ML20132G174
| ML20132G174 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/20/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20132G119 | List: |
| References | |
| 50-317-96-08, 50-317-96-8, 50-318-96-08, 50-318-96-8, NUDOCS 9612260194 | |
| Download: ML20132G174 (14) | |
See also: IR 05000317/1996008
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
License Nos.
DPR-53/DPR-69
Report Nos.
50-317/96-08; 50-318/96-08
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Licensee:
Baltimore Gas and Electric Company
Post Office Box 1475
Baltimore, Maryland 21203
Facility:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Location:
Lusby, Maryland
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Dates:
October 20,1996 through November 30,1996
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Inspectors:
J. Scott Stewart, Senior Resident inspector
Henry K. Lathrop, Resident inspector
Fred L. Bower lil, Resident inspector
Dave Silk, Emergency Preparedness Specialist, Region I
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Approved by:
Lawrence T. Doerflein, Chief
Reactor Projects Branch 1
Division of Reactor Projects
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9612260194 961220
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ADOCK 05000317
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EXECUTIVE SUMMARY
Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Inspection Report Nos. 50-317/96-08 and 50-318/96-08
This integrated inspection report includes aspects of BGE operations, maintenance,
engineering, and plant support. The report covers a six week period of resident inspection.
Plant Operations
Operator response to an automatic reactor trip was prompt and very good. The
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operators were effective in stabilizing the plant and ensuring safety. The
emergency operating procedures were accomplished satisfactorily and operators -
appropriately placed the plant in the hot standby condition.
BGE personnel appropriately conducted an event review following the reactor trip
and determined that the event had no consequence to public health and safety. The
cause of the reactor trip was understood and corrected prior to unit restart. The
unit was returned to full power operation without complication,
The inspectors considered the actions of operations personnel and the related
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activities of engineering and maintenance personnel to be both prompt and
appropriate following a control element assembly drop during testing.
During a control room panel walkdown the inspectors noted that power to the 1 A
emergency diesel generator annunciator status panel had apparently been lost.
When informed, operations personnel determined that power was available;
however, the indicating light had burned out. The bulb was replaced and panel
indication was restored to normal.
Very good questioning attitudes and safety perspectives were noted during a
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meeting of the Offsite Safety Review Committee (OSSRC), particularly regarding the
effectiveness of Unit 2 post-trip review, the timeliness of review of failed
surveillance test data by the Plant Operations and Safety Review Committee
(POSRC), and BGE's actions to respond to the recent 10 CFR 50.54(f) letter
concerning the adequacy and availability of design basis information. Overall, the
level of review and member participation met the OSSRC responsibilities.
Maintenance
The inspectors reviewed a number of surveillance tests and found that'the testing
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was performed safely and in accordance with proper procedures. The inspectors
noted that an appropriate level of supervisory attention was given to the testing
depending on its sensitivity and difficulty,
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Executive Summary (cont'd)
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BGE found fuel oil in earth samples taken in the vicinity of the 21 fuel oil storage
tank. To determine the extent of suspected leakage, BGE unearthed the lines in the
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vicinity of the oil samples and conducted both a visual piping examination and a
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series of hydrostatic tests. No piping leaks were found. The inspectors considered
the BGE activities prudent in ensuring fuel oil system integrity.
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Enaineerina
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The inspectors found that BGE did not have a criticality monitoring system and
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emergency evacuation procedures for the new fuel storage area and the spent fuel
pool as required by 10 CFR 70.24(a). The issue was a violation of NRC
requirements.
The inspectors found that there was an apparent lack of ownership for the electrical
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cable separation barriers and that a related modification package initiated in 1990
remained open. The issue is unresolved pending further NRC review.
Plant Support
Regional inspectors verified that BGE has maintained on-shift dose assessment
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capability supported by appropriate procedural guidance.
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TABLE OF CONTENTS
EX EC UTIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TA B LE O F C O NT ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
Summ ary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1. O p e r a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
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Conduct of Opera tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
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01.1 General Comincats (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
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01.2 Reactor Trip Due to Feedwater System Malfunction . . . . . . . . . . 2
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Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
07.1 Offsite Safety Review Committee . . . . . . . . . . . . . . . . . . . . . . . 3
I I . M a in t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
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M1
Conduct of Maintenance
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M 1.1 Routine Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . 4
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M 1.2 Suspected Fuel Oil leak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
M1.3 Routine Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . 4
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Ill. Engineering
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E2
Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . 5
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E2.1
Special Nuclear Material Criticality Monitors . . . . . . . . . . . . . . . . 5
E2.2 Electrical Separation Barriers . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
I V . Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
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Miscellaneous EP Issue
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Licensee On-Shift Dose Assessment Capabilities
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V. M a nag e m e nt M e eting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
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Exit Me eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
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Management Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
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Review of UFS AR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
ATTACHMENTS
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Attachment 1:
Partial List of Persons Contacted
Inspection Procedures Used
items Opened, Closed, and Discussed
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List of Acronyms Used
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fleport Details
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Summarv of Plant Status
Unit 1 began the inspection period at full power. On October 20, a control element
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assembly, (control rod) dropped fully into the core. Operators reduced reactor power to
90 percent, recovered the assembly, and restored power to 100 percent. Unit 1 remained
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at full power for the remainder of the report period.
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Unit 2 started the inspection period at full power. A reactor trip occurred on November 17,
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1996 due to a feedwater system component malfunction. Unit 2 was restarted on
November 19, returned to full power on November 21, and remained at full power for the
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remainder of the report period.
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1. Operations
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Conduct of Operations '
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01.1 General Comments (71707)
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Overall, plant operat ons were conducted with a proper focus on nuclear safety.
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On October 20, during the conduct of routine control element assembly (CEA)
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testing, a shutdown CEA dropped fully into the reactor core. Using the appropriate
abnormal procedure, reactor operators reduced reactor power to 90 percent and
stabilized the plant. Engineering and maintenance personnel were informed and the
dropped CEA was tested, baluation of the test data showed nothing abnormal,
and the rod was returned to the normal full out position. Reactor power was then
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restored to full power. An issue report was generated to ensure a review of the
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event. The review suggested that the probiern leading to the dropped rod may have
involved the test circuitry. Subsequently, engineering personnel initiated actions to
replace the Master Bypass Switch with a new, more reliable, model. The inspectors
considered he actions of operations personnel and the related activities of
engineering and maintenance personnel to be both prompt and appropriate.
During a contrr* room panel walkdown on November 12, the inspector noted that
power to the 1 A emergency diesel generator annunciator status panel had
apparently been lost. When informed, operations personnel determined that power
was available; however, the indicating light had burned out. The bulb was replaced
and panel indication was restored to normal. Neither the inspector nor BGE could
determine how long the indicating bulb had been burned out prior to the
observation.
' Topical headings such as 01, M1, etc., are used in accordance with the NRC standardized
reactor inspection report outline found in MC 0610. Individual reports are not expected to
address all outline topics.
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01.2 Reactor Trio Due to Feedwater System Malfunction
a.
inspection Scoce
The inspectors reviewed the circumstances associated with the Unit 2 reactor trip
that occurred on November 17,1996.
b.
Observations and Findinas
During a feedwater system transient Unit 2 tripped automatically from 100% power
due to low water level in the 21 steam generator. The plant responded as
designed. Following the reactor trip, the 22 steam generator feedwater pump
(SGFP) tripped on high discharge pressure and the 21 SGFP ramped to idle. The
21 SGFP did not trip because the high discharge pressure signal was not of
sufficient duration to lock in the trip circuitry. Operators restored steam generator
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water level using the auxiliary feedwater pumps. Operators promptly stabilized the
unit in Mode 3 (hot standby).
BGE formed a significant incident findings team to investigate the cause(s) of the
trip and recommend corrective actions. The preliminary findings indicated that the
21 feedwater regulating valve (FWRV) positioner mechanical feedback spring
retaining screw had broken, allowing the spring to relax. This failure caused the
valve positioner to sense the position of the FWRV as full open. The system
responded by closing the FWRV to re-establish the demanded position. Due to the
relaxed spring, FWRV position continued to indicate full open even as the positioner
drove the FWRV full closed. The transient resulted in a low water levelin the 21
steam generator, causing an automatic reactor trip.
The transient occurred in about 12 seconds which prevented manual action to
control steam generation level. Operators responded to the trip by implementing
emergency operating procedure EOP-0, " Reactor Trip." All safety functions were
met during the transient. The emergency operating procedures were accomplished
satisfactorily and there were no complications which delayed stabilizing the plant in
Mode 3. Subsequently, reactor operators transitioned to operating procedure OP-4,
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" Plant Shutdown From Power Operation to Hot Standby."
Accompanying the reactor trip was an expected rapid closure of the 22 FWRV,
which caused the 22 SGFP to trip on high discharge pressure. A few minutes after
the trip, control room operators noted that the 22 SGFP appeared to have reset
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itself and the steam admission valve indicated open. Following procedures, the
operators attempted to trip the SGFP from the panel but no response was obtained
from the trip pushbutton. An operator was dispatched to locally trip the pump.
However, on receipt of the original trip signal, the pump turbine governor valve had
closed as designed. The governor valve remained closed and the pump remained
stopped throughout the event. On control room direction, the equipment operators
isolated steam to the 22 SGFP. Troubleshooting disclosed that two relays in the
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feedpump control circuitry were degraded, the net effect being a false reset signal
that opened the steam admission valve. The relay failure had no effect on plant
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conditions during the event and did not delay recovery activities. The relays were
replaced and the circuitry tested satisfactorily prior to reactor restart.
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The FWRV positioner spring retainer for both the 21 and 22 FWRV were replaced.
' The failed component from the 21 FWRV was metallurgically evaluated and the root
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cause was determined to be torsional overload that likely occurred during
installation during the spring 1995 refueling outage. . A sampling of new spring
retainers ~were evaluated and no deficiencies were found. Plant issue reports were
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written to document and resolve the identified deficiencies. Unit 1 FWRVs will be
inspected at the next availability.-
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Unit 2 was returned to critical operations on November 19. The reactor was
returned to full power on November 21.
c.
Conclusions
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Operator response to the automatic reactor trip was prompt and very good. The
operators were effective in stabilizing the plant and ensuring safety. The inspectors
considered the operator re.sponse, well controlled, using correct procedures, and
effective in ensuring the plant remained in a safe condition.
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BGE personnel appropriately conducted an event review and determined that the
event had no consequence to public health and safety. The cause of the reactor
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trip was understood and corrected prior to unit restart. The unit was returned to full
power operation without complication.
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Quality Assurance in Operations
07.1 Offsite Safety Review Committee
The inspectors atten' ded portions of the Offsite Safety Review Committee (OSSRC)
meeting on November 21. The OSSRC composition and agenda were in compliance
with the requirements of Quality Assurance Policy, Addendum 18-1, Review
Functions of the OSSRC. The agenda included a review of plant status, significant
safety issues,10 CFR 50.59 evaluations, and proposed changes to the operating
license, including the planned change to improved Technical Specifications. A very
good questioning attitude and safety perspective were noted, particularly regarding
effectiveness of Unit 2 post-trip review, the timeliness of review of surveillance test
data by the Plant Operations and Safety Review Committee (POSRC), and BGE's
actions to respond to the NRC 10 CFR 50.54(f) letter concerning the adequacy and
availability of design basis information. Overall, the level of review and member
participation met OSSRC responsibilities.
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11. Maintenance
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Conduct of Maintenance
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M 1.1 Routine Maintenance Observations
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Using Inspection Procedures 62703,62707, and 61726, the inspectors observed
the conduct of maintenance and surveillance testing on systems and components
important to safety. The inspectors also reviewed selected maintenance activities
to assure that the work was performed safely and in accordance with proper
procedures. The inspectors noted that an appropriate level of supervisory attention
was given to the work depending on its priority and difficulty. Maintenance
activities reviewed included:
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MO 1199505993
Replace 11 Plant Air Compressor
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MO 0199602195
Excavate at 21 Fuel Oil Storage Tank to Determine Source of
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Fuel Oil Leak
MO 2199604661
Overhaul 24 Circulating Water Pump
M1.2 Susoected Fuel Oil Leak
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On November 11, the inspectors were informed by BGE that the Maryland
Department of the Environment has been notified of a suspected underground leak
of fuel oil from the 21 fuel oil storage tank. Ground samples taken while drilling
test wells during a non-safety related anodic protection activity indicated that fuel
oil had contaminated soil in the vicinity of the fuel oil supply lines to the emergency
diesel generators. To determine the extent of the suspected leakage, BGE
unearthed the lines in the vicinity of the oil samples and conducted both a visual
piping examination and a series of hydrostatic tests. No piping leaks were found.
BGE then concluded that the leakage was from an external source, possibly
overflow of a fuel oil sump some time in the past. While the lines were uncovered,
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BGE conducted an evaluation of the protective wrap for the piping and replaced
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wrap that had degraded. The inspectors considered the BGE activities prudent in
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ensuring fuel oil system integrity.
M1.3 Routine Surveillance Observations
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The inspectors witnessed and reviewed selected surveillance tests to determine
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whether approved procedures were in use, details were adequate, test
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instrumentation was properly calibrated and used, technical specifications were
satisfied, testing was performed by qualified personnel, and test results satisfied
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acceptance criteria or were properly dispositioned.
The surveillance testing was performed safely and in accordance with proper
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procedures. The inspectors noted that an appropriate level of supervisory attention
was given to the testing depending on its sensitivity and difficulty. Surveillance
testing activities that were reviewed are listed below:
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STP-0-OOS-1 AFW System Performance Testing
STP-0-073F Boric Acid Pump Performance Test
STP-0-OO8A 1 A EDG and 4KV Bus Test
111. Enaineerina
E2
Engineering Support of Facilities and Equipment
E2.1
Soecial Nuclear Material Criticality Monitors
As a follow-up to a review conducted by the NRC Office of Nuclear R3 actor
Regulation, the inspectors conducted a review of BGE's compliance wPh or
exemption from the criticality monitoring requirements of 10 CFR 70.24(a).
At the time of this inspection, BGE was not in compliance with 10 CFR 70.24(a)
because BGE did not have a monitoring system that would energize clearly audible
alarms if accidental criticality occurred in each ares in which licensed quantities of
special nuclear material was handled, used, or stored. Specifically, the new fuel
storage area and the spent fuel pool were affected arsas. BGE also did not have
emergency procedures for these areas.
An exemption from the requirements of 10 CFR 70.24(a) was granted with the
issuance of Calvert Cliffs SNM License Nos. SNM-1364 and SNM-1624; however,
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the exemption was not specifically carried forward with the issuance of the Part 50
Operating License. Pursuant to the requirements of 10 CFR 70.14(a) and 10 CFR
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70.24(d), BGE has requested a permanent exemption from the criticality monitoring
requirements of 10 CFR 70.24(a) by a letter to the NRC dated August 19,1996.
The inspectors concluded that the BGE failure to have a criticality monitoring
system and emergency procedures for the new fuel storage area and the spent fuel
pool was a violation of 10 CFR 70.24(a). (VIO 50-317&318/96-08-01)
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E2.2 Electrical Seoaration Barriers
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a.
Insoection Scope
The inspectors reviewed safety related cable separation to assure that requirements
were met for redundant channels that support protective functions during
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postulated events.
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b.
Findinos and Observations
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The inspectors reviewed the electrical separation criteria in FSAR chapter 8.5,
" Separation Criteria," and design document E-406, " Design and Construction
Standards for Cable and Raceway," and walked down the penetration rooms with
Plant Engineering and project engineering personnel. Separation barriers were
required when three feet horizontal or five feet vertical cable tray separation as
specified in the Final Safety Analysis Report (FSAR) were not maintained.
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On October 9, the inspectors identified damaged and apparently missing electrical
separation barriers in the 45 foot electrical penetration rooms. The following
apparent deficiencies vxre identified:
Cable tray ZA1 AE70 (senaration group 1) had a broken marinite board / separation
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barrier on top and did not have horizontal separation from tray ZB1 AE66 (separation
group 2)
Cable tray ZD1 AE80 (separation group 1) had a missing marinite board / separation
barrier and needed to be extended to provide separation from ZE1 AE83 (separation
group 2)
Cable tray ZA2AE74 (separation group 1) had a broken and missing marinite
board / separation barrier on the bottom and did not have separation from trays
ZB2AE66 and ZB2AE63 (separation group 2)
BGE entered the deficiencies in the corrective action system as issue reports. The
inspector also identified several other potential separation issues. BGE personnel
indicated that an extensive review of and repair to separation barriers were
performed by a plant modification that started in 1990 and an associated project
plan to correct electrical cable separation deviations and anomalous conditions
identified as a result of corrective actions for NRC Notice of Violation 50-317/89-
27-05. The BGE personnel indicated that records from this project and plant
drawings would need to be researched to resolve the potential issues.
During the review of the plant drawings for the cable trays associated with the
deficiencies, BGE personnel identified that the drawings for Unit 2 had not been
updated to reflect the as-built conditions for changes made to the separation
barriers in 1990. An issue report was generated by BGE to address this issue.
The inspector reviewed the drawings with BGE personnel and noted that the
drawings also did not appear to have been updated to reflect the as-built conditions
for a second area common to Units 1 and 2. Therefore, based on the extent of the
deficiencies, the inspector questioned the adequacy of the plant drawings to reflect
as-built conditions. The inspector also reviewed some of the original project records
related to the walkdown, engineering evaluation, and repair of separation barriers
for the project plan. The inspector noted that these records appeared to be quality
records that were not stored in the records vault and were not available for general
use by engineering personnel. The inspector also questioned whether the 1990
modification documentation had been closed since the effort was completed in
1994. BGE personnel responded that the modification was one of several older
modifications that had not been closed out.
c.
Conclusions
The inspectors concluded that there was an apparent lack of ownership for the
electrical cable separation barriers and the modifications performed in 1990. This
issue will remain open pending further NRC review of (1) BGE's corrective actions
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for NRC violation 50-317/89 27-05, for barrier deficiencies and to assess whether
or not additional deficiencies exist; (2) BGE actions to update the configuration
control' drawings for the Unit 2 separation barrier installation changes; (3) the
significance of the BGE actions to complete the closeout of the 1990 modification
package; and (4) whether appropriate administrative controls have been applied to
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the storage and retention of the applicable project records. (URI 50-317&318/96-
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IV. Plant Support
P8.
Miscellaneous EP issue
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P8.1
Licensee On-Shift Dose Assessment Capabilities
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On the week of September 30,1996, a region-based inspector conducted an
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in-office telephone interview with BGE to complete NRC Temporary
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Instruction (TI) 2515/134, " Licensee On-Shift Dose Assessment
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Capabilities". The goal of the instruction was to gather information on the
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BGE capabilities to perform on-shift dose assessment. It was determined
that BGE does have on-shift dose assessment capability supported by
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appropriate procedural guidance. The inspector concluded that BGE met
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NRC requirements to be able to perform dose assessment at all times.
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V. Mananoment Meetinas
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Exit Meeting Summary
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During this inspection, periodic meetings were held with station management to
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discuss inspection observations and findings. On December 19,1996, an exit
meeting was held to summarize the conclusions of the inspection. BGE
management in attendance acknowledged the findings presented.
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Management Meeting Summary
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On October 23,1996, Richard Crienjak, Acting Deputy Director, Division of Reactor
Projects, Region I toured the facility and held discussions with BGE staff as part of
the NRC preparations for the Systematic Assessment of Licensee Performance
(SALP) process.
On October 24, and 25,1996, Hubert Miller, the Regional Administrator, NRC
Region i toured Calvert Cliffs .md held general discussions with plant personnel.
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On November 15,1996, NRC Commissioner Edward McGaffigan toured Calvert
Cliffs and held general discussions with plant personnel.
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Review of UFSAR Commitments
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A recent discovery of a licensee operating its facility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a
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special focused review that compares plant practices, procedures and/or parameters
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to the UFSAR description. While performing the inspections discussed in this
report, the inspectors reviewed the applicable portions of the UFSAR that related to
the areas inspected to verify that the UFSAR wording was consistent with the
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observed plant practices, procedures and/or parameters, inconsistencies were
noted concerning electrical cable separation as discussed in Section E2.2.
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ATTACHMENT 1
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. PARTIAL LIST OF PERSONS CONTACTED
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P. Katz, Plant General Manager
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K. Cellers, Superintendent, Nuclear Maintenance
K. Nietmann, Superintendent, Nuclear Operations
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P. Chabot, Manager, Nuclear Engineering
T. Camilleri, Director, Nuclear Regulatory Matters
B. Watson, General Supervisor, Radiation Safety
C. Earls, General Supervisor, Chemistry
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L. Gibbs, Director, Nuclear Security
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T. Sydnor, General Supervisor, Plant Engineering
T. Forgette, Director - Emergency Preparedness
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M. Polak, Supervisor - Engineering Assessment Unit
C. Cruse, Vice President - Nuclear Energy
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NRC
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Edward McGaffigan, Commissioner
Richard Crienjak, Acting Deputy Director, Division of Reactor Projects, Region i
Hubert Miller, Regional Administrator, Region I
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INSPECTION PROCEDURES USED
IP 62707: Maintenance Observation
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lP 71707: Plant Operations
IP 93702: Prompt Onsite Response to Events at Operating Power
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Reactors
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IP 61726: Surveillance Observations
IP 37550: Engineering
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IP 37551: Onsite Engineering
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IP 71750: Plant Support Activities
IP 83750: Occupational Exposure
IP 92700: Onsite Followup of Written Reports 7f Nonroutine Ever.ts
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at Power Reactor Facilities
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iP 92902: Followup - Engineering
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lP 82701: Operational Status of the Emergency oreparedness
Program
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2
ITEMS OPENED. CLOSED, AND DISCUSSED
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Opened
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50-317&318/96008-01
Cable Separation Issues
50-317&318/96008-02
Failure to Monitor for Criticality of New Reactor
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Fuel
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LIST OF ACRONYMS USED
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Steam Generator Feedwater Pump
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FWRV
Feedwater Regulating Valve
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OSSRC
Offsite Safety Review Committee
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POSRC
Plant Operations Safety Review Committee
Final Safety Analysis Report
Updated Final Safety Analysis Report
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TI
Temporary Instruction
IFl
Inspector Followup Item
Systematic Assessment of Licensee Performance
Unresolved item
KV
Kilovolts (1,000 volts)
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MO
Maintenance Order-
Surve!Ilance Test Procedure
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lP
Inspect on Procedure
CFR
Code of rederal Regulations
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