IR 05000285/1989026
| ML20247M084 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/24/1989 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20247M009 | List: |
| References | |
| 50-285-89-26, NUDOCS 8908010433 | |
| Download: ML20247M084 (21) | |
Text
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g n... ' ' g b APPENDIX C U.S. NUCLEAR REGULATORY COMMISSION E REGION IV-t NRC Inspection Report: '50-285/89-26 Operating License: DPR-40 Docket: 50-285 , h-
- Licensee: Omaha Public. Power District (0 PPD)
L 444 South'16th Street Mall Omaha, Nebraska-68102-2247 L Facility Name: FortCalhcunStation(FCS) L Inspection At: FCS, Blair, Nebraska Inspection Conducted: June 1-30, 1989 Inspectors: P. H. Harrell, Senior Resident Inspector T. Reis, Resident Inspector R. P. Mullikin Project Engineer Approved: - 7/2V!f7 ' T. F. Westerman ipf) Project Section B Date Divisfon of Rea Wojects Inspection Summary Inspection Conducted June 1-30, 1989 (Report 50-285/89-26) Areas Inspected: Routine, unannounced inspection including review of ' previously identified items; licensee event report followup; operational safety verification; plant tours; monthly maintenance observations; monthly surveillance observations; security observations; radiological protection ! observations; in-office review of periodic, special, and nonroutine event reports; and followup of onsite events.
Results: During this inspection period, the inspectors reviewed the areas discussed below. The discussion provides an overall evaluation of each area.
The inspectors reviewed the actions taken by the licensee in response to previously identified items and licensee event reports. Based on reviews of - the actions taken by the licensee, it appeared that, in the majority of instances, the licensee had appropriately implemented both the short-and long-term actions to prevent recurrence of the identified problems.
In one instance of excessive vibration, noteo by the inspector, the licensee failed to promptly analyze the potential significance of the vibration. A deviation f rom commitments in the Updated Safety Analysis Report resulted f rom this (paragraph 3.e).
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During the. review of the actions taken by the licensee in response to problems l with the, emergency lighting system, as discussed in paragraph 5, the licensee failed to maintain an hourly fire patrol for the cable spreading room. The , commitment to. maintain the fire patrol is documentedLin Licensee: Event ] L Report 89-014.- Therefore, the failure to maintain the fire patrol appeared to deviate from a commitment made to NRC. Nomally, a notice of deviation would , be issued; however, the licensee identified the problem and took appropriate ' g ' actions to correct the problem and to prevent recurrence. Therefore, a P deviation was not issued.
During observations of activities and evolutions performed by the operations ' staff, the inspectors noted no problems with the performance of the staff.
It appeared.that the licensee's operations staff performed their duties in a professional manner to ensure safe plant operation.
I The inspectors performed numerous tours of the plant during this inspection r ' period. During the tours, no problems were noted.
It appeared that the licensee has increased the quantity and quality of their tours to identify . potential nonconforming items. The inspectors did note that the licensee 'is continuing the ongoing efforts to upgrade plant appearance and housekeeping.
Maintenance and surveillance activities were observed by the inspectors during this inspection period. During observation of these activities, the inspectors . noted that the activities were performed in a professional manner. During observations of. surveillance activities, the inspector encountered a' deficient procedure caught by the astuteness of the shift supervisor and an example of technicians' perfoming a test and failing to document and report a' deficiency.
The' latter is a violation of the NRC requirements (paragraph 8.a).
During observations of the activities and tasks performed by security and Tealth physics personnel, the inspectors noted that these personnel performed their duties in a professional manner. No observations or concerns were fdentified with these activities during this inspection period. Two security events occurred that. involved'the access control of packages and the control of safeguards information. These events will be reviewed by security specialists from the. Region IV office.
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Persons Contacted TJ. Adams, Reactor Engineer j s D. Andes, Nuclear Safety Review Group
- R.. Andrews, Division Manager, Quality and Environmental Affairs
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- J. Bobba, Supervisor, Radiation Protection C. Brunnert, Supervisor, Operations Quality Assurance
< D. Ferris, Licensing Engineer-
- J. Fisicaro', Manager, Nuclear Licensing and Industry Affairs
- S. Gambhir, Division Manager,. Production Engineering Division
- J. Gasper, Manager,7 Training R. Jaworski, Manager, Station Engineering J. Kecy, Supervisor, System Engineering
'*R. Kellogg, Supervitor,'Special Servicet.
T. Mathews, Station Licensing Engineer-.
- D. Matthews, Supervisor, Station Licensing
- T. 'McIvor, Manager,. Nuclear Projects
' *K. Miller, Supervisor, Maintenance
- K. Morris,' Division Manager, Nuclear Operations
- W..Orr, Manager, Quality Assurance and Quality Control
- G. Peterson', Mana~ger, Fort Calhoun Station
- A. Richard, Assistant Manager, Fort Calhoun Station
- J. Sefick, Manager, Security Services
- C. Simmons, Station Licensing Engineer
- F. Smith, Plant Chemist:
- J. Tills,. Assistant Manager, Fort Calhoun Station D. Trausch, Supervisor, Operations S.: Willrett, Supervisor, Administrative Services
- Denotes attendance at the monthly exit interview.
The. inspectors also contacted other plant personnel, including operators, technicians, and administrative personnel.
2.
Flant Status , During this inspection period, the plant operated continuously at 100 percent power. No plant perturbations or demands for safeguards equipment were experienced.
, 3.
Review of previously Identified Items (92701 and 92702) a.
(Closed) Open Item 285/8909-08: Personnel walking on safety-related piping.
This open item involved an observation by the inspector that an individual was walking on the steam supply line to the auxiliary _ _ _ - _ _ -
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i .feedwater pump. The line is 2 inches in diameter. The inspector's concern was that the piping could be overstressed due to the weight ! of the individual.
l i To address this concern, the licensee performed an engineering evaluation. On March 7, 1989, the licensee issued an evaluation that stated the piping wcs not damaged due to the additional weight of the j individual. (In addition, the evaluation also stated that personnel J could walk,on piping 2 inches.or greater in diameter without damaging
- the piping or pipe supports. The licensee's conclusion was based on Level-D allowables contained in the ASME Code. The NRC's Office of
' Nuclear Reactor Regulation (NRR) reviewed the evaluation performed by ! "the licensee. NRR concurred with the licensee's conclusions.
r The licensee addressed the guidelines for personnel walking on piping in memoranda dated March 7 and 15, 1989. The memoranda provided the ' basic rules, established by' engineering evaluation, to all personnel.
In addition,- the licensee issued memoranda, dated March 15 and - April 18, 1989, to address the related issue of walking in cable t rays.- These memoranda provided the basic rules to plant personnel for using cable trays as a work platform for personnel and/or equipment.
The inspector reviewed the memoranda issued by the licensee to address the concerns discussed in this item. The memoranda appeared to contain the necessary information to ensure protection of plant piping and cable trays during maintenance activities.
Based on the actions taken by the licensee and the reviews performed by the inspector and NRR, it appeared that the licensee had aderluately addressed this open item, b.
(Closed) Severity Level IV Violatisn 285/8913-05: Inadequate procedure to address the loss of component cooling water (CCW).
This violation involved the licensee's failure to issue a procedure ! In the event CCW is lost, the to adequately) address the loss of CCW. system is used as a cooling mediu raw water (RW normally cooled by the CCd system. The RW system overboard discharge Monitors (y monitored for radioactivity by Radiation is normall RM) P,M-056A and RM-056B.
If the RMs are not operable and the RW system is being used as the cooling medium, the potential exists for an unmonitored radioactive discharge to the environment.
Procedure A0P-11, " Loss of Component Cooling Water," did not provide instructions as to the actions to be taken to ensure that an unmonitored release did not occur.
To address the procedural inadequacy, the licensee issued a revision to Procedure A0P-11. The revision provides instructions for sampling the RW system effluent if RM-056A and RM-056B are inoperabl,
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The inspector reviewed the revision made to A0P-11 and it appeared to be adequate.. The inspector also reviewed other RM installations to verify that the appropriate procedures adequately addressed the actions to be taken if the RM was out of service. fio other problems were identified.
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(Closed) Open Item 285/8913-06: Inclusion of RM-056A in.
Procedure A0P-11.
This item involved actions to be taken by the licensee to include
RM-056A in Procedure AOP-11. The inclusion of RM-056A is necessary to describe the actions to be taken in the event RM-056A is inoperable.- As discussed in paragraph 3.b above, the licensee has appropriately revised Procedure A0P-11 to. include RM-056A.
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(Closed) Open Item 285/8913-07: Revision of Procedure 50-G-66 to include independent verification of the issuance of security badges.
This open item involved actions to be taken by the licensee to proceduralize a requirement to provide an independent check of all documentation prior to the issuance of a security badge. The actions were taken by the licensee after issuance of the wrong type of security badge to an individual.
The licensee revised Procedure 50-G-66, " Control of Key Card Badges," to establish a requirement for performance of an independent check of - each individual'.s documentation prior to issuance of a security badge.. The inspector reviewed the actions taken by the licensee and it appeared that the actions were adequate.
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(Closed) Open Item 285/8917-02: Excessive vibration on 1-inch steam drain lines.
! During a' tour of the plant on April 21, 1989, the inspector noted considerable vibration on the 1-inch piping from each ma'.n steam line . ' below the high-pressure turbine to Valves MOV-CV-2, MOV-CV-4, SPDV-3, and SPDV-4. The concern was brought to the attention of the secondary systems lead engineer. On April 24, 1989, a memorandum was generated from systems engineering to design engineering requesting an analysis of the condition. Additionally, systems engineering evaluated the piping supports currently in place and identified a defective pinned support which laterally supported a connecting main steam line. The steel support exhibited extreme wear from apparent years of vibration. The support was replaced and the vibration was dampened to a small degree; however, excessive vibration was still apparent.
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. a 4- '6 .a The inspector noted that' steam lines in question are not safety
grade 1but.their failure may cause a challenge to safety systems. As ' .of:the close of this inspection period, design engineering had not . completed its analysis of the vibration. The inspector considers ' the ;1ack of analysis' to be' inadequate corrective actions with respect (to an identified. deficiency which could cause a challenge < to nuclear ~ ' . safety systems.
.The piping, which appears to be inadequately supported, is ' categorized as Seismic Class II in the Updated Safety Analysis > ' > Report (USAR).
In accordance with the' USAR, Class II equipment and . components are. required to conform to applicable industry design Jcodes~and standards. No special provisions are made against seismic n , L effects.; At the time of design and construction, the' applicable industry code for such piping was USAS B31.1-1967, " Power Piping."
Section 121.1.4 of USAS B31.1-1967 indicates that power plant piping for. steam-service of 1-inch nominal pipe size should be supported at a maximum span of 9 feet in the absence of a preexisting alternate design analysis verifying acceptability.
In deviation from the above, the piping in question was found to be supported at spans of 12 feet and had no retrievable design analysis verifying its acceptability. This is an apparent deviation.
(285/8926-01)- f.
(Closed) Severity Level IV Violation 285/88201-01(ViolationA): ) I Inadequate procedure-for shift turnover.
This violation involved an inadequate procedure for shift turnover due to-Procedure 50-0-29, " Conduct of Operations." not specifying that Forms FC-9S, " Shift Turnover Log," and FC-95A, " Shutdown Shift i ! Turnover Log," had to be completed.
The licensee revised Procedure S0-0-29 to require that Forms FC-95 and FC-95A, as appropriate, be completed prior to shift turnover.
The procedure also requires that the oncoming operations shift sign the turnover log prior to relieving the onshift crew to verify that the log has been read.
The inspector reviewed Procedure 50-0-29 to verify the adequacy of the procedure changes made by the licensee. The inspectors have also monitored the content of the turnover log for accuracy during . routine visits to the control room.
In all cases during these ! , visits, the inspectors noted that the log appeared to be accurate.
! No problems were noted during review of this item, g.
(Closed) ~ Severity Level IV Violation 285/88201-02 (Violation B.1): Not making log entries for a Technical Specification (TS) limiting condition for operation (LCO).
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-7 ' ,, This,violetion was related to the failure to make a log entry for a TS LC0'when the motor-driven fire pump was out of service. The entry was 'not properly logged due to a personnel error.
' The licen'see has stressed the importance, to all licensed operators, ~ f the procedural requirement to place the appropriate entries into o ,the station log. The operators reviewed Procedure 50-0-24, " Log - Entries," to reestablish the types of entries required.
The inspectors have reviewed the station log, Form FC-95 and Form FC-95A, on a routine basis to verify that the log entries accurately reflect the plant status. During the numerous reviews, no problems were noted.
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(Closed)UnresolvedItem 285/88201-08: Plantreviewcommittee(PRC) review of temporary modifications (TMs) within 14 days.
This. item was related to th failure of the PRC to review the j installation of TMs within 14 days of the completion of installation.
i To address this item, the licensee revised Procedure 50-0-25, ' " Temporary Modification Control," to require that the PRC review all TMs. prior to installation. The procedure did not previously include a requirement for a PRC review of TMs.
In addition to a requirement for the normal review of TMs, Procedure 50-0-25 also states that the shift supervisor may authorize installation of a TM without prior PRC approval in cases of an emergency. The procedure states that PRC approval, in this case, shall be obtained within 48 hours.
The inspector reviewed Procedure S0-0-25 to verify that the procedural ! controls established by the licensee were adequate.
In addition, the inspector reviewed.the TM log to verify compliance with the procedural _ l requirements. No problems were noted.
4.
Licensee Event Report (LER) Followup (92700) Through direct observation, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine
that deportability requirements were fulfilled and immediate corrective action and corrective action to prevent recurrence had been accomplished in accordance with the TS.
The LERs listed below are closed: ! 88-025 Failure to Maintain a Continuous Fire Watch 88-036 Failure to Maintain a Continuous Fire Watch j 89-001 Loss of Shutdown Cooling a 89-006 Failure to a Perform Surveillance Test within Specified Frequency j j , A discussion of-the review performed by the inspectors for each LER is j , provided below: f-l '
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LERs 88-025 and 88-036 reported the failure to maintain a continuous i fire watch. The TS requires a continuous fire watch be established when a. fire barrier is. inoperable and the fire detectors on both ! ' sides of the barrier are inoperable. These LERs involved the emergency diesel generator rooms.
In one case, the rollup door between the diesel rooms was left open and, in the other case, a fire " damper was rendered inoperable due to scaffolding being put through it.
i To prevent future recurrences of this nature, the licensee revised Procedure 50-0-38, " Fire Watch Duties and Turnover Procedures," and Procedure S0-G-58, " Control of Fire Protection Impairments." These j procedure revisions should reduce occurrences of missed fire watches.
j These LERs are considered closed.
) i b.- LER 89-001 reported a-loss of shutdown cooling event which occurred while the plant was in a refueling shutdown on January 8,1989.
Shutdown cooling (SDC) flow was terminated when the SDC primary
1solationvalves.(HCV-347andHCV-348)wentclosed. The valves j received a signal to close when instrumentation and control (I&C) personnel simulated a pressure signal as part of Calibration Procedure CP-SCMM-A, "Subcooled Margin Monitor A Calibration Procedure." Valves HCV-347 and HCV-348 are designed to close when the pressure transmitters sense a pressure greater than 235 psig increasing. This interlock is provided to prevent overpressurization of tte SDC system.
As immediate corrective action, control room operators, who readily recognized the loss of flow, placed the operating SDC pump in a pull-to-lock condition. Control room personnel, aware of the calibration procedure which was being conducted, contacted the l technician in charge of the procedure to determine the extent of his activities. The technician was directed to cease further testing.and remove the pressure signal from the circuit. Once the pressure signal was removed, the operators were able to open HCV-347 and HCV-348. The SDC pump was then restarted and SDC flow was restored.
Based on the review of the official control room log, flow was lost for approximately 5 minutes.
The inspector reviewed the condition of the core and the time interval of the loss of SDC flow at the time of the event and concurs with the licensee's assessment that, under the conditions, the 5-minute loss , of flow was insignificant with respect to potential fuel damage. A conservative calculation indicates the core temperatures would have taken 230 minutes to reach saturation under the actual initial conditions.
The plant review committee convened to review the event and determined the root cause to be a procedural and scheduling i' deficiency.
It was concluded that it was nonconservative to allow this calibration procedure, as well as other calibration _______ -
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procedures (CP) which could result in the activation of the relay to close HCV-347'and HCV-348, to be performed when SDC was required.
The< licensee investigated and found two other cps with the same vulnerability. To prevent reoccurrence of the~ event, changes have been issued to these procedures to prevent performance during times when SDC is required.
The inspector has reviewed the procedure changes and held discussions with the system engineer related to the extent of his investigation.
The procedure changes appeared to remove the potential for having these calibrations performed while SDC is required.
~ Based on the licensee's investigation, the procedure changes made, and the review of the changes and investigation by the inspector, it appeared that the licensee has taken appropriate actions to prevent reoccurrence of the loss of shutdown cooling. This LER is considered closed.
c.
LER 89-006 reported an event where a TS-required surveillance test for the battery-powered smoke detectors located in the control room cabinets was not performed within the specified frequency. The surveillance test was to be performed in accordance with the instructions provided in Procedure ST-FD-2, " Battery Powered Smoke Detectors." This procedure contains two separate parts. One part is st heduled to be performed in June and the other part in December.
By completion of each part, the licensee complies with the TS requirement to perform the test semiannually.
Each procedure states that the actions shall be completed annually.
Due to workload considerations, the licensee opted to utilize the 25 percent grace period allowed for performance of this test.
l Instead of rescheduling the test based on the semiannual requirement, the test was rescheduled based on the annual frequency stated on the test procedure. By using the wrong basis, the surveillance test was performed 10 days late.
To resolve this apparent discrepancy, the licensee revised Procedure ST-FD-2 to state that the frequency of performance is semiannual. This procedure change will ensure that the test is performed at the correct frequency.
The inspector reviewed the actions taken by the licensee and noted no problems. The inspector reviewed a sampling of other similar types of surveillance tests and did not identify other tests that had the same type of problem. This LER is considered closed.
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l l, ., - Based on the reviews performed by the inspectors, as described above, it appears that the licensee took appropriate actions in response to the fidentified events to provide timely corrective actions and implementation < ..of controls to prevent recurrence of the event.
No violations or deviations were identified.
5.. Operational Safety Verification (71707) The inspectors conducted reviews and observations of selected activities to verify that facility operations were performed in conformance with the requirements established under 10 CFR, the licensee's administrative , procedures, and~the TS. The. inspectors made several control room observations to verify the following: Proper.. shift staffing was maintained and conduct of control room
personnel was appropriate.
3-L0perator' adherence to approved procedures and TS requirements was ' + evident.. Operability of reactor protective system, engineered safeguards
equipment, and the safety parameter display system was maintained.
" If not, the appropriate TS LC0 was met.-
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Logs,~ records, recorder traces, annunciators, panel indications, and switch positions complied with the appropriate requirements.
- Proper return to service of ' components was performed.
- Maintenance work orders (MW0) were initiated for equipment in need of
maintenance.' - ' Management personnel toured the control room on a regular basis.
Control room access was. properly controlled.
- Control room annunciator status was reviewed to verify operator
awareness of plant conditions.
Mechanical and electrical temporary modification logs were properly
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Engineered safeguards systems were properly aligned for the specific
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plant condition.
The inspector reviewed the licensee's Safety Analysis for Operability (SAO) 89-009 that deals with inadequate emergency lighting in various areas of the plant.. The lighting is needed in the event a fire, in the control room or cable spreading room, necessitates the evacuation of the control room. The SA0 was based on certain interim . ! _ _ _ _ _. _ _ _ _ _ _ _ _ _. _ _ _ _ _.. _ _ _ _ _
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measures being taken until a modification is completed by December 31, 1989. The modification will provide emergency lighting in those areas that are deficient. These interim measures include an hourly fire watch patrol in the cable spreading room, adding of flashlights and batteries near the alternate shutdown panel, routine maintenance on these flashlights and batteries, and applicable procedure changes.
The inspector aiid NRR reviewed the SA0 and determined the licensee's approach to this concern to be acceptable until permanent emergency lighting is )rovided. Also, the interim measures stated in the SA0 were verified to Je implemented.
On June 19, 1989, the licensee issued LER 89-014 to formally report to the NRC the problems identified with the emergency lighting system. A review of the licensee's actions to address tne emergency lighting problems will be performed during a review of the.LER.
Subsequent to the verification of the interim measures by the inspector, the licensee identified a problem with a commitment made to establish and
maintain an hourly fire patrol in the cable spreading room. During a review of the hourly fire watch logs on June 26, 1989, the fire protection engineer noted that the fire watch had been removed since June 23, 1989.
Upon discovery, the hourly fire watch patrol was immediately reestablished.
The licensee determined that the reason the fire watch was terminated was due to inadequate communications between the operations and security shift ' supervisors. To address this problem, the licensee revised Form FC-1040, " Fire Impairment Log," as an interim measure, to provide instructions to , the operations shif t supervisor when a fire watch may be secured. As a long-term measure, the licensee plans on revising Procedure S0-G-58, " Control of Fire System Impairments," to specify that only the fire protection system engineer will be able to remove a fire watch requirement from the fire watch patrol log. The licensee stated the revision to Procedure 50-G-58 would be issued in the near future.
Normally, the licensee would be cited for a deviation from a commitment made to the NRC for not maintaining fire watch as stated in LER 89-014.
However, in this case, the problem was identified and corrected by the licensee and actions will be taken to prevent recurrence. Therefore, a deviation was not issued. The inspectors will continue to review the status of hourly fire watch patrols in the future to verify compliance with the appropriate regulations.
No violations or deviations were identified.
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Plant Tours (71707) The inspectors conducted plant tours at various times to assess plant and equipment conditions. The following items were observed during the tours: ~, .m__ - _ _
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, m '* [Generalplantconditions,includingoperabilityofstandbyequipment, Lwere satisfactory.
, , . Equipment was.being maintained in proper condition, without fluid
leaks and excessive vibration.
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' Valves and/or switches for safety-related systems'were in the proper-
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e Plant housekeeping and' cleanliness practices were observed, including
' no fire hazards-and the control of combustible material.
~ Performance of work activities was in accordance with approved , . procedures.
- Portable-gas cylinders were properly stored to prevent possible missile hazards.
Tag out of equipment was performed properly.
- *- Management personnel-toured the operating spaces on a regular basis.
During tours of the plant, the inspectors did not identify any .nonconfonning conditions.g.The inspectors did note that the licensee 'is continuing with the plant; appearance upgrade project. The material condition.of the~ plant, including housekeeping, was determined to be E satisfactory.
No violations or de'viations were identified.
~ .7.. Monthly Maintenance Observations (62703) The inspectors observed selected station maintenance activities on safety-related systems and components to verify the maintenance was conducted in accordance with approved procedures, regulatory requirements, " I and the TS. The following items were considered during the observations: . ! The TS LCOs were met while systems or components were removed from
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Approval's were obtained prior to initiating the work.
~ Activities were accomplished using approved MW0s and were inspected,
as applicable.
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returning components or systems to service.
< l' Quality control records were maintained.
- Activities were accomplished by qualified personnel.
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- Parts and materials used were properly certified.
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Radiological and fire prevention controls were implemented.
The inspectors observed the following maintenance activities: Calibration of the breaker for Safety Injection Pump SI-2A (CP-SI-2A)
Troubleshooting of RW Pump AC-10A (MWO 893477)
A discussion of each item is provided below: a.
On June 27, 1989, the inspector observed the performance of the calibration of the breaker for Safety Injection Pump SI-2A. The calibration was performed in accordance with Procedure CP-SI-2A, "SI-2A Breaker." The technician performing the calibration was using a procedure that had not been previously used for performance of this evolution. The procedure had been recently revised by a licensee contractor to upgrade the procedural content. The procedure upgrade wts performed in an effort to provide maintenance personnel with useable procedures. This procedure will be included as part of the comprehensive upgrade of all safety-related procedures being performed by the Nuclear Project Group (Project 1991).
During performance of the calibration, the technician noted that he could not perform the procedure as written. The instructions provided by the procedure were confusing and, at times, impossible to perform. When the technician realized that the procedure was inadequate, he immediately stopped the calibration of the breaker.
The licensee initiated a formal change to Procedure CP-SI-2A to i provide the appropriate instructions. The change was instituted by discarding the revised procedure and reissuing the previous revision of the procedure. The technician used the reissued procedure to successfully complete calibration of the breaker.
The inspector reviewed the reissued procedure to verify that the procedure provided adequate guidance. Based on this review, it appeared that the guidance was adequate.
b.
On June 27, 1989, the inspector observed maintenance personnel remove and replace RW Pump AC-10A with a spare pump. The maintenance activity was performed in accordance with MWO 893477 and , Procedure MP-AC-10. " Removal, Installation, and Testing of the Raw ' Water Pumps," that was attached to the MWO. The MWO was issued to determine the reason for excessive vibration of Pump AC-10A.
Based on the observations made by the inspector, it appeared that I the maintenance personnel performed their duties in a professional j manner. The maintenance personnel followed the procedure, as written, used the appropriate rigging techniques when removing and replacing j i ' l l
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(the pump; and' stopped, maintenance activities to obtain technical ' assistance when concerns were identified that were not specifically , T addressed'by the procedure., .During. removal of Pump' AC-10A, maintenance personnel' noted that one - Ldisc from'the pump discharge check valve (RW-125) had come loose and.
twas' lying'in the pump discharge piping.- Valve RW-125 is.a.
,* double-disc check valve-constructed with a disc attached to each side of a center pin.4.The licensee removed Valve RW-125 from the system, inspected the valve, and determined that no spare parts were available to repair the ' valve.- The licensee reassembled the piping without. reinstalling Valve RW-125.
' SA0 89-010,," Raw Water System Check. Valve RW-125." was generated by - the licensee to address continued plant operation without-Valve RW-125 installed.- The SA0 concluded that no unresolved safety -. question existed; however, actions by the licensee were required to
ensure that Pump AC-10A would be placed in the proper configuration j when the pump was not operating to ensure no system backflow. The
licensee revised the procedures' listed below to provide instructions i to plant personnel on what actions to take: Procedure A0P-18. " Loss of Raw Water"
Procedure 01-RW-1, " Raw Water System-Normal Operation" -
Procedure E0P-20 " Functional Recovery Procedure" l " SAO-89-010 was reviewed by the -resident inspectors, NRR, and the regional office. No problems were noted during the reviews.
In addition, the NRC resident inspectors reviewed the procedure changes made by the licensee and verified that the appropriate operations-l personnel.had received training on the procedure changes. No problems . were identified during these reviews.
J In addition to observation of the performance of maintenance activities.
the inspector also reviewed the completed documentation of the MW0s listed below: stem Channel C wide range nuclear Repair of the reactor protection sy(MWO 893188) j
instrumentation indication lights . Investigation of a possible leaking cell on Battery 2 (MWO 893327) " i ' Repair of the pretrip light on the Channel C Trip Unit 4 indicator
'*: . light (MWO 891797) Repair of,the fuel oil storage tank foot valve (MWO 892287) .
) Troubleshooting'of Emergency Diesel Generator 2 governor (MWO 892680) '* Completed MW0s 891797, 8.92287, 892680, 893188, and 893327 were audited by " the inspector to< ensure compliance with the licensee's recently l , I --- _ _ _ - _ '
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- implemented improve'd Maintenance Procedure S0-M-101, " Conduct of:
' Maintenance." The inspector found instances where-the MWO forms were-o' .not filled out in their entirety. The omissions were insignificant in. . that the~ appropriate information could be found elsewhere in the completed
- package in each case;- The inspector relayed this finding to licensee
. management who appeared to already be aware of the problem. The omissions.
' Twere attributed by the licensee to be unfamiliarity of the craft with the 7: new system. Training'is currently ongoing and supervision is more closely l, monitoring completed documents.. F The inspector notes that the newly implemented ' technical work documents L: Lprovide for more thorough documentation which allows for logical tracking of the maintenance tasks from inception to completion. With the exception of the insignificant omissions, all of the' criteria listed'in the n . . introduction 'of this paragraph for which the documents were examined were met.
140 viblations or deviations were ' identified.
8.
- Monthly -Surveillance Observations (61726)
The inspector' observed selected portions of the performance of TS-required surveillance testing on' safety-related systems and components. The inspector verified the following items during the testing: Testing was performed by qualified personnel using approved
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procedures.- ,
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zTest instrumentation was calibrated.
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The TS LCOs were met.
Removal and restoration of the affected system and/or component were
. accomplished.
. Test results conformed with TS and procedure requirements.
- Test results were reviewed by personnel other than the individual
directing the test.
Deficiencies identified during the testing were properly reviewed and
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resolved by appropriate management personnel.
Test was performed on schedule and complied with the TS required ~ * - frequency.
The inspector observed the following surveillance test activities. The procedures used for the test activities are noted in parenthesis: RW inservice pump test (ST-ISI-RW-3)
Inservice testing of Valves HCV-385 and HCV-386 (ST-ISI-SI-1) '* - _ _ _ _ _ _ _ _ _ _ _ -
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.16 - ! / ! 'A' discussion ~of each surveillance observed is provided below: l . . i > a.- On June 26, 1989, the inspector witnessed selected portions of the l . performance of. Procedure ST-ISI-RW-3,~" Raw Water Inservice Pump j Test.". While attempting to obtain and record the discharge' pressure j from RW' Pump ~ AC-10D, the I&C technician noted that the instrument
line was clogged with sand and an accurate reading could not be l obtained. The NRC inspector noted that a safety-related pressure transmitter-(PT) was plumbed in parallel with the pressure gage. The i PT is designed to provide annunciation in the control room upon a , ' 1ow-pressure condition.- lThe inspeAtor witnessed the I&C technician' disassemble and flush .the instrumentation. tubing..'A root isolation valve was taken back to the' shop.to be flushed more thoroughly. The disassembly and flushing lof'the instrumentation lines 'and components was performed without , ,docume_ntediinstructions.
- Af terwards, the inspector. reviewed the completed test procedure
and found no comments were made indicating the deficient condition or the corrective ~ action taken. The sand in the instrument line to the i pressure transmitter could have resulted in inaccurate control room ' -indication.c , Criterion XVI of Appendix B~ to 10 CFR Part 50 and ANSI 18.7-1976, Section 5.2.11. state, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
. The Fort Calhoun Station Quality Assurance Plan, Section 10.4, ' implements this requirement and states, in part, that deficiencies shall be documented and brought to the attention of the responsible manager and the Manager-Quality Assurance and Quality Control.
Contrary.to the above, on June 26, 1989, I&C technicians were l ebserved to have encountered plugged instrumentation lines affecting a safety-related pressure transmitter, a deficiency, during the performance of Procedure ST-ISI-RW-3, " Raw Water Pump Inservice Test." The technicians were observed to have disassembled, flushed, and reinstalled the tubing without documenting or reporting the deficiency. This is an apparent violation.
(285/8926-02; b.
On June 26,1989,'the inspector witnessed the shift supervisor refuse to perform a scheduled Surveillance Test ST-ISI-SI-1, " Functional Testing of. Valves HCV-385 and HCV-386". The shift supervisor ' indicated to the system engineer that the test would place the plant in a nonconservative condition. The valves are installed in series on the recirculation line, common to all safety-injection pumps, designed to provide minimum flow for an actuated pump operating against a closed system. The test would have required the valves to be shut for 1 hour. Thus, in the event of a safety-injection 1L - __ -
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actuation signal, the high-and low-pressure safety inje;: tion pumps - would start and run without a recirculation path. This action could . cause serious pump damage.
It would take operator action to reopen , Valves HCV-385 and HCV-386.
' j The inspector found the shift supervisor's astuteness in recognizing { lthe adverse operational configuration in which the test would place j ' the plant as noteworthy. The shift supervisor felt it was too much of a risk to depend on operator action to open these valves in the event of an emergency.
The inspector was aware that this was a recently developed test but examined the records to determine if. the test had been previously performed.
It was found that it had been performed for the first time on March 27, 1989. The inspector examined the completed test record to see if any special provisions were made during the performance of the test to account for the nonconservative operational configuration in which the plant had been placed. None were documented.
' The inspecter then interviewed the shift supervisor who authorized the performance of the March 27 test.
It was found that he was well aware of the ramifications of having HCV-385 and HCV-386 closed and, to the best of his recollection, he discussed the abnormal situation ~w ti h his staff prior to commencing the test. Unlike the shift supervisor of June 26, 1989, he was willing to assume the risk of depending.on operator action to open the valves in the event of a > safetyeinjection actuation signal.
In both cases. the shift supervisors were aware of the risks associated with the performance of the test and took the actions they felt ' appropriate to disposition'_the test. The risk associated with ' the test'is~ not highlighted or even mentioned in the procedure. Of - concern here, is the fact that the PRC approved the performance of this test, apparently without recognizing the risk associated with " it. This is another example at the FCS of the support staff issuing l weak procedures and depending on the talent of its operational stef' to perform the task correctly.
The licensee has determined that ASME Code Section XI does not ,1 reouire testing of these valves on a quarterly basis end, therefore, I will change their testing frequency to cold shutdown versus
quarterly. During cold shutdown, the risk associated with this testing will be minimized. The licensee has committed to make a submittal of this inservice inspection (ISI) plan to NRR for review by July 20, 1989, which is when the quarter plus the 25 percent grace period will expire.
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In addition to observation of the performance of surveillance tests, the ) inspector also reviewed the completed documentation of the surveillance { tests listed below. The reviewt were performed to verify the adequacy of the completed tests.
Control element assambly check (ST-CEA-1) . Control room filter circuit operction (ST-CRV-1) j
Station battery charger check (ST-DC-2) i
, Channel. B safety-injection actuation sigral test (ST-ESF-2) ! Pressurizer pressure low-sigr.al channel check (ST-ESF-1)
< i Safety-injection valves quarterly-inservice testin (ST-ISI-SI-1)
Power-range safety channels monthly test (ST-RPS-1
High containment pressure monthly channel check (ST-RPS-8)
No problems were identified during review of the documentation.
9.
Security Observations (71707) The NRC inspectors. verified that the physical security plan was being ! implemented by selected observation of the following items: The security organization was properly manned.
- Personnel within the protecced area (PA) displayed their
identification badges.
, ! Vehicles were properly authorized, searched, and escorted or
controlled within the PA.
Persons and packages were properly ::iesred and checked before entry
into the PA was permitted.
The effectiveness of the security program was maintained when security equipment failure or impairment required cortpensatory measures to be employed.
The PA barrier was maintained and the isolation zone kept free of
transient material.
The vital area barriers were maintained and not compromised by
breaches or weaknesses.
Illumination in the PA was adequate to observe the appropriate areas .at night.
Security monitors at the secondary and central alarm stations were
functioning properly for assessment of possible intrusions.
During this inspection period, two security events occurred. One event was related to the failure to maintain control of safeguards information L
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> +c . These events'will be reviewed by security specialists from the Region IV
office 3 . ,
, , . . JNo'violationsordeviationsiereidentified.
10. Radiologicei 'Prbtection Observations.(71707) ~ L.
The inspectors verified that selected activitief * the liransee's ! sradiological protection" program were implemented in conforece with'the facility-policies and procedures. and in compliance with regulatory . . requirements. The activities. listed below were observed and/or reviewed: HealthTphysics (HP)' supervisory personnel conducted plant tours to ~
. check on activities in progress.
" R .HP' technicians were using calibrated instrumentation.
- Radiation work permits' contained the appropriate information to
. ensure that work was' performed in a safe and controlled manner.
Personnel in radiation controlled areas (RCA) were wearing the
required personnel' monitoring equipment and protective clothing and a ' were properly frisked prior to exiting an RCA.
Radiation and/or contaminated areas were properly posted and
controlled based on the activity levels within the area.
No violations or deviations were identified.
11...In-Office Review of Periodic, Special, and Nonroutii e Event Reports (90712 and 90713) 'In-office review of periodic, special, and nonroutine event reports was .- -performed by the NRC inspectors to verify the following, as appropriate:. . ' Correspondence included the information required by appropriate NRC
requirements.
Test results and supporting information were consistent with design
predictions and specifications.- Planned corrective actions were adequate for resolution of identified ~* ,3 problems.
Whether or not any information contained in the correspondence report
- should be classified as an abnormal emtrrence or additional reactive inspection.
Correspondence did not contain incorrect, inadequate, or incomplete
information.
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- - - , -- - - Jy%, - ' ' @{#*(.[r , .; y,. 20: , The inspectors reviewed the following correspondence: Resolution' of Four HEDs ~ Contained in the' Detailed Cointrol Room Design - L* { ~ Review,: dated Nay. 26,1989 2 Feedwater Valve HCV-1386 Inoperable Due to Maintenance Program
y< Deficiency:(LER89-012),datedJune2,1989 , 1988 Steam Generator Eddy Current Test Report, dated June 9,1989
, , . Serhor.' Reactor Operator' Eligibility Requirements. Published in ' '* ' <n NUREG-1021, dated June 12, 1989 y ,
NRC Backfit Questionnaire,. dated June 9,1989 Inaccurate'. Information on Medical. Form NRC-396, dated June 14, 1989' '* ', Fort Calhoun Statiidn Radiation' Protection Enhancement Program,
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Bimonthly Status Report, dated June 15, 1989 Schedule fer Completion ~of the'10 CFR 50.63 Station Blackout Coping
Assessment. dated June 15,'1989 Special *ROort on.Inoperability of a-Fire Barrier, dated' June 15, ' *' < 1989-c .
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Monihly 0perations Report for May 1980,.. undated D May Monthly Operating Report, dated June 15, 1989 'NRC Bulletin 89-01, Failure of Westinghouse Steam Generator Mechantcal Tube Plugs,' dated June 15, 1989 Questions on Completion Status of Certain SPDS Items, dated June 19
1989 Safety Enhancement Program, Monthly Status Report, dated June 20,
1989-Response'to Questions on OSAR 88-36, "Effect of HPSI Header " . Crossconnect Valves on Hot Leg Injection,"' dated June 16, 1989 L n Re @onse to NRC Questions Resulting from the Operational Safety Team
Inspection, dated June 15,-1989 Missed Te dc Gas Monitor Surveillance Lue to Personnel Errors
l_ (LER89-06),datedJune 16, 1989 . Auxiliary Feedwater Panel Instrumentation Outside Design Basis (LER 89-014), dated June 19, 1989 , ..
.. _ _ _ _ - o ,[ , "
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! Updated Statu. of Various Regulatory items, dated June 16, 1989
No violations or deviatiors were identified.
12. Followup on Onsite Events, (93702) On June 13,1989,-thelicenseeperformedaspecialprocedure(SP-FW-12)to i test the steam-driven auxiliary feedwater pump (FW-10). The results of the' test were to be used to upgrade the surveillance test program for.this pump by developing a means to manually control pump speed so that the pump could be tested at the same speed each month. The test also was intended , - to establish reference values for speed and steam bowl pressures for future trending and to determine the setting of the speed limiter governor.
The initial run of the test found that varying the air signal at the output of the diff erential pressure transmitter had no effect on pump speed. The licensee determined that the air controller was faulty, and
that by removing air to the controller, the pump would go to its upper q limit (7725 rpm et 1210 psig discharge pressure).
l ' The licensee performed an evaluation and determined that no unreviewed safety question existed for re,aoving the air supply to the controller.
The NRC's concern was that, had the pump been needed during a design basis accident, the pump may not have been operable before the air was removed.
An inspector performed a special onsite inspection of this event during the period June 22-23,1o89. The details of this inspection are documented in NRC Inspection Report 50-285/89-27.
i 13. Exit Interview i The Mpectors met with Mr. K. J. Morris / Division Manager, Nuclear Operations) and other members of the licensee staff on July 7,1989. The meeting attendees are listed in paragraph 1 of this inspection report. At this meeting, the inspectors sunrnarized the scope of the inspection and the findings. The licensee did not ident.ify any proprietary information to j the inspectors.
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