ML20199G750
| ML20199G750 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/23/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20199G672 | List: |
| References | |
| 50-324-97-13, 50-325-97-13, NUDOCS 9802040338 | |
| Download: ML20199G750 (50) | |
See also: IR 05000324/1997013
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U. S. NUCLFAR REGULATORY COMMISSION
REGION 11
Docket Nos: 50-325, 50 324
Report No: 50-325/97-13. 50-324/97 13
Licensee: Carolina Power & Light (CP&L)
Facility: Brunswick Steam Electric Plant, Units 1 & 2
Location: 8470 River Road SE
Southport, NC 28461
Dates: November 9 - December 27, 1997
Inspectors: C. Patterson Senior Resident Inspector
E. Brown Resident inspector
G. Guthrie, inspector in Training
J. Coley Reactor inspector (M1.3. M8.6)
J. Lendhan. Reactor Inspector (E1.1. E1.4. E5.1. E8.3.
E8.4. E8.5)
C. Doutt. Senior Instrumentation and Controls
Engineer. Office of Nuclear Reactor Regulation
(E1.1. E1.2. El.3)
G. Wiseman. Reactor Inspection (F2.1. F2.2. F2.3,
F3.1. F5.1 F6.1. F7.1)
Approved by: M. Shymlock. Chief. Projects Branch 4
Division of Reactor Projects
9802040330 900123
G ADOCK 05000324
Enclosure 2
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EXECUTIVE SUMMARY
Brunswick Steam Electric Plant. Units 1 & 2
NRC Inspection Report 50 325/97 13. 50-324/97-13
This integrated inspection included aspects of licensee operations,
engineering, maintenance, and plant support. The report covers a 6-week
period of resident inspection; in addition. It includes the resu'ts of
maintenance, engineering, and fire protection ir,pections by regional and
headquarters inspectors.
Operations
e The inspector concluded that u.e cold weather program has been
satisfactorily implemented. Adequate contingency plans and operator
checks for proper operation of the systems were noted in the procedures.
Section 01.1).
- The inspector concluded. from a safety system walkdown, that the
Containment Atmospheric Dilution system was being maintained as designed
(Section 02.1).
- The clearance reviewed was prepared. authorized, and implemented in
accordance with procedure (Section 02.2),
e The inspector concluded that the Plant Nuclear Safety Committee meeting
provided an effective review of Unit I readiness for restart (Section
07.1).
e Inspe.; tor review determined that clearance records were not retained in
accorcance with Technical Specifications (TS). The failure to maintain
clearance records in accordance with TS was a violation (Section 07.2).
- The control of a short duration mid-cycle o:tage was excellent (Section
07.3).
- Licensee investigation determined that removal of the IB Reactor
feedwater Pump at too high a power level caused larger than expected
level transients. These transients combined with the improper
functioning of the level contacts in the Reactor Recirculation Run back
logic circuitry, resulted in the November 5-6. 199/ run backs (Section
08.3).
- The inspector concluded that the licensee's control of the 2C and 20
electrical bus maintenance was weak because they did not recognize DG in
oberabilityconditionsduringtheimplementationoftt.eirclearance
( ection 08.4).
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Maintenance
e Movement of the spent fuel shi) ping cask was perforrxo in accordance
with methodology approved by t1e NRC in a letter dated December 2, 1997.
Adequate supervisory oversight was present during movement of the cask
(Section M1.1).
- The inspector observed performance of calibration of two Reactor Core
Isolation Cooling (RCIC) pressure switches. The work activities were
completed without any identified questions or concerns (Section M1.2).
- Maintenance activities observed relating to equipmert qualification of
electrical equipment were found to be conducted in a thorough and
effective manner (Section M1.3).
. A violation was identified for a preventive maintenance procedure not
indicating specific E0 requirements. This omission resulted in
deficient Nelson flame seals in motor control centers not being detected
during scheduled preventive maintenance activities (Section M1.3).
- The licensee continues to struggle with proper dispositioning of
abnormal indications. The failure to maintain the Daily Surveillance
Report in accordance with procedure was a violation. Abnormal values
observed fer the Steam Jet Air Ejector radiation monitor and subsequent
test indicated potential fuel failure for Unit 1 (Section M3.1),
- The licensee identified that the Unit 2 Core Spiay sparger differential
alarm setpoints were outside of the TS allowable range. The cauce was
attributed to voiding of the sparger nozzles similar to the phenomenon
identified previously on Unit 1. The alarm setpoints were adjusted and
the associated documentation was updated (Section M8.5).
-Engineerino >
+ An additional example of a violation was identified for an inadequate
procedure for the conduct of E0 maintenance (Section E1.4). Two
inspector followup items were identified to review revisions to
instrument setpoint procedures and to review terminal block leakage
current evaluations (Section El.1 and Section E1.4).
- A weakness was identified regarding a procedure reference to a drawing
for accident temperature data which was not available for use and
wording inconsistencies in the procedure (Section E1.1).
- The licensee was making progress in resolution of the technical issues
and closure of CRs and JCOs (Section E1.4). The licensee training and
qualification for E0 personnel meets NRC requirements (Section E5.1).
Instrument setpoint calculations were technica ly adequate and complied
with NRC requirements (Section E1.2).
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Plant Support
. The ins)ector determined that each of the locked high radiation area
doors w11ch were checked were locked. lhe ins)ector concluded that the
licensee is satisfactorily controlling locked ligh radiation areas in
the plant (Section Rl.1).
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The inspector determined that several poor radiological work practices
existed in a radioactive material storage area (Section Rl.2).
The inspector found the status and condition of the protected area fence
i to be satisfactory (Section S2.1).
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Corrective maintenance on degraded fire protection systems was
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accomplished in a timely manner. The maintenance and material condition
of the fire protection equipment and features were satisfactory
(Section F1.1).
. The inspector concluded that silicone foam penetration seal field
verificction documentation was maintained by the licensee. The
inst 311ation and repair procedures for penetration seals provided
adequate guidance to ensure that materials were installed per design
requirements. However, the designs were not supported by seal testing
documentation, vendor data and inspection criteria, installer
qualification and training records, and engineering evaluations that
satisfy the guidance of Generic Letter 86-10 for deviations from the
fire barrier configuration qualified by tests (Section F2.2).
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The inspector concluded that fire door surveillance procedures and
acceptance criteria for verification of fire door clearances were in
accordance with National Fire Protection Association (NFPA) guidance.
However, an updated Final Safety Analysis Report (UFSAR) discrepancy
associated documentation of fire door and frame evaluations was
identified (Section F2.3).
. General housekeeping was satisfactory. Fire retardant plast.ic sheating
and film materials were being used. Lubricants and oils were properly
stored in approved safety containers. Controls for combustible gas bulk
storage and cutting and welding operations were being enforced.
Controls were being properly maintained for limiting transient
combustibles in designated separation zones and other restricted plant
areas (Section F3.1).
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The fire brigade organization and qualification training .act the
requirements of the site Procedures. Fire brigade turnout gear and fire
fighting equipment were being properly maintained (Section F5.1).
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The coordination and oversight of the tacility's fire protection program
had been reassigned from the previous Loss Prevention Unit organization
to shift. Operations. The new organizat.onal structure met NRC
guidelines and the licensee's fire protection program requirements
(Section F6.1). .
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. The 1997 Nuclear Assessment Section assessment of the facility's fire
protection program was comprehensive and was effective in identifying
fire protection program performance deficiencies to management. Planned
corrective actions in response tc the audit issues were substantial and
included a fire p.'otection reorganization (Section F7.1).
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ReDort Details
~ Summary of Plant Status
Unit I returned to power o)eration on November 14. 1997, following a mid-cycle
outage that began on Novem)er 5. 1997, to remove leaking fuel assemblies. Two
leaking fsel assemblies were identified and removed during the mid cycle
outage. However, indications of a potential fuel leaker remained after the
unit returned to full power operation. At the end of the report period the
unit had been on-line 42 days.
Unit 2 operated continuously during this report period. At the end of the
report period the unit had been on-line continuously for 59 days.
Due to concerns about the control room dose, the licensee imposed an
administrative limit on lodine until a Technical Specification (TS) amendment
submitted was a) proved. The licensee made a orocedure change to
Administrative procedure 0Al-81. Water Chemistry Guidelines, setting the limit
at 0.1 microcurie per gram dose equivalent L 'ine 131 compared to the TS value
of 0.2 microcurie per gram. Also, the licet ;e has been providing weekly
water chemistry data to NRR and the Resident Inspector for review. None of
the data reviewed has exceeded the administrative limit.
Due to a reconstitution of the Environmental Qualification (EO) program and
items identified, there are 12 of 24 Justification for Continued Operation
(JCO) that remain open for both units. The following provides the status of
the EQ JCOs and associated Engineering Service Requests (ESRs):
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1) ESR 97-00087. E0-Type JC0 for Improperly Configured Conduit Seal.
2) ESR 97-00574 Greyboot Connectors.
3) ESR 97-00329 (old ESR 96-00625). E0 Type JC0 for EQ Fuses Without
a Qualification Data Package (00P).
4) ESR 97-00289. Post A cident Sampling System (PASS) Valve Limit
Switch Panel Wiring.
5) ESR 97-00238. JC0 for Standby Gas Treatment Motor Operated Valve
(MOV) Position Indicator Rheostat.
6) ESR-97-00534. GE c' Type Terminal Strips.
7) ESR 97-00513. In-b Drywell Electrical Penetrations.
8) ESR 97-00535. Target Rock Solenoids TB Spray.
9) ESR 97-00449, Degraded Junction Boxes.
19) ESR 97-00250. Conduit Union in EQ Boundary.
11) ESR 96-00425. Evaluation of E0 sealants.
12) ESR 97-00523. High Pressure Coolant Injection Auxiliary Oil Pump
Motor Unit 1.
0P10
13) ESR 97-00446. GE Radiation Detectors. closure date to be
determined (TBD).
14) ESR 96-00503. Associated Circuit E0. closure date TBD.
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15) ESR 97-00330 (ola ESR 96 00501). Motor Control Center (MCC) E0 was
closed by the licensee, but was reopened - closure date TBD.
16) ESR 96-00426. Evaluation Quality class and E0 classification of
PASS valves was scheduled for completion June 6, 1997. but closure
date is TBD.
17) ESR 97-00529. Failure of Unit 1 Drywell Motor, closure date TBD.
18) ESR 96 00587 PASS Valves, closure date TBD.
19) ESR 96 00627 ODP for Marathon 300 Terminal Blocks was scheduled
for completion December 31, 1937 but revised to August 1. 1997,
but closure date is now TBD.
20) ESR 97-00229. JC0 for GE Condition Report (CR) 151 B Terminal
Blocks was scheduled to be completed September 1, 1997, but
closure date is now TBD.
21) ESR 97-00256. Main Steam Insulation Valve Hiller Aci . tor JCO. was -
scheduled for completion September 2, 1997. but closure date is
now TBD.
22) ESR 97-00343. Qualification of Kulka Model 600 Terminal Blocks was
scheduled for completion September 1. 1997, but closure date is
now TBD.
23) ESR 97-00435. MCC Fittings, closure date TBD.
24) ESR 97-00602. Solenoid Valve Field Wiring, closure date TBD.
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In summary Unit I returned to power operation following completion of a mid-
cycle outage. Unit 2 o)erated continuously; however there were 12
outstanding JCOs in the E0 area for both units.
I. Ooerations
01 Conduct of Operaticns
01.1 Cold Weather Preparation
a. Insoection Scone (71714)
The inspector reviewed the licensee's cold weather program to determine
whether it had been effectively implemented.
b. Observations and Findinas
The inspector reviewed the licensee's cold weather 3rogram for adequacy
and implementation by reviewing their Cold Weather 3111 and Freeze
Protection Procedure. Operating Instruction 001-01.02: Fire Protection
Procedure 0FPP-024. Freeze Protection of Fire Suppression System; and
Preventive Maintenance Procedure OPM-HT001. Preventive Maintenance on
Plant Freeze Protection and Heat Tracing. The inspector determined that
the procedures were adequately implemented. Additionally, the
procedures were adequately employed on multiple cold weather days. as
observed by the inspector.
The inspector conducted a walkdown of plant syn. , which were exposed
to cold weather. Systems which were heat traced were observed for
adequacy. The inspector looked for systems that did not have cold
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weather. heat trace installed. The inspector determined that the
operation of the Makeup Water Tank system heat trace was not controlled
by any procedure. The licensee stated that this heat trace system was
being controlled b," operator knowledge only. The licensee initiated a
procedure change request to place this heat trace system into their cold
weather procedures. The inspector noted on the Unit 2 Condensate
Storage Tank. High Pressure Coolant Injection (HPCI)/ Reactor Core
Isolation ~ Cooling (RCIC) level switch vent line that a six inch portion
of the lagging was missing at the top of the vent line and that the tin
shielding was missing around the lagging at an elbow on the vent line.
The lagging was wetted and degraded at the elbow. The inspector
discussed these two items with the licensee. The licensee did not
warrant these deficiencies as requiring corrective action. The
inspector did not find other systems requiring heat trace that were not
heat traced based on present system conditions and projected use of the
systems observed.
c. Conclusions
The inspector concluded that the cold weather program has been
satisfactorily implemented. Adequate contingency plans and operator
checks for proper operation of the systems were noted in the procedures.
02 Operational Status of Facilities and Equipment
02.1 Containment Atmosoheric Dilution (CAD) System Walkdown
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a. Insoection Scope (71707)
On December 10. 1997, the inspector performed a walkdown of the CAD
system in the Nitrogen and Off-Gas Services Building.
b. Observations and Findinos
The CAD system is described in Updated Final Safety Analysis Report
(UFSAR) Section 6.2.5. Combustible Gas Control in Containment. The CAD
system provides long-term nitrogen makeup after a Loss of Coolant
Accident (LOCA). This function is accomplished by vaporizing liquid
nitrogen and feeding it into containment as required to maintain an
oxygen concentration at or below five percent. The system is designed
to Engineered Safety Feature (ESF) standards, all equipment for CAD
service is designed with suitable redundancy and interconnections such
that no single failure of an active component will render the system
inoperable. This equipment includes one liquid nitrogen storage vessel.
two electric vaporizers, two flow-regulating stations. flow and
temperature indicators. and appropriate redundant valves and
interconnecting piping.
The inspector traced the system piping in the Nitrogen and Off-Gas
Services Building. The configuration was compared to plant drawing
0 02560. Containment Atmospheric Control System. The configuration was
found to be like the plant drawing. The inspector observed an inch of
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frost on the outside of the piping insulation on both sides of valve
HV-11. This valve is a manual isolation between the nitrogen tank and
an 85 pound pressure regulating valve.
The inspector questioned why the frost was on the line. The licensee
stated that the 90 pound relief valve setpoint was near the controlling
pressure of the 85 pound regulator and some nitrogen was venting off.
The redundant pressure regulating valve was isolated and it's isolation
valve (HV-12) was closed. The inspector questioned by keeping HV-12
closed, if the system was single failure proof. The licensee initiated
CR 97-04128. CAD Tank Isolation Valve, to address this issue, The
licensee concluded that no automatic action was required to address a
LOCA. Manual alignment of the pressure regulator was acceptable since
this was a long term post-LOCA action.
c. Conclusions
The inspector concluded, from a safety system walkdown, that the CAD
system was being maintained as designed.
02.2 Clearance Verification
l a. Insoection Scoce (71707)
The inspector reviewed the tagout for the Unit 2 Residual Heat Removal
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(RHR) system to verify proper clearance preparation, authori7 n. and
implementation,
b. Observations and Findinas
On December 10. 1997, the inspector performed verification of the proper
alignment and tagging of clearance 2-97-1781 on the Unit 2 RHR System.
All accessible components were verified to-be in the proper position
with the appropriate tags in place. The inspector reviewed Nuclear
Generation Group Standard Procedure OPS-NGGC-1301. Equipment Clearance.
The clearance package was adequately prepared, authorized, aad
implemer.ted. The inspector subsequently verified proper clearance
removal for those accessible components.
c. Conclusions
The clearance reviewed was prepared, authorized, and implemented in
accordance with procedure,
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07 Quality Assurarm in Operations
07.1 Restart Plant Nuclear Safety Committee (PNSC)
a. Insoection Scone (71707)
On November 11 and 12. 1997, the inspector attended the Unit 1 PNSC
restart assessment following a mid-cycle outage to replace two leaking
fuel assemblies,
b. Observations and Findinos
On November 11, 1997. PNSC was convened to review Unit I readiness for
restart. The committee reviewed the fuel sipping results and core
reload. Other maintenance activities during the outage were also
reviewed.
The meeting was conducted in accordance with TS with attendance by all
primary members, with no alternates. The meeting provided a thorough
discussion of all agenda items. The PNSC Chairman concluded that the
discussion of recirculation pump runbacks that occurred on November 5.
1997, during removal of the reactor feed pumps during the planned
shutdown was not complete. This item was statused as a restart
constraint requiring another PNSC review prior to restart. Noteworthy
in the review was the risk assessment review conducted for a failed
Control Rod Drive (CRD) pump. During the mid-cycle outage one of the
two CRD pump motors failed. The Probabilistic Safety Analysis (PSA)
person attended the comnittee meeting and presented the results from
running the risk assessment model considering failure of both CRD Jumps.
This risk was determined acceptable based on other TS required higi
pressure injection sources such as HPCI and RCIC.
On November 12. 1997, the inspector attended a second meeting. In this
meeting discussion was held regarding the problem with run backs and it
was concluded that this was due to a design deficiency that was already
corrected and installed on Unit 2 and scheduled for Unit 1 at the time
of the next refueling outage,
c. Conclusions
The inspector concluded that the PNSC meeting provided an effective
-review of Unit I readiness for restart.
07.2 Retention of Clearance Records
a. Insoection Scope (71707)
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The inspector reviewed whether configuration management documents,
specifically ciearances, were retained in accordance with TS 6.10. This
specification requires that facility records be retained in accordance
with the American National Standards Institute (ANSI) N45.2.9-1974
Collection. Storage, and Maintenance of Quality Assurance Records.
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b. Observations and Findinas
During ins)ector review of clearance errors which resulted in damage to
the Unit 23 recirculation pump seals, the licensee was unable to locate
a clearance hung to facilitate repairs on the recirculation motor oil- -
cooler. Tha clearance. 2-97-1531. was hung _resulting in a configuration
change for the B recirculation pump, but no maintenance on the system
was performed. The clearance was removed from the field, thus restoring
the system, and " rolled back" to allow use at a later date.
Subsequently, a scheduler requested the clearance be deleted due to the
repair activities being complete and approved without need for the
clearance boundary. As a result of the deletion of the clearance, no
record of the change in plant configuration was retained.
The inspectoi : viewed TS 6.10. UFSAR Section 1.8. Regulatory Guide
1,88, and ANS1 N45.2.9-1974. fhe inspector questioned the correctness
of not retaining the clearance. Since a configuration change did occur
despite the recirculation motor cooler activities not needing the cooler
isolated. Nuclear Records Management Procedure ORMP-001. Indexing of
Plant Records. defined those records required to be retained to satisfy
the 0A requirements stated in ANSI N45.2.9-1974. Discussion with the
licensee revealed that the records required to be retained did not
include clearances. The inspector reviewed the Nuclear Generation Grou)
Standard Procedure OPS-NGGC-1301. Equipment Clearance, and the Brunswicc
Required Records List. Neither document required that clearances be
retained.
TS 6.10 requires facility records shall be maintained in accordance with
ANSI N45.2.9-1974. ANSI N45.2.9-1974, in Section 3.2.7. Retention of
Records. states that Appendix A to the standard defined the types of 0A
records and the recommended retention periods. The failure to maintain
data sheets or logs on equipment alignment consistent with ANSI N45.2.9-
1974 is a violation. This violation is identified as VIO 50-325
(324)/97-13-01. Failure to Retain TS Required-0A Record.
c. Conclusion
Inspector review determined that clearance records were not retained in
accordance with TS. The failure to maintain clearance records in
accordance with TS was a violation.
07.3 Mid-Cycle Outaae (71707)
a. Insoection Scope
The inspector reviewed the mid-cycle outage activities to remove the
leaking fuel assemblies.
b. Observations and Findinas
Unit 1 was returned to power operation on November 14. 1997. This
completed a mid-cycle outage in eight days. The unit was shutdown.
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leaking fuel assemblies identified, removed, fuel reloaded and returned
to power o)eration. This short duration outage was the quickest on
record. T11s was accomplished with plant personnel without any major
problems. This outage was planned and controlled similar to a regular
refueling outage.
c. Conclusions
The control of a short duration mid-cycle outage was excellent.
08 Miscellaneous Operations Issues (92700, 92901)
08.1 (Closed) Unresolved Item (URI) 50-325/96-15-01: Vessel Disassembly
Without Secondary Containment.
During a refueling outage, the reactor vessel head and steam
dryer /separatorr assemblies were removed from the reactor vessel without
secondary containment integrity (SCI) established. This issue was
reviewed by the NRC Office of Nuclear Reactor Regulation. It was
determined that the removal of the nead and assemblies without SCI
established were not activities prohibited by TS 3.6.5.1. The potential
! for load handling accidents was a safety cuestion that has been reviewed
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by the NRC. However, maintenance of SCI curing vessel disassably was a
logical extension of the defense-in-depth ap3 roach used in addressing
the heavy loads issue and encouraged by the 4RC. The licensee's action
in proceeding with vessel disassembly was not conservative. The
licensee implemented controls during the Unit 2 refueling outage to
maintain secondary containment operable during vessel disassembly. This
issue was thoroughly evaluated as part of the licensee's Safe Shutdown
Risk Management Assessment.
08.2 (Closed) Violation V10 50-325(324)/97-02-01: Locked Valve Out of
Position
The licensee's response to this violation was dated May 5, 1997, and was
accepted by the NRC in a letter dated May 23. 1997. The corrective
actions described in the response letter were verified as complete by
the inspector. This violation is closed.
08.3 (Closed) URI 50-325/97-12-03: Recirculation Pumo Run backs
On November 5. 1997, the licensee began a c0ntrolled shutdown for the
Unit 1 forced outage in order to replace leaning fuel bundles. During
the shutdown. Unit I received two recirculation pump run backs to the 45
percent limiter. During the second run back the five percent buffer
region was entered and exited in accordance with procedures.
) Subsequently. no other transients or run backs were ercountered while
removing the Reactor Feedwater Pumps (RFPs) from service. The licensee
preliminarily attributed the first run back to a malfunction of the 1B
discharge check valve causing diversion of the 1A RFP through the 1B
discharge valve to the main condenser. The final analysis was provided
in the root cause analysis for CR 97-3917. Unit 1 Plant Transients While
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Removing a Reactor Feed Pump from Service. The inspector reviewed the
analysis and noted that the root cause attributed the run backs to the
removal of the RFPs at too high of a power level and a design problem in
the a) plication of the Metal-On-Silicon Field Effect Transistor (MOSFET)
switcl. The MOSFET was used in the 45 percent recirculation pump run
back logic to indicate the below 182 inches reacter water level contact
which is one of two contacts required to initiate the run back.
Reactor water level perturbations are expected during the removal of the
RFPs from service: however the magnitude of these perturbations seen for
these events were outside of the operators expectations. The root cause
analysis stated that removal of the RFP at 65 percent power was
inappropriate in that 65 percent during this evolution has changed since
power uprate. Before power uprate. RFPs were removed from service 3er
10P-32, Condensate and Feedwater System Operating Procedure, at or )elow
65 percent. Under current conditions 65 percent is approximately
equivalent to 68 percent power pre-uprated power. The analysis
attributed the magnitude of the perturbations to removal at too high of
a power level. In addition, the licensee determined that when the first
RFP was taken out of service, the less than 20 percent RFP flow contact
for the 18 pump was made up and with the MOSFET improperly indicating
below 182 inches water level the run backs were received. The design of
the MOSFET causes the contact to not be able to properly position itself
u'aon loss of the constant voltage supply. Therefore interruptions in
tle voltage will cause the MOSFET contact to not function as designed.
The second Run back was also attributed to the MOSFET. The licensee
intends to replace the MOSFETs in the next Unit 1 outage, The inspector
noted that the MOSFETs had already been replaced in Unit 2.
The licensee is reviewing plant operation to determine the appropriate
power level for removal of the RFPs from service. Based on licensee
satisfactory comaletion of the investigation into the cause for the
multiple run bac(s on November 5-6, 1997 this item is closed.
08.4 (Closed) URI 50-325(324)/97-12-04: Diesel Doeration Low Voltace Auto
Start Defeated
The inspector reviewed the licensee's root cause investigation CR 97-
03683, 4KV Bus 2C/2D Clearances. The licensee's investigation
determined that the number 3 diesel generator (DG) undervoltage relay
had been disabled in the same manrer as the number 4 DG during similar
maintenance activities on different days.
The inspector verified that the licensee did not exceed TS action,
limiting condition for operation, or time requirements for both
electrical bus maintenance activities. The inspector found that, on
October 9. 1997, the plant was under a TS action statement requirement
per TS 3.8.2.1. to restore the inoperable bus to operable within 8
hours, or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The electrical
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bus was not restored, in this case, for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 58 minutes. This
plant condition was not recognized as a problem until the root cause
investigation was performed. The root cause investigation was found to
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be adequate. The ins)ector concluded that the licensee *s control of the
2C and 2D electrical aus maintenance was weak because they did not
recognize that the DG would be inoperable during the implementation of
their clearance. This item is closed.
II. Maintenance
M1 Conduct of Maintenance
M1.1 Spent Fuel Cask Movement
a. Inspection Scooe (62707)
The inspector observed transfer of the spent fuel shipaing cask from the.
117 foot elevation to the transport v'hicle and from t1e transfer
vehicle to the 117 foot elevation of the Unit 1 Reactor Building. s
b. Observations and Findinas
On December 8. 1997, the inspector observed the removal of the spent
fuel shipping cask, with fuel in the cask from the 117 foot to the 20
foot elevation in the Unit 1 Reactor Building. On December 15, 1997,
the inspector observed shipping cask movement, without fuel in the cask,
from the 20 foot elevation to the 117 foot elevation in the Unit 1
Reactor Building. During both evolutions the cask was transferred with
the valve box covers removed while being moved by the non-single failure
proof yoke. Approval for use of a non-single failure proof yoke for
movement of the cask with the valve covers removed was granted to the-
l licensee by the NRC in a letter dated December 2, 1997. Upon reaching
the transfer vehicle on December 8. 1997. the cask was wiped down to
reduce contami.1ation levels. During both movements the inspector noted
that the area was adequately posted for the radiological conditions
I present and i ealth pnysics personnel were present. The inspector noted
that adequate maintenance supervisory oversight was present for both
cask movements.
Subsequent surveys of the cask after removal from the Reactor Building
revealed that the shipment exceeded required limits. This event was
captured in CR 97-4161. S)ent Fuel Cask (IF-300). The cask was returned
to the Reactor Building w1ere additional decontamination was conducted.
The licensee attributed the contamination levels seen to leaching of the
contamination due to changing temperatures and weather conditions.
c. Conclusions
Movement of the spent fuel shi) ping cask was performed in accordance
with methodology approved by t1e NRC in a letter dated December 2. 1997.
Adequate supervisory oversight was present during movement of the cask.
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10
M1.2 RCIC Turbine Exhaust Diaphraam High Pressure Instrument Calibration
a. Insoection Scoce (61726)
The inspector observed the performance of Maintenance Surveillance Test
2MST-RCIC230. RCIC Turbine Fxhaust Giaphragm High Pressure Instrument
Channel Calibration, for the pressure switches 2-E51-PSH-N012A and 2-
E51-PSH-N012C.
b. Observations and Findinas
On December 24. 1997, with Unit 2 at 100 percent power the inspector
observed the channel calibration for RCIC pressure switches 2 E51-PSH-
N012A and 2-E51-PSH-N012C. The inspector verified that duriug the
performance of this channel calibration that HPCI and Automatic
Depressurization System (ADS) were o)erable and that no othar work
activities were being conducted whic1 could cause an inadvertent
isolation. This test verified that, upon sensing of a high pressure
condition between the t'arbine exhaust dia)hragms, an isolation signal is
sent in accordance with TS 4.3.2.1 and Ta)les 3.3.2-2(4.b.6) and 4.3.2-
1(4.b.6)
The inspector reviewed the work request / job order (WR/J0) AKNU 19 and
the governing procedure 2MST-RCIC230. The procedure in use was verified
to be the correct revision and the test instrumentation in use was
within the allowable calibration duration. The inspector observed the
' procedure in use at all work locations and adequate communication was
maintained throughout the test. The work observed was completed
satisfactorily with no observed concerns.
c. Conclusions
The inspector observed performance of cal:uration of two RCIC pressure
switches. The work activities were completed without any identified
questions or concerns.
M1.3 General Comments
a. Insoection Scone (62700)
The inspector examined the following work activities involving EQ
electrical equipment to verify maintenance implementation of EQ
requirements.
WR/JO 97-ALVT-002 Verified Calibration of Unit 1 Loop B Residual
Heat Removal (RHR) Service Water Pressure Switches Tag No.
1-SW-PS-1176 B and 1-SW-PS-11760
WR/JO 97-AGDR-002 Verified Calibration of Unit 1. Loop A. RHR Flow
Transmitter (1-E11-FT-N015A). Converter (1-E11-FY-5119A). Square
Root Converter (1-E11-FY-K600A)
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11
.
WR/JO 97-AAAS-002 Unit 2. Loop B. RHR Breaker Test in compartment
DM 5 of GE IC 7700 Series MCC 2XB-2 Division Il
b. Observations and findinos
The above work was ,m cformed with the work packages present and in
active use. Technicians were skillful, experienced, and knowledgeable
of their assigned tasks. However, on December 10, 1997, while observing
Instrumentation and Control (I&C) maintenance personnel perform work
activities in accordance with WR/JO 97-AAAS-002, the inspector noted
that one of the multiple cable electrical penetrations in the top of MCC
2-2XB-2 did not have Nelson flame guard putty on the inside surface as
required by Maintenance Procedure OMMM 016. Environmental Qualification
Maintenance Program. Revision 4. to properly seal the penetration. The
inspector examined the putty installation on the top of the MCC cabinet
for each of the penetrations and found the putty seal severely damaged
on a second multiple cable penetration. In addition, cables were loose
in both of the multiple cable penetrations. The applicable
Environmental Otalification Data Package (ODP). ODP 67, requires missing
or disturbed Nelson putty seals to be repaired or replaced. However,
the PM procedure used to maintain and inspect the MCC's (PM Procedure
OPM-MCC002. Revision 7. PM of GE Motor Control Centers and Switchboards)
did not have inspection requirements or acceptance criteria to ensure
that putty seals were properiy sealing the cabinets. On September 17.
f 1997, a three-year PM conducted on MCC 2-2XB-2 would have identified
l this discrepancy had procedure OPM-MCC002 included the acceptance
criteria for the Nelson flame seal putty. A subsequent inspection
performed on December 11. 1997 by the licensee, of 22 MCCs found an
additional three MCC cabinet penetrations with damaged Nelson putty
seals. In addition. 15 3ercent of the cables inspected in cabinet
penetrations had putty w1ich appeared not to fully adhere to the cable
in some areas. Failure of the procedure to implement E0 requirements
for Nelson autty seals is identified as VIO 50-325(324)/97-13-02.
Inadequate 3rocedure for the Conduct of E0 Preventive Maintenance.
c. Conclusions
Maintenance activities observed related to E0 of electrical equiament
were found to be conducted in a thorough and effective manner, iowever,
a violation was identified for a PM procedure not indicating specific E0
requirements. This omission resulted in deficient Nelson flame seals in
MCCs not being dettcted during scheduled PM activities.
M3 Maintenance Procedures and Documentation
M3.1 Steam Jet Air Eiector Off-Gas Radiation Monitor increase
a. Inspection Scoce (61726)
The inspector reviewed selected sections of Operating Instruction 101-
03.1. Control Operator Daily Surveillance Report to ensure that
i
1
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12
appropriate and prompt actions were taken to address abnormal TS
surveillance values,
b. Observations and Findinos
On December 2. 1997. Unit 1 was in mode 1 at 100 percent power. The
inspector reviewed the daily surveillance report as contained in
Attachment 1 to 101-03.1 for November 30 through December 1. 1997. The
inspector noted that the values for the Steam Jet Air Ejector (SJAE)
i
off-9as radiation monitors on aage 26 were between 1570 and 1780
millirem per hour (mR/hr) whici was greater than the T3/ Operating Limit
- value of 1000 mR/hr. The SJAE off-gas radiation monitors provide for
the detection of fuel element failures. The radiation levels are
recorded in 101-03.1 to provide an indication whether SJAE off-gas
radiation levels are approaching the alarm setpoint, which serves to
ensure that dose rates for gaseous effluents do not exceed the limits
l
prescribed in TS 3.11.2.1. Dose Rate.
l The inspector reviewed the associated procedures, work tickets, and
discussed the abnormal values with the licensee. Step 4.2 c 'f 001-03.1
required the control operator to red circle all values wt are not
within required limits. The inspector noted no indication on the
attachment or in the operator logs that action had been taken or was
expected to be performed to address the out-of-range values. Subsequent
reviews of the daily log entries by the inspector indicated continual
abnormal values and no red circles. These failures were recorded in CR
97-4136. Daily Surveillance Report. The failure to red circle values
not within required limits is a violation. This violation is identified
as VIO 50-325/97 13-03. Failure to Note Abnormal TS Surveillance Values.
CR 97-4100. Questioned OG Data / Fuel Leak indicated that on December 3,
1997, a step increase of approximately 200 mR/hr was seen on the
radiation monitor Subsequent sample results have shown an increase in
the Sum of Six value ano changes in the fuel reliability index which are
signs of potential fuel failure. In addition, the inspector noted that
incorrect sensitivities were used during the November 25, 1997.
adjustment of the SJAE radiation monitor alarm setpoilts. This was
documented by the licensee in CR 97-4046. SJAE Rad Mci. sensitivities.
CR 97-4180 SJAE rad monitor setpoints, addressed coordination problems ',
between the Operations procedure used to request new radiation monitor
setpoints, the Environmental and Radiological Control (E&RC) proced ce
that calculates the new setpoint, and the Maintenance procedure that
installs the new setpoints. By the time the radiation monitor setpoints
were ready to be installed the new values needed to be recalculated.
The inspector determined as a result of the cited failure and the three
additional CRs mentioned previously, that control and monitor'.ng of the
alarm setpoint was poor. Previous instances of failing to properly
disposition abnormal values were recorded by the NRC in Inspection
Re) ort (IR) 50-325(324)/97-12, when inadequate corrective action was
tacen for abnormally high drywell temperature. Tne abnormal temperature
resulted in exceeding the calculated environmental limits for ten
snubbers in the drywell.
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13
c. Conclusions
The licensee continues to struggle with proper dispositioning of
abnormal indications. The failure to maintain the Daily Surveillance
Report in accordance with procedure was a violation. Abnormal values
observed for the Steam Jet Air Ejector radiation monitor and subsequent
test indicate potential fuel failure for Unit 1.
M8 Miscellaneous Maintenance Issues (92902)
M8.1 (Closed) Licensee Event Reoort (LER) 50-325(324)/96-017-00: Invalid
Loss of Coolant Accident Locic Actuation
The invalid LOCA. initiation signal occurred during installation of test
equipment to support surveillance testing. P16nt systems responded as
designed. The initiation signal resulted in the following actuation:
Automatic start of emergency DGs 1.2.3. and 4.
Automatic start of Unit 1 Core Spray (CS) pump 1A.
Automatic start of Unit 2 Nuclear Service Water (NSW) pump 2A.
Unit 1 Grou) 10 division 1 actuation.
Closure of Jnit 1 Reactor Building Closed Cooling Water heat
exchanger Service Water isolation valve.1-SW-V106.
0)ening of NSW header to vital header isolation valve. 1-SW-V117.
, Slutdown of 1A and 10 Unit 1 drywell coolers ;
1
Corrective actions, described in the LER. were reviewed and verified by
the inspector. -These included: appropriate administrative action with
the involved technician; briefing of maintenance 1&C technicians on this
event; providing maintenance I&C personnel managements expectations ft
the restart of surveillance tests after problems have been encountered;
restricting the use of Simpson Model 260 Voltage Ohm Meters (V0Ms) for
circuit checks specified in maintenance surveillance tests: developing
training to enhance technician knowledge of the effects of test
equipment misalignment: and revising maintenance procedures to preclude
similar events.
This event did not violate TS. This LER is closed.
M8.2 (Closed) LER 50-325/97-009-00: Missed Increased Frecuency Inservice
Testino Recuirement
The American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code.Section XI, 1980 Edition through Winter 1981. Addenda
Section IWV-3414(a), requires an increase in test frequency in the event
an increase in stroke time of 25 percent or more from the previous test
is observed. Contrary to this requirement, the test frequency was not
increased as required. The required testing was missed by about two
weeks. Upon discovery. the valve was tested and the stroke time was
within the previous value and the test met the ASME Section XI
requirements.
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14
The corrective actions to prevent recurrence of this event. described in
the LER. were reviewed and verified by the inspector. Administrative
controls have been revised to ensure completed test results are reviewed
-in a timely manner and changes in test frequency are promptly initiated.
This event did not violate TSs. This event had minimal safety
significance from a-valve operability viewpoint since the retest of the
valve showed it was operable, ASME Section XI provides an intermediate
condition that allows continued operation without need for immediate
corrective action. From an administrative view, trending valve stroke
times is an imaortant indication of valve performance. Corrective
action taken s1ould improve this situation. This LER is closed.
M8.3 FClosed) LER 50-325/97-001-00: Rod Block Monitor Surveillance
.
nadeauacy
'
A discovery that the surveillance procedure fer testing the rod block
monitor (RBM). did not contain the pro 3er s 4
Ncessary to ensure ;
testing of the RBM instrument channel 3 int '
- tion, This condition
has existed since November 1996 for Unit 1, ma December 1996 for
'
Unit 2. Upon discovery, the correct tests were performed on both units
which indicated that the equipment was in calibration and capable of
performing its safety function.
The error was attributed to an inadequate administrative review of ;
reformatting changes made in September 1996. The surveillance procedure
changes were being upgraded in accordance with the generic procedure
writers guide. However, these changes did not insert the proper steps
to test the RBM inop instrument channel B.
' Corrective actions, described in the LER. were reviewed and verified by
-the inspector. The inspector determined that this event did not violate
TS since only the test for channel B was missed. The situation was
corrected within the allowable time specified by TS 3/4.3.4.
The-results of the RBM inop functional tests performed on toth units
upon discovery, indicated that the equipment was in calibration and
capable of performing its intended safety function. This LER is closed.
M8.4 (Closed) LER 50-325(324)/95-022-00: HPCI System Discharae Flow Element
Gasket Leak
During performance of a post maintenance test on the HPCI system. the
discharge flow element flanged gasket developed a 5 to 10 gallons per
minute (gpm) leak. Several other problems were also observed with
system operation.
Investigation revealed that undersized flange studs had been originally
installed on the flow element flange, allowing the Flexitallic gasket to
be installed off center. The off centered gasket degraded during the
post maintenance test. This condition existed on both units and
prompted declaring a potential failure of the HPCI system to ]erform its
intended safety function. With the HPCI system inoperable tie TS
U
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15
oermitted continued reactor operation provide 1 the ADS. CS system, and
RCIC were operable. This event was withir, Me TS requirement.
Corrective measures as described in the LER were reviewed and verified
by the inspector. This LER is closed.
M8.5 (Closed) Ins)ection Follow-un item (IFI) 50-325/97-05-02: Abnormal CS
Soarcer Brea t Detector Indication
(Closed) VIC 50-325/97-06-03: Inadeauate CS Surveillance Procedure
.(Closed) LER 50-325/97 02: Core Soray Header Differential Pressure
Instrumentation InoDerable
On March 9.1997, en auxiliary o)erator (AO) was verifying
instrumentation indications in tie Unit 1 Reactor Building. The A0
observed.that the reading displayed for 1-E21-PDS-N004A. Core Spray Line
Break Indicator, was not within TS 4.5.3.1.2.c.2 requirements. This
)ressure switch functioned to detect a break in the CS piping located
l 3etween the vessel and the shroud. The differential pressure (dP)
sensor measures the pressure across the core. Due to the addition of
'
L the drop from the steam separator, any break in the line would cause the
l -
indicated pressure drop to increase which would cause a more positive
indicated dP. The out of tolerance condition had existed since
November 1996 as stated in LER 50-325/97-02. During review of the
associated surveillance procedures, the inspector determined that actual
verification of the CS sparger alarm setpoint in relation to the
" normal" indicated instrument pressure was not being performed.
L
I
Themfore. the licensee could not evaluate whether the alarm setpoint
was within the " normal" TS range. This nonconformance resulted in VIO
50 325/97-05-02. Inadequate CS Surveillance Procedure.
The licensee performed reviews of data collected nonroutinely during
1995-1996 and in ESR 97-181 calculated a " normal" value for setpoint
verification in the related surveillance procedures. The licensee
subsecuently changed the alarm setpoints and updated the affected
procec ures. Additionally. the licensee performed a review of the TS and
determined that appropriate logging of required TS values was being
accomplished. During the refueling outage for Unit 2 from Se]tember to
October 1997 the licensee, with prior NRC approval, uprated t1e 100
percent _ rated thermal power 5 percent. The licensee included
verification of CS sparger dP " normal" values as part of the uprate
test program performed in accordance with S)ecial Procedure 2SP-97-204.
Unit 2 Power Jprate Data Collection. The cleck served to record the CS
sparger shutdown values.
The inspector reviewed ESR 97-634. ESP-97-204. CR 97-3870. LER 50-
325/97-02, and other related documentation. The inspector verified that
routine recording. upon entering mode 1. of the CS sparger dP was
incorporated into 0)erating Instruction 001-03.3. Auxiliary Operator
Daily Surveillance Report for both units. CR 97-3870. Core Spray Leak
Detection, documented the discovery on October 29, 1997 by an AD, that
4
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16
the 2-E21-PDS-N004A. CS A Loop Leak Detection, was outside of its
specified range. The instrument was declared inoperable and an LC0 was
entered. The licensee determined the new CS dP range in ESR 97-634
Core Spr 3y Loop Line Breuk Detectio , Allowable Range Change. The new
alarm setpoints were implemented and integrated into the affected
surveillances. 3rocedures, and design documents. Based on completion of
the review of t1e TS for other " normal" values not properly trended,
adjustment of the dP alarm setpoints*to bring the setpoints into
rvpliance with TS. and the institution of routine monitcring of the CS
.qarger " normal" values these items are closed.
M8.6 (Clos (d) VIO 50-325(324)/97-02-04: Failure to Imolement the
Renuirements of (a)(1) and (a)(2) of 10 CFR 50.65. The Maintenance Rule
This violation reported that all historical data since July 10. 1993.
had not been obtained to establish baseline system / structure / component
(SSC) performance, validate scoping, and set initial condition (a)(1)
and condition (a)(2) in the case of the reactor protection system (RPS),
Only corrective work. requests / job orders had been used for initial
determination of functional failures. Therefore, instrument out-of-
calibration data had not been reviewed for the period of July 10. 1993
through October 30. 1995. As an action related to Maintenance Rule
implementation. Procedure OMMM-004. PM. was revised on October 30. 1995,
to require that out-of-calibration data be evaluated for Maintenance
Rule functional failure applicability. However, this requirement only
collected subsequent instrument out-of-calibration data.
As corrective action for this violation, the licensee reviewed all
available instrument out-of-calibration data for the RPS and other
components / systems which support the Maintenance Rule functions.
Functional failures identified were evaluated against performance
criteria to determine whether (a)(1) status should be assigned.
Although six condition reports were issued to evaluate additional
functional failures, no system was required to be classified (a)(1)
based on this review. The inspector reviewed the licensee's corrective
actions and held discussions with a)plicable management and engineering
personnel concerning this issue. T1e inspector concluded that the
licensee had taken the necessary corrective action to correct the
deficient condition and had taken appropriate corrective action to
prevent its recurrence. This item is closed.
III. Enaineerina
El Conduct of Engineering
El.1 Review of Enaineerina Procedures
a. Insoection Scoce (37550)
The inspectors reviewed the licensee's procedures which control the
environmental qualification program.
. _ _ _ _ _ _ _ _ _ _ _ _ -
4
4
17
b. Observations and Findinas
- The inspectors reviewed the procedures listed below which control
various activities related to the environmental qualification 3rogram to
determine if the procedures implement the requirements of 10 C:R 50.
Appendix B. and 10 CFR 50.49. The following procedures were reviewed:
EGR-NGGC 0005. Engineering Service Requests. Rev 6. dated
Septembe" 5. 1997
EGR-NGGC-0007. Maintenance of Design Documents, Rev. 2. dated
August 22, 1997
'
EGR-NGGC 0153. Engineering Instrument Setpoints. Rev. 3. dated
- August 22. 1997
l EGR-NGGC-0156. Environmental Qualification of mlectrical Equipment
l Important to Safety. Rev. 4. dated October 8.1997
ENP-13.6 Equipment Data Base System. Control and Revision
Rev. 12. dated June 25. 1997
MCP-NGGC-401.. Material Acquisition (Procurement Receiving, and
Shipping). Rev. 3. dated August 26, 1997
The inspectors verified that the procedures provided adequate
instructions for establishing, maintaining and implementing the
requirements of'10 CFR 00.49 except for the issues discussed
below.
Section 9.6 of procedure EGR-NGGC-0156 provided the guidance for
maintaining E0 qualification data packages (ODPs). The procedure
specified that changes to ODPs are to be captured using the ESR
process. The procedure required that ODPs were to be periodically
updated as necessary to maintain auditability, to incorporate new
'
requirements, to meet plant specific requirements, ard to keep the
number of outstanding-changes at a reasonable level. However
5
procedure EGR-NGGC-0156 did not specify a clear time requirement
for updating the CDPs. The inspectors also determined that
procedure EGR-NGGC-0007 did not provide any requirements for
updating ODPs. The failure to s]ecify specific criteria in
procedures could result in the 0)Ps becoming unauditable which is
contrary to the requirements of 10 CFR 50.49. The failure to
maintain and u]date the ODPs was one of the causes of the
violation whic1 resulted in the civil penalty identified in NRC
Inspection Report (IR) 50-325(324)/96-14. The failure to
establish clear, definite requirements for updating ODPs was
identified as a violation example at the Shearon Harris Nuclear
Plant in NRC IR 50-400/97-12. Since all Brunswick 00Ps are being
revised and updated at the current time, a violation was not
identified for this issue during the current inspection. The
licensee's corrective actions for the Harris plant will resolve
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18
this problem since the Harris. Brunswick, n.d H. B. Robinson
plants use the same corporate EGR-NGGC ?,ocedures.
Procedure EGR NG D 0153 provides the methodology to establish
instrument setpoint margins sufficient to account for various
instrument uncertainties and environmental effects including
temperature, pressure, radiation, seismic, and insulation
resistance errors
Although procedure EGR-NGGC-0153 provided guidance on the
treatment of environmental effects, the inspectors noted that in
the discussion of temperature effects, the applicability of vendor 3
worst case performance specifications to plant specific conditions i
was not clear. The inspectors also noted that requirements for
seismic effects in procedure EGR-NGGC-0153 were not clear
regarding t6 match / confirmation of vendor profiles to plant
specific [ les or configuration,
in addition, the inspectors noted that procedure EGR-NGGC-0153
referenced Drawing 0-03056. Service Environment Chart Normal &
Accident Conditions. Units 1 & 2. for information on accident
temperature data to be used in instrument setpoint calculations.
The inspectors determined that-Drawing D-03056 was " frozen" on
December 12. 1996, and was not available for use. The reason for
removal of Drawing 0-03056 from use was documented in CR 96-04002
which identif9d the need to revise. and update Drawing D-03056-to
incorporate f icironmental data from the Reactor Building
Environmentai Renort (RBER), Revision 5. The inspectors noted in
review of calculations initiated since December 1996, the RBER
was referenced for temperature profiles in the re:ctor building.
The licensee indicated that a revision to EGR-NGGC-0153 will be
initiated to resolve inconsistency in wording regarding the
application of accident temperature / seismic effects to make it
clear that vendor test results would fully envelope site specific
profiles unless an evaluation has been aerformed to evaluate the-
differences. Additional guidance will 3e included to characterize
the requirements for engineering reviews of test-data to ensure
seismic and environmental profiles are bounding for site specific
conditions. The licensee indicated procedure EGR-NGGC-0153 will
also be revised to either remove D-03056 as the reference for
temperature data and replace it with the appropriate reference
(the RBER) . or to correct the drawing.
The inspectors also identified that procedure EGR-NGGC-0153 unde-
Section 9.5.1. Calibration Errors, was not clear regarding
instrument calibration surveillance requirements for as-left, as-
found or leave-alone zone tolerances. The licensee indicated
that procedure EGR-NGGC-0153. Section 9.5.1. would be revised to
clarify these requirements to indicate that calibration tolerances
are the defined limits, above and below a desired value, within
which an instrument loop signal may vary and not require
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adjustment. Licensee engineers stated that calibration tolerances
are understood to be "as-left" values.
The inspectors will review Procedure EGR-NGGC-0153 in a future
inspection to followup on these issues. An ins)ector followup
item (IFI). 50-325(324)/97-13 06. Revisions to )rocedure EGR-NGGC-
0153, was identified to the licensee pending further review by
liRC.
c. Conclusions
With the exception of the issues discussed above, the inspectors
concluded that the licensee's procedures for implementation of the
Environmental Qualification com) lied with the requirements of 10 CFR
50.49 and 10 CFR 50. Appendix 3. An IFI was identified to review
procedure EGR-NGGC-0153 to verify that the licensee incorporates the
above comments and clarifications. The reference to a " frozen" drawing
to obtain accident temperature data and the wording inconsistencies
discussed above were identificd to the licensee as a weakness.
El.2 Review of Instrument Setooiit Calculations
a. Insoection Stone (37550) ,
The inspectors reviewed randomly selected instrument setpoint
calculations to deternine the adequacy of the licensee's calculations.
b. Observations and Finninos
The inspectors reviewed the instrument setpoint calculations
listed below and verified that the calculations were completed in
accordance with NRC requirements. The inspectors verified that
the calculations incorporated industry standards. Updated Final
Safety Analysis Report commitments. Technical S)ecification
requirements, and recommendations contained in iRC Regulatory
Guides. Calculations reviewed were as follows:
-
-Calculation OE41-0036. Power Uprate HPCI Steamline Flow High
Uncertainty and Scaling Calculation.
-
Calculation ORWCU-0010. U1/U2 RWCU Flow Accuracy
Calculation. Units 1 and 2 RWCU Differential Flow Leak
Detection / BESS I&C.
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Calculation 0821-0068. Power Uprate Main Steam Line Flow
High Setpoint Uncertainty and Scaling Calculation.
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Calculation 0-01534A-297. Insulation Resistance Degradation
Calculation.
From review of System Description SD-01.2. Reactor Vessel
Instrumentation. and the Safety Evaluation by the Office of
, _ _ _ _ _ _ _ _ _ a
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20
Nuclear Reactor Regulation. Conformance to Regulatory Guide 1.97
Revision 2. Brunswick Steam Electric Plant. Units 1 and 2. Dated
May 14. 1985. the inspectors concluded that these calculations
were. typical. The instrument setpoint calculations typically
considered 140 F as the maximum temperature in the calculations.
From review of the calculations, the inspectors determined that
instruments that perform a safety function are analyzed for a LOCA
environment in the reactor building. The calculations showed that
instrument uncertainties considered instrument temperature effects
for a maximum temperature of 140' F which is bounding for the
analyzed LOCA environment.
The inspectors also determined that instruments relied upon to
mitigate the effects of a high energy line break (HELB) were also
evaluated by the licensee. For this instrumentation,
environmental uncertainties-for a harsh environment were not
required to be considered since the instrumentation function would
occur before the reactor building temperature )rofiles listed in
p the Reactor Building Environmental Report (REBR) Revision 6.
dated November 5. 1997, would reach 140 F and affect instrument
performance. The ins)ectors noted that abnormal temperatures were
not discussed in the-RBER. Discussions with licensee engineers
disclosed that the design base accident event is based on an
initial building environment airspace temperature of 104 F. The
building temperatures ace measured and recorded daily by plant
operators in accordance with procedure numbers 101-03.4.1 and 201-
03.4,4. Unit 1 and 2 Control Operator Daily Check Sheets.. The
= operators are required to contact the duty engineer when the
reactor building temperature exceeds 104 F so that engineering
can perform an assessment of the effects of temperature on
environmental qualification.
-The inspectors noted that calculations for instrumentation which
mitigates a HELB demonstrated that the instrument and associated
equipment would not be exposed to a harsh environment before the
instrumentation performed its safety function. In the instrument
calculations reviewed by the inspectors instrument setpoints were
based on a maximum temperature of 140 F (non-steam environment).
Although allowances were not made for a harsh environment. a
seismic allowance was included in the calculations.
Review of the temperature profiles as shown in the Brunswick
Reactor Building Environmental Report showed that the actuation
isolation signal would occur before exceeding the temperature
allowances assumed in the setpoint uncertainty calculations. An
exce) tion was the High Pressure Coolant Injection (HPCI) line
breat in the steam tunnel where the temperature profile showed
that 140 F would be exceeded for ap3roximately 2.5 seconds before
the isolation trip _ signal occurs. iowever this instrumentation
would remain operable based on thermal delays. However, the HPCI
isolation function would most likely be initiated by temperature
.
21
sensors in the steam tunnel or HPCI room which would occur
imediately with no time delay.
The inspectors concluded that the instrument setpoint calculations
complied with NRC requirements and were technically adequate.
Review of the calculations showed that environmental effects,
j- specifically accident temperature, were correctly evaluated in the
calculations,
c. Conclusions
The inspectors concluded that the licensee's calculations were
technically adequate and complied with NRC requirements. The
inspectors concurred with the licensee's conclusions that the
setpoints for instruments relied upon to mitigate the effects of a
KLB did not require inclusion of uncertainties for a harsh
environment since the instruments perform their ft..iction before
being effected by the harsh environment. Setpoints for
instruments required for LOCA effects include the appropriate
environmental uncertainties.
-El.3 Enaineerina Service Reaucst (ESR) 97-00426
a. Inspection Scoce (375501 '
The inspectors reviewed ESR 97-00426 which was prepared to address
questions on instrument setpoints.
b. Observations and Findinas
A review of procedures and various documents by an independent
consultant resulted in questions involving environmental effects
including uncertainties on instrument accuracy. These guestions .
were dccumented in an E-mail message dated June 20, 1997 Subject: '
E0 and Instrument Accuracy. The licensee addressed the referenced !
memo in Engineering Service Request ESR 97-00426. Revision 0.
-dated September 18. 1997. ESR 97-00426 documents the evaluation
completed by the licensee to address environmental effects on-
instrumentation. The inspectors noted that the licensee response
did not address the questions in the June 20, 1997 E-mail message
point by point. but provided an evaluation that was more generic
in nature. The inspectors noted that ESR 97-00426 was an
engineering disposition (ED) type ESR. as defined in procedure
EGR-NGGC-0005. The use of this type ESR to respond to the E-mail
cuestions was appropriate since the ESR only communicated existing
cesign requirements, did not produce design output, and did not
change existing engineering documents.
The ESR concluded that instruments that aerform a safety function
are analyzed for a LOCA environment in t1e reactor building. The
instrument uncertainties consider -instrument temperature ef fects
for a maximum temperature of 140"F which is the maximum bounding
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22
temperature for the analyzed LOCA environment. The inspectors
noted that the word minimum had been incorrectly used in the
fourth line, third paragraph in Section 2.0 of the ESR. The
licensee stated that they will correct this error when the ESR is
revised. as discussed below.
ESR 97-00426 also concluded that harsh environmental effects have
been appropriately accounted for in safety related uncertainty
calculations. The ESR concluded that the isolaticr. aquence for a
HELB due to main steam line break. reactor core isolation cooling
l steam-line break, high pressure coolant injection steam line
break, cr a piping failure in the reactor water cleanup system is
such thtt the isolation function will occur before the
instrumentation is exposed to harsh environmental effects. This
conclusion was based on the instrumentation being able to perform
its safety function prior to the temperature exceeding the
temperature allowance assumed in the setpoint calculations. For
area temperatures exceeding the setpoint temperature uncertainty
allowance, the use of emergency operating procedures (EOPs),
operator action, and local temperature instrumentation would
mitigate the event and provide the actions to determine and/or
maintain. reactor level during a LOCA or HELB.
When temperatures exceed the temperatures (140 F) assumed in the
setpoint calculations, plant operation is controlled through the
' COPS. A review'of E0P-03-SCCP Revision 5. Secondary Containment
Control Procedure, and 2EOP-LPC Revision 1. Level / Power Control,
shows that high area temperatures are an entry condition into
secondary containment control procedure E0P when area temperatures
exceed the maximum safe operating value requiring manual reactor
sCrdm.
E0P-03-SCCP Revision 5. refers the operators to Caution 1 to
determine reactor level instrumentation operability. A review of
Caution 1 disclosed that vessel level wide range instrumentation ;
8B21 - LI - R604A/604B and C32 - PR - R609 are not to be used when
secondary containment temperature exceeds 140 F. This exclusion
was because the reference leg and associated instrumentation for
these loops are in secondary containment. E0P Caution 1 then
)rovided compensation data for the remaining level instrumentation
]ased on drywell tem]erature, reactor saturation limit, and
reactor pressure. iowever, for secondary containment
temperatures above 140 F. Caution 1 instrumentation may not be
o)erable with instrumentation exposed to temperatures greater
tlan 140*F during an event. In cases when vessel level can not
adequately be determined, the E0Ps direct the operators to
depressurize by initiating ADS and flood the vessel using low
pressure emergency core cooling systems.
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c. Conclusions
The inspectors concluded that the licensee adequately addressed
the questions in the June 20. 1997 E-mail message regarding
instrument and E0 accuracy. However, the licensee stated that
_
they will revise F.SR 97 00426 to address each question and
recommendation ir. the E-mail message point by point to further
clarify their response to the concerns / issues raised in the
June 20, 1997 E mail message.
El.4 Environmental Qualificat%1
,
a. Insnection Scooe (37550.92903)
The inspectors reviewed the licensee's corrective actions for the
Environmental Qualification (FO) program, in response to findings
l identified during Self-Assessment numbers 95-0041 and 96-0271 and
the violations identified in NRC IR 50-325(324)/96-14.
b. Observations and Findinas
1) Review of E0 Equipment Data Base
The licensee's corrective actions to resolve the discrepancies in
the E0 program identified by NRC (See IR 50-325 324/96-14)
include corrections to and updating of the Equipment Data Base
System (EDBS). Numerous errors in EDBS had been identified and
corrected by the licensee since the inspection findings were
identified in IR 50-325(324)/96-14. The errors in EDBS were .
identified during E0 equipment walkdowns and review of various !
data bases. In addition, numerous errors were identified in the
EQ zones listed in EDBS for the location where various components
were installed. These primarily occurred at. zone boundaries and
were being resolved during review of walkdown data.
The requirements for. recording and correcting E0 data in EDBS was
s)ecified in- CP&L procedures EGR-NGGC-0156 and ENP-33,6. The
-c1anges to EDBS to correct errors were processed using Form 100 of
ENP-33.6. The Form 100 was design verified in the E0 unit and was
then forwarded to appropriate personnel for entry into EDBS. All
EDBS data entries made were independently verified by personnel in
the Configuration Management group in the Design Control Unit.
The independent verification was performed to minimize o-
eliminate data entry errors. Additional corrections to EDBS were
ongoing to incorporate E0 walkdown ins)ection results and the
revisions to EQ qualification data paccages.
The inspectors reviewed some randomly selected revisions to EDBS
identified as a result of the E0 corrective actions and verified
the EDBS data had been corrected. The inspectors also discussed
the program for control of changes to EDBS with various licensee
personnel who perform the day to day system revisions. These
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24
discussions disclosed that these individuals were cognizant of the
requirements for controlling and making corrections to EDBS.
2) Review of Qualification Data Packages
The inspectors reviewed a draft cop.f of Revision 4 of ODP No. 49.
titled. " Qualification Data Package For NAMCO EA180 Series Limit
Switches" to determine if it adequately demonstrated environmental
qualification for the safety related NAMCO switches for use inside
the drywell in accordance with 10 CFR 50.49 and appropriate
licensee E0 Prccedures. The package addressed the following:
qualification level (0588 Cat. I); tag numbers of equipment
covered in the QDP: test report aaplicability; similarity of test
specimens to installed equipment: E0 parameters. temperature,
pressure, relative humidity, radiation, chemical spray,
submergence; cualified life: E0 maintenance requirements; test
anomalies; anc operating experience items.
During review 3f the Draft ODP. the inspectors identified the following
questions / comments:
.
The text in the CDP indicates that there were five anomalies in
.
'
Qualification Test Report (OTR) 130 but only four anomalies were
discussed in the ODP.
l .
Attachment 2 to the ODP included a calculation for qualified life
l of the limit switches which was not signed as reviewed.
- Differences were noted in the system component evaluation
worksheets (SCEW) for the same limit switches in the different
units.
- Data was missing from some of the SCEW sheets. That is, there
were blanks on the data sheets. For example, data on accuracy was
left blank.
Some components were specified with Anaconda flex and others just
stainless steel flex conduit. Additionally, only certain
components were specified for weep holes.
- Page 49 section 4.1 Installation requirements indicates that the
conduit seal may not be necessary for those limit switches
installed in the Reactor Building. This requirement should be
clear and should specifically list those limit switches which
require conduit selling to ensure qualification.
- Page 13 lists the 16 Namco EA180 limit switches which had been
installed. However only 14 were considered qualifieo by this ODP.
Unit I limit switch tag numbers 1821-ZS-5373 and 1B21-ZS-5374 were
excluded from the E0 requirements by ESR-97-00431. The Unit 2
equivalent switches were not discussed in the ODP.
. _ _ _ _ .
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,
25
- In Section 2 of the 00P it was stated that it was a good
maintenance practice to lubricate the NAMCO limit switches.
however. lubrication was not specified in Section 4 of the ODP
which lists recommended maintenance practices.
In Section 4.2 of the ODP it was stated that the switches can be
refurbished. However, a statement was made on page 21 that
qualified replacement part kits were no longer available.
- A reference was made to abnormal temperatures on page 38 of the
ODP. However, abnormal temperatures were not included in DR 227.
- The inspectors questioned apparent inconsistencies between
activation energies and aging methods discussed in referenced
qualification test reports (OTRs).
The licensee indicated that these comments would be evaluated by
the E0 group and if appropriate, addressed in Revision 4 of the
QDP when it is completed.
The inspectors reviewed a draft copy of Revision 7 of ODP-67
General Electric Company IC 7700 Series Motor Control Centers for
BNP. The GE MCCs. located or, the 20, 50, and 80 foot elevations
of the Units 1 and 2 Reactor Buildings, are subject to harsh
environments resulting from postulated design basis accidents and ;
have a safety function to mitigate the consequences of these F
accidents. The MCCs were qualified in ODP-67.
A series of similarity analysis were performed to demonstrate
similarity between the tested configuration and supplied. The
inspectors reviewed portions of DR 232. "Nutherm Report No. CPL-
7806R. Qualification Test Results Applicable to Brunswick Nuclear
Power Plant Safety-Related GE 7700 MCCs." Revision 0, dated
June 30, 1997 which dccumented the similarity analysis. Section 2 of
DR-232 contains a discussion on the similarity analysis between the
components tested by NUTHERM and those installed in the Brunswick MCCs.
The similarity discussion covers fuses, stab assemblies, control
transformers control and power wiring, overload heaters. overload
relays, terminal boards, starters and contactors, molded case circuit
breakers. circuit protectors. disconnect switches. potentiometers, and
indicating lights. The similarity analyses were based on the similarity
analyses contained in DR 1.1. GE Company NEDC-30696-P. May 1985. MCC
Oualification Test Report Phase Il for CP&L Brunswick Plant, or were
devices which could be directly linked to a test specimen and did not
require a similarity analysis. Based on review of DR-232 NRC concluded
that NOTHERM was able to establish that the com3onents they tested were
in the same family as those provided by GE in t1e MCCs. This review was
also dccumented in IR 50-325(324)/97-09.
A draft copy of Revision 0 of ODP 99. R. G. Laurence Series 500
and 600 Solenoid Valves was reviewed. The inspectors verified
that similarity analysis was included in the ODPs.
>
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26
3) Review of EO Walkdown Data
The inspectors reviewed E0 walkdown data which document inspection
of E0 equipment-in the Unit 2 MSIV pit and drywell, and Unit 2
reactor building. Tha E0 walkdowns were performed in accordance
with CP&L Special Procedure OSP-96-014. EQ Equipment Field
Verification. The pyrpose of the walkdowns was to verify the
accuracy of the manulacturer/model number listed in the licensee's
data bases and to verify the equipment installed orientation and
configuration were in accordance with the E0 qualification
documentation. The ins)ectors reviewed walkdown records for scram '
pilot' solenoid valves, 1AMC0 limit switches, temperature elements,
excess flow check valves, and pressure switches. The walkdown
data was recorded on field inspection data sheets which were'then
converted into an electronic data base. The inspectors verified
that discrepancies identified during the walkdowns were documented
either on a work request (WR/J0) for repair, or in a condition
re) ort (CR). The ins)ectors reviewed completed WR/JO numbers 97-
AF JR1, 97-AFUR2, 97- A UR3, and 97-AFUR4. These WR/J0s document
drilling of weepholes in junction boxes in the Unit 2 MSIV pit to
resolve a moisture intrusion issue. These boxes are associated
.
with limit switches for the Unit 2 main steam isolation valves.
L The completed WR/J0s showed that the weepholes were drilled to
resolve the concerns. The inspectors did not identify any
- discrepancies in the records reviewed.
4) Review of Environmental Qualification Condition Reports
The inspectors reviewed the licensee's corrective c.,ctions to
L disposition the CRs listed below. These CRs were initiated by the
licensee to-document and disposition nonconforming items whicn
were identified during the ongoing E0 reconstitution project. The
nonconforming items were identified as a result of E0 equipment
walk h ns, review cnd updating of E0 equipment ODPs, omissions
from the original program, or changes to the operating
environment. The CRs reviewed were as follows:
CR 97-02015
The licensee initiated CR 97-02015 on June 6. 1997 to document and
disposition deficiencies that had been identified by the
licensee's training staff during observation of simulator training
when the fire protection system had not been isolated within the
15 minute time period after initiation of a HELB specified in
31 ant o)erating 3rocedures. The 15 minute time period is the
) asis w1ich esta)lished flood '.evels for E0 e
and north and south RHR and core spray rooms.quipment in the HPCI
Review of closure
for CR 97-02015 disclosed that the licensee concluded that the
issue has been adequately addressed by operator training,
primarily through critiques which were held following the
completion of the simulator training to discuss deficiencies noted
during the training. In response to the CR. Action Items were
Y
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27
'
assigned to the Operator Training group to incorporate the basis
,
for the need to isolate the fire protection system into training
materials. However, review of the training records on June 12,
1997, by personnel from the E0 group resulted in additional
questions regarding the licensee s corrective actions. The
records reviewed by the E0 personnel indicated that during
'
simulator training, approximately 10 to 20 percent of the
operators were failing to enter AOP-05,0, Radioactive Spills. High
. Radiation, and Airborne Activity, or were entering the AOP late
- (after 15 minutes). The inspectors made an indepen6nt review of
, the training records reviewed by the E0 personnel. This review
disclosed that the records the E0 personnel reviewed on June 12,
1997 were for the six month
02015 (January - June 1997) The .
period prior to reviewed
inspectors initiation of CR 97-
training
records for July - September, 1997 and noted significant
improvement in this area, although the HELB scenario was not
included as part of the simulator training exercises in this time
period. The training scenario did include a torus leak which
required entry into A0P-05.0.
I
The inspectors noted that the concern regarding flooding of
_
instruments could also be caused by other accidents such as pipe
L breaks in the service water or Reactor Building Closed Cooling
l Water (RBCCW). Operator actions in these cases would be directed
e by E0P-03 SCCP Secondary Containment Control Proccdure (SCCP),
based on high water leve'is in the HPCI and north and south RHR and
core spray rooms. An uttry into E0P-03-SCCP would also result
from flooding in these same rooms caused by activation of the fire
protection system. As aaditional followup on this issue, the
inspectors observed simulator training scenarios performed on
December 3 and 17, 1997. Included in the scenario was a RCIC
steam line break (HELB) and activitation of the fire protection
system. Both crews participating in the training scenario
isolated the fire protection system within the 15 minute time
period. The inspectors also questioned some randomly selected
reactor operators regarding the need for entry into A0P-05.0
following a HELB. The operators were cognizant of the basis of
the actions in A0P-05.0 (need and reason for isolating the ' ire
protection s
CR 97-02015.ystem) and were familiar with the problem addrc ses by
The inspectors verified the action items associated with the CR
were completed. CR 97-02015 was closed on December 11. 1997.
CR 97-01841. 97 02025. & 97-02408 These CRs documented various
issues regarding possible effects of moisture on E0 equipment. CR
97-0184) was initiated to document the effect of spray from the
fire protection system on E0 equipment in the reactor building.
The licensee has resolved all the issues associated with this CR
except for drilling of weepholes in junction boxes whicn may be
affected by the water spray. Licensee engineers are currently
- preparing instructions and procedures for completing this work.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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28
The problem documented in CR 97-02025 concerned an issue which had
been the subject of IE Circular 79-05. Moisture Leakage in
Stranded Wire Conductors, which was issued by NRC on March 20.
1979. This affects Patel seals which were used to seal some
stranded wire conductors in instrument circuits. CR 97-0?408
documents several other moisture intrusion issues. The immediate
corrective action taken to resolve these issues, as documented in
CR 97-02408 was to hire an outside consultant to address the
issues. The consultant has reviewed many of the issues documented
in CR numbers 97-01841, 97-02025. and 97-02408 and made
recommendations, some of which have been implemen.ed. The
consultant also addressed another issue in the CRs involving
current leakage in control circuit and the possible impact on ODPs
and E0 of equipment. This concern was the effect of moisture
intrusion through stranded wire conductors, sealed with Patel
seals, which could result in leakage currents in instrument
circuits. ESR 97 00440 was issued for the 120 volt AC circuits and
ESR 97-00441 for DC circuits. These ESRs are currently being
reviewed by licensee engineers. The current leakage issue was
also applicable to questions raised regarding the NAMCO limit
switches. The inspectors will review the licensee's evaluation of
current leakage and its ap311 cation to evaluation of E0 equipment
in a future inspection. T11s was identified to the licensee as
IFl 50 325(324)/97-13-07. Review Technical Evaluation of Current
!
Leakage and the Effect on EQ Equipment. pending further review by
NRC.
The licensee also aerformed an evaluation of the potential for
moisture wicking t1 rough Patel seals. This evaluation was
i documented in ESR 97-00423. 03erability Evaluation - Wicking.
Review of the ESR disclosed t.at the licensee performed a detailed
evaluation of the Patel seals by comparison of the installations
at Brunswick with the configurations tested by NRC at Sandia
Laboratorics (NUREG/CR 0699. Jublished March.1979). The
licensee's conclusions were t1at the design function of the
instellea equipment will not be effected by moisture intrusion
through the stranded wire. The ESR was based on a review of the
duration of the design accidents and the resulting leakage
currents caused by moisture intrusion into limit switches.
Further review of this ESR will be performed as part of IFI 50-325
(324)/97-13-07, discussed above.
CR 97-02016 & 97-02074
CR numbers 97 02016 & 97-02074 were initiated to document issues
involving NAMCO limit switches. The following issues were
identified in the CRs:
- Inability to identify the date of manufacture of the switches
since the codes for date of manufacture were painted over.
__ ________ ____
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I
29
- Potential for paint to impair the operability of the switches.
The concern was that paint on the roller arms would impair
mechanical function of the switches.
'
- Chemical reaction between paint and internal switch components
would cause corrosion of switches, leading to failure of the
switches.
- Use of incorrect qualification test reports (0TRs) in the
qualification test reports which qualified the switches.
- Effect of current leakage on switch operability.
A total of 14 NAMCO limit switches were covered under the E0
program. These switches were installed during modifications
completed in 1983 and 1984 The licensee has determined that none
of the switches were purchased or manufactured prior to 1980.
Therefore, the concern raised by IE Bulletin 79-28. Possible
Malfunction of NAMC0 Model EA 180 Limit Switches at Elevated
Temperatures, would not apply to the switches installed at
Brunswick, Review of the licensee's response to IEB 79-28
disclosed that none of the potentially defective switches had been
purchased by the Brunswick site.
Review of the i1censee's corrective actions completed to date
disclosed that the following actions have been completed:
The licensee has identified the date of manufacture for most of
the NAMCO limit switches. Additional manufacture dates may be
identified when the Unit 1 walkdowns are completed during the
Spring 1998 refueling outage. However, the licensee has
conclusively determined that none of the switches would be
affected by the defects identified in IEB 79-28.
.
The switches were stroked in accordance with frequencies per the
Technical Specifications which demonstrates that the mechanical
function of the switches had not been impaired by the paint.
- The paint has been tested. The test results show the
not cause corrosion or deterioration of the switches paint would
. The ODP. has been revised to incorporate the correct OTRs. The
ODP. ODP 49, was still in draft.
. The current leakage issue has been evaluated " ESR numbers 97-
00440 and 97-00441, which are currently being reviewed by licensee
engineers.
The licensee subsequently has determined that the switches were
still within their qualified ;'fe. No equiament operability
issues related to tv.e NAMCO ilmit switches lave been identified.
_
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30
[R 97 02367
This CR was initiated on July 3, 1997 to document the failure to
initiate CRs for nonconforming items, specifically, MCC door
gaskets and non standard Raychem splices identified as a
violation by NRC during an inspection documented in NRC 1R 50-325
(324)/97 08.- The licensee's corrective actions included
completion of a review of all the E0 walkdown data sheets to
identify any nonconforming equipment. Additional corrective
actions included training of personnel in the E0 group regarding
the corrective action program and assessment of the effectiveness
of the corrective actions. These correcthe actions were also
associated with other similar corrective action CRs. such as CR
97 01972 and CR 97-02465. The inspectors reviewed the completed
corrective actions and concurred with closure of CR 97 02367. The
CR was closed on December 14. 1997.
CR 97-02465 and 97-02672
This CR wac initiated on July 15, 1997, to document concerns on EQ
operability determinations. This CR referenced CR numbers 97-
01841, 97 02025. and 97 02408. discussed above, which involve
moisture intrusion issues. As a result of the concerns raised in
CR 97 02465, the E0 group presented an action plan to resolve the
moisture intrusion issues (CR 97 02465) to the plant nuclear
safety committee. Although, further review showed the operability
determinations for the three CRs were correct, the root cause
analysis concluded that there were other problems which resulted
in CR 97-02465.
The root cause of CR 97-02465 was attributed-to weak E0 project
management. The root cause/ event review for the CR listed the
causal-factors indicative of weak E0 3roject management to be poor
communications within the E0 group, tie site position that E0
problems were primarily docunitation problems, and a poor
corrective action culture within the E0 group. The poor
corrective action culture was evidenced by corrective action items
which were routinely extended, overdue, or completed late: failure
to prepare JCOs: numerous CRs written against the E0 grou) for
improper corrective actions: and closing CR action items )y other
action items without completing the corrective actions. A
violation of NRC requirements was identified in IR 50 325, 324/97-
12 for failure of the licensee to implement their corrective
action program.
The licensee's corrective actions to address the issues raised in
CR 97-02465 included increased management oversight aerforming a
review of the E0 project schedule to complete the higlest priority
work activities first, conducting more frequent E0 group meetings
to improve communications within the E0 group, transferring some
E0 group functions from the Design Control l%1t to a site
organization. and performance of an effer' ve. dss review of the
.
.
.
31
completed corrective actions. The CR was closed on December 17,
1997. The inspectors reviewed the completed corrective actions
.and concurred with closure of the CR. The ins)ectors concurred
with the licensee's conclusions that the opera]ility
determinations for the three referenced CRs were appropriate. NRC
will perform review of the liccasee's actions to correct the l
violation in future inspections,
CR 97-02672, which was inniated on August 5. 1997, indicated that
the Supervisor comments listed in CR 97 02465 were a misstatement
of the consensus of opinion of individuals which met to discuss CR
97-02465. Review of CR 97-02672 disclosed that the CR did not
raise any new issues or conceriis which had not been addressed by
CR 97 02465. CR 97-02672 was closed on December 17, 1997. NRC
concurs with the licensee's conclusions and closure of the CR.
CR 97 4059
This CR was initiated on December 2, 1997, to document concerns
and questions on ESR 97-00426. The questions involved
appropriateness of E0P actions, the need to include evaluation of
drywell instrumentation in tic ESR, and various questions on
instrument setpoints. The 1 Lensee completed a review of the
questions raised in the CR and concluded that the ESR had
addressed these issues, or the issues were beyond the scope of the
ESR, For exam)le, appropriateness of E0P actions were approved by
NRC for all BW1s and do not involve instrument setpoints. There
are no instruments in the drywell which provide signals for
automatic actuation. The inspectors reviewed the licensee's
responses to the questions in the CR and concurred with the
licensee's conclusions that no new corrective actions were
required to resolve the concerns / questions raised in CR 97-04059-
which had not been previously resolved.
5) Review of Environmental Qualification Requirements in
Procurement Practices
Th'e inspectors reviewed CP&L procedure MCP-NGGC-0401, Material
Acquisition (Procurement. Receiving, and Shipping). Revision 4,
dated August 26, 1997. This procedure specifies the instructions
for procurement of safety related materials for use in CP&L
nuclear plant. The inspectors noted that the requirements for
obtaining reviews by E0 engineers is specified in the procedure.
Discussions with licensee engineers and review of previous
revisions of the procedure disclosed that the procedure had been
revised to strengthen the need for the E0 review in Revision 2 of
.MCP-NGGC 0401, effective April 15. 1997. Revision 2 added
- - - - -
requirements that components that require environmental
qualification:be reviewed by the E0 group.
During review of CRs. the inspectors identified several examples
of acceptance of materials / equipment by procurement engineering
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
32
for use in E0 installations which were based on test reports which ;
had not been reviewed by the E0 group -These were documented in
'CR numbers 97-01970 and 97-03036, Several additional examples of
discrepancies in documents prepared by procurement engineering
which affected E0 equipment were also identified during review of
procurement specifications and other documents su * as material
evaluations. These discrepancies were documented in CR 97-04035
which tas initiated on November 25, 1997. The review of
procurement documents was being performed as part of the
corrective actions to address the E0 program discrepancies
identified in IR 50-325(324)/96 14. This was listed as Commitment *
- 4 in the licensee's December 19, 1996 Reply to Notice of ,
Violation,
6) Equipment Lubrication Requirements
The inspectors reviewed CP&L procedure MMM-053. Equipment
Lubrication Application Guidance and Lubricant Listing,
Revision 6 dated November 11, 1997. This procedure provides a
listing of plant equipment with recommended lubricants to be used,
guidelines for lubrication of plant equipment, and lubricant
sampling methods. The inspectors identified the following issues
after reviewing the procedure:
-
ODPs 26, 68, and 88 were not referenced in procedure MMM-
053. These ODPs cover environmental qualification of
Reliance electric motors.
- Document References corresponding to above ODPs were not
referenced.
- The types of lubricant specified fo, the Reliance motors in
procedure MMM 053 differ from those listed in the ODPs 26
and 68.
-
Procedure MMM 053 permits maintenance to change the
lubricant without obtaining engineering review or approval.
Discussions with licensee engineers disclosed the CR 97-04015 was
initiated on November 20, 1997, to document the fact that the
procedure permits changes to lubricants without performance. of an
engineering review. Action Item 40 to CR 97-02627 was issued to
document a similar issue. This action item was closed by CR 97-
04015.
The inspectors determined that the licensee had not evaluated that
the type of lubricants (Mobil) specified-in procedure MMM 053 for
Reliance electric motors differed from those listed in ODP 26 and
68, Review of ODP 26. Revision 1. Joy Fan /Peliance Electric
Company, Class 1E Continuous Duty, 20 HP, and ODP 68, Revision 5.
-Standby Gas Treatment System - Fair Company Filter Unit and
Control, showed that the electric motors were both qualification
_ _ _ _ _ - _ _ _
.
33
'
tested using Chevron SRI 2 grease. The impact of using a
dif ferent type of grease to lubricate the motors on the
environmental qualification testing of the motors had not been
documented by the licensee. The licensee initiated CR 97 04064 to
document the fact that substitution of alternate lubricants had
not been evaluated by E0 engineers. The failure to establish
maintenance procedures appropriate to the circumstances for
performing maintenance was identified to the licensee as another
example of violation item 50 325(324)/97-13-02. Inadequate
Procedure for the Conduct of E0 Preventive Maintenance.
c. Conclusions
1
One violation example was identified regarding an inadequate E0
maintenance procedure for lubrication of E0 electric motors. Two
inspector followup items were identified to followu) on revisions
to instrument setpoint procedures and to review leacage current
calculations. The licensee was making progress in resolving and
closing CRt identified by the E0 group. As of the inspection
dates, no 0DPs had been issued.
E5 Engineering staff Knowledge and Qualification
E5.1 Trainino and Qualification of E0 Personnel
a. Insnection Scone (37550)
The inspector reviewed the licensee's program for training and
qualification of personnel in the E0 task force. including both
CP&L and contract engineers,
b. Observations and Findinos
The requirements for performance of E0 equipment walkdowns are
specified in CP&L Special Procedure OSP-96-014. E0 Equipment Field
Verification. The prerequisite in procedure OSP 96-014 for
individuals performing the walkdowns was to read the procedure.
The licensee qualified a number of individuals to perform the
field walkdowns through a training program conducted in accordance
with CP&L procedure TI-100. Conduct of Training. These
individuals included Instrumentation and Control technicians.
contract engineers, and personnel assigned to the E0 group who
were qualified E0 engineers. The training for the qualified E0
engineers consisted of reading the procedure. orientation and on-
the-job training to become familiar with the walkdown and data
s gathering process. For other personnel, the training included
reading of the procedures, formal classroom lectures.
demonstrations, performance of practical exercises, and on-the-job
training. The walkdown group supervisor performed a detailed
review of the result < of practical exercises and data gathered <
during initial walke is prior to signifying the individuals were
s. _ __
34
qualified to perform walkdowns. The training provided for the
walkdcwn personnel exceeded the procedural requirements. The E0
walkdown grou) supervisor stated that the level of training
provided to t1e walkdown personnel war to assure that the walkdown
results were very accurate and to preclude the need for repeat
work. The inspectors revieweJ the training records for the
walkdown personnel and verified that they had been trained in
accordance with the licensee's program. The inspectors noted that
the experience level for the walkdown personnel varied from a
recent graduate engineer to individuals with more than 20 years of
experience. The inspectors reviewed the walkdown inspection
records prepared by various individuals in the walkdown group and
noted that the original walkdown records were complete and
accurate, with some exceptions. Discussions with the walkdown
group supervisor disclosed that corrections noted on the records
were the result of reviews perfnrmed to resolve discrepancies in
the records. The changes were made as a result of additional
walkdown inspections which were doc'mented in the records. In one
case, an individual was terminated for failure to perform the
walkdowns and complete the walkdown records properly. This
individual's work was reviewed by the licensee and corrected where
necessary.
The inspectors also reviewed the training and qualification
records for E0 technical personnel. These records included
previous work experience, education and training, and CP&L
specific training applicable to the E0 project. This training
included E0 technical reviewer, E0 design verifier E0
calculations, and E0 ESR originator. The inspectors also
questioned the manager of the E0 group concerning work assignments
within the E0 grou). That is, assignment of specific activities
to individuals wit 1 previous experience in a particular area of
specialization, such as review of requirements for qualification
of motors or specific types of instrumentation. The E0 group
manager has recently are)ared a directory of all engineers working
within the E0 group w11c1 lists each engineer's experience and
what work activities they have completed for the E0 project at
Brunswick. The purpose of this directory was for the engineers
within the group to know who has worked on various problems and
issues so they could obtain assistance from these individuals when
they become involved with similar technical issues. The directory
was distributed'to personnel in the E0 group. The E0 group
manager provided a copy of the directory to the inspectors and
discussed the basis for the various work assignments within the
group.which were based on the past work experience of the E0
technical personnel,
c. Conclusions
The inspector concluded that the licensee's program for training
and qualification of E0 engineers meets NRC requirements.
,
.
.
35
E8 Hiscellaneous Engineering Issues (37551, 92903)
E8.1 (Closed) URI 50-325(324)/97-08-04: Control of Ecuioment Data Base
System (EDBS) Information
The licensee issued CR 97-02400. Non Validated EDBS Information,
concerning rc, tine use of non-validated EDBS information. This wes
associated with VIO 50 325(324)/97-08 03. Safety Relay Setting Change
Made as Pen and Ink Changes to Procedure. The licensee replied to this
violation on September 2. 1997. The reply discussed licensee corrective
'
action regarding the use of EDBS. Likewise. the licensee responded on
November 26. 1997 to VIO 50-325/97-11-01. Failure to Initiate Alternate
Safe Shutdown Impairment. addressed corrective action fcr use of an EDBS
non validated field for determination of an Alternate Safe Shutdown
impairment. Plant procedure OENP-33.6. Equipment Data Base System
Control and Revision, provides instructions for control of EDBS
information. Color coding of fields in the electronic database
represent the various types of data present. This procedure provides
direction that certain types of data are not to be used until verified.
Accordingly two previous violations address the use of non-verified
EDBS information. The licensee corrective actions for these violations
are being completed. The requirements for the control of information
are in procedure OENP-33.6. Previous items address the concern of this
URI. therefore this item is closed.
E8.2 (Closed) LER 50-325(324)/97-04: Soent Fuel Shionina Cask Handlina
Activities
This report documented the discovery by the licensee that the heavy load
analysis as described in tne UFSAR did not completely bound movement of
the shiroing cask from the primary lift to the secondary lift with the
valve box covers removed. It was determined that movement of the cask
with a non single failure proof yoke and less than full cask integr'ty
constituted an unreviewed safety question (US0) in accordance with the
requirements specified in 10 CFR Part 50.55 The failure to obtain
prior approval for a previously unanalyzed condition was determined in
IR 50-325(324)/97-12 to be a violation. In a letter to the NRC dated
August 6. 1997, the licensee requested a license amendment for review of
the US" The licensee re evaluated findings relative to the 30 foot
dro: ~cident and qualified the transfer yoke using guidance provided in
NUR b 0612. Control of Heavy Loads at Nuclear Power Plants. This
evaluation contended that a fuel shipping cask drop event was not
credible. therefore operation with less than full cask integrity was no
longer a problem due to acceptable redundancy in the lifting yoke. In a
letter to the licensee dated December 2. 1997, the NRC accepted the
licensee determination that operation with the valve covers removed
would not compromise the health and safety of the public due to
acceotable redundancy of the lift devices. Based on the acceptance by
the NRC of the licensee's evaluation and issuance of the enforcement
action as described in IR 50-325(324)/97-12 this item is closed.
.- . ._
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.
4
36
E8.3 (Closed) Inspector Followun item 50-325(324)/96-14-05.'Effect of EO
Accuracy on Instrument Setooint Calculations.
'
Review of procedures and various documents by an independent *
contJ1 tant had resulted in a number of questions regarding the
~
effect of environmental effects (uncertainties) on instrument
accuracy The questions / concerns were documented in an E mail *
message dated June 20, 1997. subjert E0 and Instrument Accuracy,
in order to address the issues raised in the June 20 E mail
message, a review of instrument setpoint calculations was
performed by licensee instrumentation and controls (l&C)
engineers. . The review was documented in ESR 97-000426, which was
'
discussed in paragraph El.3. above. The inspectors also reviewed
various instrument setpoint calculations (documented in paragraph
"
E1.2. above) and determined that E0 accuracy has been aroperly
considered in the instrument setpoint calculations. T1e
ins)ectors had no further questions regarding instrument setpoint
metloaology or accuracy at this time.
E8.4 JClosed) Violation item 50-325(324)/97-02-08. Failure to Imolement ao
nsoection Procram for Safety-Related Miscellaneous Structural Steel
The licensee responded to this violation in letters dated
April 30. 1997, and June 26. 1997 Subject: Reply to Notice of
Violation. The licensee's corrective actions included revision of
Specification 248-107 and review of other specifications to assure OC
inspection criteria required by applicable codes and standards
referenced in the UFSAR had been included in the specifications.
Specifications reviewed included the following: 248-117 - Installation
of Piping Systems: 048 012 - Installation of Electrical Cables: 006 001
- Design. Testing & Inspection of Concrete Mixes. Concrete Materials and
High-Strength Bolts: 005-005 - Design. Testing, & Inspection of Concrete
Mixes. Concrete Materials: 013 001 - Concrete Work: and 018-002.
Miscellaneous Steel. Additional corrective actions included inspection
of a sample of safety related high strength bolts installed using
Specification 248-107. The inspectors reviewed the results of the
structural steel inspectior.s which were documented in ESR 97-00085.
-
hiscellaneous Structural Steel Connection Inspections. The licensee
also revised procedure MMP-013. to incorporate the specification 248-107
changes and trained OC. engineering and planning personnel on the
changes to specification 248-107 which now require additional QC
inspections. The inspectors reviewed records which documented
inspections performed for selected USl A-46 modifications completcd on
Unit 1 during the Fall.1997 refueling outage and verified the
structural steel inspections were completed in accordance with the
revised procedures.
4
Ee.5 (Closed) Violation item 50-325(324)/97 08-07. Failure to initiate
Condition Reports to [,0cument Nonconformina E0 Items
I
. The licensee
September 2. 199 reshonded
Subject: Reply to this violation
to Notice of Violation. in a letter The dated
,
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37
licensee's corrective actions included training of E0 personnel on
the corrective action program, a review of che E0 walkdown data
sheets to identify any potential nonconforming conditions which
had not been previously identified and dispositioned, and
organizational changes to improve management o"ersight in the E0
group. CR 97-02367 was initiated by the licensee on July 3. 1997
to document and disposition the two s)ecific examples of failure
to initiate CRs identified by NRC. Tie inspectors ceviewed the CR
closecut records (CR was closed on December 14, 1997) and the
licensee's corrective actions and verified that the actions were
completed in accordance with the licensee's violation response.
IV. Plant SuppEt
R1 Radiological Protection and Chemistry Controls
RI.1 Use of locks to Control Access
a. Insnection Stone (71750)
The inspector verified a selected sampling of doors required to be
locked, by plant TSs and procedures, fc r the purpose of radiation
protection,
b. Observations and Findinas
The inspector reviewed Environmental & Radiological Control 0E&RC-0040.
Control of Locked High Radiation and Very High Radiation Areas, to
determine the controls used to lock high radiation area doors and
barriers. The inspector located a sampling of the locked high radiation
area doors specified in OE&RC-0040 and tested them to ensure that they
were locked. The ins)ector found that all the locked high radiation
doors tested were locced,
c. Conclusions
The ins)ector determined that each of the locked high radiation area
dcors w11ch were checked were locked. The ins)ector concluded that the
licensee is satisfactorily controll1ng locked ligh radiation areas in
the plant.
R1.2 Radioactive Material Controls
a. insoection Scqoe (71750)
The inspector conducted a housekeeping tour of radioactive material
storage areas located in outside areas within the protected area,
b, Observations and Findinas
The inspector found several poor radiological work practices in the
radiological material (RAM) storage area located aojacent to the
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.
38
Radiological Maintenance Service Building in the northwest corner of the
p.*otected area. A bucket containing scaffolding brackets was half
filled with water and was labeled as radioactive material. The label
identified the brackets as contaminated. This practice had the
possibility of allowing the potentially contaminated water to cause a
spread of contamination in an RAM storage area. There was also
scaffolding identified as radioactive lying unprotected on a wooden
pa l l e'. .
The ~icensee conducted a walkdown of this area and the radiological
service building, and identified multiple conditions requiring action.
These items were identified in CR 97-04122. Nonconforming Material
Condition,
c. Conclusions
The inspector determined that several poor radiological work practices
existed in a radioactive material storage area.
S2 Status of Security Facilities and Equipment
c2.1 Plant Access Control and Physical Barriers
a. Inspection Scone (71750)
The inspector verified the status and condition of the protected area
fencing,
b. Qbser"ations Jnd Findinas
The inspector performed a walkdown of the protected area fence. The
fence was inspected for integrity such as corrosion on the posts, gaps
in the fence, and general adequacy. The inepector noted no
deficiencies,
c. Conclusions
The inspector found the status and condition of the protected area fence
to be satisfactory.
F1 Control of Fire Protection Activities
F1.1 Operability of Fire Protection Facilities and Eauioment
a. Ipsoection Scone (64704)
The inspector reviewed the operation's fire protection daily impairment
reports on the facility's fire protection systems and features, and
inspected these items to determine the performance trends and the
material conditions of this equipment.
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4
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39 ,
b. - Observations end Findinas !
A review of the Loss Prevention Unit daily Impairment Reports for
-December 8 - 11, 1997.- indicated that the following fire-protection
,
components or systems for safety related areas were out of service: ,
.
'
fire Protection-System ~ Number of Imoairments
Thermo-Lag Fire Barriers 2
Fire Doors 6
'
, Cable Coating- 1
- -
Fire: Detection System - 3 1
Fire Suppression System 3
The inspector noted that a number of- fire doors were out of service.
This high number was attributed to the current DG building fire door
corrective action (door replacement and repairs) that was in process for
discrepancies identified during a June 1997 licensee self assessment of
the fire protection program.- Appropriate compensatory measures had been
-
-1mplemented for the fire protection features which were out of service.
The impairment status report provided the licensee with a good means of
identifying out-of-service fire protection equipment and provided status
-
for compensatory measures that were implemented. The corrective
maintenance on degraded fire protection systems was accomplished in a
-
timely manner,
,
During the plant tours the inspector noted that the maintenance and
material condition of the fire protection equipment were satisfactory.
c. Conclusions
'
Correstive maintenance on degraded fire protection systems was
accomplished in a. timely manner.>The maintenance and material condition
,
of the fire protection equipment and features were satisfactory.
-
,
F2 Status of Fire Protection Facilities and Equipment
F2.1 E3ssive Fire Barriers
.
Fire barriers ~ include penetration seals. wraps, walls. structural member--
fire resistanticoatings.. doors, dampers. etc. Fire barriers are used to
-prevent the spread of fire and to protect redundant safe shutdown
equipment. Laboratory testing of fire barrier materials is done only on
a-limited range of test assemblies. In-)lant-installations can vary
-,
from the tested configurations. -Under tie provisions of Generic Letter
(GL) 86-10. Implementation of Fire Protection Requirements, licensees
are permitted to develop engineering evaluations justifying such
deviations.
w -, ,, - . . . -. - - . - - ,. . -
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'
40
2.2 Silicone foam Penetration Seals
a. Inspection Stone (64704)
The inspector reviewed the fire barrier ,,ilicone foam penetration seal
design end testing. The inspector compared as-built fire barrier
silicone foam penetratioh seals to fire endurance test configurations to
verify that the as-built penetration seals reviewed were qualified by
appropriate fire endurance tests, representative of, and bounded by, the
design and construction of the fire endurance test specimens. During
plant walkdowns the inspector observed the installation configurations
of selected fire barrier silicone foam 3enetration seals to unfirm that
the licensee had established an accepta)le design basis for those fire
barriers used to separate safe shutdown functions.
b. Observations and Findinas
The inspector reviewed the fire barrier seal design and testing for six
- of ten fire barrier silicone foam seal penetrations, Additional reviews
I are documented in NRC 1Rs 50-325(324)/92-31, 93 08. and 93-38.
The inspector reviewed Brunswick Specification No. 118 003, Revision 7.
Selection and Installation of Fire Barrier Penetration Seals: Corrective
Maintenance Procedure OCMP-010, Revision 2, Installation of Fire
Barrier, Pressure Boundary Penetration and Water / Moisture Seals: Fire
Protection Procedure FFP-015. Revision 23, Fire Barrier Penetration Seal
Work Control: Periodic Test OPT-34.6.7.12. Revision 3. Fire Barrier
Penetration Seals: and the Fire Hazards Analysis (FHA) for the location
and description of fire areas: and assessed the licensee's supporting
technical justification and any available engineering evaluations for
the sampled silicone foam type oenetration seals,
The inspector's review focused on verifying that the following design
and installation paramaters for the as-built configurations were
adequately bounded and justified by the licensee's engineering
evaluations:
. penetration opening sizes
e thermal mass of penetrating items
e clearances of penetrating items
e unexposed surface temperatures
The insoector found that penetration seal field verification
documentation was maintained by the licensee. However, the seal
installers * qualification and training records were not readily
available for review. Although the installation and repair procedures
for penetration seals provided adequate guidance to ensure materials
were installed per design requirements, the inspector could not verify
that the established surveillance recuirements included vendor
recommendations for inspection and icentification of silicone foam seal
aging and shrinkage.
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'
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41
The licensee was unable to locate the penetration seal testing
documentation and the vtador data for the tested prototype
configurations or GL 8610 engineering evaluation documentation that
evahated the adequacy of the deviations from a tested fire barrier
contiguration. This does not satisfy the guidar.ce of GL 8610. The
i
licensee stated that industry documentation is available to support
silicone foam penetration seal installations at Brunswick but the
.tiformation was maintained at other Carolina Power and Light (CP&L)
sites.
The penetration seal testing documentation, vendor data and inspection
criteria, installer qualification and training records, and evaluations
of deviations from tested fire barrier configurations will be reviewed
during a subsequent NRC inspection. This is identified as IFl 50 325
(324)/97-13 04. Review of Licensee Records and Engineering Evaluations
to Establish the Fire Resistant Capabilities of Fire Rated Silicone foam
Penetration Seals,
c. Conclusions
The inspector concluded that silicone foam penetration seal field
verification documentation was maintained by the licensee. The
installation and repair procedures for penetration seals provided
adequate guidance to ensure that materials were installed per design
requirements. However, the designs were not supported by seal testing
documentation, vendor data and inspection criteria, installer
qualification and training records, and engineering evaluations that
satisfy the guidance of GL 8610 for deviations from the fire barrier
configuration qualified by tests.
F2.3 Fire Doors
a. Insnection Scone (64704)
The inspector reviewed UFSAR Section 9.5.1.4.3.4.b. Fire Doors, and
performed plant walkdowns to verify that the UFSAR wording was
consistent with the observed plant installation configurations for
selected fire doors installed in fire barriers used to separate safe
shutdown functions.
b. Observations and Findinas
The UFSAR St.ction 9.5.1.4.3.4.b. Fire Doors, states that doors and
frames are either listed by a national testing laboratory or are
constructed similar to listed doors and frames. All doors and frames
have been evaluated to assure satisfactory ratings. Results are
documented in the FHA. During the review of the FHA the inspector
identified that, while evaluations of fire doors and frames existed. the
-licensee failed to document their results in the FHA. which is section
9.5.1.5 of the UFSAR.
1
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_ _ _ _ _ _ - _
42
After discussions with the licensee. CR 97-04103 was issued to track the
l failure to provide the results of fire door evaluations in the FHA.
This UFSAR discrepancy was identified by the inspector and is discussed
in Section F2.4.
A review of the surveillance ins)ection and testing procedures for fire
doors was performed to confirm tlat the licensee specified fire door
clearance acce)tance criteria was in accordance with the guidance of
National Fire )rotection Association (NFPA) 80. Standard for Fire Doors
and Fire Windows. On December 10. 1997. the inspector observed ongoing
door replacement and repair activities for fire doors in the DG
building. No discrepancies were identified,
c. Conclusions
I
The inspector concluded that fire door surveillance prc:edures and
acceptance criteria for verification o' fire daor clearances were in
accordance with NFPA quidance. Howevr a UFSAR discrepancy associated
documentation of fire door and frame eu.uations was identified.
F2.4 UFSAR Review
A recent discovery of a licensee o)erating the facility in a manner
contrary to the UFSAR description lighlighted 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 the observed plant
practices, procedures, and/or parameters.
The inspector reviewed UFSAR Section 9.5.1.4.3.4.b, Fire Doors, as part
of the fire protection program review activiti u , An inconsistency was
noted in that the licensee failed to document the results of evaluations
of fire doors and frames in the FHA which is section 9.5.1.5 of the
UFSAR. This issue is discussed in Section F2.3. This item will be
identified as part of URI 50-325(324)/97-13-05. UFSAR Discrepancy Fire
Doors.
F3 Fire Protection Procedures and Documentation
F3.1 Fire Protection Procedures
a. Insoection Scone (64704)
The inspector evaluated the adequacy and implementation of the
licensee s Eire Protection Program described in the UFSAR and in Plant
Operating Manual Fire Protection Procedure OPLP 01. Revision 6. Fire
Protection Program Document. In addition a comparison was made of the
program to selected NRC Safety Evaluation Reports which ap3 roved the
station fire protection program. The inspector reviewed t7e following
>
procedures for compliance with the NRC requirements and guidelines:
.
. . . . - .. ~ .
..
.
.
. _ . . . . .
.
.
.
.
.
43
-
OPLP-01. Revision 6. Fire Protection Program Document
-
0FLP-01.1. Revision 12. Fire Protection Commitment Document
-
OPLP-01.2 Revision 10. Fire Protection System Operability.
Action, and Surveillance Requirements
-
FPP 005. Revision 15. Fire Watch Program
-
FPP-008. Revision 24. Fire Protection Weekly inspection
-
FPP 013. Revision 25. Transient Fire Load Evaluation
-
FPP 014. Revision 17. Control of Combustible. Transient Fire loads
!
and Ignition Sources
Plant tours were also performed to assess procedure complianc.e.
b. Obji.ervations and Findinas
The listed procedures were issued to implement the facility's fire
protection program. These procedures contained requirements for program
administration, controls over combust 1 oles arid ignition sources, fire
watch duties and training, and operability requirements for fire
i
protection systems and features. The 3rocedures were well written and
met the licensee's commitments to the 1RC.
General plant walkdown inspections were perfoimed by the inspector to
verify: acceptable housekeeping; compliance with the ]lant's fire
prevention procedures such as control of transient com)ustibles:
operability of the fire detection and suppression systems: emergency '
lighting: and installation and operability of fire barriers, fire stop
and penetration seals (fire doors, dampers, electrical penetration
seals, etc.),
c. Conclusions
General housekeeping was satisfactory. Fire retardant plastic sheeting
and film materials were being used. Lubricants and oils were properly
stored in approved safety containers. Controls for combustible gas bulk
storage and cutting and welding operations were being enforced.
Controls were being properly maintained for limiting t' alsient
combustibles in designated separation zones and oth' restricted plant
. areas.
F5 Fire Protection Staff Training and Qualification
F5.1 EireBrioade
a. Insoection Stone (64704)
,
,-- _____.
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m
. 1
1
44
The inspector reviewed the fire brigade organization and training
program for compliance with the NRC guidelines and program requirements.
'
b. Observations and Findinos
'
The organization and training requirements for the plant fire brigade
were established by Fire Protection Procedure 0FPP-051. Loss Prevention
Emergency Response 0ualification/ Training and Drill Program. The fire
brigade for each of five shifts was composed of an operations support
fire protection technician shift incident commander (fire brigade
leader) and at least four additional brigade members consisting of
Auxiliary Operators. Chemistry Technicians and Maintenance personnel.
Each operations shift also had a Senior Reactor Operator / Reactor :
Operator Fire Brigade Advisor assigned to respond tr ' ires with the fire
brigade.
As of the date of this inspection, there were a total of 48 fire brigade
members 26 from operations and 22 from E&RC and Maintenance on the
pic t fire brigade. The inspector verified that sufficient shift
personal were available to staff each shift's fire brigade with at
least five qualified fire brigade members.
A review of the training records for the fire brigade members indicated
that the training, drill, respiratory and physical examination
requirements for each active member were up to date and met the
established site training requirements.
Fire Briaade Ecuioment:
The fire brigade turnout gear and a fire response vehicle and trailer
with fire brigade equi) ment was stored in the Operations / Fire Protection
equipment building. T1e_ inspector's inventory of the fire brigade
equipment indicated that a sufficient number of turnout gear, consisting
of coats, pants, boots, helmets, etc. , was provided to equip the fire
brigade members expected to respond in the event of a fire or other
emergency. The fire brigade turnout i., ear and fire fighting equipment
were being properly maintained,
c. Conclusions
The fire brigade organization and qualification training met the
-requirements of the site procedu.m . Fire brigade turnout gear and fire
fighting eouipment were being properly maintained.
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45 j
F6 Fire Protection Organization and Administration
F6.1 Fire Protection Mananement and OraanizatioD
a. Inspection Scope (64704),
The licensee's management and administration of the facility's fire
protection program were reviewed for compliance with the commitments to
the NRC and to current NRC guidelines.
b. Observations and Findinos
During this report period the licensee reassigned the responsibility ior
the administration and implementation of the fire protection program
from the previous Loss Prevention Unit (LPU) to the Operations Shift and
Support organizations. The LPU organization was dissolved.
The designated onsite manager responsible for the administration and
implementation of the fire protection program was the Operations
Manager, This responsibility had been delegated to the Operations
Support Superintendent. The Operations Support Superintendent was
responsible for the station fire protection program, fire protection
surveillance testing of fire protection systems and equipment, and
ensuring that the aopropriate fire prevention procedures and fire
b:'igade programs were implemented. A Fire Protection Program
C0ordinator reported to the Operations Support Superintendent.
Maintenarice of the 31 ant fire protection equipment was performed by the
Maintenance Unit. cire protection related training was planned and
conducted by the Brunswick Training Se: tion. Coordination of the
station's fire protection program commitments and engineering functions
was provided by a fire protection system engineer in the Brunswick
Engineering Support Section,
c. Conclusions
The coordination and oversight of the facility's fire protection program
had been reassigned from the previous LPU organization to Shift
Operations. The new organizational structure met NRC guidelines and the
licensee's fire protection program requirements.
F7 Quality Assurance in Fire Protection Activities
F7.1 Fire Protection Audits
a. Insoection Scope (64704)
The following audit report and the plant response to the issues were
reviewed:
- Nuclear Assessment Section (NAS) Report B-FP-97-01. Brunswick Fire
Protection and Loss Prevention Unit Assessment, dated
August 1. 1997.
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b. Observations and Findinas
The licensee's Nuclear Assessment Section performed an assessment of the
fire protection program and LPU on June 16-27. 1997. The report for
this assessment was Re) ort No. B FP-97 01. The assessment team
determined that the LPJ fire prevention and fire response activities
were adequate; however, its implementation of the fire protection
)rogram was ineffective based on a number of program elements found to
)e below acceptable standards. Findings from these assessments were
categorized as strengths, issues, or weaknesses. The assessment report
identified six program element issues and one weakness.
The inspector reviewed the final audit report, the licensee's response
to the identified issues. the planned corrective actions, and the NAS
evaluation of the response adequacy.
This NAS assessment of the facility's fire protection program was
comprehensive and effective in identifying fire protection program
performance deficiencies to management. The audit team identified
deficiencies in LPU'c management oversight of fire protection
procedures, training, problem identification, procedure performance
standards, corrective actions, and personriel safety. Corrective actions
in response to the identified issues were substantial and included a
fire protection reorganization to integrate the former LPU organization
into the shift Operations and Operations Sup) ort organizations under
direct management of the Operations Support Manager,
c. Conclusions
The 1997 Nuclear Assessment Section assessment of tite facility's fire
protection program was comprehensive and was effective in identifying
fire protection program performance deficiencies to management. Planned
corrective actions in response to the audit issues were substantial and
included a fire protection reorganization.
V. Manaaetment Meetinas
XI Exit Meeting Summary
The inspector presented the inspection results to members of licensee
management at tN conclusion of the ins)ection on January 8,1998. Post
inspection briefings were conducted on )ecember 12, 1997. The licensee
acknowledged the findings presented. The licensee stated that they had
not determined if clearance records were required QA records.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
A. Brittain. Manager Security
M. Christinziano, Manager Environmental and Radit lon Control
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W. Dorman. Supervisor Licensing and Regulatory Programs
N. Gannon. Manager Maintenance
J. Gawron. Manager Nuclear Assessment Section
S. Hinnant. Vice President. Brunswick Steam Electric Plant
K. Jury. Manager Regulatory Affairs
R. Krich, Chief Engineer. Nuclear Engineering Department
B. Lindgren. Manager Site Su) port Services
J. Lyash. Manager Brunswick Engineering Support Section
R. Mullis. Manager Operations
Other licensee employees or contractors included office, operation,
maintenance. chemistry, radiation, and corporate personnel.
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INSPECTION PROCEDURES USED
IP 37550: Engineering
IP 37551: Onsite Eng11eering
IP 61726. Surveillance Observations
IP 62700: Maintenance Program implementation
IP 62707: Maintenance Observations
IP 64704: Fire Protection
IP 71707: Plant 0)erations
IP 71714: Freeze )rotection
IP 71750: Plant Support Activities
IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power
Reactor Facilities
IP 92901: Followup - Operations
IP 92902: Followup - Maintenance
IP 92903: Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-325(324)/97-13-01 VIO Failure to Retain TS Required QA Record (Section
07.2)
50 325(324)/97 13-02 VIO Inadequate Procedure for the Conduct of E0
Preventive Maintenance (Section M1.3, El.4.b.6)
50 325/97-13-03 VIO Failure to Note Abnormal TS Surveillance Values
(Section M3.1)
50 325(324)/97-13-04 IFl Review of Licensee Records and Engineering
Evaluations to Establish the Fire Resistant
Capabilities of Fire Rated Silicone foam
Penetration Seals (Section F2.2)
50-325(324)/97-13-05 URI UFSAR Discrepancy Fire Doors (Section F2.4)
50 325(324)/97-13 06 IFl Revisions to Procedure EGR-NGGC-0153 (Section
El.1)
50-325(324)/97-13-07 IFl Review Technical Evaluation of Terminal Block
Current Leakayc and the Effect on EQ Equipment.
(Section El.4.b.4)
Closed
50-325/96-15-01 URI Vessel Disassembly Without Secondary Containment
(Section 08.1)
50-325(324)/97 02-01 V10 Locked Valve Out of Position (Section 08.2)
50-325/97 12 03 URI Recirculation Pump Run back (Section 08.3)
)
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50-325(324)97-12-04 URI Diesel Generator Low Voltage Auto Start Defeated
(Section 08.4)
50 325(324)/96-017-00 LER Invalid Loss of Coolant Accident (Section M8.1)
50_-325/97_009-00 LER Missed Increased Frequency inservice Testing
Requirement (Section M8.2)
50-325/97-001-00 LER Rod Block Monitor Surveillance inadequacy
(Section M8.3)
50-325(324)/95-022 00 LER High Pressure Coolant injection System Discharge
Flow Element Gasket Leak (Section M8.4)
'
50 325/97-05-02 IFl
Abnormal CS Sp)arger Break Detector Indication
(Section Md.5
50 325/97-05-03 VIO Inadequate CS Surveillance Procedure (Section
M8.5)
50 325/97-02 LER Core Spray Header Differential Pressure
Instrumentation Inoperable (Section M8.5)
50-325(324)/97-02-04 VIO Failure to implement Requirements of the
Maintenance Rule (Section M8.6)
50-325(324)/97-08-04 URI Control of EDBS Information (Section E8.1)
50-325(324)/97-04 LER Spent Fuel Shipping Cask Handling Activities
(Section E8.2)
50-325(324)/96-14-05 IFI Effect of EQ Accuracy on Instrument Setpoint
Calculations (Section E8.3)
50-325(324)/97-02-08 VIO Failure to Implement an Inspection Program for
Safety-Related Miscellaneous Structural Steel
(Section E8.4)
50-325(324)/97-08 07 VIO Failure to Initiate Condition Reports to
Document Nonconforming EQ ltems (Section E8.5)
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