IR 05000324/1998003
| ML20216H332 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/13/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20216H295 | List: |
| References | |
| 50-324-98-03, 50-324-98-3, 50-325-98-03, 50-325-98-3, GL-89-10, NUDOCS 9804210113 | |
| Download: ML20216H332 (40) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-325. 50-324 License Nos:
50-325/98-03. 50-324/98-03 Licensee:
Carolina Power & Light (CP&L)
Facility:
Brunswick Steam Electric Plant. Units 1 & 2 Location:
8470 River Road SE Southport. NC 28461 Dates:
February 1 - March 14, 1998 Inspectors:
C. Patterson. Senior Resident Inspector E. Brown. Resident Inspector
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E. Guthrie. Inspector in Training
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J. Coley. Reactor Inspector (Sections 05.1. M1.1)
P. Kellogg Reactor Inspector (Sections E1.1. E8.4)
D. Jones. Reactor Inspector (Sections R1.1 - R1.3.
R8.1)
M. Holbrook, Consultant. INEEL Accompanying Personnel: T. Scarborough, Nuclear Reactor Regulation Approved by:
M. Shymlock Chief. Projects Branch 4 Division of Reactor Projects
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Enclosure
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9804210113 980413
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PDR ADOCK 05000324 O
PDR l
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EXECUTIVE SUMMARY Brunswick Steam Electric Plant. Units 1 & 2 NRC Inspection Report 50-325/98-03, 50-324/98-03 This integrated inspection included aspects of licensee operations, maintenance, engineering. and plant support.
The report covers a 6-week period of resident inspection; in addition. it includes the results of a maintenance, motor-operated valve. and health physics inspection by regional inspectors.
Operations The inspector concluded that the High Pressure Coolant Injection (HPCI)
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system was in standby and that the material condition of the system was satisfactory.
However, the inspector identified a weakness in the licensee configuration control program for not identifying a valve i
installed on the HPCI turbine stop valve operating cylinder (Section 01.1).
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The inspector concluded that current procedures do not control the use I
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of the Emergency Rod In Notch Override switch.
The licensee was responsive to this issue (Section 01.2).
j The inspector determined that personnel clearance tags were used in an e
application permitted by plant procedures (Section 02.1).
The inspector concluded that the Standby Liquid Control safety related
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tagout was implemented satisfactorily. (Section 02.2).
The inspector concluded that the diesel generator (DG) electrical
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lineups and plant drawings were consistent with the DG Motor Control Center (MCC) 120 volt AC (VAC) distribution panels. Four breakers located on the DG MCC 120 VAC distribution panels were incorrectly labeled as compartment heaters instead of spares (Section 02.3).
An exercise scenario, and a subsequent critique and training. on the
plant simulator which required an operations crew to demonstrate their ability to respond to plant transients and don self-contained breathing apparatus while the control room emergency ventilation system was undergoing modifications, was performed in an effective and thorough manner (Section 05.1).
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i Maintenance Maintenance activities observed were conducted in an effective and
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thorough manner by knowledgeable and skilled technicians (Section M1.1).
The surveillance and work activities observed by the insoector were
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performed satisfactorily, Procedures. and drawings used lad been verified as current. Satisfactory supervisory oversight was present (Sections M1.2, M1.3, M1.4).
The inspector concluded that the programs governing control of component
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. leakage and the maintenance backlog were ineffective in the timely-resolution of a degrading secondary containment penetration link seal.
The failure to promptly correct this situation has resulted in a spread of contamination into clean areas, and the establishment of long-term contaminated areas (Section M1.5).
The inspector concluded that the " red-line~ changes made to vendor
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procedures were. consistent with the requirements of licensee procedures and Technical Spec 1'ications (TSs). The inspector identified that improvements were needed in monitoring training certification of (
contractor qualification (Section M71).
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Enaineerina The inspectors reviewed the implementation of the licensee's generic
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L letter 89-10 Motor-Operated Valve program.
Based on completed and planned licensee actions. the inspectors concluded that the licensee had made adequate progress on the implementation of.the program and had met
.the intent of GL 89-10 (Section E1.1).
The inspector concluded that operator training for the oncoming crew on-
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the compensatory measures during upgrades of the Control Building l
Ventilation System was not timely (Section E2.1).
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.The inspector concluded that the licensee had performed an adequate load e
l analysis for the radwaste shipping containers and concluded that the storage of these containers was acceptable (Section E2.2).
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A non-cited violation was identified for failing to perform a 10 CFR 50.59 review when downgrading the quality classification of the Control Building air-conditioning units (Section EB.3).
Plant Sucoort The licensee's radiation control practices for Radiation Control Area p'
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ingress and egress control, personnel dosimetry, radiological surveys
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and postings, and container labeling were consistent with the licensee's Radiation Control and~ Protection Manual-and relevant sections of 10 CFR 20 (Section R1.1).
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l The licensee was generally effective at identifying and correcting l
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radiological control related deficiencies in accordance with the
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corrective action management program (Section R1.2).
The licensee effectively implemented training and qualification for
Radiation Protection and Chemistry Technicians in accordance with the training program procedural requirements (Section R1.3).
l The inspector concluded that radiological work activities in a
contaminated and high radiation area were properly controlled (Section R3.1)
The inspector concluded that security provided proper access control
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into the protected area (Section S2.1)
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l Reoort Details Summary of Plant Status Unit 1 operated continuously during this report period. At the end of this report period the unit had been on-line continuously for 121 days.
Unit 2 operated continuously during this report period. At the end of the report period the unit had been on-line continuously for 146 days.
Due to concerns about the control room dose, the licensee imposed an administrative limit on Iodine until a Technical Specification (TS)
amendment submitted was approved.
The licensee made a procedure change to Administrative Procedure 0Al-81. Water Chemistry Guidelines, setting the limit at 0.1 microcurie per gram dose equivalent Iodine 131 compared to the TS value of 0.2 microcurie per gram.
Also, the licensee 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.
I Due to a reconstitution of the Environmental Qualification (E0) program l
and items identified, there are 10 of 26 Justification for Continued 0)eration (JCO) that remain open for both units.
The following provides t1e status of the EO JCOs and assc iated Engineering Service Requests (ESRs):
Closed 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). EQ Type JC0 for E0 Fuses Without a Qualification Data Package (0DP)
4)
ESR 97-00289, Post Accident Sampling System (PASS) Valve Limit Switch Panel Wiring 5)
ESR 97-00238 JC0 for Standby Gas Treatment Motor Operated Valve (MOV) Position Indicator Rhecstat 6)
ESR 97-00534 GE EB-5 Type Terminal Strips l
7)
ESR 97-00513. In-Board Drywell Electrical Penetrations 8)
ESR 97-00535. Target Rock Solenoids Terminal Block Spray 9)
ESR 97-00449. Degraded Junction Boxes 10)
ESR 97-00250, Conduit Union in E0 Boundary 11)
ESR 96-00425. Evaluation of E0 sealants 12)
ESR 97-00523, High Pressure Coolant Injection (HPCI) Auxiliary Oil l
Pump Motor Unit 1
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13)
ESR 97-00446. GE Radiation Detectors 14)
ESR 97-00229. JC0 for GE Condition Report (CR) 151 B Terminal Blocks 16)
ESR 96-00503. Associated Circuit E0 l
Doen 17)
ESR 97-00330 (old ESR 96-00501). Motor Control Center (MCC) E0 was closed by the licensee, but was reopened - closure date to be determined (TBD)
18)
ESR 97-00529. Failure of Unit 1 Drywell Motor closure date TBD 19)
ESR 96-00627. ODP for Marathon 300 Terminal Blocks the closure date is TBD 20)
ESR 97-00256. Main Steam Isolation Valve (MSIV) Hiller Actuator JCO. was scheduled for completion September 2. 1997. but closure date is now TBD 21)
ESR 97-00343. Qualification of Kulka Model 600 Terminal Blocks was scheduled for completion September 1. 1997. but closure date is now TBD 22)
ESR 97-00435. MCC Fittings, closure date TBD 23)
ESR 97-00602. Solenoid Valve Field Wiring. closure date TBD 24)
ESR 97-00710 Lubricants for Joy / Reliance Fan Motors - closure date TBD 25)
ESR 98-00093. Weed Resistance Temperature Detector Electric Termination Sealing - closure date TBD 26)
ESP,98-00091. E0 of Transformer Indication Lights - closure date TBD In summary. Unit 1 and 2 operated continuously. However, there were 10 outstanding JCOs in the E0 area for both units.
I. Operations
Conduct of Operations 01.1 Hiah Pressure Coolant Iniection (HPCI) System walkdown
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a.
Insoection Scone (71707)
The inspector conducted a walkdown of the Unit 1 HPCI system on
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February 25. 1998. using Inspection Procedure 71707: the Unit 1 Operating Procedure 10P-19. High Pressure Coolant Injection System Operating Procedure, electrical lineup and valve lineup.
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Observations and Findinas The inspector verified that the HPCI system was in standby.
The inspector found no electrical breakers or switches out of position in the Control Room. Electronic Equipment Room. Cable S) reading Room. or
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the Reactor Building.
The inspector found none of t7e accessible valves i
that were observed in the system to be out of position. The inspector verified the valve lineup in the main control room was according to procedure.
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The inspector observed the material condition of the HPCI system to be
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l satisfactory. The inspector noted, however that three valves on the turbine were missing name tags.
This condition was discussed with the
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system engineer. The system engineer found that two of the valves had tags attached but were lodged under the insulation which was covering the piaing.
The valves were 1-E41-V80 and 1-E41-V81. The third valve was a aleeder valve located on the operating cylinder of the turbine trip throttle valve 1-E41-V8. The ins)ector asked the system engineer why this valve was not identified on t1e HPCI system valve lineup.
The engineer informed the inspector, subsequent to a review of the drawings for the turbine stop valve, drawing (DWG) 1-FP-9527-5337, that the bleeder valve was not identified on the drawing and therefore should not be installed.
This drawing for the turbine stop valve indicated that the port should have a plug installed..The engineer noted that the Unit 2 HPCI system did not have this valve installed. Additionally, the licensee identified. after they conducted walkdowns of the HPCI and Reactor Core Isolation Cooling (RCIC) systems. that a drain valve was found not tagged on the Unit 1 RCIC system and two vahes on each units HPCI barometric condenser sight glass gage isolation valves were not I
tagged.
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The licensee initiated CR 98-00506. HPCI V8 Oil Vent Valve, which addressed the configuration control problem with having the unidentified valve installed.
The licensee initiated a Deficier.cy Log Entry (DLE)
98D00800 to have the valve removed.
The licensee did not have an operability concern, since the valve was not leaking and had been installed during multiple surveillance tests without problems.
The licensee did not know why the valve was installed in the system.
Minor oil leakage was noted from connections in the turbine oil system, but it did not affect system operation and oil catch measures were present to contain the leakage.
Not identifying the type valve installed is identified as a weakness in the licensee configuration control program, c.
Conclusions The inspector concluded that the HPCI system was in standby and that the material condition of the system was satisfactory.
However, a weakness was identified in the licensee configuration control program for not identifying the valve installed on the HPCI turbine stop valve operating cylinder, 01.2 Use of the Emeraency Rod in Notch Override Switch
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a.
Insoection Scope (71707)
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The inspector reviewed the licensee's use of the Emergency Rod In Notch Override switch.
This review was prompted by a control rod mispositioning event, which occurred at another Boiling Water Reactor (BWR) facility, while using the same type of switch.
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b.- Observations and Findinas On February 26. 1998 the inspector reviewed the licensees use of the Emergency Rod In Notch Override switch. The review concentrated on the use of this switch to continuously withdraw control rods.
The inspector reviewed the Reactor Manual Control System (RMCS) Operating Procedure.
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-10P-07, section 5.1 Continuous Control Rod Withdrawal. This procedure described three initial conditions required to withdraw control rods continuously. The operating procedure allowed the continuous withdrawal of a rod from any starting position to any intermediate position.
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precautions or limitations were written-in the procedure describing L
concerns for mispositioning a rod due to overshooting the required position.
Use of the Rod Movement Switch alone in the " Rod In Notch" or
" Rod Out Notch" position allows for movement of the control rod one
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l notch at a time. The inspector noted that by using the Rod Movement l
Switch, when the control rod was near a required position, rod travel would be limited to one notch movements, thus minimizing the possibility of overshooting a specified rod position.
l The inspector discussed the use of the override switch with a Reactor-Operator (RO). The R0 described current problems with the Unit 1 RMCS and 3ast problems with Unit 2 RMCS where using the Rod Movement Switch in t1e out direction would sometimes move the control rod two notches
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This condition had been captured as a deficiency in valve timing on the Hydraulic Control Units needing repair during the~
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next refueling outage for Unit 1.
The o)erators compensated for this condition by maintaining their hand on t1e Emergency Rod In Notch Override switch during all rod movements to anticipate this problem and position the override switch to " Rod In Notch" to insert the rod to the
required position before it is out of Josition due to a double notch problem..The control rods that have t11s condition are caution tagged to alert the operator. The operator described the practice of keeping their hand on the Emergency Rod In Notch Override switch, for continuous rod movements from any starting position to any intermediate position, as normal practice.
The inspector reviewed Periodic Test (PT) procedure OPT-14.1 Control Rod Operability Check and noted that this procedure allowed the operator to immediately reposition the control rod to the required position in the event that a control rod moves two notches when only one notch movement was expected.
Review of Abnormal Operating Procedure (A0P)
-0A0P-2.0. Control Rod Malfunction /Misposition, immediate actions stated that in the event a control rod was found out of position any further power changes were to be stopped and the Nuclear Engineer contacted for
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further control rod movement instructions.
Since the licensees control rod operating. procedures do not restrict the use of the Emergency Rod In Notch Override-switch for rod movements the operator could potentially use this switch during any control rod withdrawal.
If the override switch was used during the OPT-14.1 rod operability check and the control rod was inadvertently withdrawn past (i.e. greater than 2 notches) the required position this condition would be considered a Control Rod Mispositioning event and thus the operator would not be l
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allowed to immediately restore the rod to the required position. This scenario is similar to a rod mispositioning event which occurred at another BWR facility during the time of this inspection period.
The inspector discussed the above observations with licensee management.
l They agreed that the possibility existed for a control rod mispositioning event, based on the controls that were currently in place.
The inspector was informed that the licensee was planning to develop a team that was gcing to assess reactivity management at Brunswick.
The team was going to be called the Reactivity Event Review Team (RERT).
Based on the inspector's discussion with Brunswick
management, the licensee was going to provide this concern to the RERT for evaluation of their procedures and controls regarding control rod positioning.
Management informed the inspector that they would discuss
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this concern with the Shift Supervisors to heighten the awareness of the
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crew to help prevent mispositioning errors until the appropriate actions are taken by management to address the concern.
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Conclusions I
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The inspector concluded that current procedures do not control the use of the Emergency Rod In Notch Override switch.
The licensee was responsive to this issue.
01.3 Soecial UFSAR Review A recent discover of a licensee operating the facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures, and/or parameters to the UFSAR descri)tions.
While performing the inspections discussed in this report. tie inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected.
The inspectors verified that the UFSAR wording was consistent with the observed plant practices, procedures, and/or parameters.
The inspector reviewed the UFSAR section 7.7.1.8.2.1.b.
Operation of the RMCS. This section described the use of the override switch to continuously withdraw rods. The inspector concluded that operation of the RMCS was consistent with plant practices and procedures.
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Operational Status of Facilities and Equipment 02.1 Personnel Clearance Taas a.
Inspection Scooe (71707)
The inspector reviewed a clearance tag associated with the Control Building Ventilation Upgrades.
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.b; Observations and' findinas-On February 20, 1998, the ins)ector conducted a routine tour of the Control Building Ventilation Room.- The. inspector observed two clearance tags on the floor.
No one was seen in the area. The tags were handwritten identification label and were identified as 98-DTS-0001.
Valves SV-1 and LV-1 were indicated as closed with a person's name and phone extension number.
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The inspector went to the Operations Work Control Center (WCC) to review
.the clearance tag'index.
No clearance record could be found with this type number. The WCC contacted the' individual indicated on the clearance tags and were noted to be Personnel Clearance tags. This type l
of clearance was permitted-by procedure OPS-NGGC-1301. Equi) ment
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. Clearance.
However, this type of clearance was to be used )y licensee
personnel company-wide and had very restrictive conditions that
. permitted its use.
No equipment under a limiting condition for
operati.on (!-CO) could have a personnel clearance tags.
The~ inspector
questioned the application of these tags. The licensee stated that the tags were used on a temporary condensing unit after it had been charged
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with. refrigerant. _The unit was then lifted onto the Control Building roof. The licensee further stated that the tags were removed from the
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L unit but fell out of a workers pocket.
The-inspector reviewed the equiament clearan'ce 3rocedure and the usage
.of the tags was permitted for t1e application tlat the licensee discussed.
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Conclusions The inspector determined that personnel clearance tags were used in an application permitted by plant procedures.
02.2 Standby Liauid Control (SLC) System Taaout Review
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a.
Insoection Scooe (71707)
The-inspector reviewed the SLC safety system tagouts for proper implementation,
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Observations and Findinas i-On February 26, 1998 the inspector reviewed the clearance CL 1-98-00252 used to perform calibration maintenance procedure Calibrate 1-C41-XY-M600B Scuib Valve Continuity Indicator 1-C41-M006B. The inspector verifiec that the clearance process was adhered to 3rior to the commencement of the calibration and subsequent to tie completion of the calibration. The clearance Japerwork was reviewed after the tags were hung. : The inspector noted t1at section 1.3, of the Operations Clearance Form Attachment B, was not filled out. This section s]ecified the Technical Specification (TS) reference that was used w1en preparing the
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clearance. The inspector asked the operators why this was not filled l
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out and was told that this is normally not filled out since the TS number was referenced later during the process.
Leaving Attachment B blank on the clearance form did not affect the outcome of this clearance.
The inspector verified that the correct LCO was entered to initiate the clearance.
The inspector verified that the clearance tags were hung on the correct equipment and that the components were positioned properly. The inspector verified that the system was restored properly when the clearance was removed and that the clearance form was filled out correctly.
The inspector observed the final clearance package as it would be filed as a Quality Assurance record and noted no discrepancies except that section 1.3 was not filled out specifying the TS reference as discussed before.
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Conclusions The inspector concluded that the SLC safety related tagout was implemented satisfactorily.
02.3 Diesel Generator (DG) Motor Control Center (MCC)-120 Volt Alternatina Current (AC) Distribution Panel Breaker Status a.
Insoection Scooe (71707)
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The inspector performed a tour of the DG building and systems per Inspection Procedure 71707. and reviewed breaker positions on several 120 VAC Motor Control Centers (MCC) and their labeling.
b.
Observations and Findinas On March 4,1998 the ins)ector was
)erforming a tour of the DG building.
The inspector reviewed tie circuit areaker positions on several 120 VAC distribution panels.
Several breaker positions did not agree with other similar panels.
The inspector was informed that the loads on these breakers were disconnected and should have been identified as spare breakers. A work order was initiated to have the breakers labeled as spares.
The plant electrical drawings indicated that the loads had been disconnected.
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Conclusions The inspector concluded that the diesel generator electrical lineups and plant drawings were consistent with the DG MCC 120 volt AC distribution panels. Four breakers located on the DG MCC 120 volt AC distribution
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panels were incorrectly labeled as compartment heaters instead of i
spare Operator Training and Qualification 05.1 Verification of Control Roos Ooerator Simulator Trainina Usina Self Contained Breathina Aooaratus (SCBAs)
a.
Insoection Scooe (92901)
In a letter dated November 6. 1997, as supplemented by letter dated January 28, 1998, the licensee submitted a recuest for license amendments applicable to Brunswick Units 1 anc 2.
Based upon use of certain compensatory actions to assure accomplishment of design basis functions. the proposed amendments revised, for a limited Jeriod of
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time, the design basis qualification for the Control Room Emergency Ventilation System (CREVS) and CREVS instrumentation. The proposed changes were requested to support modifications to upgrade the control room air conditioning (AC) equipment and some supporting components.
The AC system was required to prevent the failure of safety-related j
equipment and to ensure control room habitability following certain design basis events.
The planned modifications would affect the seismic integrity of the control room envelope and could not be completed within the allowed outage times specified in the TS.
Without approval of the proposed license amendments, im)lementation of the control room upgrades would require shutting down bot 1 Brunswick units. The amendment provides for the use of temporary AC equipment and ductwork barriers which do not fully meet the design basis for CREVS during certain external events (e.g.. earthquakes, tornadoes and hurricanes.
radiological sabotage and missile hazards).
Compensatory actions would be taken by the licensee to minimize the risk under the temporary amendment and to assure necessary functions could be accomplished.
On February 6. 1998. NRC issued Amendment No. 191 to Facility 0)erating License No. DPR-71 and Amendment No. 222 to Facility Operating icense No. DPR-62 for Brunswick Steam Electric Plant. Units 1 and 2.
respectively.
The amendments consisted of changes to the TS and associated Bases and were approved on a one-time-only basis to support modifications upgrading the CREVS during the period of February 6. to May 1. 1998.
One compensatory action committed to by the licensee prior to the commencement of the CREVS modifications, was that one operations crew would participate in an exercise scenario on the plant simulator which would require them to demonstrate their ability to respond to plant transients and don SCBAs while the control room emergency ventilation system was undergoing modifications.
Lessons learned from that exercise would be provided to each operating crew.
On February 12. 1998, the inspector observed a simulator scenario where an operations crew responded to plant transients and don SCBAs.
The simulator scenario was a large break of Main Steam Line ~0~ in the turbine building and a seismic event resulting in an automatic reactor scram and Group 1 isolation.
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b. Observations and Findinas During the simulator scenario the crew recognized and responded to the
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seismic event annunciator. Shutdown of the plant was directed when the l
seismic acceleration was determined to be greater than 0.089 During the crews preparations for the required plant shutdown, a large break on Main. Steam Line D" in the turbine building occurred.
This resulted in an automatic reactor scram and Group 1 isolation. As the crew was
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stabilizing th plant, a control room high radiation alarm was received.
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The crew donned the SCBAs within 10 rinutes utilizing the " buddy" system l
(two operators assist each other). While the crew continued their efforts to stabilize the plant, environmental and radiation control t-(E&RC) personnel were directed to sample the control room atmosphere for l'
indications of radioactivity. Upon notification that the control room
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environmental ventilation intake plenum blanks were intact and the control room airborne radioactivity levels were acceptable, the reactor L
pressure vessel (RPV) level and pressure were stabilized and the crew removed the SCBAs.
The inspectors observed a subsequent critique which was held with a l
large contingency of operation personnel to discuss " lessons learned" during the simulator scenario.
In addition to the " lesson learned" discussions, engineers provided training for the operators on how the modifications would be performed, the sequence to which certain equipment would be operational and how and when the LCO would be entered. This instruction was excellent and questions raised by the operators were dealt with in a thorough and effective manner.
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Conclusions The exercise scenario on the plant simulator which required an operations crew to demonstrate their ability to respond to plant
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transients and don self-contained breathing apparatuses while the control room emergency ventilation system was undergoing modifications was successfully performed. A subsequent critique and training given by engineering on the control room emergency ventilation modification was conducted in an effective and thorough manner.
Miscellaneous Operations Issues (92901)
L 08.1- (Closed) Violation VIO 50-325(324)/97-11-02: PNSC Ouorum Too Many Alternates This TS violation was that a Plant Nuclear Safety Committee (PNSC)
meeting was conducted with more than two alternates The licensee responded to the violation on November 16, 1997. The violation occurred due to including the Manager-Outage and Scheduling and the Manager-Training as members on the list comprising the PNSC These members were not from the functional areas specified in TS 6.5.3.3.
fhe licensee revised their list of required members. A review of past PNSC quorum requirements since February 8,1996, was conducted to determine if an adequate _ safety review was conducted. The inspector reviewed the
licensee's response to the violation and corrective action.
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violation is closed.
II. Maintenance M1 Conduct of Maintenance M1.1 Observation of On-line Maintenance Activities a.
Insoection Scooe (62700)
The inspector examined portions of the following on-line maintenance activities to verify that the maintenance activities were being conducted in a manner that resulted in reliable and safe operation of the plant.
PM ALWY 007. Unit 1. Electrical Protection Assembly (EPA) Breakers
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5 & 6 removed from service to perform Maintenance Surveillance
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Test 1-MST-RPS21SA I
PM AFUW 003. Unit 2. Replacement of inline filters for the 2-CAC-
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AT-4409-FLT-1A & 2A j
PM AFLM 001 and Trouble Ticket 97AIWW1. Unit 2. Performed VOTES
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test on 2-E41-F079-MO.
PM-AFLJ 001. Unit 2. Performed VOTES test on 2-E41-F042-M0 Per
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OSPP-M0009 b.
Observations and Findinas The above work was performed with the work package present and in active use.
Technicians were skillful, experienced and knowledgeable of their assigned tasks. The EPA breakers were successfully calibrated in accordance with 1MST-RPS21SA. The filters in panel 2-CAC-4409 were successfully re) laced and leak tested.
In addition, the portions of the motor operator DMs and Votes testing, observed by the inspector, were performed in a very effective manner by knowledgeable and highly trained technicians, c.
Conclusions Maintenance activities observed by the inspector were conducted in an j
effective and thorough manner by knowledgeable and skilled technicians.
M1.2 Remote Shutdown Panel (RSDP) Trio Channel Calibration
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a.
Insoection Stone (61726)
The inspector observed a portion of the Unit 1 Maintenance Surveillance Test 1MST-RCIC270. RCIC High Water Level RSDP Trip Channel Calibration.
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Observations and Findinas On February 18, 1998, the inspector observed part of the performance of 1MST-RCIC270.
This surveillance test verified the operability of reactor water level monitoring instruments located on the remote shutdown panel in accordance with TS 4.3.5.3.
The inspector noted that the procedure and associated plant drawings in use had been properly verified as correct. The inspector verified that the test equi) ment used was within the current calibration cycle.
The inspector o] served concurrent verification during component positioning.
The return to service of the level transmitter was considered a high risk evaluation due to the sharing of the transmitter variable leg with reactor scram and grou) 2. 6. and 8 primary containment isolation system components.
During t1is high risk evolution satisfactory supervision was evident.
The inspector verified that the surveillance was scheduled and performed at the frequency required by TSs.
Review of the unit logs. indicated that the correct LCO had been properly entered within the time allowed.
and that the surveillance was completed satisfactorily.
M1.3 Reclaucment of RHR Discharae Pressure Switch a.
Insoection Scone (62707)
The ins 3ector observed portions of the replacement of the ASCO Tri-point Unit 1 Residual Heat Removal pump discharge pressure switch.
This replacement was required for environmental qualification.
b.
Observations and Findinas On March 4, 1998, the inspector observed the removal of the 1-E11-N016A.
The inspector noted that the procedures and the associated plant drawings in use had been properly verified. The test equipment used was within the current calibration cyclo The inspector observed concurrent verification during the lifting of the pressure switch leads, valve positioning and pressure switch remova.
The inspector noted during the maintenance activity that the technicians generally used good radiation controls.
A bag was used to catch the contaminated water from the system and the technician working on the pressure switch was in protective gloves. The inspector questioned the lack of gloves by an additional technician aiding in the removal of the contaminated switch.
The technician immediately corrected the oversight.
M1.4 Maintenance Conclusions
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The surveillance and work activities observed by the inspector were performed satisfactorily.
Procedures, and drawings used had been verified as current.
Satisfactory supervisory oversight was present.
M1.5 North Residual Heat Removal (RHR) - 17 Elevation Cleanliness a.
Insoection Scone (62707)
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The inspector performed routine tours of the Reactor Building observing general plant cleanliness and equipment conditions.
b.
Observations and Findinas On February 5.1998, during a routine tour of the Unit 2 Reactor Building, the inspector discovered a large 3001 of water and a brown liquid in the north Residual Heat Removal (lHR) room on the -17 foot (ft.) elevation.
The leakage was located around the 2A RHR pump and a nearby instrument rack.
The leakage covered a considerable portion of the floor.
The inspector observed that the pool was enclosed within the existing contaminated area boundary.
This area had been roped off due to previous leaks which had existed for a long time.
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The inspector questioned the licensee concerning the pool of water on the -17 ft. elevation.
The licensee cleaned the water, the brown substance-. and installed a portable spill dike to prevent the spread of water outside of the contaminated area.
This issue was described in CR 98-297. Rattle Spaces Need Pumping.
The CR investigation determined that the leakage was from a degraded secondary containment penetration link seal. Additional conversations with the licensee determined that a similar condition had been identified on January 26. 1998, by the Unit 2 Health Physics technician in CR 98-161. N. RHR Trench.
This CR identified that the link seal leakage was " sending potentially contaminated materials into the clean area at Instrument Rack #18.
acrossthecleanareaandintoanothercontaminationarpaattheRHR pumps where contamination levels are 160.000 dpm/100 cm... This is an as low as reasonably achievable (ALARA) concern, safety slip hazard, housekeeping eyesore, and a source of radwaste".
The CR indicated that a work ticket had been generated in March 1995. but no action had been taken to fix the seal.
The licensee indicated that both CR 98-161 and CR 98-297 had been voided since a work ticket had already been written on the leaking link seal.
The inspector questioned voiding CR 98-161 due to not all the issues identified being addressed by the issuance of work request / job order (WR/J0) 95-ACMR1. The licensee reopened CR 98-161 to address the lack of timeliness in the scheduling of the link seal repair.
The inspector reviewed WR/JO 95-ACMR1.
The ins)ector noted that the work ticket was the sixth oldest unrestrained worc ticket for Unit 2 and was one of three tickets for leaking link seals on a. list of the top 20 oldest unrestrained work tickets. Through discussions with the licensee the inspector determined that the ticket had not been worked due to the high area dose rate in the area. The ins)ector questioned how the ticket would ever get worked, since not worcing the ticket resulted in continued seal leakage which increased the contaminated area where the seal was leaking. The licensee indicated that the )rograms governing plant leaks and maintenance backlog reduction were )eing reconstructed.
A new initiative was being developed to reduce the average age of the work ticket backlog In addition, the responsibility for plant leakage has recently been trcisferred from E&RC to Operation c.
Conclusions The inspector concluded that the programs governing control of component leakage and the maintenance backlog were ineffective in the timely resolution of a degrading secondary containment penetration link seal.
The failure to promptly correct this situation has resulted in a spread of contamination into clean areas, and the establishment of long-term contaminated areas.
M7 Quality Assurance In Maintenance Activities M7.1 Failed Fuel Insoections - Unit 1 a.
Insoection Scoce (62707. 37551)
As a result of issues at another nuclear facility concerning control of vendor activities, the ins)ector reviewed applicable procedures. ESRs.
and training records for t1e licensee's failed fuel inspection activities.
b.
Observations and Findinas On February 3.1998, the licensee informed the inspector that failed fuel inspection activities for two fuel assemblies had been postponed until certain administrative controls were put in place to assure the assumptions made in the UFSAR were not exceeded.
The inspector discovered that issues were raised over the acceptability of leaving a fuel assembly in the fuel preparation machine (FPM) overnight, the adequacy of criticality controls and the procedures controlling the inspection activities.
The licensee informed the inspector that safety reviews were completed verifying the seismic qualification of the FPMs and for those vendor procedures used to perform the failed fuel inspection.
The inspector reviewed the training records for the vendor personnel conducting the inspections. The inspector noted that the licensee in Engineering Procedure OENP-57. Fuel Assembly Inspection. Testing, and Repair. had verified the vendor personnel's qualification. The inspector noted that two individual's training had not been reevaluated after three years as required by American National Standard Institute /
American Society of Mechanical Engineers (ANSI /ASME) NOA-1-1979. Quality Assurance Program Requirements for Nuch ar Power Plants.
The inspector discussed this observation with the licensee and appropriate documentation was subsequently obtained from the vendor. The licensee documented this oversight in CR 98-289. GE Qualifications Expired.
The l
licensee indicated that changes wccid be made to OENP-57. to require l
verification that the vendor personnel's qualifications were current.
i
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_ - _ _ _ _ _ _
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
_--_ - _ ______-__-_--_________-____
L f
The: inspector reviewed the following licensee and vendor procedures:
Administrative Procedure 0AP-004. Temporary Changes to Procedures
l Nuclear Generation Group Standard Procedures. EGR-NGGC-003. Design
Review Requirements GE Test Plan & Procedures, Brunswick 1 Failed Fuel Examination.
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Post-Mid Cycle 11 Outage GE Nuclear Energy. 246-GP-01. Fuel Bundle Upper Tie Plate Removal /
Replacement and Individual Rod Handling The inspector noted that the two vendor procedures contained " red-line" or " pen and ink" changes.
The changes imposed requirements not previously contained in.the procedures. The licensee indicatod-that the changes were made to enhance the administrative controls to ensure the assumptions made in Chapter 15 of the UFSAR were met during vendor inspection activities.
The inspector questioned the validity of the vendor having permission to make intent temporary changes to vendor procedures used for safety-related activities. The licensee indicated that per OENP-57. the vendor could make changes to vendor procedures provided the change was more conservative and a licensee's owner review.
was conducted.
The-inspector reviewed the 10 CFR 50.59 screen. The 10 CFR 50.59 screen reviewed the technical validity of the added information. - Discussions with the licensee and the vendor revealed that a safety review was
performed on the information added to the procedure.
Additional licensee review determined that the information added would not have introduced a situation not previously considered.
c.
Conclusion The. inspector concluded that the " red-line" changes made to vendor procedures were consistent with the requirements of licensee procedures and TSs. The inspector identified that improvements were needed in
- monitoring training certification of contractor qualification.
M8 Hiscellaneous Maintenance Issues (92902)
M8.1 (Closed) Licensee Event Report LER 50-325/97-006-00: Engineered Safety Feature Actuation Due to Loss of Emergency (E) Bus E-2 This event was discussed in Inspection Report (IR) 50-325(324)/97-08.
No new? issues were revealed in the Licensee Event Report (LER). This LER is closed.
M8.2 (Closed) Violation VIO 50-325(324)/97-08-02: Failure to Verify / Check
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E-bus Relay Operability The inspector verified the corrective actions described in the-licensee's response letter, dated September 2. 1997, to be adequate and complete.
No similar problems were identified. This violation is close ;
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'15'
III. Enoineerina El Conduct of Engineering El.1 Generic Letter (GL) 89-10 Procram Imolementation
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a.
Insoection Scooe (Temocrary Instruction 2525/109)
GL 89-10. Safety-Related Motor-Operated Valve Testing and Surveillance.
requested the implementation of a program to ensure that safety-related
motor-operated valves (MOVs) were capable of performing their design-I basis functions, GL 89-10 and its seven supplements provided recommendations for this program which included establishment of MOV switch settings, design-basis testing to demonstrate MOV capability, and actions to ensure that MOV design-basis capabilities are maintained.
The NRC staff documented the completion status of Brunswick's MOV program in IR 50-325(324)/97-11 and concluded that the licensee had not satisfactorily implemented GL 89-10.
The licensee responded to the l
' issues identified in IR 50-325(324)/97-11 with planned corrective actions that were documented in the CP&L (licensee) letter BSEP-97-0408.
. dated October 2. 1997.
The main purposes of this inspection _were to review: 1) the licensee's response to IR 50-325(324)/97-11. and 2) the Brunswick MOV program implemented in response to GL 89-10.
During the course of the
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= inspection, numerous licensee documents and several MOV setup calculations were reviewed.
In addition. the inspectors conducted a detailed review of the following MOVs:
1/2821-F019-Main Steam Line. Drain Outboard Isolation Valves 1/2E11-F016A/B RHR Drywell Spray Outboard Isolation Valves 1/2E11-F028A/B RHR Suppression Pool Discharge Isolation Valves 1/2E41-F006 High Pressure Coolant Injection Valves 1/2G31-F004 RWCU Inlet Outboard Isolation Valves 1/2SW-V102 Conventional / Nuclear Header Cross-Tie Valves b.
Qbservations and Findinas The licensee's guidance for the Brunswick MOV program was documented in CP&L Procedure EGR-NGGC-0203. Motor-0perated Valve Performance Prediction, Actuator Settings and Diagnostic Test Data Reconciliation.
Revision 4. dated December 15. 1997. Brunswick-specific engineering documents contained the justifications for MOV program assumptions, including justifications for vaive factors, load sensitive behavior, and stem friction coefficient assumptions.
-MOV Ooerability Assessments L
'IR 50-325(324)/97-11 noted that the licensee had initiated an effort to completely revise Brunswick's MOV setu) calculations. This revision was
. performed to incorporate MOV setu) metlodology imarovements identified during an inspection at another C)&L facility (Ro)inson) and to account
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for more conservative actuator efficiency assumptions. 'The licensee completed these revisions prior to this inspection.
i-L The licensee's application of. the program-justified requirements and more conservative assumptions caused numerous MOVs to have current thrust settings outside the revised setup windows.
In response.
Brunswick personnel developed operability assessments for the 98 MOVs
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l identified to have this condition. The following are examples of the l
methods used by the licentee to demonstrate MOV operability:
For some MOVs. the licensee relied on the close tor bypass setting to ensure that the valve will close.que switch
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At Brunswick.
96 percent of the closing stroke is used for the bypass setting to allow the licensee to take full credit for the actuator's degraded
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voltage motor capability.
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In a few instances. the licensee used actual measured stem
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friction coefficients or packing loads in lieu of program-assumed values in demonstrating adequate MOV capability.
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The licensee identified several MOVs with torque switches set h
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above.the revised actuator capability under worst-case degraded
voltage conditions..In these cases, the valves are able to L
achieve their close safety function, but the motor might stall and become overheated if the torque switch does not open.
The.
licensee performed a thorough review of the affected valves'
safety functions to determine if the valves are needed to reopen for a subsecuent safety function.
If so, the licensee then developed acditional justifications for operability; However, if the valve did not need to reopen, the licensee considered it acceptable to allow the valve to stall in its safety function j
position.
For those MOVs that had torque switches set above the motor
capability but must open later in an accident, the licensee applied its " modified Commonwealth Edison (Comed) method" to demonstrate capability above the motor's nameplate rated output.
i With this approach, the licensee compared the locked rotor torque values obtained from motor testing documented in Comed's White Paper 125 to the locked rotor torque values indicated by each
' motor's generic performance curve.
These differences were statistically analyzed to determine an appropriate, reduction to apply to a given motor's locked rotor torque The licensee also replaced the standard degraded voltage squared term in Limitorque's sizing equation with an exponent of 2.25.
The inspector reviewed the operability assessments for the Brunswick MOVs and discussed each assessment with licensee personnel.
The inspector did not identify any immediate o)erability concerns, but provided comments to licensee personnel.
or example, the licensee will need to provide additional justification if its " modified Comed method" i
is used to generally determine available motor torque.
Also, the i
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licensee should focus its operability assessments on available test data rather than using qualitative arguments when valve-specific test data exist.
In response to the inspector comments. the licensee strengthened the operability assessment for some MOVs and specified modification schedules for certain marginal MOVs in its letter (BSEP 98-0058) dated March 20, 1998.
CP&L Resoonses to NRC Insoection Reoort 50-325(324)/97-11 Subsequent to IR 50-325(324)/97-11. the licensee's letter (BSEP-97-0408)
dated October 2,1997, identified 32 items that needed to be addressed in its MOV program to meet GL 89-10.
The licensee's CR 97-02911 documented the completion of these items. The following sections discuss the 32 items:
(1)
(Closed) MOV Caoability Tab _le:
During the inspection covered by IR 50-325(324)/97-11. the ins)ector found it difficult to clearly determine the capability of tie Brunswick GL 89-10 MOVs.
During this ins)ection. the licensee provided a table that included all MOVs in 3runswick's GL 89-10 program and specific information on their capability.
The inspector found the table to be accurate and useful in achieving the goals of the inspection.
This item is closed.
(2)
(Closed) VOTES Open Uncertainty Review:
During The inspection covered by IR 50-325(324)/97-11, it was found that the licensee had not accounted for the potentially large VOTES diagnostic equipment uncertainty that occurs if open thrust measurements are not within the calibration range.
Since then, the licensee reviewed its diagnostic test data and revised the setup calculations to use the correct open uncertainties.
During this
inspection, the inspector checked the uncertainties used in several setup calculations. The inspector found the correct diagnostic uncertainty values being used. This item is closed.
(3)
(Closed) Basis for Comoletion Letters: The licensee submitted i
letters notifying the NRC of the implementation of the GL 89-10
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program for Brunswick Units 1 and 2 on June 21. 1995, and April 12, 1996. respectively.
During The ins)ection covered by IR 50-325(324)/97-11. the inspector questioned t7e basis for those GL 89-10 com)letion letters.
In response, the licensee indicated that the ) asis for the letters was completion of MOV testing and preparation of MOV setup calculations using the best available information at that time.
Subsequently, additional information
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became available to the licensee and the industry which potentially affected the MOV setup calculations. The licensee i
l then developed the CP&L GL 89-10 Corporate Improvement Plan and the 1997 Improvement Plan & Schedule for the Brunswick MOV Program to address these concerns.
During this inspection, the inspector reviewed the licensee's actions with respect to evolving industry MOV information and concluded that the licensee has taken (or is
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taking) adequate actions to address these items.
This item is closed.
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-(4)
(Closed) Justify 0.7 Wedae Gate Valve Factor Assumotion: During
-The inspection covered by IR 50-325(324)/97-11. it was noted that
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the' licensee had selected a valve factor of 0.7 for wedge gate valves that was not based on valve-specific test data. -Since.
l then, the licensee conducted an industry survey to obtain test
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data for wedge gate valves.
With respect to Anchor / Darling flexwedge gate valves, the licensee
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obtained Comed White Pa)er 154 to help justify its 0.70 valve
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factor assumption for tiose valves.
In addition, the licensee l~
updated its valve factor justification (BNP-MECH-MOV-VF) to L
separate its valves by valve size and American Nuclear Standards Institute (ANSI) pressure classes to identify those groups that included in-316nt test results,. Based on Brunswick and industry test. data, tie inspector concluded the licensee had adequately
supported use of a 0.70 valve factor for its Anchor / Darling L
flexwedge gate valves.
l With respect to Velan solid-wedge gate valves (size 2 inches and smaller). the licensee originally assumed a valve factor of 0.70 l
for these-valves.
In res3onse to IR 50-325(324)/97-11. the l
licensee reviewed the Comed White Paper 172 on miscellaneous valve data and noted that the highest " Bounding Group"ll industry test valve factor was l
1.067.
Based on the Comed information and overa data, the licensee considered a valve factor of 1.0 to be reasonably bounding for use in the setup.calculacions for these valves.
The inspector determined this approach to be acceptable.
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This item is closed.
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(5)
(Ocen) Justify Globe Valve Factor Assumotions:
During The
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inspection covered by IR 50-325(324)/97-11. it was noted that the i-licensee assumed a valve factor of 1.10 for non-dynamically tested globe valves. The licensee considered this valve factor supported by its overall in-plant globe valve test results.
However, the licensee's test data did not include all globe valve types at
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Brunswick.
In response to IR 50-325(324)/97-11. the licensee i
conducted an industry survey to obtain test information for non-testable globe valves used at Brunswick. The survey results were documented in CR 97-02911. Task 7.
After review of the survey results, the inspector determined that it was not clear if all of Brunswick's globe valve designs had been discussed with other i
licensees.
In its letter (BSEP 98-0058) dated March 20, 1998 the licensee committed to conduct an additionai industry survey for globe valves to obtain more specific test information based on valve type., size. rating, and manufacturer by July 1. 1998. With the typical acceptability of a 1.10 valve factor for globe valves.
the inspector considered it acceptable to close the NRC staff review of the Brunswick GL 89-10 program with the licensee's commitment to strengthen its survey to confirm its valve factor L
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E l^
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assumption for globe valves.
This item will be identified as the
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i first example of inspector followup item IFI 50-325(324)/98-03-01.
Completion of MOV Program Followup Items.
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(6)
-(Closed) Justify Stem Friction Coefficient < 0.20 Assumotions:
In L
Inspection Report 97-11 the inspector noted that the licensee had i
assumed a stem friction coefficient for-certain MOVs less than the 0.20 program assumption.
In response to IR 50-325(324)/97-11. the
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licensee committed to revise the stem friction coefficient r
assumptions for MOVs 1-E11-F047A. 2-E51-F007. and 2-G31-F001 to a l
value of 0.20 to agree with the program assumptions. The licensee also committed to revise the stem friction coefficient
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~ justification to 2nsure that less conservative assumptions are
3roperly justified and controlled in the MOV setup calculations.
)uring this inspection, the inspector reviewed the current setup calculations for MOVs 1-E11-F047A. 2-E51-F007. and 2-G31-F001, and i-verified that a stem friction coefficient of 0.20 was used. These l
revisions were documented by CR 97-02911. Task 8.
The licensee l
also had made changes to the stem friction coefficient
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justification (BNP-MECH-MOV-SF) to ensure that stem friction coefficient values less than 0.15 are used only with specific l-justification, and to apply a default value of 0.20 when test data are not available.
These actions were documented by CR 97-02911.
l Task 9.
This item is closed.
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I (i)
(Closed) Acolication Factor Assumotions for DC-Powered MOVs:
In
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IR 50-325(324)/97-11. the.. inspector noted that the licensee had l
not verified the acceptability of an application factor of 1.0 in
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its sizing calculation for DC-powered motor actuators.
In
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response to IR 50-325(324)/97-11. the licensee committed to revise the DC-powered MOV setup calculations to require the use of a more typical 0.9 application factor. The licensee also committed to
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revise EGR-NGGC-0203 to allow a 1.0 application factor only for ac-powered MOVs in accordance with Limitorque Technical Update 93-03.
During this inspection, the ins)ector verified that the specified change was made to Rev. 4 of EGR-NGGC-0203, as documented by CR 97-02911, Task 10. Completion of setup calculation revisions was documented by CR 97-02911. Task 42. The inspector also reviewed MOV setup calculations for DC-powered MOVs 1/2E41-F001. 1/2E41-F003. 1/2E51-F008, and 1/2G31-F004 and
i verified that an application factor of 0.9 was used.
This item is closed.
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(8)
(Closed) Justify Data Extracolation Guidance:
During the
,
l inspection covered by IR 50-325(324)/97-11. it was found that
)
Section 9 of EGR-NGGC-0203 did not include EPRI's latest recommendations for identifying the valve disc loads that are necessary to ensure that test results can be reliably extrapolated.
In response, the licensee revised this document to L
reference EPRI's guidance, as indicated by CR 97-02911. Task 11.
During this inspection, the inspectors verified that Section 9,1.22 of EGR-NGGC-0203 required that dynamic test conditions meet
!
EPRI's extrapolation guidance and, if not, further justification of test conditions would be necessary. This item is closed.
(9)
(Closed) EGR-NGGC-0203 Toroue Switch Peoeatability Correction:
During the inspection covered by IR 50-325(324)/97-11, it was found that the thrust margin calculation contained in Section 9.6.2.6 of EGR-NGGC-0203 did not include consideration of torque switch repeatability.
In response, the licensee committed to-revise this document and to revise Procedure OMMM-030, Motor Operated Valve Analysis Tracking, and Trending Program, to be consistent with EGR-NGGC-0203.
During this inspection, the inspector verified that EGR-NGGC-0203 was revised to include torque switch repeatability as documented by CR 97-02911. Task 12.
Also, the licensee revised Procedure OMMM-030 to require that differential pressure test evaluations be performed in accordance with the guidance provided in EGR-NGGC-0203.
This action was documented by CR 97-02911, Task 43.
This item is closed.
(10)
(Closed) Seat Area Determination for Globe Valves:
In IR 50-325(324)/97-11, the inspector found that licensee personnel had not reviewed Brunswick's globe valve population to identify valves that may exhibit guide-based behavior.
This' review was needed to ensure that adequate disc area terms are used in the thrust calculations.
In response to IR 50-325(324)/97-11, the licensee revised Section 9.1.15.2 of EGR-NGGC-0203 to reference EPRI's guidance for determining the proper disc area. During this inspection, the inspector verified that the revised MOV setu)
calculations used disc area terms based on this guidance.
T11s item is closed.
(11) _(Closed) Insoection of Actuators Utilizina Extended Ranae Values:
During the inspection covered by IR 50-325(324)/97-11. it was noted that the licensee did not include specific guidance in its actuator maintenance inspection procedures for MOVs with extended thrust ratings.
In response to that finding, the licensee revised its procedure OPM-M0504. Mechanical Inspection and Lubrication of Limitorcue Operators.
During this inspection, the inspector reviewec and found the procedure change to be acceptable. This item is closed.
(12)
(Ocen) DC-Powered MOV Stroke Time Test Methodoloav:
During the inspection covered by IR 50-325(324)/97-11, it was found that the licensee's method for determining DC-powered MOV stroke time did not account for the motor load profile that would be present under design-basis conditions.
In response to IR 50-325(324)/97-11. the licensee performed comparisons between its method and a Boiling Water Reactor Owners Group (BWROG) method for four MOVs under differing load assumptions and found that neither methodology was more conservative in all cases.
This analysis was documented in CR 97-02911. Task 15.
During this inspection, the inspector noted that, although the licensee's methodology was more conservative than the NRC staff's independent calculations, the licensee's
method predicted the same stroke time for differing load assumptions, rather than predicting longer stroke times as load increased.
In response, the licensee initiated CR 98-00541 to address the independence of the stroke time from loading conditions indicated by its DC-?owered MOV stroke time methodology.
In its letter (BSEP 97-0408) dated October 2. 1997, the licensee had committed to complete the consideration of its DC-powered MOV stroke time methodology by September 1, 1998.
With the margin provided by the licensee's DC-powered MOV stro b time methodology the inspector determined that the NRC staff review of GL 89-10 could be closed with the licensee's plans to further address the effect of load on its DC-powered MOV stroke time prediction.
This item will be tracked by an inspector followup item.
(13)
(Closed) GL 89-10 Sucolement 3 Desian Basis Differential Pressure:
During the inspection covered by IR 50-325(324)/97-11, questions were raised regarding the differential pressure values assumed for MOVs within the scope of Su)plement 3 to GL 89-10 which must close under High Energy Line Breac (HELB) conditions. After further discussions with licensee personnel, the inspector considered these questions to be resolved.
This item is closed.
(14)
(Closed) Response to Information Notices 96-48 and 97-07: During the inspection covered by IR 50-325(324)/97-11, it was not clear that the licensee had responded to Information Notices 96-48.
Motor-0perated Valve Performance Issues, and 97-07. Problems Identified During Generic Letter 89-10 Closecut Inspections. _ In response, the licensee documented its review of these notices in CR 97-02911. Task 18.
During this inspection, the ins)ector determined that the license has adequately addressed t1e information in those notices.
This item is closed.
(15)
(Closed) Valve 2-E51-F046 Stall Event: On June 8. 1994. Reactor Core Isolation Cooling (RCIC) Valve 2-E51-F046 tripped on thermal overload during an open stroke.
During the inspection covered by IR 50-325(324)/97-11. it was found that the licensee had not performed a weak link analysis to determine if any damage to the actuator or valve had occurred.
Subsequent licensee analysis determined that the valve weak link limits may have been exceeded.
During this inspection.'the inspector noted that the valve had been disassembled and inspected, and that no problems had been identified as documented in CR 97-02911. Task 19. This item is closed.
(16)
(Ocen) Core Soray System Valve 2-E21-F0318:
As documented in LER 2-97-002, the licensee declared Core Spray System Valve 2-E21-F031B inoperable when it determined that the torque switch settings for this MOV were incorrect.
Further analysis by the licensee later determined that the settings were adequate on a short-term basis.
However, the plant technical specifications state that this valve must close for a containment isolation
22 function.
During this inspection, the licensee provided the inspectors with a revised operability assessment which indicated that the valve was operable but degraded.
In its letter (BSEP 98-0058) dated March 20, 1998, the licensee committed to upgrade and i
test this valve.
With the current capability of this MOV. the inspector determined that the NRC staff review of GL 89-10 could be closed based on the licensee's commitment to Upgrade and test this MOV. This item will be tracked by an inspa cor followup item.
(17) (Closed) Planned MOV Modifications Schedule:
During the inspection covered by IR 50-325(324)/97-11. the licensee did not have available a specific up-to-date MOV modification schedule.
During this inspection, the inspector reviewed the modifications planned for the next unit refueling outages.
In response to inspector questions, the licensee described a more detailed modification schedule with specific actions for certain MOVs then provided in its letter (BSEP 98-0058) dated March 20, 1998.
This item is closed.
(18) (Closed) MOV Post Maintenance / Modification Testino: The issues raised in this item are addressed under other items as follows:
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Valve 2-E51-F064 Stall Event (Item 15): Valve 1-G31-F004 Packing Adjustment (Item 29): Valve 2-E11-F009 Stall Event (Item 31): Post
Maintenance Test Requirement Following Valve Packing Adjustments (Item 32).
This item is closed.
(19) (00en) Ball Screw Sten Nut Rate of Loadina and Efficiency Justification:
During the inspection covered by IR 50-325(324)/
97-11, it was found that the licensee had assumed that ball-screw stem nuts were immune to stem friction coefficient changes and load sensitive behavior. The licensee did not have in-plant test data to support this assumption, but used vendor-provided stem factor information.
In its letter (BSEP 98-0058) dated March 20.
1998, the licensee committed to resolve these questions by performing tests of 8 MOVs that use ball-screw stem nuts.
Based on vendor and industry information on ball-screw stem nut performance, the inspector determined that the GL 89-10 review
could be closed with the licensee's commitment to conduct plant-specific tests on its ball-screw stem nut actuators.
This item j
will be identified as the second example of inspector followup
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item IFI 50-325(324)/98-03-01. Completion of MOV Program Followup Items.
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(E0) (Closed) Stub-ACME Threaded Stem Rate of Loadino and Coefficient of Friction Plan: During the inspection covered by IR 50-325(324)/
97-11 it was noted that the licensee assumed that the program r
default of 0.20 for stem friction coefficient was applicable to MOVs with stems that have stub-ACME threads.
However, the licensee's stem friction coefficient justification did not contain data obtained from in-plant testing of valve stems with stub-ACME threads.
In response to IR 50-325(324)/97-11. the licensee i
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i conducted an industry survey of available stub-ACME thread information. Other licensees indicated that they had not observed any significant difference in performance of MOVs with stud-ACME threaded stems as compared to MOVs having stems with general purpose ACME threads.
Further, the Robinson plant indicated that it will be conducting additional dynamic tests with Teledyne strain gages of stub-ACME threaded stems. This information will be reviewed by Brunswick personnel when available.
Based on the
information provided by other licensees and the planned testing at l
Robinson, the inspector considered this item closed.
(21) (Closed) Globe Valve Rate of Loadina and Coefficient of Friction Plan:
During the inspection covered by IR 50-325(324)/97-11 the inspector noted the licensee's load sensitive behavior analysis contained a limited amount of globe valve data.
In response, the licensee reviewed globe valve fluid conditions and piping configurations at Brunswick to evaluate potential differences in globe valve load sensitive behavior (as documented by CR 97-02911,
,
Task 25). Other than MOVs 1/2E21-F015A/B, the licensee expected
{
its globe valves to exhibit load sensitive behavior similar to its gate valves because the globe valve fluid conditions were detennined to not be severe. The licensee dynamically tested MOVs
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1/2E21-F015A/B and used actual load sensitive behavior data in its setu) calculations for these valves.
Further. the licensee will be o)taining additional test information on its globe valves such as described in its letter (BSEP 98-0058) dated March 20, 1998.
This item is closed, (22) (Closed) Reactor Recirculation System Valves 1/2E32-F031A/B Desian-Basis Differential Pressure:
During the inspection covered by IR 50-326(324)/97-11, it was noted that the licensee's design-
)
basis documents identified a closing design-basis differential pressure of 65 psid for the Reactor Recirculation Pump Discharge
'
Valves 1/232-F031A/B.
In response to IR 50-325(324)/97-11, the licensee conducted further review of the design-basis fluid conditions and MOV capability as documented in BNP-MECH-B32-DP
Evaluation. and BNP-MECH-B32-F031A/B Evaluation.
In this re-
evaluation, the licensee provided additional support for its design-basis differential pressure assumption for these valves.
The licensee predicted the closing thrust requirements using the EPRI PPM hand-calculation methodology for these Anchor / Darling
double disc gate valves.
The licensee also indicated that the l
MOVs have a closing capability based on this methodology up to approximately 100 psid.
The inspector did not identify any concerns with the licensee's evaluation of these MOVs.
However, the licensee will be expected to address the conditions and limitations (such as consideration of potential blowdown effects)
described in the NRC Safety Evaluation (SE) (dated March 15. 1996)
This item is close l 24-(23) (00en) Anchor /Darlina Double Disc Gate Valve Factors:
During the inspection covered by-IR 50-325(324)/97-11. the inspector found the licensee to not have-taken sufficient action to provide an adequate basis for its valve factor assumption for Anchor / Darling-double disc gate valves.
In response to-IR 50-325(324)/97-11 the licensee performed an industry survey of valve performance for
,
Anchor / Darling double disc gate valves.
The licensee determined that the EPRI MOV PPM was the most appropriate method to establish design-basis requirements for this valve design. Therefore, the licensee re-performed the setup calculations for its Anchor /
Darling double disc gate. valves using the EPRI PPM.
During this inspection the inspector noted that the licensee had not. formally reviewed the NRC SE conditions and limitations associated with use of the EPRI PPM.
In response, the licensee committed in its letter (BSEP.98-0058) dated March 20. 1998, to address the EPRI PPM conditions and limitations for Anchor / Darling double disc gate valves hy August 1. 1998.
With the licensee *s use of the EPRI PPM in the setup calculations for its Anchor / Darling double disc gate valver, the ins)ector considered the licensee to have adequately justified its t1 rust prediction for these valves sufficient to close the NRC review of GL 89-10 at Brunswick with the licensee's commitment to address the conditions and limitations in the NRC SE on the EPRI PPM. This item will be identified as the third example of ins)ector followup item IFI 50-325(324)/98-03-01.
Completion of 10V Program Followup Items.
(24) (Closed) Butterfly Valve Toraue Reauirements:
During the inspection-covered by IR 50-325(324)/97-11. it was noted that the
'
licensee did not have in-plant test data to support use of the manufacturer's torque requirements for its b;tierfly valves.
In.
response to IR 50-325(324)/97-11. the licene was able to obtain limited. test information from other facili' :
but was unable to obtain Fisher or PosiSeal test data. The licensee determined that the EPRI PPM could not be directly applied to the Jamesbury
,
design, and that the EPRI PPM had only limited applicability to its Fisher and PosiSeal designs. Therefore, the licensee plans to install Teledyne smart couplings which will allow direct force measurements to be acquired under dynamic conditions. As
,
described in the licensee's letter (BSEP-98-0058) dated March 20.
1998, testing is scheduled to be performed on several butterfly valves to confirm the adequacy of the manufacturer's predicted torque requirements. With the margin currently available for these MOVs above the manufacturer's torque predictions and the plans for additional dynamic testing of several butterfly valves, f-the inspector considered this item to be closed.
[
(25) (Closed) MOV Self Assessment Issues:
During the ins)ection
"
I covered by IR 50-325(324)/97-11. it was noted that t1e licensee had not incorporated self assessment findings into the 1997
,
Improvement Plan & Schedule.
During this inspection, the inspector reviewed the revised Improvement Plan & Schedule and
determined that the previous findings had been incorporated as
l
'
l
well as a number of additional assessment findings that had been identified by the new MOV program manager.
These and the previous findings had been assigned to specific individuals and were being tracked by the corrective action program.
The inspector concluded l
that the licensee's ongoing actions with. respect to assessment
'
findings were adequate. This item is closed.
(26) (Closed) MOV Setuo Calculation Revision:
In IR 50-325(324)/97-11.
,
l it was indicated that the licensee was revising its setup calculations.for all Brunswick GL 89-10 MOVs to incorporate improvements and new information.
Since The inspection covered by IR 50-325(324)/97-11. the licensee completed its revision of the MOV setup calculations to incorporate the improvements and to
!
account for more conservative actuator efficiency assumptions.
Further. the licensee recently added several MOVs to the Brunswick
,
GL 89-10 program to resolve issues that were documented in an NRC l
i SE (dated October 16. 1996) on the scope of the GL 89-10 program at the Hatch Nuclear Plant.
During this inspection, the inspector
.
reviewed a sample of the revised calculations and found them to be
!
complete and accurate.
Further. the inspector verified that setup calculations had been prepared for MOVs 2-B32-V22. 2-E11-F004D. 2-
'
E11-F0208, 2-RCC-V28. and 2-RC-V52 which were recently added to the Brunswick GL 89-10 program.
The inspector considered this I
item closed.
(27)
(Closed) GL 89-10 ScoDe Chances:
This item was a duplication of Item 30.
This item is closed.
(28)
(Closed) Periodic Verification of MOV Settinas: The licensee's l
plans for periodic verification of MOV settings were discussed in IR 50-325(324)/97-11 (Section E1.1.b.4).
The licensee's long-term l
MOV periodic verification program will be reviewed under GL 96-05.
Periodic Verification of Design-Basis Capability of Safety-Related
!
Motor-Operated Valves.
This item is closed for GL 89-10.
(29)
(00en) RWCU System Valve 1G31-F004 Packino Adiustment:
On October 23. 1996, the licensee re-torqued the packing gland nuts
installed on Reactor Water Cleanup (RWCU) System Valve 1-G31-F004 i
to 33 feet-pounds to stop a minor packing leak.
The licensee did
'
not perform a post-maintenance thrust verification prior to i
returning the MOV to service.
The licensee stroked the valve L
electrically with no problem identified.
The licensee determined that diagnostics measurements could not be reliably obtained during the test because of the high stem temperature.
The licensee relied on information from the packim manufacturer that this adjustment would not significantly chane the packing load.
In response to inspector questions, the lice ".,ee committed in its letter (BSEP 97-0408) dated October 2. 1997, and renewed in its letter (BSEP 98-0058) dated March 20. 1998, to perform a thrust verification during the spring 1998 refueling outage.
With the information from the packing manufacturer and the valve stroke test performed by the licensee, the inspector determined that the
F
GL 89-10 review could be closed with the licensee's commitment tio verify the packing load during the next refueling outage. This item will be identified as the fourth example of inspector followup item IFI 50-325(324)/98-03-01. Completion of MOV Program-Followup Items.
(30)
(Closed) GL 89-10 MOV Scooe Validation: During the inspection covered by IR 50-325(324)/97-11. the inspector questioned the omission of certain MOVs from the Brunswick GL 89-10 program.
In response to IR 50-325(324)/97-11. the licensee conducted a review of the NRC SE (dated October 16, 1996) on the scope of.the GL 89-10 program at the Hatch Nuclear Plant.
Based upon this review.
the licensee added 47 MOVs-(that had either been previously removed or never included) to'its GL 89-10 program.
The licensee removed one valve from its GL 89-10 program based on its review.
During this inspection, the inspector reviewed a sample of the tiOV additions to (and the deletion from) the GL 89-10 progr6m.
The inspector did not identify any concerns with the scope of the 1-licensee's GL 89-10 program.
This item is closed.
(31)
(Closed) Residual Heat Removal System Valve 2-E11-F009 Stall Event.
On March 3. 1996, the motor for RHR Shutdown Cooling Suction Valve 2-E11-F009 failed during a closing' stroke due to hydraulic locking of the actuator spring pack.
During the inspection covered by IR 50-325(324)/97-11. the inspector found that the licensee had not performed an overload analysis at the time of the event to determine if the actuator or valve was-l damaged. 'The licensee subsequently determined that the valve's
'
yoke clamp structural limit might have been exceeded. The licensee inspected the valve yoke and yoke clamp of MOV 2-E11-F009 on September 24, 1997. for cracking, looseness and general degradation.
No problems were identified. This item.is closed.
(32)
(Closed) Post-Maintenance Test Reauirements Followino Valve Packina Adiustment: In IR 50-325(324)/97-11. the inspector raised a concern regarding procedural controls to provide adequate post-maintenance testing following valve packing adjustments.
In res)onse to IR 50-325(324)97-11. the licensee modified Procedure OPL)-20. Post-Maintenance Testing Program, to require a thrust L
verification after packing adjustments prior to returning a valve I
to service. This action is documented in CR 97-02911. Task 27.
The licensee also conducted a review of packing adjustments performed in the last 2 years to verify that a post-maintenance thrust verification was performed.
With the exception of MOV l
.1G31-F004 (discussed Item 29 above), tests had been performed in l
all cases.
During this inspection, the inspector reviewed the l-licensee's actions regarding this item and considered them to be adequate.
This item is close c.
Conclusions In its letter (BSEP 98-0058) dated March 20. 1998, the licensee committed to: (1) implement MOV modifications according to the schedule in the Brunswick MOV Improvement Plan. (2) conduct differential-pressure tests of 27 gate, globe, and butterfly valves according to the schedule in the Brunswick MOV Improvement Plan. (3) conduct testing on 8 MOVs to address ball-screw stem nut load sensitive behavior and efficiency issues. (4) verify the packing loads for MOV 1-G31-F004 by direct force measurement during the spring 1998 outage. (5) address the conditions and limitations for Anchor / Darling double disc gates in the NRC SE on the EPRI PPM by August 1. 1998. (6) conduct an additional industry survey to obtain valve-specific globe valve data by July 1.1998, and (7) adjust the torque switch for MOV 2-E11-F024B to reduce excess seating torque during the next Unit 2 refueling outage. The licensee committed to provide the ongoing status of these actions by January 29, 1999, and a full completion status by January 31, 2001. The NRC staff review of the GL 89-10 program at Brunswick is being closed based on the completed and scheduled work, including the commitments in the licensee's letter (BSEP 98-0058) dated March 20, 1998.
The completion of the commitments in the licensee's letter (BSEP 98-0058) dated March 20, 1998, and the closure of the specific remaining items described above will be tracked as Inspector Followup Item IFI 50-325(324)/98 03-01. Completion of MOV Program Followup Items and may i
be reviewed during a future inspection.
E2 Engineering Support of Facilities and Equipment E2.1 Control Buildino Ventilation Modifications i
a.
Insnection Scooe (37551)
The inspector reviewed installation activities for the upgrade to the Control Building Ventilation System
,
b.
Observations and Findinas The licensee obtained NRC approval for a one time TS amendment to implement the modifications. The modification involved two major tasks.
The first task consisted of installing temporary barriers in the Control Building ventilation ductwork to allow Seismic Qualification Utilities Group required structural upgrades on the ductwork at the evaporation coils. A total of 16 days was allowed for this task.
The second task consisted of removing and replacing the three air conditioning (AC)
l condensing units.
A total of nine weeks of operation on temporary AC i
units was allowed until the new 0-class AC units were installed and l
operational.
After approval of the NRC one-time TS amendment. the licensee entered the 16 day LC0 and a nine week LCO on February 16. 1998.
The inspector routinely toured the work activity areas and moriitored the licensee status for completion of the modification.
i
28 On February 17, 1998, the inspector attended the 7:10 a.m. shift turnover meeting conducted in the control room for the crew that had just assumed operation of the )lant.
During the meeting it was discussed that training would 3e conducted at 8:00 a.m. that morning.
This operations crew had been off work for several days and had not conducted training concerning the modification and compensatory measures.
Some of the com)ensatory measures were the unique LCO requirements, special checcs on the operator rounds, and lessons learned from SCBA training.
The inspector discussed with licensee management that the training should be conducted orior to taking control of plant operations. The licensee reviewed the crew training conducted for all crews due to the modifications. One additional crew that was to start work that evening had not completed training. The licensee kept the day shift on shift until the evening shift had completed the training.
c.
Conclusions The inspector concluded that operator training for the oncoming crew on the compensatory measures during upgrades of the Control Building ventilation system was not timely E2.2 Structural Load of Concrete Shinoino Containers a.
Insoection Scooe (37551)
On February 23, 1998, the inspector reviewed the load analysis for the radwaste concrete shipping containers stored between the DG Building and Radwaste Loading Dock.
b.
Observations and Findinas During routine tours of the facility, the inspector observed the storage of containers on a concrete pad.
The containers are about 10 feet in diameter and weigh around 100.000 pounds.
There have been as many as twenty containers stored in this area. The inspector questioned the licensee regarding a load analysis due to underground piping in the area.
The licensee's res)onse was that CR 97-01453 was written on April 21, 1997, to address t1e same question.
The licensee determined that there was no structural loading concern.
This was based on the fact that the concrete pad was an eight inch thick reinforced pad.
The soil beneath the pad had been compacted to support a minimum load bearing of 3000 pounds per square foot (psf) and the concrete pad had a greater loading ability.
The calculated distributed load for the containers was about 1.300 psf or much less than 3.000 asf.
The elevation of the concrete pad was 19 feet 6 inches.
The higlest elevation of buried commodities in the area was 16 feet 10 inches for an electrical duct bank. The electrical conduits are encased in a concrete duct bank. The highest piping was 14 feet and covered by at least five feet of soil.
The licensee concluded that it was acceptable to store the containers in this locatio The inspector reviewed CR 97-01453. the licensee's analysis, plant drawing F-03343 (Electrical Underground Duct Runs), and F-02198 (Yard Piping Composite Plan).
c.
Conclusions The inspector concluded that the licensee had performed an adequate load analysis for the radwaste shipping containers and concluded that the storage of these containers was acceptable.
E8 Miscellaneous Engineering Issues (92903)
E8.1 (Closed) Violation VIO 50-325(324)/97-05-05: Timeliness of Operability and Reportability Determination On March 10, 1997 the licensee was notified by the BWR vendor of an error associated with the Unit 1 Cycle 10 Supplemental Reload Licensing Report.
The vendor had failed to inform the licensee that a TS amendment was needed to impose requirements to maintain fuel integrity during the continuous withdrawal of a high worth control rod. A report in accordance with 10 CFR 50.72 wt.s issued after discussions between the inspector and the licensee.
A violation was issued for the failure to promptly initiate a CR for a potential reportable event. The licensee responded to this violation in a letter dated June 11. 1997. A lack of knowledge by a cor) orate office engineer of the corrective action requirements for t1e prompt initiation of a CR to begin the i
reportability assessment was identified as the root cause.
Contributing
)
factors to this included the failure by the onsite engineering personnel to verify the existence of a CR and that a review for reportability had been completed for the failure to request a TS amendment.
The inspector reviewed the root cause and associated corrective actions.
Based on satisfactory completion of training on the requirements of the corrective action program for both corporate and onsite organizations involved, this item is closed.
E8.2 (Closed) Violation VIO 50-325(324)/97-08-01: TS/LC0 Administration In June of 1997 the licensee was assessing the operability of the
)
hydrogen / oxygen analyzer thermo-electric cooler units due to a 10 CFR 21 (Part 21) notice. The Part 21 suggested that the environmental qualified life for the cooling units be reduced from 40 to 2 years based on galvanic corrosion problems.
A violation was issued in IR 50-325 l
(324)/97-08. for the failure of the licensee to properly satisfy system
.
l operability concerns in accordance with plant procedure.
The licensee l
responded to this violation in a letter dated September 2. 1997.
'
l In April 1997 the Unit 1 Senior Reactor Operator (SRO) permitted the performance of a surveillance which initiated a chlorine detector isolation signal.
The signal also rendered the control room radiation and smoke protection functions of the Control Room Emergency Ventilation System (CREVS) inoperable.
The SR0 failed to log entrance into the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LCO.
Review of this event by the inspector and the licensee led to
L
!
l l
the discovery of multiple instances where inaccuracies, errors, and oversights were made concerning entrance into the correct LCO for inoperability of the CREVS. A violation was issued for the failure to properly disposition equipment operability concerns.
The licensee performed a stand down to reinforce management expectations regarding operability determinations, and Operating Instruction 001-1.08. Control of Equipment and System Status was revised to require the
!
concurrence of a second SRO for " logbook" LCOs. The inspector reviewed the Operating logs and determined that " logbook" LCOs were receiving concurrence by a second SRO.
Based on the satisfactory completion of the associated corrective actions this item is closed.
[
E8.3- (Closed) Unresolved Item URI 50-325(324)/97-08-09: 50.59 Review for Control Building Air-Conditioning Quality Classification Downgrade l
This issue was identified by the licensee during a Control Building HVAC SSFI completed June 27, 1996.
The Control Building HVAC air-conditioning units were downgraded from quality class ~0" to "non-0" without a 10 CFR 50.59 review.
However, the unit did remain operable after the classification downgrade. The licensee will install a new "0" class air-conditioning systems as part of the Control Building l
Ventilation System modification discussed in this IR. The non-re)etitive, licensee identified and corrected violation is identified as a
ion-Cited Violation, consistent with Section VII.B.1 of the NRC
!
Enforcement Policy.and is designated NCV 50-325(324)/98-03-02, Control Building Air-Conditioning Units Quality Classification.
E8.4 (Closed) Violation VIO 50-325(324)/97-11-04: Inadequate Corrective Actions for MOVs This violation concerned the identification of inadequate corrective action to address MOV 3rogrammatic and valve-specific problems-.
The violation identified t1ree examples.
!
l In the first example, the violation noted that the licensee had not resolved inadequacies in its justification for the valve factor assumption for Anchor / Darling double disc gate valves or fcr the torque requirements for its butterfly valves.
In response to this violation example, the licensee revised its setup calculations for Anchor / Darling double disc gate valves to apply the EPRI PPM methodology.
As discussed
'
L in Item 23 above, the licensee has committed in its letter (BSEP 98-
>
0058) dated March 20, 1998, to address the conditions and limitations in the NRC SE on the EPRI PPM.
Further, as discussed in Item 24 above, the licensee surveyed other licensees to obtain information on butterfly valve torque requirements and has committed in its letter (BSEP 98-0058)
dated March 20, 1998, to conduct additional dynamic testing of several butterfly valves to confirm the manufacturer's torque predictions.
In the second example the violation indicated that the licensee had relied on thrust measurements obtained from VOTES diagnostic equipment without addressing the potential uncertainties that occur if thrust l-i l
_ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ - _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _
. - _ _ _ _ _ _ _ -
31 measurements in the open stroke direction are not within the calibration range. As discussed in Item 2 above, the licensee has revised its MOV setup calculations to incorporate the correct uncertainty values.
l l
In the third example, the violation noted that the licensee had not evaluated stall events of MOVs 2-E11-F009 and 2-E51-F046 for possible structural damage to valve components.
As discussed in Item 31 and 15.
respectively, the licensee has inspected MOVs 2-E11-F009 and 2-E51-F046 and identified no structural damage.
I Based on the above discussion. the inspector concluded that the licensee's completed and planned corrective actions. including the
.
commitments in its letter (BSEP 98-0058) dated March 20. 1998, were l
adequate to close this violation. The completion of the commitments described in the licensee's letter (BSEP 98-0058) dated March 20. 1998, and the items described in this report will be tracked by the above identified inspector followup item IFI 50-325(324)/98-03-01. Completion of MOV Program Followup Items.
l IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Occuoational Radiation Exposure Control Proaram l
a.
Insoection Scoce (83750)
The inspector reviewed implementation of selected elements of the licensee's radiation protection program pertaining to control of l
occupational radiation exposure.
The review included com)arison of observed radiation control practices to descriptions of t1ose controls l
delineated in the licensee's Radiation Control and Protection Manual and i
to relevant sections of 10 CFR 20.
b.
Observations and Findinas
!
The inspector conducted tours of the Radiation Control Area (RCA) to observe radiation control practices including access controls to the RCA. use of dosimetry, radiation and contamination area surveys and postings, container labeling. and Jersonnel monitoring at the RCA exit portal.
Individuals entering the RCA were observed to be wearing dosimetry in an appropriate manner.
Thermoluminescent dosimeters (TLDs)
'
were used as the primary device for monitoring personnel radiation exposure.
In addition. alarming electronic dosimeters were used for l
monitoring the accumulated dose and the encountered dose rates during i
each RCA entry. The electronic dosimeters were set to alarm at administrative limits established for the specific Radiation Work Permit (RWP) under which the RCA entry was being made.
The electronic dosimeters were also used to log individuals on to the Radiation
'
Information Management System (RIMS) prior to their entry to RCA.
That system required individuals to acknowledge their dose and dose rate l
l l
_
limits for their RWP before granting access to the RCA.
As the individuals exited the RCA the accumulated dose and encountered dose rate information was transferred from the electronic dosimeters to the RIMS data base in order to track individual exposures.
The inspector determined that the licensee's radiation control practices for access to the RCA and for use of dosimetry were consistent with the Radiation Control and Protection Manual.
During the course of the inspection the inspector observed several individuals entering the RCA and determined that the above RCA access controls were followed.
The inspector also noted that personnel exiting the RCA routinely surveyed themselves for contamination using personal contamination monitors (PCMs).
The inspector reviewed survey maps which depicted the radiation and contamination levels at various locations within the RCA. Those maps were maintained at the RCA entrance for use by Radiation Protection personnel when briefing workers on the radiological conditions in their work areas.
The licensee provided the inspector with copies of maps for several areas selected by the inspector for use in verification of recorded dose rates and contamination levels.
At the inspector's request, a licensee Radiation Protection staff member performed dose rate and contamination surveys at several locations selected by the inspector from the provided maps.
The inspector verified that the survey instrument readings were consistent with the dose rates and contamination levels recorded on the area survey maps.
The inspector also verified that the postings for radiation and contamination levels in the selected areas were consistent with the area survey ma)s. During tours of the RCA the inspector also observed that containers ) earing radioactive materials were properly labeled.
c.
Conclusions Based on the above reviews and observations. the ins)ector concluded that the licensee's radiation control practices for RCA ingress and egress contrcl. 3ersonnel dosimetry, radiological surveys and postings, and container la)eling were consistent with the licensee's Radiation Control and Protection Manual and relevant sections of 10 CFR 20.
R1.2 Self-Assessment Proaram a.
Insoection Scope (83750)
The inspector reviewed the licensee's program for identifying and correcting deficiencies related to the control of radiation exposure and radioactive materials.
The licensee *s follow-up activities for a selected number of licensee identified issues were reviewed for consistency with the licensee's corrective action management program.
b.
Observations and Findinas The inspector recuested a listing of the CRs issued during the previous six months for acverse conditions related to radiological controls.
From that listing the inspector selected six CRs for review of the
I L
licensee's follow-up corrective actions. CR 97-04122 was issued for nonconforming material conditions in and around the Radioactive Material Container Storage Building (RMCSB).
The inspector toured the area.in and around the RMCSB and observed that the deficiencies noted in the CR had been corrected but housekeeping in that area could have been-improved.
CR 97-04112 was issued as a result of two NRC cited violations, failure to control a ~ locked high radiation area in accordance with procedures and failure to issue a CR for that event.
The inspector toured the Instrument Calibration Room and observed that
,
l lock and key controls had been implemented in order to control access to l
the room as Locked High Radiation Area when the calibration source was in use. Through discussion of this. issue with Radiation Protection supervisors and technicians, the inspector determined that licensee l
management 1ad held staff meetings to re-emphasize expectations regarding the importance.of timely documentation of adverse conditions hnd to determine whether supervisors were aware of other adverse conditions which had not been properly documented. As a result.
CR 97-04164 was issued to document that CRs had not been issued for five additional adverse conditions related to radiological controls for
'
labeling, posting. boundaries and gates.
During discussion of this
,
issue. Radiation Protection personnel indicated that those conditions had been corrected on-the-spot.
The inspector toured the areas involved and observed that the adverse conditions did not exist at the time of the tour.
CRs 98-00426 and 98.-00427 were recently issued to address problems related to the effects of high background radiation levels on survey instruments used for free release of equipment and materials from the RCA. At the time of this inspection not all of the action items.for these CRs had completed but the inspector did observe that survey procedures were in the process of being revised and that the licensee was collecting reference materials to establish the technical basis for the allowable instrument background count rate.
CR 98-00490 was recently issued to document that two items were found in the spent fuel pools which had not been tagged and that the tags for some other items in the pools did not have all the information required by procedure.
The CR indicated that immediate corrective action was taken to correct the deficiencies at the time of observation.
The ins)ector toured the spent fuel pool area and verified that the two items lad been properly tagged. The inspector also noted that the tags for two other items did not have the date and responsible person recorded in the spaces
)rovided.
The licensee indicated that inventory records were currently 3eing reviewed to find that information.
From the sample of six CRs selected for review. the inspector determined-that identified adverse conditions were being documented in accordance with plant program procedure OPLP-04 Corrective Action Management and a)propriate actions-were being taken to correct identified deficiencies.
W1en it was found that CRs had not always been issued for problems corrected on-the-spot. management expectations regarding identification.
i documentation and resolution of adverse conditions were re-emphasized to l
the staff.
!
l
.i
c.
Conclusions Based on the above reviews and observations, the inspector concluded that the licensee was generally effective at identifying and correcting radiological control related deficiencies in accordance with the corrective action management program.
Rl.3 Trainina and Qualification Proaram a.
Inspection Scooe (83750)
The inspector reviewed implementation of the licensee's training and qualification program for Radiation Protection and Chemistry
.echnicians.
The program was implemented through Training Instruction OTI-100 Conduct of Training.
The training records for selected technicians were evaluated against specific elements of the program described in the implementing procedure.
b.
Observations and Findinas The records for two recent shipments of spent fuel were reviewed by the inspector in order to identify the Radiation Protection Technicians involved in providing support for the shipments.
The inspector verified that the training records for four selected technicians included Qualification Checkout Cards for providing Health Physics (HP) support for spent fuel shipments and determined that the individuals identified in those Qualification Cards as Trainers and Evaluators were included on the approved list of Designated Evaluators / Trainers.
Chemistry logs were reviewed by the inspector in order to identify Chemistry Technicians who had recently performed chemical analyses. The inspector verified that the Chemistry Qualification Matrix indicated that two selected technicians were qualified to perform the reported chemical analyses listed on the Chemistry logs.
Three recently completed Qualification Cards were selected by the inspector for comparison to the Chemistry Qualification Matrix.
The ins)ector verified that the matrix accurately indicated the analyses that t1e technicians were qualified to perform.
c.
Conclusions Based on the above reviews, the inspector concluded that the licensee effectively implemented training and qualification for Radiation Protection and Chemistry Technicians in accordance with the training program procedural requirement l
R3 RP&C Procedures and Documentation l
R3.1 Work in Hioh Radiation Area a.
Insoection Scone (71750)
On March 3, 1998, the inspector observed work activities in a high radiation area for proper work practices.
b.
Observations and Findinas The inspector observed work activities on Unit 1 RHR loop A.
The area around the RHR pumps is designated as a contaminated and high radiation
{
area.
The inspector observed the personnel performing the work. control of items into and out of the area, and personnel removal of protective clothing upon exiting the area.
These activities were checked for i
conformity to the Special RWP R98-1016 covering the work activities.
No discrepancies were noted, q
c.
Conclusions The inspector concluded that radiological work activities in a contaminated and high radiation area were properly controlled.
R8 Hiscellaneous RCaC Issues (92904)
R8.1 (Closed) Violation V10 50-325(324)/97-09-05:
Failure of Radiation Protection Technicians to Complete Required Task Qualification Requirements Prior to Independently Performing a Task The inspector verified that the corrective actions delineated in the licensee's reply to the Notice of Violation, dated October 15, 1997, had been completed.
Documentation to support completion of those actions was included in the records file for CR 97-02329.
The licensee's corrective actions included counseling the individuals involved and holding staff meetings with Radiation Protection technicians and supervisors to review qualification status and to ensure qualification completion arior to performance of assigned activities.
The inspector discussed tlis issue with selected technicians and supervisors and those selected individuals indicated that departmental management had re-emphasized their individual responsibilities for remaining cognizant of the tasks for which they were qualified to perform.
The licensee's corrective actions also included the performance of a root cause analysis for the violation and disseminating the results to other departmental managers.
Documentation for distribution of the root cause analysis was included in the CR file. The inspector noted that the root cause analysis determined that the technicians task assignments were based on a review of training records instead of qualification records.
To correct that practice, the licensee placed notebooks. with the technicians task qualification records, in the Radiation Protection supervisors office area and at the entrance to the RCA where task assignments are usually made.
This violation is close S2 Status of Security Facilities and Equipment S2.1 Plant Access Control (71750)
i a.
Insoection Scooe On February 24, 1998, the ins ector observed the security force controlling access into the p ant.
b.
Observations and Findinos j
The inspector observed the plant security force process plant personnel into the 3rotected area. All activities were properly monitored.
Items passing t1 rough the x-ray machine were stopped and searched if a questionable image was seen.
I While the inspector was between the full length turnstile and waist high turnstile, a licensee personnel entered the turnstile area behind the
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inspector.
The full length turnstile was promptly locked and the second
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individual in the turnstile area removed.
The security guard counseled the individual concerning improper entry into the area.
{
c.
Conclusions The inspector concluded that security provided proper access control into the protected area.
V.
Manacement Meetinos XI Exit Meetino Summary The inspector presented the inspection results to members of licensee
management at the conclusion of the inspection on March 14, 1998.
Post inspection briefings were conducted on February 13 and March 6, 1998.
The licensee acknowledged the findings presented.
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l PARTIAL LIST OF PERSONS CONTACTED Licensee
'
A. Brittain. Manager Security
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M. Christinziano Manager Environmental and Radiation Control W. Dorman Supervisor Licensing and Regulatory Programs N. Gannon. Manager Maintenance J. Gawron. Manager Nuclear Assessment Section M. Herrell Manager. Training S. Hinnent. Vice President. Brunswick Steam Electric Plant K. Jury. Manager Regulatory Affairs B. Lindgren. Manager Site Support Services J. Lyash. Plant General Manager G. Miller. Manager Brunswick Engineering Support Section R. Mullis. Manager Operations i
I Other licensee employees or contractors included office, operation.
i maintenance, chemistry. radiation, and corporate personnel.
I NRC E. Brown J. Coley E. Guthrie D. Jones P. Kellogg C. Patterson T. Scarbrough M. Shymlock
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INSPECTION PROCEDURES USED f 37551:
Onsite Engineering u
j'
IP 61726:
Surveillance Observations-l IP 62700:
Maintenance. Implementation j
i-IP 62707:
- aintenance Observations L
r IP 71707:
Plant Operations i
iP 71750:
Plant Support Activities l
IP 83750:
Occupational Radiation Exposure-
!.
IP 92901:
Followup - Operations l
IP 92902:
Followup - Maintenance l
IP 92903:
Followup - Engineering (
'IP 92904:
Followup - Plant Support Temporary Instruction 2515/109:
Safety-Related Motor Operated Valve i
Testing and Surveillance
,
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ITEMS OPENED AND CLOSED Ooened
'50-325(324)/98-03-01 IFI Completion of MOV Program Followup Items (paragraph E1.1)
!
50-325(324)/98-03-02 NCV Control Building Air-Conditioning Units Quality Classification (paragraph E8.3)
Closed 50-325(324)/97-11-02 VIO PNSC Quorum Too Many Alternates (paragraph 08.1)
50-325/97-006-00 LER Engineered Safety Feature Actuation Due to Loss of Emergency Bus E2 (paragraph M8.1)
50-325(324)/97-08-02 VIO Failure to Verify / Check E-Bus Relay Operability (paragraph M8.2)
'50-325(324)/97-05-05-VIO Timeliness of Operability and Reportability r
J Determination (paragraph E8.1)
50-325(324)/97-08-01 VIO TS/LCO Administration (paragraph E8.2)
!
l 50-325(324)/97-08-09 URI 50.59 Review for Control Building Air-Conditioning Quality Classification Downgrade i
(paragraph E8.3)
50-325(324)/98-03-02 NCV Control Building Air-Conditioning Units Quality Classification (paragraph E8.3)
50-325(324)/97-11-04-VIO Inadequate Corrective Actions for MOVs (paragraph E8.4)
F L
i l
50-325(324)/97-09-05 VIO Failure of Radiation Protection Technicians to
!
Complete Required Task Qualification Requirements Prior to Independently Performing a Task (paragraph R8.1)
,