawbridge, Director of Services l
M. Dugan, NES Manager Field Operations
*
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P. Falman, Auxiliary System Engineering Supervisor ,
R. Godbout, Instrument and Controls Supervisor i
*
A. Giotas, Chemistry Supervisor '
*
J. Grillo, Oversight Manager
R. Gwinn, Program Support i
G. Kann, Program Support Supervisor
G. Kline, Technical Support
*
D. Kochman, Senior Engineer, Licensing i
*
J. Kwasnik, Senior Radiation Scientist !
W. Leland, Chemistry and HP Manager
*
R. Litman, Chemistry Department Supervisor
J. Peterson, Maintenance Manager
T. Pucko, Regulatory Compliance Engineer ;
J. Rafalowski, Chemistry and HP Project Supervisor
J. Savold, Senior l&C Technician !
*
J. Sobotka, Regulatory Compliance Manager
E. Soretsky, Technical Projects Supervisor
*
G. St. Pierre, Operations Manager ,
*
B. Roach, Maintenance Manager j
*
L. Tardif, Senior Chemist
M. Toole, instrument and Controls Supervisor
*
L. Walsh, Station Staff
K. Whitney, Senior ISI Engineer ;
(*) Denotes those present at the exit meeting on April 15,1997.
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. Yankee Atomic Nuclear
S. Volk, NDE Level lil
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HEC !
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J. Macdonald, Sr. Resident inspector l
D. Mannai, Resident inspector
W. Olsen, Resident inspector
.
,
D. Silk, !
J. Brand, Reactor Engineer
A. Lohmeier, Senior Reactor Engineer -
J. McFadden, Radiation Specialist
A. DeAgazio, Project Manager ,i
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INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering l
IP 40500: Effectiveness of Licensee Controls in identifying, Resolving, and Preventing :
Problems
IP 61726: Surveillance Observation
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IP 62707: Maintenance Observation
IP 71707: Plant Operations ;
IP 71750: Plant Support Activities
IP 73051: Inservice inspection - Review of Program
IP 73753: Inservice inspection
IP 83729: Occupational Exposure During Extended Outages
IP 83750: Occupational Radiation Exposure
IP 84750: Radioactive Waste Treatment, and Effluent and Environmental Monitoring IP
IP 92700: Onsite Followup of Written Reports of Nonroutine Events at .
Power Reactor Facilities
IP 92902: Followup - Engineering
IP 92903: Followup - Maintenance
IP 93702: Prompt Onsite Response to Events at Operating Power Reactors
ITEMS OPENED, CLOSED, AND DISCUSSED
Ooened:
IFl 50-443/97-03-01 Review root cause of inadvertent steam generator drain-down.
(Section 02.2)
VIO 50-443/97-03-02 Failure to take adequate corrective actions for use of pressure
tubing. (Section M2.1)
IFl 50-443/97-03-03 Review station procedure revision process. (Section E2.1)
IFl 50-443/97-03-04 Review root cause evaluation for LLRT failure. (Section E2.2)
VIO 50-443/97-03-08 Designated Vehicle Left Unattended with Keys in the ignition 1
and Running (Section S1.2). l
Closed:
NCV 50-433/97-03-05 Failure to promptly report LLRT f ailure in accordance with 10
CFR 50.72. (Section E2.2) '
i
LER 96-009-00, Missed Surveillance PCCW Rate of Change Monitor Alarm. This is closed.
l (NCV 50-443/97-03-06) i ! LER 97-005-00, Misposition of Main Steam Line Radiation Monitors. This item is closed.
(NCV 50-433/97-03-07)
- Discussed: None
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( LIST OF ACRONYMS USED
ACR Adverse Condition Report
ALARA As low As is Reasonably Achievable
ANSI American National Standards Institute
ASME American Society of Mechanical Engineers
ARMS Area Radiation Monitoring System
AVB Anti-Vibration Bars
B&PVC Boiler and Pressure Vessel Code
CEDE Committed Effective Dose Equivalent
CFM Cubic Feet per Minute
CFR Code of Federal Regulations
CMMEB Civil, Mechanical, Metallurgical Engineering Branch
CTB Containment Building
DAC Derived Air Concentration i
DRP Division of Reactor Projects
DRS Division of Reactor Safety
E/C Eddy Current
ED Electronic Dosimeter
EPRI Electric Power Research Institute
FSB Fuel Storage Building
HEPA High Efficiency Particulate
HP Health Physics
HPS Health Physics Society
HPSTID Health Physics Study / Technical Information Document
HRA High Radiation Area
IP inspection Procedure
iPM Installed Personnel Monitor
IR Intermediate Range
ISI Inservice Inspection
La Maximum Allowable Containment leakage Rate
LHRA Locked High Radiation Area
LLRT Leak Rate Test
MREM Millirem
MSL Main Steam Line
MT Magnetic Particle Test
NDE Non-destructive Examination
NRC Nuclear Regulatory Commission
NVLAP National Voluntary Laboratory Accreditation Program
ODCM Offsite Dose Calculation Manual
OR05 Refueling Outage 5
PAB Primary Auxiliary Building
PC Protective Clothing
PCR Personnel Contamination Report
PDT Primary Drain Tank
PT Penetrant Test
l OA Quality Assurance
QC Quality Control
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RCA Radiologically Controlled Area
i RCP Reactor Cooling Pump
RCS Reactor Coolant System
RHR Residual Heat Removal
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RHS Reheat System ;
RMS Radiation Monitoring System '
l ROR Radiological Occurrence Report
RP&C Radiological Protection and Chemistry
RSC Radiation Safety Committee
RWP Radiation Work Permit ,
- RWST Refueling Water Storage Tank l
!
SAM Small Article Monitor i
scfh Standard Cubic Feet per Hour '
SG Steam Generator .
TEDE Total Effective Dose Equivalent !
'
TLD Thermoluminescence Dosimeter
TOFD Time of Flight Diffraction j
'
TS Technical Specifications
UFSAR Updated Final Safety Analysis Report
UT Ultrasonic Test
l VCT Volume Control Tank
WBC Whole Body Count l
WPB . Waste Processing Building 1
WRGM Wide Range Radioactive Gaseous Monitor l
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