IR 05000461/1987036
| ML20236Y387 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 12/04/1987 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236Y367 | List: |
| References | |
| 50-461-87-36, NUDOCS 8712110289 | |
| Download: ML20236Y387 (22) | |
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V. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-461/87036(DRP)
Docket No. 50-461 License No, NPF-62 Licensee:
Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name: Clinton Power Station Inspection At: Clinton Site, Clinton, IL Inspection Conducted: October 14 through November 25, 1987 Inspectors:
P. Hiland S. Ray R. Hasse
R. C. Knop, Chief ] #
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4 Approved By:
Projects Section IB Date Inspection Summary Inspection on October 14 through November 25, 1987 (Report No. 50-461/87036(DRP))
Areas Inspected:
Routine, unannounced rafety ispection by the resident inspectors and region-based inspector of licensee action on previous inspection findings; offsite review committee; onsite fellowup of written reports of nonroutine events at power reactor facilities; operational safety verification; engineered safety feature system walkdown; monthly maintenance observation; monthly surveillance observation; training effectiveness; Startup test witnessing; onsite followup of events at operating reactors; and management meeting.
Resulta Of the 10 areas inspected, one violation was identified in the area of onsite followup of events. This violation is receiving licensee management attention.
In addition, one violation of Technical Specifications was identified in the area of onsite followup of events for which a Notice of Violation was not issued in accordance with 10CFR2, Appendix C, Paragraph V (failure to perform shif tly channel check - Paragraph 11.b.(1)).
8712110289 871207 PDR ADOCK 05000461 G
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DETAILS'
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' Personnel Contacted-Illinois Power Company (IP)
- K. Baker, Supervisor I&E Interface
- G. Bell, Director, Material Management
- R. Campbell, Manager - QA-
. E. Corrigan, Director Quality Engineering'and Verification
- J. Fertic, Director,. Quality Systems.& Audits
- R. Freeman, Manager, NSED
- W. Gerstner, Executive Vice President
- K. Graf, Director - Operations Monitoring. Program
- D. Hall, Vice President, Nuclear
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D. Hillyer, Director - Plant Radiation Protection D. Holesinger, Assistant Plant Manager
- E. Kant, Director - Design Engineering
- A. Mcdonald, Director - Nuclear Program Assessment
- J. Miller, Manager, Scheduling & Outage Management
- J." Perry, Manager - Nuclear Program Coordination F. Spangenberg, Manager - L&S
- J. Weaver, Director - Licensing
- J. Wilson, Manager - CPS
- R. Wyatt, Director - Nuclear Training Department Soyland/WIPCO J. Greenwood, Manager Power Supply Nuclear Regulatory Commission
- P. Hiland,. Senior Resident Inspector, Clinton
- S. Ray,' Resident Inspector, Clinton
- R. Knop, Chief, Section 1B, Region III R. Hasse, OPS, Region III
- Denotes those attending the management meeting on November 24, 1987.
- Denotes those attending the monthly exit meeting on November 25, 1987.
The inspector also contacted and interviewed other licensee and contractor personnel.
2.
Previously Identified Items (92701) (92702)
a.
(Closed) Open. Item (461/87031-02):
Periodic Operability Testing of
.the Entire Scram Discharge Volume System.
This item referred to Temporary Instruction 2515/90 which provided guidance on inspection of the Scram Discharge Volume (SDV) system to ensure compliance with commitments concerning Multiplant Action (MPA) Item B-58.
One criterion of the inspectica required that
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operability of the entire system be demonstrated periodically. The original. inspection noted that Clinton did not perform such a test,
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nor was it required by Clinton Power Station Technical Specifications.
This item remained open pending resolution by N3R.
The inspector discussed this item with NRR (Mr. Kenneth Eccleston)
on October 15,'1987.
Based on that discussion, a review of Clinton's Safety Evaluation Report (SER), and those of other recently licensed BWR-6 plants, the inspector determined that periodic testing of the entire SDV system was not required..In Clinton SER, NUREG-0853, Supplement No. 1. Section'4.6, the staff concluded that the design of'the SDV was acceptable.
This item is closed, b.
(Closed) Violation (461/87006-02(0RS)):
Failure To Prepare E0Ps In Accordance With The P-STG and Control Changes.
Previous licensee actions on this ite.n were reported in Inspection Report No. (461/87030(DRS).
The inspector reviewed licensee actions resulting from'the deficiencies described in that report. These deficiencies. involved differences between the P-STG and the E0Ps.
These deficiencies had been corrected. The inspector also reviewed'
the results of the licensee's " Tabletop" review of the previous revision of E0Ps.4401.01, " Level. Control Emergency" and 4402.01,
" Containment Control - Emergency" against the current revisions of these procedures and the carrent revision of the P-STG. The other E0Ps were' reviewed during the previous inspection. No further discrepancies were found. The " Tabletop" review items examined by the inspector had been incorporated into the current revisions. On this basis, this item is considered closed.
c.
(Closed) Violation (461/87030-05):
Failure to Follow Maintenance Work Instructions Resulting in a Reactor Scram. During the conduct of maintenance / calibration work being performed on the reactor recirculation (recirc) flow control systems, maintenance technicians failed to verify the Hydraulic Power Unit for the recirc valve being worked was locked out as required by the work instruction.
The licensee responded to the Notice of Violation via IP letter U-601057, dated October 3, 1987, in a timely manner. The licensee reported the resultant Reactor Protection System actuation in LER 87-042-00 as documented below in paragraph 4.c.
The licensee's investigation concluded that this violation was not.
the result of program or procedural problems; but was directly attributable to personnel error.
The inspector reviewed the licensee's corrective action for this violation in conjunction with the licensee's corrective action to LER 87-042-00. The inspector verified through review of training records that appropriate site personnel had been trained on lessons learned from LER 87-042-00 as stated in the licensee's response to this violation. Based on completion of corrective actions, this item is closed.
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d.
(0 pen) Open Item (461/87032-02): Material Control Conditions in Containment.
During this inspection period, the licensee implemented Administrative Procedure CPS No. 1050.02, Revision 0, dated October 16,1987, " Foreign Material Exclusion in the Containment and Drywell". The procedure established accountability for material entering the containment and drywell and emphasized the need to limit the amount and types of material brought into those areas and the need to promptly remove materials that are no longer needed.
It also discussed the need to secure material in place that is located in the suppression pool swell volume.
The licensee also took additional actions to inform plant personnel via a Plant Manager's memorandum of the importance of material control in containment and the implementation of the foreign material exclusion procedure. A training video tape was also made available.
The inspector acknowledged the efforts discussed above and noted a general compliance with the procedure and a significant improvement in containment material control conditions.
Since the plant was shutdown during most of this inspection period, with a significant amount of work being performed in the containment and drywell, the routine condition of the containment during ope ation could not be determined.
This item will remain open pending additional observations by the inspector during power operations, e.
(Closed) Open Item (461/85005-04):
Verify Modification of Scram Discharge System to Meet Acceptance Criteria.
This item referred to a licensee commitment in the Safety Evaluation Report (SER), Section 4.6, to modify the Scram Discharge System (SDV) to meet the criteria enumerated in the NRC generic study, "BWR SCRAM Discharge System Safety Evaluation", dated December 1, 1980.
By letter dated December 3, 1981, the licensee provided a point-by point description of the design of the SDV against the criteria of the generic study.
In Supplement 1 to the SER, the NRC staff concluded that the design of the SDV was acceptable.
The inspector reviewed IP memorandom Y-807?6 dated May 13, 1986, which reported that the SDV modifications we a complete and IP memorandom JAM 86-242 dated July 3,1986, which slated that the preoperational tests on the SDV were successfully comp ~eted.
Inspection Report 50-461/87031, Paragraph 4, documented the results of an inspection of the SDV system in which one item was left open.
That item was closed in paragraph 2.a. above.
Based on the results of the inspections previously documented as stated above, this item is closed.
No Violations or Deviations were identified.
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3.
Offsite Review Committee (40701)
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During this report period, the inspector observed the activities of the I
licensee's Nuclear Review and Audit Group (NRAG).
The purpose of this I
inspection was to verify the licensee's compliance with CPS Technical Specification 6.5.2 which detailed the responsibilities and requirements of the offsite review committee (i.e. NRAG).
The inspector reviewed the following NRAG meeting minutes to determine if required reviews and followup action items were performed as required by technical specifications:
NRAG Meeting Minutes Dated 86-02 April 22, 1986 86-03 May 13, 1986 86-04 June 10, 1986 86-05 July 30, 1986 86-06 September 3, 1986 86-07 October 22, 1986 86-08 December 2, 1986 87-01 January 16,-1987 87-02 February 3, 1987 87-03 March 3, 1987 87-04 May 7, 1987 87-06 September 1, 1987 87-07 September 29, 1$87 The inspector attended an NRAG meeting conducted.on October 29, 1987.
The meeting agenda included:
a review of previous NRAG meeting minutes; a review of open action items; a discussion of LERs; subcommittee reports; and a review of proposed changes to the Clinton Technical Specifications.
Based on the review performed on past NRAG meeting minutes and the inspector's observations at the October 29 NRAG meeting, the inspector concluded that.the activities of the Offsite Review Committee were being conducted in accordance with the technical specifications.
No Violations or Deviations were identified.
4.
Onsite Followup Of Written Reports Of Nonroutine Events At Pow Reactor Facilities (92700)
For the LERs listed below, the inspector performed an onsite followup inspection of the LERs to determine whether response to the events were adequate and met regulatory requirements, license conditions, and commitments and to determine whether the licensee had taken corrective actions as stated in the LERs.
a.
(Closed) LER No. 86-006-01 (461/86006-L1): Automatic Initiation of Essential Service Water Due to Transient Pressure Drop In Nonessential Service Water.
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The original version of this LER was closed in Inspection Report 50-461/87002, paragraph 5.a.(1).
At the time of that review, the
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inspector also reviewed the revision but inadvertently did not list it in the inspection report. This LER is closed.
b.
(Closed) LER No. 86-025-01(461/86025-L1): Automatic Actuation of Shutdown Service Water Pump "B" and Other Various Division II Equipment Due to Utility Electrical Technician Error.
The original version of this LER was closed in Inspection Report 50-461/87013, paragraph 3.
The narrative of the inspector's comments indicate that the inspector also considered the additional information in the revised LER and it also should have been closed at that time. This LER is closed, c.
(Closed) LER 87-042-00 (461/87042-LL): Automatic Actuation of the Reactor Protection System Due to Utility Personnel Error. While performing calibration on the reactor recirculation flow control valves, maintenance technicians failed to inform the control room operators of steps being performed contrary to the requirements of the procedure in use.
This event was previously documented in Inspection Report 50-461/87030, paragraph 11.b.(4).
A Notice of Violation was issued (87030-05) for which the licensee's corrective action are documented above in paragraph 2.c. of this report.
The inspector reviewed training records to verify that corrective actions as stated in the LER were completed.
Based on the inspector's review, corrective actions as stated in this LER were completed and this item is closed.
a.
(Closed) LER 87-044-00 (461/87044-LL):
Violation of Plant's Technical Specifications Due to Utility Personnel Error Resulting from a Deficient Surveillance Procedure.
This event was originally reported as discussed in Inspection Report 50-461/87031, paragraph 10.c.(13)
At the time that inspection was conducted it was thought that shiftly channel checks on the Average Power Range Monitor (APRM) Neutron Flux High channel were being done
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only daily instead of shiftly as required by Technical Specification Table 4.3.1.1-1, Iter
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This was due to a typographical error made while preparing.evision 22 to CPS No. 9000.01D001, Control Room Operator Surve.ilance Log - Mode 1, 2, 3 Data Sheet.
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Subsequent licensee investigation into this event determined that another step in CPS No. 9000.010001 met the requirements of Technical Specification Table 4.3.1.1-1, Item 2.a for Modes 1, 2, and 3, and that this step was performed shif tly as required.
However, it was determined that the channel check being performed in Modes 4 and 5 in accordance with CPS No. 9000.01D002, Control Room Operator Surveillance Log - Mode 4/5 Data Sheet, was not sufficient to meet the definition of a channel check in technical specifications.
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The inspector verified by review c ? plant procedures and other
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records, that the corrective actions of LER 87-044-00 had been
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critique were also verified complete with the exception of a QA
Department 100% verification of the Master Surveillance List against j
technical specifications. That verification is expected to take approximately three years to complete and is being tracked by the licensee's Commitment Tracking System under CCT No. 046857.
This event was identified as a violation (461/87031-13) which met the criteria of 10CFR2, Appendix C, paragraph V, for which a Notice of Violation was not issued.
The licensee was asked to respond generically to several identified violations of technical specifications in Inspection Report 50-461/87031. That response was provided by letter U-601040 on September 21, 1987, and will be
evaluated separately. This LER is closed.
l e.
(Closed) LER 87-050-00 (461/87050-LL): Automatic Actuation of the Reactor Protection System Due to Utility Non-Licensed Operator Error
Resulting From a Deficient Procedure. Deficiencies in the operating l
procedure for shif ting the Steam Jet Air Ejectors (SJAE) resulted in a loss of condenser vacuum and turbine trip. A reactor scram
occurred, as designed, on receipt of the main turbine trip.
The licensee's corrective action for this event included revising Operating Procedure CPS No. 3215.01, "Off Gas (OG)" and l
recalibrating condenser vacuum instrumentations.
In addition, the i
calibration frequency for the condenser vacuum instruments was increased from an eighteen to a twelve month frequency.
l The inspector reviewed revision 8 of CPS No. 3215.01, dated September 22, 1987, and noted that the changes stated in the licensee's corrective action to this LER had been incorporated.
The inspector reviewed completed maintenance work request (MWR) C-51004 which documented the calibrations performed on the condenser vacuum instruments.
The inspector noted that the periodicity of required instrument calibrations for the condenser vacuum instruments was l
revised to a twelve month frequency.
Based on the inspector's
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confirmation that corrective action was completed as stated in the LER, this item is closed.
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(Closed) LER 87-051-00 (461/87051-LL):
Violation of Plant's Technical Specifications Due to Utility Personnel Error Resulting
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from a Procedural Deficiency.
This event was initially reviewed as documented in Inspection Report 50-461/87030, paragraph 11.b.(11). At the time of that inspection, this item was left unresolved (461/87030-09) pending further investigation by the licensee.
In Inspection Report 50-461/87032 the unresolved item was changed to one of six examples of a violation (461/87032-01A).
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During this report period, the inspector verified through review of CPS No. 3002.01, revision 9, Heatup and Pressurization, and IP
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letter U-601048, dated October 30, 1987, Proposed Amendment to Facility Operating License NPF-62, as well as other records, that the corrective actions as described in LER 87-051-00 had been completed. Additional generic corrective actions resulting from the 11 cense 3's response to violation 461/87032-01A will be evaluated separately.
This item is closed.
g.
(Closed) LER 87-052-00 (461/87052-LL):
Engineered Safety Feature (ESF) actuation due to Reactor Core Isolation Cooling (RCIC) Inboard Isolation Valve Closure.
During troubleshooting of the Division II RCIC steam flow indication, maintenance technicians caused an automatic isolation signal to be generated by throttling a transmitter isolation valve.
The licensee determined the root cause of this event was personnel error due to the maintenance technicians failure to recognize the high potential for generating an isolation signal while performing the troubleshooting activity. The inspector reviewed training records to verify corrective action as stated in this LER were completed.
In addition, the inspector reviewed Nuclear Station Engineering Department (NSED) Procedure A-16, "NSED Action Plan",
revision 0, dated November 5, 1987. That review indicated that the licensee had developed a controlled process for activities such as troubleshooting.
Based on the inspector's confirmation that corrective actions as stated in this LER were completed, this item is closed.
h.
(Closed) LER 87-057-00 (461/87057-LL): Missed 8 Hour Verification of Offsite Power Breaker Lineups During Diesel Generator Inoperability Due to Line Assistant Shift Supervisor Oversight.
This LER was initially reviewed as documented in Inspection Report 50-461/87032, paragraph 10.b.(10). At the time of that inspection, this LER remained open pending the inspector's verification of corrective actions.
During this report period, the inspector verified through review of training and other records that operations department personnel had been provided awareness training on this LER.
In addition, this event was one of six examples of a violation (461/87032-01E) for which the licensee was required to respond describing any additional generic corrective actions.
Based on the above verifications that corrective actions as described in LER 87-057-00 had been completed and the tracking of any additional corrective actions in the inspector's followup to the violation, this item is closed.
1.
(Clored) LER 87-058-00 (461/87058-LL):
Missed Containment Airlock Door Test Due to Line Assistant Shift Superv'sor Misinterpretation of the Technical Specifications.
This event was initially reviewed as documented in Inspection Report 50-461/87032, paragraph 10.b.(11).
During this report period, the
inspector verified through review of training and other records that
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k the lessons learned from this LER had been incorporated into the operator requalification training program.
In addition, the
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Licensing and Safety Department provided the Operations Department with a formal interpretation of the technical specification pertaining to the containment airlocks which explains the actions l.
required to maintain one door OPERABLE while the other is l
In addition, this event was one of six examples of a violation
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(461/87032-01F) for which the licensee was required to respond I
describing any additional generic corrective actions.
Based on the above verifications that corrective actions as described in LER l
87-058-00 had been completed and the tracking of any additional l
corrective actions in the inspector's followup to the violation, this LER is closed.
No violations or deviations were identified.
5.
Operational Safety Verification (71707)
The inspector observed control room operations, attended selected pre-shift briefings, reviewed applicable logs, and conducted discussions with control room operators during the inspection period.
The inspector verified the operability of selected emergency systems and verified tracking of LCOs.
Routine tours of the auxiliary, fuel, containment, control, diesel generator, turbine buildings and the screenhouse were conducted to observe plant equipment conditions including potential for fire hazards, fluid leaks, and operating conditions (i.e., vibration, process parameters, operating temperatures, etc). The inspector verified that maintenance requests had been initiated for discrepant conditions observed. The inspector verified by direct observation and discussion with plant personnel that security procedures and radiation protection (RP) controls were being properly implemented.
Inspections were routinely performed to ensure that the licensee conducts activities at the facility safely and in conformance with regulatory requirements.
The inspections focused on the implementation and overall effectiveness of licensee's control of operating activities, and the performance of licensed and nonlicensed operators and shift technical advisors.
The following items were considered during these inspections:
Adequacy of plant staffing and supervision.
Control room professionalism including procedure adherence, operator attentiveness and response to alarms, events, and off normal conditions.
Operability of selected safety-related systems including attendant alarms, instrumentation, and controls.
Maintenance of quality records and reports.
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The licensee completed their power ascension test program 'during this report period.
Following the last major test, Generator Load Rejection (see paragraph 10. below), the licensee entered a planned maintenance outage for the primary purpose of conducting scheduled surveillance.
The maintenance outage was completed late in the report period and the reactor was brought critical on November 22,-1987.
No violations or deviations were identified.
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Engineered Safety Feature System Walkdown (71710)
The inspector performed a complete walkdown of the Division I Low Pressure Coolant Injection (RHR) system during the report period to verify.the system status. At the time the walkdown was performed, the licensee had identified Division I RHR system as an operable Emergency Core Cooling system meeting all the requirements of the plant's technical
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specification.
For the purpose of this walkdown, the inspector utilized the following system drawings and checklists contained in the system operating and surveillance procedures.
CPS No. 3312.01V001, revision 5, RHR Valve Lineup
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CPS No. 3312.01V002, revision 3, RHR Instrument Valve Lineup CPS No. 3312.01E001, revision 5, RHR Electrical Lineup
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CPS No. 9961.050001, revision 23, Containment and Drywell Test
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Connection, Vent and Drain Valves - Monthly CPS No. 9961.050002, revision 23, Containment and Drywell Test
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Connection, Vent and Drain Valves - Quarterly P&ID M05-1075, sheet 1, revision AA
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.P&ID M05-1075, sheet'4, revision S
For the inspection performed, the following. attributes were observed:
Hangers and supports were made up properly and aligned
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correctly.
Housekeeping and cleanliness were adequate.
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Valvas in the system were installed correctly and did not
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exhibit leakage or mechanical problems.
(One pipecap was not installed and is discussed below).
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the vicinity.
Major system components were properly labeled, lubricated and
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cooled, and no leakage existed.
l Interiors of electrical and instrumentation cabinets were free
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of debris, loose material, uncontrolled jumpers, with.no evidence of rodents.
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Instrumentation was installed, functioning, and exhibited normal expected values.
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Instrument calibration dates were current.
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Support systems essential to system actuation or performance were operational.
Valves were in their proper positions and locked where required
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and system lineups matched plant drawings.
(Exceptions are discussed below).
a.
During the walkdown, the following discrepancies were noted with the valve lineup procedures:
(1) CPS No. 3312.01V001, RHR Valve Lineup, listed the desired position of most containment isolation vent, drain, and tes'
valves as " closed". CPS No. 9061.050001 and 9061~.05D002 liss these valves as required to be " locked closed". All the valves appeared to be locked closed.
The valves noted were:
1E12-F331A, F229A, F107A, F418, F419, F414, F415, F334A, F335A, F420, F421, F365A, F366A, F058A, F349A, F432A, F433A, and F056A.
This same discrepancy had been noted on previous inspections and was being tracked as Open Item 461/87030-02.
(2) 1E12-F042A, LPCI From RhR-A Shutoff Valve, had a "LLRT In Progress" tag on it but the LLRT was not in progress.
(3)
1E12-F059A was locked closed but was listed only as " closed" on both CPS No. 3312.01V001 and CPS No. 9061.05D001.
(4) Valve IE12-F102 was listed as being on the 702' elevation of the RHR HX A Room on CPS No. 3312.01V001, but was actually located on the 737' elevation.
(5) The pipe cap was missing downstream of valves 1E12-F347A and 1E12-F348A.
(6) CPS No. 3312.01E001, RHR Electrical Lineup required that the breaker on MCC 1A2 for valve 1E12-F052A be " locked off".
Actually, the breaker was danger tagged "off" but not locked.
The corresponding breaker in the B RHR system was danger tagged
" locked off".
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For items (1) thrtugh (6) discussed above, the inspector noted that those discrepancies did not impact the operability of the Low Pressure Coolant Injection system.
b.
In conjunction with the above, the inspector reviewed the results of current surveillance performed on the A LPCI system to verify technical specification requirements were met. The following surveillance test results were reviewed.
Surveillance No.
Title Frequency Test Date CPS No. 9053.01 Low Pressure Coolant Monthly Nov. 12, 1987 Injection (LPCI) A/B Operability Checks CPS No. 9053.04 Residual Heat Removal Quarterly Oct. 20/21 (RHR) A/B/C Valve 1987 Valve Operability Checks CPS No. 9053.05 RHR/LPCS Valve Cold Oct. 22, 1987 Operability Shutdown (Shutdown)
CPS No. 9053.07 RHR Pumps A,B,C Quarterly Aug. 29, 1987 Operability Test The inspector concluded that the A LPCI system was operable based on direct field observations of the above lineups and inspection attributes.
In addition, the inspector's review of current
,e surveillance tests for the A LPCI system indicated that the plant's technical specifications were being met.
No violations or deviations were identified.
7.
Monthly Maintenance _Ob crvation (62703)
Selected portions of the plant maintenance activities on safety-related systems and components were observed or reviewed to ascertain that the activities were performed in accordance with approved procedures, regulatory guides, industry codes and standards, and that the performance of the activities conformed to the Technical Specifications.
The inspection included activities associated with preventive or corrective maintena1ce of electrical, instrumentation and control, mechanical equipment, and systems.
The following items were considered during these inspections:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibration was performed prior to returning the components or systems to service; parts and materials that were used were prnperly certified; and maintenance of appropriate fire prevention, radiological, and housekeeping conditions.
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The inspector observed / reviewed the following work activities:
Maintenance Work Request No.
Activity C-38147/C-48984 MSIV Repair (CPS No. 8216.01)
C-39340 Butt Splice Inspection / Repair C-39341 Butt Splice Inspection / Repair No violations or deviations were identified.
8.
Monthly Surveillance Observation (61726)
An inspection of inservice and testing activities was performed to ascertain that the activities were accomplished in accordance with applicable regulatory guides, industry codes and standards, and in conformance with regulatory requirements.
Items which were considered during the inspection included whether adequate procedures were used to perform the testing, test instrumentation was calibrated, test results conformed with technical specifications and procedural requirements, and that tests were performed within the required time limits.
The inspector determined that the test results were reviewed by someone other than the personnel involved with the performance of the test, and that any deficiencies identified during the testing were reviewed and resolved by appropriate management personnel.
The inspector observed / reviewed the following activities.
Surveillance / Test Procedure No.
Activity CPS No. 9861.02, Revision 28 MSIV Leak Rate Tests (Appendix J)
l CPS No. 9861.02, Revision 28 Equipment Hatch Leak Rate Test
(Appendix G5)
l CPS No. 9861.02, Revision 28 Upper Air Lock Seal Leak Rate
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(Appendix G2)
Test
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l CPS No. 9053.05 RHR Operability Check l
i CPS No. 9080.03, Revision 21 Divisior. II '.oss of Power I
(TPD 87-1635)
No violations or deviations were identified.
9.
Training and Qualification Effectiveness (41400 & 41701)
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The effectiveness of training programs for licensed and nonlicensed personnel were reviewed by the inspector during the witnessing of the licensee's performance of routine surveillance, maintenance, and operational activities and during the review of the licensee's response j
to events which occurred during the months of October / November 1987.
Personnel appeared to be knowledgeable of the tasks being performed.
l No violations or deviations were identified.
10.
Startup Test Witnessing and Observation (72302)
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During the report period, the inspector witnessed the performance of
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Startup Test Procedure (STP)-27-6, " Generator Load Rejection". This was the last ma.jor test in the power ascension program.
The inspector determined by direct observation that licensee operating and test personnel were knowledgeable in their individual roles and
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responsibilities. Adequate communications were established and (1 maintained throughout the tests.
Prior to, during, and subsequent to the subject tests the inspector verified the following:
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Crew requirements were being met as defined in plant procedures, and staffing satisfied requirements of technical specification regarding licensed operators.
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The proper versions of the test procedures were in use and were being followed. All referenced procedures had been reviewed and approved.
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Each of the prerequisites had been satisfied.
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Changes or revisions to the test procedures were properly reviewed and approved.
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Data sheet entries were legible and recorded in permanent ink.
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Review of the test results will be conducted during a future inspection.
No violations or deviations were identified.
11. Onsite Followup of Events at Operating Reactors (93702)
a.
General The inspector performed onsite followup activities for events which occurred during the inspection period.
Followup inspection included one or more of the following:
reviews of operating logs, procedures, condition reports; direct observation of licensee actier,s; and interviews of licensee personnel.
For each event, the
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inspector reviewed one or more of the following:
the sequence of
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actions; the functioning of safety systems required by plant
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conditions; licensee actions to verify consistency with plant
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procedures and license conditions; and attempted to verify the nature of the event. Additionally, in some cases, the inspector verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel errors and were taking or had taken appropriate corrective actions. Details of the events and licensee corrective actions noted during the inspector's followup are provided in paragraph b. below.
b.
Details (1) Instrument Channel Check Missed Due To Incorrect Temporary Procedure Change [ ENS No. N/A]
On October 14, 1987, the licensee discovered that a Procedure Deviation for Revision (PDR) issued on October 13, 1987, for CPS No. 9001.010001, CR0 Surveillance Log Mode 1, 2, and 3 Data Sheet had incorrectly deleted a shiftly channel check of containment pressure instruments required by Technical Specification 4.3.9.1-1.1.6.
As a result, the shiftly channel check was not performed on second shift (1600-2400) on October 13, 1987, or on third shift (0000-0800) on October 14, 1987.
The cause of the incorrect PDR was that an operator reviewing a revision to CPS No. 9001.010002, CR0 Surveillance Log Mode 4/5 Data Sheet, in which a channel check of containment pressure instruments was being deleted, incorrectly assumed that the channel check of containment pressure instruments should also be deleted from the mode 1, 2, and 3 shif tly surveillance. The mode 1, 2, and 3 requirement referred to different instruments and a different technical specification.
This fact was not discovered during the independent technical review or during the management review / approval of the PDR.
During the investigation of this event, the licensee also discovered that other aspects of the PDR did not meet the requirements of CPS No. 1005.07, " Temporary Changes to Station Procedures and Documents".
This temporary change method only permits changes to correct procedure deficiencies that are technical or safety in nature and which could adversely impact plant safety or availability during performance of the procedure.
Failure to perform the shiftly channel check on containment pressure instruments required by Technical Specification 4.3.9.1-1.1.6 is a licensee identified violation (461/87036-01)
which meets the criteria of 10CFR2, Appendix C, Paragraph V; consequently, no Notice of Violation will be issued, and this j
matter is considered closed.
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The licensee reported this event in LER 87-061-00 submitted on November 9, 1987.
The LER will be reviewed separately.
(2) Main Steam Lines Fail Their Local Leak Rate Tests [ ENS No. 10466]
On October 26, 1987, the licensee identified that three of the four Main Steam Lines had failed the type "C" Local Leak Rate Tests (LLRT) on the Main Steam Line Valves (MSIV).
Repairs were made to the affected MSIVs and the LLRTs were reperformed.
All MSIVs with the exception of the "C" line inboard MISV passed the subsequent LLRTs. On November 11, 1987, the "C" MSIVs were tested a third time and passed.
The licensee reported this event in LER 87-062-00 submitted on November 24, 1987.
The LER will be reviewed separately.
(3) Partial ESF Actuation [ ENS No. 10529]
On October 31, 1987, the licensee experienced an unexpected ESF actuation during the performance of surveillance testing.
Three valves in the shutdewn cooling system (SX) actuated when an instrument card was inserted following testing.
Surveillance of the " untested i:10.1d;",t:s 17. M z e esa on one Division II containment spray logic card.
Following testing, the logic card was was inserted causing a momentary voltage of sufficient duration to actuate valves 1E]2-F0148, 1E12-F0688, and ISX0828.
However, the voltage surge was not of sufficient duration to seal-in.
At the time of event occurrence, the reactor plant was in cold shutdown and Division II was tagged out for routine maintenance and surveillance testing. After reviewing the event and in light of the fact that the system was out of service at the time of the actuation, the licensee determined that the event was not reportable.
Based on NUREG-1022, Supplement No. 1, Question 6.9, the inspector determined that the licensee's conclusion was reasonable.
(4) ESF Actuation of Division III Emergency Diesel Generator [ ENS No. 10546]
On November 2,1987, the licensee experienced an unexpected ESF actuation when the Division III Emergency Diesel auto started.
While performing a routine surveillance on the second level undervoltage relay for a degraded grid, the Division III Emergency Diesel auto started when the Division III 4160 volt bus was deenergized. The maintenance technician performing the surveillance incorrectly removed the second level undervoltage relay which caused the normal supply breaker from the RAT to i
open and prevented closure of the reserve power supply breaker from the ERAT.
Upon sensing the loss of voltage to the 4160 volt bus, the Division III Diesel auto started and closed its output breaker onto the bus as designed.
At the time of event f
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occurre'nce, the reactor plant was'in cold shutdown and Division III (High Pressure Core Spray system) was' removed from
service for routine maintenance and surveillance testing.
The licensee reported this event in LER 87-064-00. submitted on'
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November 24, 1987.
The LER will be reviewed separately.
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(5) Reactor Protection System Trip Due to Loss of Instrument Air
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[ ENS No. 10576]
On November-4, 1987, the licensee experienced an unexpected Reactor Protection System (RPS) trip during the, performance of a routine surveillance.
The RPS trip signal resulted from a high water level trip in the scram discharge instrument volume. While conducting a normal'to alternate circuit transfer on the Division II 4160 volt vital bus 1B1, power to Division II instrument air containment isolation valves IA006 and IA007 was interrupted causing these air operated valves to close.
Subsequent bleed down of the instrument air header within the containment caused the scram discharge volume (SDV) vent and drain valves to close and the scram valves to onen. The RPS trip signal was generated on high SDV water level. At the time of event j
occurrence, the reactor plant was in cold shutdown with all control rods inserted; therefore, no rod movement occurred.
The apparent cause of this event was the failure of the licensee to recognize the potential for automatic isolation of the instrument air header during the performance of the surveillance test.
A contributing cause for the event was failure on the part of the licensed Control Room operator to notice the annunciator for low containment instrument air pressure which was received several minutes before the scram valves opened.
(6) Environmental Qualification Degradation [ ENS No. 10625]
On November 9, 1987, the licensee identified deficiencies in various junction boxes that had the potential to affect the Environmental Qualification (EQ). The licensee discovered that weep holes were not present in junction boxes located in'the auxiliary, fuel, and containment buildings.
The junction boxes enclosed marathon terminal blocks that could be subject to corrosion caused by a buildup of moisture.
The presence of a weep hole prevents moisture buildup. At the time of discovery, the plant was in cold shutdown with a planned maintenance outage in progress.
The licensee completed an inspection and l
made repairs to all junction boxes prior to commencing a plant
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startup.
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l As'previously documented in Inspection Report 50-461/87026, l
paragraph 6.c.(2), a violation was issued.(87026-03b)
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time of that inspection, this was considered to De an isolated-
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occurrence. The inspector requested the licensee to include.in their response to violation 461/87026-03b the results'and corrective actions for the above event. The licensee stated j
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be included in their response'to. violation 461/87026-03b.
(7) Loss of Emergency Assessment Capability. [ ENS No. 10650]
Cn November 11, 1987, the licensee identified that their site release (SR)' computer was incapable of performing emergency -
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At the time of event occurrence, the. plant was in cold shutdown I
with planned maintenance outage in progress. With the loss of
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the SR computer, the licensee performs manual dose calculations j
in accordance with approved precedures.
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An investigation by the licensee determined that the problem
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with the emergency dose rate function was a software problem
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when data files wore fillod with test F Oy?rd st hts ihich was outside the detector's normal acceptable range.
The computer files-were purged and rebooted and a standing order
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was developed to instruct the operators how to overcome the software problem. The SR computer was declared operable on
' November 12., 1987. No stack releases occurred while the emergency dose rate function was inoperable.
The licencee made the one hour ENS notification on November 11, 1987, because the cause and extent of the EDC failure.was.
unknown and the ability to immediately assess dose for the public was affected. After the nature of the problem became
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known, the licensee determined that the event was not a major
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loss of emergency assessment capability and was therefore not j
reportable.
The inspector determined that the licensee's
conclusion was reasonable.
(8) _ Isolated Reactor Pressure Instrument [ ENS No. N/A]
On November 22, 1987, at about 11:50 p.m., the licensee noted that the recorder for Division I Reactor Vessel Pressure i
Instrument 2B21-N078A was not tracking up from 0 psig as the
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plant was heating up.
The operators had instrument technicians
. investigate the cause and they found that the instrument root valve, 1821-N078AR, was closed.
The operators immediately entered the appropriate LCOs for the inoperable channel and took actions to return the channel to operable status.
By about 2:00 a.m. on November 23, 1987, the instrument lineup was performed, and the instrument was calibrated and returned to an operable condition.
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affected the following technical specification requirements:
Technical Specification 3.3.1, Table 3.3.1-1, Reactor Protection System Instrumentation, Item 3, Reactor Vessel Steam Dome Pressure - High, Modes 1, 2 Technical Specification 3.3.2, Table 3.3.2-1, Containment and Reactor Vessel Isolation Control System Instrumentation, Item 5.e, Reactor Vessel (RHR Cut-in Permissive) Pressure - High Modes 1, 2, 3 Technical Specification 3.3.3, Table 3.3.3-1, Emergency.
Core Cooling System Actuation Instrumentation, Item A.1.c, Reactor Vessel Pressure - Low (LPCI and LPCS Injection Valve Permissive), Modes 1, 2, 3, 4, 5 Technical Specification 3.5.1.a, ECCS Division I, Modes 1,
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2, 3 Technical Specification 3.3.7.5, Table 3.3.7.5-1, Accident Monitoring Instrumentation, Item 1, Reactor Vessel Pre:sure, Modes 1, 2, 3 Technical Specification 3.3.10, Self Test System (STS),
Modes 1, 2, 3, 4, 5 During the time the instrument was isolated, the licensee entered Mode 2 with ECCS Division I inoperable. They also operated in Mode 2 for greater than one hour without the Containment and Reactor Vessel Isolation Control System
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Instrumentation channel being placed in the tripped condition.
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The cause for the valve being closed was determined to be an
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error on the part of instrument technicians who performed a response time surveillance on the instrument on November 18, 1987.
The valve was not required to be closed by the surveillance procedure, but the technicians closed the valve without authority or documentation, to provide double valve isolation while they disassembled fittings at the Instrument.
The technicians then forgot to reopen the root valve.
In addition, the critique of the event disclosed a generic lack of understanding on the part of instrument technicians of the difference between " double verification" and " independent verification" when restoring valve lineups.
Entering Mode 2 (applicable operational condition) with Division 1 of ECCS inoperable due to an inoperable Reactor Vessel Pressure - Low (LPCI and LPCS Injection Valve Permissive) channel is a violation of CPS Technical Specification 3.0.4 (50-461/87036-02(DRP)).
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(9) ESF Actuation of Reactor Core Isolation Cooling On November 24, 1987, the Reactor Core Isolation Cooling system (RCIC) isolated during the performance of a routine surveillance on the Reactor Water Cleanup system (RT).
The surveillance in progress on the RT system required the removal of leads for installation of a simulated signal.
During the lead removal process, a bundle of wires had to be moved aside to allow access to the terminals. When the bundle stas moved, one of the wires broke.
The broken lead was from a thermocouple in the RCIC equipment area ventilation delta temperature circuit and caused an isolation of the RCIC steam supply outboard containment isolation valve. At the time of the isolation, the plant was operating in mode 1 at bout 30%
power.
During the licensee's investigation of the event, it was determined that the wire bundle had to be moved aside during the performance of surveillance in the panel about 120 times per month.
The licensee was evaluating modifications to provide generic corrective actions to prevent recurrence of this type cf problem. While the RCIC system was isolated, the licensee performed valve maintenance and returned the system to operation on November 25, 1987.
One violation was identified. One violation was identified for which a Notice of Violation was not issued in accordance with 30CFR2, Appendix C, Varagraph V.
12. Special/ Management Meetings (30702)
On November 24, 1987, NRC management met with IP management at the Clinton Power Station to discuss the status of the facility, the licensee's Monthly Performance Monitoring Management Report and actions being taken to enhance the licensee's performance in several areas.
Key personnel attending the meeting are identified by (#) in paragraph 1. of this report.
The licensee discussed the activities performed during the recently completed maintenance outage. That discussion detailed their performance on maintenance work requests, plant modifications, surveillance, and preventive maintenance.
In addition, the licensee discussed actions being taken to reduce the number of Licensee Event Reports (LERs).
The licensee discussed the results of their assessment of maintenance and radiation protection activities during the outage. Areas identified by the licensee as needing improvement were discussed with regard to the recommended actions to improve performance.
Region III management acknowledged the licensee's status and plans.
The meeting concluded with a tentative agreement to meet again in January 1987, at the Clinton Power Station with a similar agenda.
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No violations or deviations were identified.
13.
Open Items
Open items are matters which have been discussed with the licensee, which
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will be reviewed further by the inspector, and which will involve some j
action on the part of the NRC or licensee or both.
One open item
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disclosed during the inspection was discussed below in paragraph 15.
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Violations For Which A " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requirement.
However,
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because the NRC wants to encourage and support licensee's initiatives j
for self-identification and correction of problems, the NRC will not
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generally issue a Notice of Violation for a violation that meets the testes of 10 CFR 2, Abendix C,Section V.A.
These tests are:
(1) the violation was identif ed by the licensee; (2) the violation would be
categorized as Severity Level IV or.V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and (5) it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
Violations of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued are discussed in Paragraph 11b.1.
15.
Exit Meetings (30703)
The inspector met with licensee representatives (denoted in paragraph 1)
throughout the inspection and at the conclusion of the inspection on November 25, 1987.
The inspector summarized the scope and findings of the inspection activities.
The licensee acknowledged the inspection findings.
The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any documents / processes as proprietary.
The inspector attended exit meetings held between Region III based inspectors and the licensee as follows:
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Inspector Date
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H. A. Walker 10/16/87 D. E. Hills 11/06/87 H. A. Walker 11/20/87 R. A. Paul 11/20/87 W. J. Slawinski 11/20/87 During the November 6, 1987, exit meeting conducted by regional specialist inspector Mr. D. E. Hills, a generic comment was made regarding the licensee's plant specific emergency procedure guidelines CPS No. 1450.00, revision 3, dated October 12, 1987.
In those staff apprcved guidelines, the use of the verb "should" appeared ir Entry Conditions (e 9. section 3.2, 5.0).
The inspector commented that the verb "Shall" might be more appropriate.
The use of the verbage should/shall in the plant specific emergency procedure guidelines is considered an Open Item (461/87036-03) pending further review by a Region III specialist inspector.
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