IR 05000461/1986072
| ML20212E487 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 12/24/1986 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20212E448 | List: |
| References | |
| TASK-B-19, TASK-OR 50-461-86-72, GL-85-13, GL-85-22, GL-86-02, GL-86-2, IEB-79-18, IEB-86-001, IEB-86-003, IEB-86-1, IEB-86-3, IEIN-86-64, NUDOCS 8701050336 | |
| Download: ML20212E487 (25) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-461/86072(DRP)
Docket No. 50-461 License No. NPF-55 Licensee: Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name: Clinton Power Station Inspection At: Clinton Site, Clinton, IL Inspection Conducted: November 4 through December 15, 1986 Inspectors: T. P. Gwynn P. L. Hiland M. Schumacher RFidA M Approved By: R. C. Knop, Chief 12/rs9r4/
Projects Section IB Da'te '
Inspection Summary Inspection on November 4 through December 15, 1986 (Report No. 50-461/86072(DRP))
Areas Inspected: Routine, unannounced safety inspection by the resident inspectors and a region-based inspector of licensee action on previous inspection findings; generic letter followup; IE Bulletin followup; licensee action on 10 CFR 50.55(e) report; licensee event report review and followup; review of allegations; training; Region III request; operational safety verification; onsite followup of events at operating reactors; emergency procedures review, and management meetin Results: Of the twelve areas inspected, no violations or deviations were identified in eleven areas. One licensee identified violation was discussed in one area (paragraph 7.a - failure to follow procedure). Inspector concerns were identified regarding control of station commitments and plant emergency off-normal procedures (paragraph 12) and control of design drawing changes (paragraph 10). These concerns were discussed with licensee management to i
assure compliance with NRC requirements and operational schedules.
l l 8701050336 861224 PDR ADOCK 05000461 Q PDR
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P DETAILS 1. Personnel Contacted Illinois Power Company (IP)
- R. Campbell, Manager - QA
- W. Connell, Manager - Nuclear Planning & Support
- G. Edgar, Attorney
- J. Fertic, Director, Quality Systems & Audits
- R. Freeman, Assistant Plant Manager, Maintenance
- W. Gerstner, Executive Vice President
- J. Greene, Manager - Nuclear Station Engineering Department (NSED)
- D. Hall, Vice President, Nuclear
- H. Lane, Manager, Scheduling and Outage Management
- D. O' Nan, Project Engineer, NSED
- J. Palchak, Supervisor - Plant Support Services
- J. Perry, Manager - Nuclear Program Coordination
- F. Spangenberg, Manager - L&S
- J. Wemlinger, Supervisor, Operations Training
- J. Wilson, Manager - CPS
- R. Wyatt, Director, Nuclear Program Assessment Soyland/WIPC0
- J. Greenwood, Manager Power Supply Nuclear Regulatory Commission - Region III
- E. Greenman, Deputy Director, Division of Reactor Projects
- T. Gwynn, Senior Resident Inspector, Clinton
- P. Hiland, Resident Inspector, Clinton
- R. Knop, Chief, Projects Section IB
- R. Warnick, Chief, Projects Branch 1
- Denotes those attending the monthly exit meeting on December 15, 1986.
- Denotes those attending the management meeting on December 1, 1986.
l The inspectors also contacted and interviewed other licensee and j contractor personnel.
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2. Licensee Action On Previous Inspection Findings (92701)
I (Closed) Open Item (461/85050-01): Training program for plant l employees to make them aware of the importance of limiting the l distribution of chemicals throughout the plant.
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The licensee completed preparation of administrative procedure CPS No.1019.05, " Control of Chemicals," Revision 1 and training lesson plan No. 10117, " Control of Chemicals", October 7, 1986. The lesson plan was designed to implement training on the procedure. Sixteen licensee personnel received training on control of chemicals on October 16, 1986. Additional personnel were scheduled for trainin The resident inspectors attended this course as part of the licensee's general employee training / radiation worker training and qualification requirements. The training given met all objectives stated in the lesson plan. This item is close b. (Closed) Open Item (461/66054-01): Containment Purge 36" Ventilation Valve Illinois Power letter U-600672 dated August 8, 1986, identified four 36" containment ventilation isolation valves that did not meet the TMI action plan requirements for containment isolation dependability. The licensee committed to lock these four valves closed until completion of plant modification #VQ-00 During this report period, the inspector reviewed completed plant modification #VQ-002, through supplement 4 dated October 8, 198 The inspector noted that post modification testing was conducted in accordance with section 7.6 of Preoperational Test Procedure (PTP)-VQ-01. Results of the post modification test confirmed the satisfactory completion of plant modification #VQ-002. This item is close c. (Closed) Open Item (461/86064-01): The licensee scheduled an audit of the CPS Emergency Off-Normal Procedures in response to IE Information Notice 86-64. This item was opened pending NRC review
! of the results of the licensee's audi The licensee provided the results of an audit conducted by a joint team composed of Licensing department, Plant Staff Technical department, and Nuclear Training department personnel to the resident inspector for review. The details of that review and the licensee's audit results are discussed in paragraph 12.c. of this report. This item is close d. (Closed) Open Item (461/85039-23): The public address system (Gaitronics) which includes the site-wide warning (siren) system must be completely installed and operationa This item was previously inspected as documented in Inspection Report 50-461/86060, paragraph 2.1. At the conclusion of that inspection, the licensee had provided interim guidance to site personnel concerning actions to be taken when alarms were heard and the accompanying directions (announcement) could not be heard; a maintenance work request had been initiated for gaitronics equipment at the remote shutdown panel; and the licensee was preparing a root cause analysis to assure that the cause of observed gaitronics system problems was determined and correcte .
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The inspector reviewed the results of the licensee's root cause analysis and observed portions of the extensive maintenance work undertaken to correct multiple problems with the gaitronics syste The inspector reviewed revised procedure CPS No. 3842.01, " Plant Communications Alarm Test", and noted that the procedure now required weekly verification of communications equipment operability from both the main control room and the remote shutdown panel. The inspector reviewed CPS No. 2107.01, "Gaitronics Verification",
revision 0, and noted that the procedure provided for a monthly sampling of gaitronics units to verify operability such that once every six months, all gaitronics units would be verified operabl This procedure was intended to assure the reliability of the gaitronics system equipmen The inspector observed the conduct of a weekly communications system alarm test wb'ch the licensee determined was satisfactory. Several areas in the plant were observed not to have audible sound levels for the plant siren and/or accompanying announcement. The licensee evaluated each area and determined that either audibility was not required in the area or an outstanding maintenance work request would provide correction of the deficiency. Overall, the gaitronics system appeared to be functioning properly. The inspector notedc that additional licensee action is required to assure the audibility of plant communications once the station is in operation per their commitments to IE Bulletin 79-18. The inspector will review the licensee's actions under that Bulletin in a separate inspectio This item is close No violations or deviations were identifie . Generic Letter Followup (92703) (Closed) Generic Letter 85-22 (461/85022-HH): Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage. This generic letter discussed the potential for thermal insulation blockage of suction strainers in the suppression pools of boiling water reactor This generic letter was reviewed in Inspection Report N /86037. At the time of that inspection, the licensee had defined specific changes to plant procedures that would assure debris control near fluid intakes is considered when reviewing design change During this report period, the inspector reviewed changes made to Nuclear Station Engineering Department (NSED) Procedure D.22,
" Engineering Interdisciplinary Review and Impact Assessment for Plant Modifications", revision 6, dated April 30, 1986. In addition, the inspector reviewed changes made to Administrative Procedure CPS No. 1005.06, " Conduct of 10 CFR 50.59 Reviews", revision 3, dated September 3, 1986. The inspector verified the changes identified by the licensee were incorporated in the above procedures
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and concluded that the changes made adequately addressed the concerns identified by Generic Letter 85-22. This item is close b. (Closed) Generic Letter 86-02 (461/86002-HH): Technical Resolution of Generic Issue B-19, Thermal Hydraulic Stability. This generic letter provided the NRC staff's conclusion on the technical evaluation of topical reports submitted by General Electric and Exxon. BWR owners were requested to review the need for technical specifications in light of the approved stability criteri This generic letter was reviewed in Inspection Report N /86037. At the time of that inspection, the inspector confirmed the licensee had received this generic letter and had commenced their internal review proces During this report period, the inspector reviewed the licensee's completed actions in response to their internal review proces The provisions of General Electric Service Information Letter (GE SIL)-380 had been incorporated into Clinton Power Station (CPS)
Technical Specification 3/4.4.1. In addition, the inspector reviewed changes made to the following plant procedures: CPS N .03, " Correction of Power Distribution Problems", revision 2, dated January 31, 1985; CPS No. 3005.01, " Unit Power Changes",
revision 3, dated June 26,1986; CPS No. 4008.01, " Loss of Reactor Coolant Flow", revision 5, dated October 10, 1986; and Nuclear Station Engineering Department (NSED) Procedure F.0, " Nuclear Fuel Management", revision 1, dated October 28, 198 The inspector confirmed that the abova changes addressed the recommendations of Generic Letter 86-02. NSED Procedure F.0 was revised to include the guidance of Generic Letter 86-02 in the performance of 10 CFR 50.59 evaluations for core reload submittal This item is close (Closed) Generic Letter 85-13 (461/85013-HH): Transmittal of NUREG-1154 Regarding the Davis-Besse Loss of Main and Auxillary Feedwater Event. The licensee's review of this Generic Letter was previously documented in Inspection Report 50-461/86037, paragraph 4. At the time of that inspection, the licensee was requested to provide the inspector with additional information concerning their Nuclear Training Department's review of Generic Letter 85-1 During this report period, the licensee provided the inspector IP memorandum Y-203739, dated December 10, 1986. That memorandum detailed the Nuclear Training Department's review and incorporation into staff training the lessons learned from the Davis-Besse even In particular, concerns identified in NUREG-1154, applicable to Clinton Power Station, were identified by the Nuclear Training Department and specific training was provided to plant operator The inspector concluded that the licensee had adequately addressed this generic letter in accordance with their established progra This item is close .
O No violations or deviations were identifie . IE Bulletin Followup (92703)(25581)
The bulletin listed below was reviewed to verify that the written response was within the time period stated in the bulletin, that the written response included the information required to be reported, and that the written response included adequate corrective action commitments based on information presented in the bulletin and the licensee's respons (Closed) IE Bulletin (461/86003-88): Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-0perated Valve in Minimum Flow Recirculation Lin The licensee received IE Bulletin 86-03 on October 3, 1986. Their written response, letter U-600737 dated November 5, 1986, indicated that the Clinton Power Station ECCS minimum flow bypass capability did not have the single failure vulnerability identified in the bulletin. Their response was provided in the time frame required by the bulletin and provided the information required by the buller.in. The licensee's response to IEB 86-03 was essentially identical to IEB 86-0 Based on inspection results documented in Inspection Report 50-461/86065, paragraph 5. for IE Bulletin 86-01, this information provided a sufficient basis for closure of bulletin 86-03 No violations or deviations were identifie . Licensee Action on 10 CFR 50.55(e) Report (92700)
(0 pen) 10 CFR 50.55(e) Report (461/86007-EE): Broken Tack Welds on Anchor Darling Globe Valve This matter was previously reviewed, as documented in Inspection Report l 50-461/86060, paragraph 3.a. That report determined that the licensee j had implemented administrative controls as an interim measure to preclude
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the potential for further degradation of those affected valves whose failure could have an adverse impact on plant operations. The administrative controls were adequate to allow fuel loading and operation up to but not including the initial criticality milestone. The licensee's interim report, letter U-600722 dated September 30, 1986, committed to complete all actions required to resolve this issue prior l to initial criticality.
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l The licensee provided a final report on this matter in their letter U-600767 dated November 26, 1986. That report stated that the failure of tack welds between the valve disc and disc nut on two valves in the l reactor water cleanup system represented a reportable (safety
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significant) deficiency that was potentially applicable to a number of additional valves installed in safety related systems at CPS.
l The inspector reviewed the licensee's letter, discussed the letter and l the corrective actions contained therein with the licensee, and observed that the administrative controls previously provided had not been removed i
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as the letter stated. Continuation of the administrative controls was appropriate since resolution of the deficiency had not been reviewed by the NR The licensee's final report stated that certain Anchor Darling valves used in throttling applications where the valve is severely throttled (i.e., less than 20% open) would be susceptible to failure similar to the failures that were reported. The licensee's vendor, Anchor Darling, evaluated operating data provided by the licensee for each affected valve and determined that only two of those valves had experienced operational conditions to date that may have resulted in degradation of the valve The other valves were determined not to have been degraded based on the operational data provided. The licensee inspected the two potentially degraded valves and took appropriate corrective action based on the inspection result The licensee's evaluation of the condition of each potentially affected valve was based on operating data obtained during preoperational testin That evaluation determined that one valve required repair and the other valves had not been degraded. The inspector noted that a total of 4 Anchor Darling globe valves at Clinton have experienced degradation and/or failure due to operation in extreme throttling applications to date (1E12-F012, IE22-F010, and IG33-F042A and B). The licensee planned to inspect 32 potentially affected valves and to review their operating history over the first operating cycle at the first refueling outage to determine the need for further repairs or administrative control However, the licensee provided no justification for operation of the valves during the first cycl In particular, the inspector was concerned that preoperational test data may not be representative of actual valve operating conditions during the first operating cycle or in the event that the plant systems are called upon to perform their intended safety function over an extended period of time (e.g., extended core cooling functions after a severe accident). The inspector requested that the licensee provide justification for operation of each of the 32 valves during the first operating cycle prior to initial criticalit This matter will be reviewed further during a subsequent inspectio No violations or deviations we ~ identifie . Licensee Event Report (LER) Review and Followup (90712 & 92700) In-Office Review Of Written Reports Of Nonroutine Events At Power Reactor Facilities (90712)
For the LERs listed below, the insroctor performed an in-office review of each LER to determine th- reporting requirements had been met; that the corrective action discussed appeared appropriate; that the information provided satisfied the applicable reporting requirements; to determine if appropriate actions had been taken on any generic issues present; and to determine if any additional NRC inspection, notification, or other response was appropriate. Where
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O determined appropriate, the LER was scheduled for onsite followup inspection or other necessary action by cognizant NRC personne (1) (0 pen) LER No. 86-006-00 (461/86006-LL): Automatic Initiation Of Essential Service Water Due To Transient Pressure Drop In Nonessential Service Wate The licensee is planning to provide a supplemental report by December 13, 1986. This matter will be reviewed further with the supplemental repor (2) (Closed) LER No. 86-010-00 (461/86010-LL): Auto Initiation of the Reactor Protection System Due to Electrical Noise in Main Steam Line Radiation Detector Channel (3) (Closed) LER No. 86011-00 (461/86011-LL): Automatic Initiation of Shutdown Service Water Due to Inadequate Surveillance Procedur An onsite followup inspection of this LER was documented by a Region III based specialist inspector in Inspection Report 50-461/8607 In addition, a violation was identified in Inspection Report 50-461/86065 (461/86065-04C) related to this even Followup of the licensee's corrective actions will be tracked by the open violatio (4) (Closed) LER No. 86-012-00 (461/86012-LL): Reactor Water Cleanup Inboard Containment Isolation Due to Transmitter >
Ventin (5) (Closed) LER No. 86-013-00 (461/86013-LL): Operator Error Resulting In RPS Actuatio A violation was identified and closed in Inspection Report 50-461/86065 (461/86065-09A) related to this even (6) (Closed) LER No. 86-014-00 (461/86014-LL): Standby Gas Treatment System Initiation Due To Jumper Installation Erro Inoffice review of this LER indicated that two format errors existed, as follows:
(a) The abstract of the LER contained details not contained in the narrr.tive description ( a common LER writers error).
(b) The narrative description did not discuss the type of personnel involved in the personnel. error as required by 10 CFR 50.73 (b)(2)(ii)(J)(2)(iv).
These matters were reviewed with the licensee's licensing personne I
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An onsite followup inspection of this LER was documented by a Region III based specialist inspector in Inspection Report 50-461/8607 In addition, a violation was identified and closed in Inspection Report 50-461/86065 (461/86065-098)
related to this even (7) (Closed) LER No. 86-015-00 (461/86015-LL): Auto Actuation of an Engineered Safety Feature Due to an Inattentive Workma (8) (0 pen) LER No. 86-017-00 and 86-017-01 (461/86017-LL):
Engineered Safety Feature Actuation Due To Spiking On Intermediate Range Monitor This LER remains open pending receipt and review of a supplemental report. The licensee's supplement was scheduled
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for submittal on December 31, 198 (9) (Closed) LER No. 86-018-00 (461/86018-LL): Reactor Water Cleanup Outboard Containment Isolation Due To Loss Of Powe The inspector verified that all C&I test personnel had been instructed to stop when confronted with a procedural step that cannot be performed precisely as writte No violations or deviations were identifie b. Onsite Followuo Of Written Reports Of Nonroutine Events At Power Reactor Facilities (92700)
For the LERs listed below, the inspector performad an onsite followup inspection of each LER to determine whether responses 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 LE (1) (0 pen) LER No. 86-004-00 (461/86004-LL): Unplanned Automatic Initiatior Of Standby Gas Treatment System Due To Inadequate Procedure This LER was previously reviewed as documented in Inspection Report 50-461/86065, paragraph 6.b. That review indicated that the LER was not clearly written, contained some inaccuracy concerning the scope of corrective actions taken, and did not provide a substantial basis for limitations placed on the scope of corrective actions taken concerning annunciation of trip signals. The inspector requested two actions by the licensee concerning this LER, as follows:
(a) That the licensee provide the basis for limiting the scope of corrective actions taken related to seal-in logic for l the five radiation monitors mentioned in the report, and (b) That the licensee consider providing a revision to the LER
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During this report period, the licensee provided their basis for limiting corrective actions taken. A review was performed by the Nuclear Station Engineering Department which indicated that the logic trip seal-in used in the five radiation monitors referred to in this LER may be used in other plant equipmen This preliminary information was provided to the inspector by the licensee on December 15, 1986. The licensee is considering the need for a revision to this LER. The inspector will review this LER further during a subsequent inspectio (2) (0 pen)'LER No. 86-007-00(461/86007-LL): Reactor Water Cleanup System Isolation Due To High Differential Flo A violation was identified in Inspection Report 50-461/86065 (461/86065-04A) related to thfs event. Followup of the licensee's corrective actions concerning the inadequate surveillance procedure will be tracked by tFe open violatio The inspector noted that this and other corrective actions discussed in this LER were pending ection at the completion of the inspectio The root cause of this LER was determined by the licensee to be the lack of a caution statement in the system operating procedure to require filling and venting of the reactor water cleanup (RT) system prior to restoration after an isolatio The inspector reviewed the applicable system operating procedure, CPS No. 3303.01, Reactor Water Cleanup System, revision 4 dated July 23, 1986 and determined that the procedure contained no instructions for filling and venting the system except for explicit instructions to vent the high points on the vertical regenerative and nonregenerative heat
[ exchangers. This matter appeared to be a deviation from the licensee's FSAR, Chapter 13.5.2.1.1 which states in part that instructions for filling and venting are delineated in the I
l system operating procedures. The licensee provided a condition report (CR) No. 1-86-09-039 dated September 4, 1986 which indicated that this condition had been identified by IPQA as a condition applicable to a number of system operating procedures. The CR did not have an approved corrective action l plan at the time of this inspection (reference Inspection Report 50-461/86065, paragraph 10). However, the licensee provided a Centralized Commitment Tracking Input Form No. 43639 dated November 20, 1986 which indicated that the specific procedures referenced by the CR would be revised to include steps for filling and venting the system by March 30, 198 The inspector discussed this matter with the Supervisor - Plant j Operations Support concerning the need for timely corrective
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l This LER will remain open pending review of an approved corrective action plan for CR 1-86-09-039 and pending licensee verification of corrective action completion fcr this LER.
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(3) (0 pen) LER No. 86-008-00 (461/86008-LL): Containment Isolation of the Instrument Air System Due to Procedural Inadequac The inspector's review of this LER indicated that a complete and accurate description of the entire event was not provided by LER No. 86-008-00. The inspector noted that while not referenced in LER No. 86-008-00, a complete description of this event included information that was described in LER N . The licensee stated that a revision to LER N will be issued and that it would correct deficiencies noted by the inspector. In addition, the revised LER N will include all the information contained in LER No. 86-008-00. The licensee stated that LER No. 86-008-00 would be withdrawn upon issuance of revised LER No. 86-009-0 This item will remain open pending the inspector's review of revised LER No. 86-009-0 (4) (0 pen) LER No. 86-009-00 (461/86009-LL): Operator Error Resulting In Technical Specification Violation As discussed in (3) above, this LER provided information that was pertinent to the event described in LER No. 86-008-0 Both of these events were documented in Inspection Report 50-461/86065, paragraph 12.b.(3). Several violations were identified in Inspection Report 50-461/86065 (461/86065-04C, 06B, and 07A, B, C) related to this event. Followup of the licensee's corrective actions will be tracked by the open violation The inspector's review of LER No. 86-009-00 identified a number of discrepancies between the licensee's description of the event and the inspector's onsite followup documented in Inspection Report 50-461/86065. In particular, LER N stated that the licensee had violated the Technical Specification for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 28 minutes. The inspector's onsite followup of the entire event (LER N and 86-009-00) identified that the licensee was in violation of the plant Technical Specifications for approximately 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> The inspector noted during discussions with the Supervisor -
Plant Operations, that the licensee was knowledgeable of all the facts surrounding the events described in LER No. 86-009-00 (and LER No. 86-0'.8-00) . The licensee agreed with the inspector that LER No. 86-009-00 (and LER No. 86-008-00) did not accurately describe this event. The licensee stated that LER No. 86-009-00 would be revised and that the related event described in LER No. 86-008-00 would be included in that revisio LER No. 86-008-00 would be withdrawn when the revised LER No. 86-009-00 was issued. This item remains open pending the inspector's review of revised LER No. 86-009-0 .
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(5) (0 pen) LER No. 86-016-00 (461/86016-LL): Operator Error Resulting in Technical Specification Violatio A violation was identified in Inspection Report 50-461/86065 (461/86065-070) related to this event. Followup of the licensee's corrective actions will be tracked by the open violatio In-office review of this LER identified one error and a lack of complete reporting of this event. The error involved referencing the wrong reporting requirements in the LER. The licensee acknowledged the error and the incomplete narrative description and stated that a revised report would be provide The inspector noted that the " Additional Information" provided by the licensee in this LER was apparently in error in that no similarity could be found between this LER and LER N .
This LER remains open pending review of the revised repor No violations or deviatioris were identifie . Review of Allegations (99014) (Closed) Allegation (RIII-86-A-0109; #194): On June 12, 1986, the resident inspector received an allegation that unqualified personnel were operating the polar crane and the fuel handling cran NRC Review and Results The need to provide training for crane operators was addressed by the NRC in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", dated July 1980. This NUREG was provided to the licensee via letter dated December 22, 1980. The licensee reviewed the guidelines of NUREG-0612 and provided several responses to NRC
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noted that IP letter U-0800, dated February 21, 1985, specifically responded to NRC questions on Crane Operator trainin In that l response, the licensee stated their intent to comply with ANSI B30.2-1976 for Crane Operator training.
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The Clinton Power Station - Final Safety Analysis Report (FSAR),
Section 13.5.2.1.1, Amendment 38, dated May 1986, stated that
" Procedures shall require crane operators who operate cranes over fuel pools be qualified and conduct themselves in accordance with ANSI B30.2-1976."
The inspector reviewed the requirements for training crane operators contained in procedure CPS No. 8106.02, " Qualification of Crane Operators", revision 1, dated March 15, 1985. Paragraph 2.1 of CPS
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No. 8106.02 stated: " Operators of cranes listed in Appendix A of this procedure shall be qualified in accordance with the requirements of this procedure". Appendix A to CPS No. 8106.02 listed the Containment Polar Crane (1HC01G) and the Fuel Building Crane (1HC07G). The inspector noted the training requirements for qualification of crane operators contained in CPS No. 8106.02 met the quidelines of ANSI B30.2 1976, Chapter 2-3, " Qualification for Operators".
The licensee initiated Condition Report (CR) 1-86-07-012 on June 21, 1986, which documented a condition similar to the allegation received by the inspector on June 12, 1986. As described in this CR, startup personnel were operating the Fuel Building Crane (IHC07G) on June 20, 1986, without having been qualified in accordance with CPS No. 8106.02. The licensee's investigation determined that startup personnel were operating the Fuel Building Crane without having completed the qualification requirements of CPS No. 8106.02. The approved corrective action to this condition report was that startup would use only " qualified" individuals for all further testing on equipment released to Plant Staf The crane operator training guidelines contained in NUREG-0612, and committed to by the licensee in their FSAR, provided one measure to assure the potential for a load drop while handling " heavy loads" is small. The term " heavy load" was defined in NUREG-0612 as follows:
"Any load, carried in a given area after a plant becomes operational, that weighs more than the combined weight of a single spent fuel assembly and its associated handling tool for the specific plant in question." Since the licensee did not receive an operating license until September 29, 1986, the term " heavy load" was not applicable to the Clinton Power Station during the time that " unqualified" personnel were known to have operated the Fuel Building Cran The inspector interviewed the cognizant Mechanical Maintenance Assistant Supervisor and four Mechanical Maintenance personnel (including two qualified crane operators) to determine if the training requirements of CPS No. 8106.02 were being adhered to subsequent to issuance of the operating license. The individuals expressed no concerns with the crane training or with adherence to crane training requirements. In addition, the inspector reviewed training records for 16 " qualified" crane operators to verify the documentation of training was adequate and the training provided was in accordance with CPS No. 8106.0 Conclusion This allegation was substantiated in that startup personnel had S operated the Fuel Building Crane on June 20, 1986, without completing the qualification requirements of CPS No. 8106.02,
" Qualifications of Crane Operators". The inspector noted that the
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licensee had taken corrective action to the violation of procedural requirements for training crane operators as documented in Condition Report (CR) 1-86-07-012. Since the violation of procedural requirements was identified by the licensee; the violation fit into a severity level IV or V; the violation was not reportable; the violation was corrected within a reasonable time frame; and a previous violation had not occurred, the inspector determined that a Notice of Violation was not warranted. The inspector concluded through interviews of licensee personnel and review of training records that the requirements of CPS No. 8106.02 were being adhered to subsequent to the issuance of the operating license. This allegation is close (Closed) Allegation (RIII-86-A-0142): An individual contacted the resident inspector and stated that computer inputs for the measuring and test equipment (M&TE) trace system were three months behind the actual work completion and, for that reason, the M&TE trace printout may not contain all applicable uses of the affected M&TE. The individual was concerned that inadequate use history an. lyses ca .;d result from this condition. The individual stated that a condi-ion .
report (CR) had been generated by the Controls & Instrumentation (C&I) calibration laboratory on this issue but necessary actions had not been taken to rectify the proble This matter was referred to the licensee by Region III for their review and evaluation by letter dated October 20, 1986. The licensee's response, letter U-600766 dated November 7, 1986, provided the results of their review and evaluation of the allegation. That response substantiated the allegation and its potential impact on the M&TE use history analysis, and provided information concerning corrective actions being taken to address the condition as documented in a licensee initiated C NRC Review and Results The inspector reviewed the licensee's letter; reviewed CR 1-86-09-137 dated September 17, 1986, and the licensee's corrective action plan; discussed the CR and planned corrective actions with cognizant licensee personnel; interviewed the C&I calibration laboratory supervisor; reviewed the CR log over the period August-September 1986; and discussed the results of this review with the licensee's quality assurance departmen The inspector determined the following:
(1) The inspector found no evidence that a CR was ever initiated by the C&I calibration laboratory concerning this matte (2) The CR referenced in the licensee's response was initiated by the Assistant Plant Manager - Mz titenance due to a backlog of use history analysis (UHA) maint. nance work requests (MWRs)
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An evaluation was performed which indicated there was no impact on plant operation in modes 4 and 5. The corrective action plan attached to the CR addressed subsequent milestones (modes 1, 2, and 3) but-was not complete at the time of this inspection. The CR referenced in the licensee's response did not explicitly address a backlog of computer inputs to the M&TE trace syste (3) Interviews of licensee personnel indicated that the backlog of M&TE trace inputs did not, by itself, represent a condition adverse to quality. The inspector was informed that personnel performing UHA for out of calibration, lost, damaged, or stolen M&TE use the M&TE trace printout and the backlog of inputs when performing UHA MWR The inspector discussed the above results with the licensee's QA department and requested that an additional review be performed to determine whether a CR had been initiated in the C&I calibration laboratory concerning this matter. The licensee's QA department interviewed 75 of 97 total C&I personnel, including technicians, foremen, and supervisors. In addition, all CRs initiated by personnel working in the C&I calibration laboratory during the period June 1 through October 1,1986, were reviewed by the licensee. The licensee concluded that they could not locate the CR referred to in the allegation and that there was no evidence of CRs being suppressed in the C&I maintenance group. The licensee substantiated that a current backlog of 600-700 MWRs awaiting input to the M&TE trace computer existed at the beginning of December 1986. This condition required a manual search be done when performing UHA MWRs which increased the time required for the UHA and increased the potential for human error. The condition did contribute to the backlog of UHA MWRs which was the subject of CR 1-86-09-13 Conclusion The allegation was partially substantiated in that there was a backlog of M&TE trace inputs which contributed to the cause of a condition identified by the licensee and reported in CR 1-86-09-13 The licensee had acted to resolve the condition once it was identified. However, the inspector noted that there was no evidence of a CR written by the C&I calibration laboratory or others on the backlog of M&TE trace inputs or of suppression of CRs in the C&I maintenance group. The condition documented on CR 1-86-09-137 was
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primarily caused by cumbersome administrative controls which were being revised at the time of the inspection. The licensee had evaluated the condition documented on CR 1-86-09-137, had determined that there was no impact on plant operation in modes 4 and 5, and had scheduled revision of the administrative control procedures to remedy the cause of the condition identified in the CR. The inspector will verify the effectiveness of the licensee's corrective actions in a subsequent inspection. This is an open item (461/86072-01). This allegation is close __ _- .
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Both allegations are close . Training (41700)
Training activities were observed during this inspection to ascertain that training was being provided in accordance with technical specifications and licensee commitments. The inspector observed the following formal training classes in progress during the report period: Precriticality Simulator Training This training was conducted in accordance with Nuclear Training Department (NTD) Lesson Plan No. SE 95057, revision 1. Participants included the Main Control Room shift crew, the Shift Test Engineer, and the Nuclear Engineer assigned. The purpose of the training was to familiarize the plant operators with specific plant evolutions performed during initial control rod withdrawal to criticality and subsequent plant heatup. The training included potential malfunctions and appropriate responses, CFR 50.59 Safety Evaluation Trainina This training was conducted in accordance with a Student Handou The purpose of the training was to familiarize personnel assigned responsibility for performing safety evaluations with plant procedures and procedure changes associated with the performance of 10 CFR 50.59 Safety Evaluation Control of Chemicals The inspectors attended this course as a required portion of the licensee's radiation worker training. This training was conducted in accordance with NTD Lesson Plan No. 10117, revision For each course observed, the inspector verified that the course was conducted in accordance with the approved lesson plan and that personnel presenting the course were knowledgeable of the course content and were responsive to student question ,
No violations or deviations were identifie . Region III Request (92701)
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The inspector reviewed the status of the licensee's actions with regard to General Electric (GE) Co. Service Information Letter (SIL) No. 445, Intermediate Range Monitor (IRM) Fuse Failure, dated July 26, 198 GE SIL 445 identified a condition whereby the IRMs may be inoperable due to a blown fuse in the -24 VDC power supply without immediate operator
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detection. This SIL was preceded by a GE Rapid Information Communication Services Information Letter (RICSIL) No. 007, dated June 26, 198 __
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The licensee received SIL 445 on September 26, 1986 and assigned the
> Nuclear Station Engineering Department (NSED) primary responsibility for *
evaluation of the SIL. Their evaluation was documented in IP letter Y-82326, dated October 20, 198 IP NSED concluded that the IRM fuse failure identified in RICSIL 007 would have no adverse safety impact on CPS. However, two of three recommendations made by GE in SIL 445 were pending action at the time of this review. The first recommendation involved evaluation of licensee procedures for establishing operability of a safety.related instrument channel after its loss. GE recommended that a channel functional test be performed following return to operability. This recommendation was pending action within the Plant
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Staff technical department. The second recommendation involved-evaluation of the need for annunciation of loss of the -24 VDC IRM power -
- supply. NSED initiated X-MOD X-NR006 to evaluate the need for such annunciation. The inspector will review the results of the licensee's i evaluations when they are complete. This is an open item (461/86072-02).-
!- No violations or deviations were identifie . Operational Safety Verification (71707)
l The inspectors observed control room operations, attended selected
- pre-shift briefings, reviewed applicable logs, and conducted discussions
- with control room operators during the inspection period. The inspectors
verified the operability of selected emergency systems and verified i tracking of LCOs. Routine tours of the auxiliary, fuel, containment, l control, diesel generator, turbine, and screenhou:e buildings 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 inspectors
! verified that maintenance requests had been initiated for discrepant conditions observed. The inspectors verified by direct observation and t
discussion with plant personnel that security procedures and radiation protection (RP) contrels were being properly implemented.
J The inspectors observed plant housekeeping / cleanliness condition No
! discrepancies were noted.
I The above reviews and observations were accomplished to verify that facility operations were conducted in conformance with the CPS technical specifications and the conditions of the operating licens During main control room observations, the inspector observed the licensee performing Surveillance Procedure CPS No. 9027.01 " Remote Shutdown Panel Operability Check", revision 20 (including TCF 86-1563),
dated October 16, 1986. While preparing to perform step 8.5 of this surveillance, operations personnel identified a discrepancy between the
! controlled-schematic drawings (E02-1AP99 sheets 13 and 38) and the actual l field installation for Diesel Generator IA output breaker relay 227X-DGKA. The licensee initiated Condition Report (CR) 1-86-11-144 to document the identified discrepancy and to provide corrective actio .- _ . _ , _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ _
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The discrepancies concerned Field Engineering Change Notice (FECN) 8189 that had made wiring changes to the subject relay (227X-DGKA) during the construction phase.at Clinton Power Station. The inspector confirmed by review of vaulted construction records, that the wiring changes directed by FECN 8189 had been completed in March 1985. The controlled schematic drawings (E02-1AP99 sheets 13 and 38) had been properly updated to reflect incorporation of FECN 8189. However, the actual "as built" condition identified during the performance of Surveillance Procedure CPS
- No. 9027.01 was not reflected in the controlled drawing The licensee documented their investigation of Condition Report 1-86-11-144 in Nonconforming Material Report (NCMR) 1-2575. That NCMR identified that two additional Field Engineering Change Notices (FECNs)
had been incorporated subsequent to the completion of FECN 8189. These additional FECNs directed wiring changes to relay 227X-DGKA which affected changes incorporated via FECN 8189. However, reference to FECN ,
8189 was not made_during the subsequent FECN incorporations nor was FECN '
8189 revised, cancelled, voided, or deleted from the licensee's Design Status System (DSS). This resulted in the controlled drawings not reflecting the actual "as-built" condition The inspector noted that the failure to reference FECN 8189 was identified by the licensee on August 7, 1986 as documented on a " Change Document Supersede / Void Request" form dated August 7, 1986. While this document identified the discrepancies, no action was initiated to correct the design drawings until November 25,1986 (ref. SLEI-20589). The failure to maintain design drawings that accurately reflect the as-built configuration was the subject of a Notice of Violation issued during this report pericd in Inspection Report 50-461/86073. The above item was considered to be an additional example of that violatio The inspector discussed this item in detail with the licensee to determine if the root-cause for the identified discrepancy had a potential impact on other design drawings. In particular, the inspector requested the licensee to determine if additional " Change Document Supersede / Void Request" forms (or similar documents) were still outstanding against design drawings. This is considered an unresolved item pending the inspector's review of the licensee's response (461/86072-03).
One unresolved item was identifie . Onsite Followup of Events at Operating Reactors (93702) General The inspector performed onsite followup activities for events which occurred during the inspection period. Followup inspection
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included one or more of the following: reviews of operating logs; l
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procedures; condition reports; direct observation of licensee actions; and interviews of licensee personne For each event, the inspector reviewed one or more of the following: the sequence i
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of actions; the functioning of safety systems required by plant conditions; licensee actions to verify consistency with plant 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) Control Room Ventilation System Shift To High Radiation Due To Radiation Monitor Failure (ENS 6856)
At approximately 4:08 a.m. CST on November 10, 1986, the control room ventilation (VC) system shifted to the high radiation mode due to upscale failure of a VC system air intake radiation monitor. The licensee carried out the technical specification action to fail the instrument downscale within 10 minutes of the event. The VC system, which is not required by technical specifications in the current plant operating mode, remained in the high radiation mode pending investigation of the radiation monitor failure. The licensee notified the NRC Operations Center of this event at about 6:49 a.m. LST. This matter will be reviewed further during review of the licensee's LE (2) ESF Actuation - Reactor Water Cleanup System Isolation At approximately 12:15 p.m. CST on November 14, 1986, the licensee notified the NRC Operations Center of a reactor water cleanup (RT) system isolation. The system experienced an isolation signal and actuation of the division 1 outboard isolation valves at about 11:15 a.m. CST due to the cutting of an electrical lead from the temperature detector (division 1)
in RT pump room A during repair of a Raychem splic Preliminary information from the licensee indicated the cause of the isolation was an inadequate maintenance work request (MWR) that failed to identify the response of the RT system to the cutting of the temperature detector lead. The MWR in use identified only an alarm and recorder actuation associated with cutting the temperature detector lead. This matter will be
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reviewed further during review of the licensee's LE (3) Control Room Ventilation System Shift To High Radiation Mode Due To Radiation Monitor Failure (ENS 7000)
At approximately 7:16 a.m. CST on November 24, 1986, the Control Room Ventilation (VC) system shifted to the high radiation mode due to upscale failure of a VC system air intake radiation monitor. The licensee carried out the technical specification action statement to fail the instrument downscale
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within 10 minutes of the event. The VC system, which is not required by technical specification in the current plant operating mode, remained in the high radiation mode pending investigation of the monitor failure. A similar occurrence of this event was previously reported by the licensee on November 10, 1986 (ENS 6856). The licensee was investigating why corrective action taken following the November 10 occurrence were not effective. The licensee notified the NRC Operations Center of this event at about 9:20 a.m. CST. This matter will be reviewed further during review of the licensee's LE (4) ENS Declared Inoperable (ENS 7006)
At approximately 2:00 p.m. CST on November 24, 1986, the licensee declared the ENS inoperable when the Control Room and Technical Support Center ENS phones failed. The licensee successfully tested the Control Room ENS phone at about 2:25 p.m. and declared it operational. Licensee investigation
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of problems experienced with the Technical Support Center ENS phone resulted in identification of the need for maintenanc The licensee notified the NRC Operations Center of this event at about 2:50 p.m. CS (5) Technical Specification Violation (ENS 7060)
At about 2:09 a.m. CST on December 1, 1986, the licensee discovered control rod 24-21 was at position 02 (1 notch withdrawn) during performance of a surveillance test. All control rods were required to be fully inserted in the current plant operating mode. The licensee inserted the control rod to position 00 and all control rods were verified to be fully inserted. The licensee determined that control rod 24-21 had e been withdrawn to position 02 at 2:04 a.m. No cause for the control rod motion was determined although personnel error was suspected. The movement of the control rod was a core alteration for which the plant was not properly aligned in accordance with the Technical Specifications resulting in a technical specification violation. This technical specification violation was reported to the NRC Emergency l Operations center via the ENS at about 3:06 This matter l will be reviewed further during review of the licensee's LER.
I (6) ESF Actuation - RPS Trip Signal (ENS 7123)
On December 7, 1986 at about 8:05 a.m. CST, the reactor protection system (RPS) received a trip signal due to high water level in the scram discharge volume (SDV) of the control rod drive system. The plant was in mode 5 with all control rods fully inserted at the time the trip signal occurred. The high water level in the SDV resulted from cleaning activities being performed to remove scale from the internals of the SDV
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piping using a high pressure hydrolaser. The licensee notified the NRC Emergency Operations Center of this event at about 4
9:15 a.m. CST on December 7, 198 Preliminary results of the s licensee's investigation of this event indicate the cause was a lack of adequate communications between the cleaning crew snd ,,
the control room supervisor. This matter will be reviewed further during review of the licensee's LE One violation was identified and reported by the licensee. This matter will be reviewed further during review of the licensee's LE ,
12. Emergency Procedures Review (42452)
This inspection completed a review (reference Inspection Reports 50-461/86048, paragraph 6.c, 50-461/86054, paragraph 5.b and 50-461/86060, paragraph 6.b.) of procedures to be used in the plant operations phase to confirm that the plant emergency procedures are prepared to adequately control safety related functions when a system or component malfunction is indicated, Applicable Requirements, Applicant Commitments, and Guidance Documents (1) 10 CFR 50 (2) Regulatory Guide 1.33, revision 2, " Quality Assurance Program Requirements" (3) ANSI N18.7-1976, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants" (4) ANSI N45.2-1977, " Quality Assurance Program Requirements for Nuclear Facilities"
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(5) CPS No. 1005.01, revision 16, " Preparation, Review and Approval of Station Procedures" (6) CPS Final Safety Analysis Report (FSAR), Chapter 1 (7) NUREG-0853, Safety Evaluation Report, and Supplements c
(8) CPS No. 1450.00, CPS Emergency Procedures Guidelines, revision (9) CPS No. 1005.01, Appendix C, Writer's Guide For Emergency Off-Normal Procedures, revision 1 '
(10) IE Information Notice 86-64, Deficiencies In Upgrade Programs For Plant Emergency Operating Procedure (11) TI 2515/79, Inspection Of Emergency Operating Procedures.
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5 Procedures Reviewed
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(1) CPS No. 4401.01, Level Control Emergency, revision s(2) CPS No. 4402.01, Containment Control Emergency, revision (3) CPS No. 4404.01, Reactivity Control Emergency, revisicn , Discussion The inspector reviewed the index of plant emergency off-normal
' procedures against the CPS Emergency Procedure Guidelines (EPGs) and verified that all applicable Emergency Off-Normal Procedures (EOPs)
had been or were being developed. One open item concerning the development c' a hydrogen control E0P was being tracked as open item-s (461/85015-07).
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The inspector then reviewed a sample of the procedure steps for each procedure selected to verify that the procedures were consistent t with the NRC approved CPS EPGs and that the procedure steps had been written in accordance with the NRC approved Emergency Procedures
., i Writers Guide. NRC Temporary Inspection Instruction (TI) 2515/79
was used for guidance during this inspection; a more detailed inspection using TI 2515/79 is expected to be conducted by a Region III based team at a later date. This review identified a number of discrepancies detailed later in this section of the repor s The inspector then reviewed the results of a licensee audit of the CPS EPGs and E0Ps, conducted in response to IE Information Notice 86-64. The audit represented a sample of each of the CPS EPGs and E0Ps, included a review of the calculational basis for the E0Ps, a review of the E0P verification and validation (V&V) program, and a review of the resolution of NRC and General Electric Company review questions and IP responses which resulted from the review and approval process. The following summarizes the significant findings of the licensee's audit:
l 0 (1) The CPS EPGs have not been updated to be consistent with the l i most recent E0P revisions. In some cases, the EPGs were revised and the LJP was no (2) Calculational results were not always accurately reflected in the E0Ps.
i (3) The CPS E0Ps do not contain a reference section as required by l the writer's guid ,
(4) V&V of the CPS E0Ps was not documented in accordance with the licensee's commitments.
l (5) Revisions to E0Ps were not V&V'd as required. (Note: This was identified in Condition Report (CR) 1-85-11-107 dated November 8, 1985).
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-4 (6) There were unresolved GE comments outstanding against the CPS EPGs/EOP The above audit findings were transmitted to the Manager - CPS for corrective action on December 8, 1986. The Licensing department recommended that the plant staff take immediate action concerning findings reflected in (1) and (2) abov In addition to the licensee's audit findings, the following items were identified during NRC review of the CPS E0Ps:
(7) In one case, required operator actions referenced were not contained in the referenced documen [ CPS No. 3314.01, Standby Liquid Control, did not contain instructions for injecting boron into the reactor vessel via the RCIC storage tank in the event the containment was inaccessible].
(8) The E0Ps were not prepared in a " style that presents information in a simple, familiar, specific,... manner" as recommended by the emergency procedures writer's guid (9) The inspector noted an inconsistent approach to the control of flow charts attached to the CPS E0P Some of the flow charts were identified as safety related with FRG review and management approval. Others were identified as non-safety related with no FRG review and a lower level of management approval. Some flow charts had been subject to an independent technical review and others had not. The E0P writer's guide implied that the flow charts were a part of the E0P but the flow charts had not received the same level of review and approval and had not been subject to V& (10) A corrective action plan had been approved for CR 1-85-11-107 on October 17,1986 (see IP audit finding (5) above). The inspector noted that some corrective actions had been completed l including revision 17 to CPS No. 1005.01 to provide for V&V of l E0P revisions. The scope of E0P revisions to which the l corrective actions in the CR applied was defined in the CR l based on the date of revision of CPS No. 1005.01. The
inspector reviewed one E0P revision package that w6s not within
- the scope of the CR and found no evidence that the E0P revision l had been processed in accordance with the revised procedure.
l (11) The IP audit finding related to the lack of a reference section in the E0Ps (see IP audit finding (3) above) was similar to a Facility Review Group (FRG) action item from a meeting held in February, 1986. The resolution of that action item resulted in a centralized commitment tracking (CCT) system item No. 40943, discussed in FRG meeting minutes for meeting 86-069 dated May 7, 1986 which stated that the E0Ps would be revised at the biennial review to include a references section. The inspector noted that the licensee did not meet this internal commitment
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O during revision of two CPS E0Ps for which credit was taken for a biennial review [ CPS No. 4401.05, revision 7 and CPS N .01, revision 5].
, While many of the discrepancies described above were minor in nature, they represented matters that required review and evaluation by the licensee and. in some cases, prompt corrective action. The licensee had' scheduled completion of their evaluation and determination of a corrective action plan for December 15, 198 The above matters are considered to be an unresolved item pending completion of the licensee's evaluation and review of their corrective action plan (461/86072-04). Results The inspector confirmed discrepancies identified in the licensee's audit and identified several additional items for evaluation by the licensee. Those items are being tracked as a unresolved item. The NRC plans a more detailed review of the licensee's E0Ps at a later dat One unresolved item was identifie . Management Meeting (30702)
On December 1, 1986, NRC management met with IP management at the Clinton Power Station to discuss matters related to Licensee Event Reports, corrective actions being taken by the. licensee, the licensee's self-monitoring methods, and to arrange for periodic (initially monthly)
NRC/IP management meetings to discuss plant performanc Personnel attending the meeting are identified by (#) in paragraph 1. of this repor The meeting focussed on management actions being taken by IP to improve their operating performance as a followup meeting to the one held on l October 22, 1986 in the Region III office (reference Inspection Report j 50-461/86065, paragraph 13).
! At the conclusion of the meeting, both parties agreed to meet periodically initially monthly) beginning in January 1987 to more closely monitor CPS progress in achieving improvement goal . Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which will involve some action on the part of the NRC or licensee or both. Two open items disclosed during the inspection were discussed in paragraphs 7.b. and 9.
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W 15. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations. Two unresolved item disclosed during this inspection were discussed in paragraphs 10. and 1 . Exit Meetings (30703)
The inspectors inct with licensee representatives (denoted in paragraph 1)
throughout the inspection and at the conclusion of the inspection on December 15, 1986. The inspectors summarized the scope and findings of the inspection activities. The licensee acknowledged the inspection findings. The inspectors highlighted the need for management attention to internal commitments and the CPS emergency off-normal procedure The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents / processes as proprietar The resident inspectors attended exit meetings held between Region III based inspectors and the licensee as follows:
Inspector (s) Date Falevits 11/20/86 McCormick-Barger, Wei;.e, O'Dwyer, and Ridgeway 12/05/86 l
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