IR 05000461/1987015

Safety Insp Rept 50-461/87-15 on 870406-0518.Violations Noted:Concerning Event Followup.Unresolved Item Noted Re Operational Safety Verification.Licensee Reviewing Unresolved Item for Needed Corrective Action

ML20214Q353
Person / Time
Site:	Clinton
Issue date:	05/29/1987
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20214Q307	List: IR 05000461/1987015 ML20214Q304 ML20214Q341
References
RTR-NUREG-0737, RTR-NUREG-737 50-461-87-15, IEIN-87-008, IEIN-87-8, NUDOCS 8706040350
Download: ML20214Q353 (26)

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, i U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-461/87015(DRP)

Docket No. 50-461 License No. NPF-62 Licensee: Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Nanie: Clinton Power Station Inspection At: Clinton Site, Clinton, IL Inspection Conducted: April 6 through May 18, 1987 Inspectors: P. Hiland C. Brown S. DuPont R. Gardner B. Hasse B. Lerch B. Little L. McGregor A. Morrongfello J. O'Dwyer B. Siegel T. Taylor

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Approved By: R. C. Kno(p), Chief Projects Section IB

&-h Date l Inspection Suninary inspection on April 6 through May 18, 1987 (Report No. 50-461/87015(DRP))

Areas Inspected: Routine, unannounced safety inspection by the resident inspectors and region-based inspectors of licensee action on previous insoection findings; review of allegations; employee concerns; licensee actdon on TMI action plan requirements; IE Infornation Notice followup; lic.insee event report review and followup; monthly maintenance observation; monthly surveillance observation; operational safety verification; modtf fcation testing; training effectiveness; onsite followup of events at Operating reactors; startup test witnessing; and management meetin Results: Of the 14 areas inspected, no violations or deviations were TdentifTe,d in 12 areas. Two violations were identified in the area of event followup. The licensee took intnediate action to correct the violation One unresolved item was identified in the area of operational safety verifi-cation. The lir;ensee is reviewing this unresolved item for needed corrective actio /06040*J50 070029 PDH ADOCK 05000461 o PDR

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DETAILS

, Personnel Contacted Illinois Power Company (IP)

• K. Baker, Supervisor - I&E Interface, Licensing and Safety (L&S)

1. T. Camilleri, Manager - Scheduling Outage and Maintenance

• 1. R. Campbell, Manager - QA

1. W. Connell, Manager - Nuclear Station Engineering Department (NSED)

1. J. Cook, Assistant Manager - Clinton Power Station (CPS)

1. R. Freeman, Assistant Plant Manager, Maintenance

• 1. D. Hall, Vice President, Nuclear '

• E. Kant, Assistant Manager, NSED J. Miller, Assistant Manager - NSED J. Palchak, Supervisor - Plant Support Services J. Perry, Manager - Nuclear Program Coordination

• A. Ruwe, Director - SOM

1. F. Spangenberg, Manager - L&S P. Telthorst, Licensing and Safety

• 1. E. Till, Director Nuclear Training J. Weaver, Director - Licensing

• J. Wilson, Manager - CPS Soyland/WIPC0 J. Greenwood, Manager Power Supply

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Nuclear Regulatory Commission

• 1. P. Hiland, Senior Resident Inspector, Clinton

1. R. Knop, Chief, Projects Section IB

1. C. Norelius, Director, Division of Reactor Projects, RIII

1. Denotes those attending the management meeting on May 15, 198 * Denotes those attending the monthly exit meeting on May 18, 1987.

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The inspector also contacted and interviewed other licensee and contractor personne . LicenseeActionOnPreviousInspectionFindings(92701)(92702) (Closed)OpenItem(461/86065-08): Action Statement No. 29 in Technical Specification Table 3.3.2-1 required clarificatio The Clinton Technical Specification (NUREG-1235) was revised to delineate more clearly the actions required relative to instru-mentation of the containment and reactor vessel control system (CRVICS). The revisions to technical specifications were reviewed by the staff as described in Supplemental Safety Evaluation Report No.8(SSER8). This item is close _ _ _ _ , _ _ _ _.___._._ -

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, * (0 pen) Open Item (461/86072-01): Control of Measuring and Test Equipment (M&TE). Effectiveness of licensee corrective action to Condition Report (CR) 1-86-09-137 was to be evaluated. During this report period, the inspector reviewed the licensee's program for calibration and control of measuring and test equipmen As described in CR 1-86-09-137, an excessive number of Use History Analysis (UHA) maintenance work requests (MWRs) had not been evaluated for impact on system operation. The licensee revised the M&TE program and issued the controlling administrative procedures CPS No.1012.01, " Control of M&TE", revision 4, and CPS No. 1512.01,

" Calibration and Control of M&TE", revision 8, on January 30, 198 The inspector reviewed the revised procedures and discussed their implementation with cognizant personnel in the licensee's M&TE calibration la Through the above discussions and review of M&TE files it appeared that the revised procedures were being properly implemente Personnel interviewed indicated that the controls now in place were effective in reducing the backlog of Use History Analysis (UHA).

However, the inspector noted that about 26 M&TE items had required a UHA, but the responsible departments were late in responding. The inspector noted that the licensee was actively pursuing the late UHAs in accordance with administrative procedure CPS No. 1512.01, paragraph 8.5.5. This item will remain open pending additional review of program effectiveness in a future inspectio c. (0 pen) Open Item (461/86074-03): Lack of formal procedures to define responsibility and the mechanisms to ensure the identifi-cation of new non-licensed operator (NLO) qualification requirements and the completion of these requirement The inspector determined that the new procedure written by the Nuclear Training Department (NTD) did not completely address the concern. Section 5.3.2.5 of NTD Procedure 2.15, revision 0, "NTD NLO Training Program", stated that the Shift / Assistant Shift Supervisor (S/ ASS) "may" revise an in-process checklist by filling out a "NLO Training Addendum Sheet", (Attachment 4 of NTD Procedure 2.15). This was only an option and does not explicitly define requirements. Further, section 5.3.2.5 did not specify who, if anyone, shall review checklists of all personnel affected by revisions, or temporary changes to the checklist requirements of NTD Procedure 2.15 and who shall ensure that a "NLO Training Addendum Sheet" shall be completed and attached to each affected checklis The licensee has been informed of the inspector's concerns and this item will remain open pending future NRC evaluation of additional corrective actions, d. (Closed) Violation (461/86060-02): Corrective actions in response to IPQA Audit Q38-86-10 and IPQA Surveillar.ce Finding M-86-005 were not effective to prevent recurrenc The licensee had identified deficiencies in the processing of Maintenance Work Requests (MWRs)

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This item was previously inspected as documented in Inspection Report 50-461/87002, paragraph 2.k. At that time, this item remained open pending the formal change to the MWR Preparation and Routing Procedure, CPS No. 1029.01. During this report period, the inspector reviewed revision 17 of CPS No. 1029.01. The formalized changes had been incorporated in section 8.2 of that procedur This item is closed, (Closed) Violation (461/86065-03): Procedure CPS No. 1016.01, CPS Condition Reports, was not followed in that corrective action plans were not approved prior to implementation; block 2 of the condition report form was not always filled out; and reviews of the condition reports did not identify and correct the violations that existe This item was previously inspected as documented in Inspection Report 50-461/87002, paragraph 2.1. At that time, this item remained open pending completion of the actions discussed in the licensee's response to the violation, IP letter U-600806, Attachment A, paragraph I During this report period, the inspector reviewed revision 16, of CPS No. 1016.0 Paragraph 8.5.2.1 was revised to require approval of the appropriate department head for the written corrective action plan before corrective action was to be implemented. In addition, IP memorandum JWW-2425-87 was issued on April 10, 1987, instructing appropriate implementing procedures in use be revised to correspond with the requirements of revision 16 to CPS No. 1016.01. The actions discussed in IP letter U-600806, Attachment A, paragraph II.a. had been completed. This item is close (0 pen) Violation (461/87006-02(DRS)): Failure to prepare Emergency Operating Procedures (EOPs) in accordance with the Plant-Specific Technical Guideline (P-STG) and control changes thereto. The inspector reviewed the E0Ps and verified that the NRC identified safety significant deficiencies had been corrected. The inspector also reviewed the results of the licensee's " tabletop" reverifica-tion of the E0Ps. Additional potentially safety significant deficiencies in the E0Ps identified by this reverification had also been corrected. All E0Ps are now technically correct and adequate; however, this item will remain open pending completion of required revisions to the P-STG identified during the NRC inspection and the licensee's " tabletop" reverificatio (Closed)UnresolvedItem(461/87006-03(DRS)): Potential unreviewed i

safety question. The inspector reviewed the licensee's safety evaluation assessing the impact of changing an entry condition in the " secondary containment control" emergency procedure. The inspector agreed that no unreviewed safety question was generate This item is close (Closed) Unresolved Item (461/87006-05 (DRS)): Licensee identified problems associated with calculations supporting the E0Ps. Seven calculation problems had been identified. The specific problems and resolutions are listed below:

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e (1) The calculated value for indicated water level and range for fuel zone 1 and 2 was different than that given in the P-STG and the values in the E0Ps did not match either set of value The problem involved a change in scale reference level between the calculation results and the P-STG and a typographical error in the P-STG. The values used in the E0Ps were correct. The typographical error in the P-STG will be corrected at the next updat (2) The graph for the suppression pool load limit given in the P-STG and E0Ps did not accurately reflect the results of the supporting calculations. The inspector verified that the graphs in the E0Ps had been revised to accurately reflect the results of the load limit calculation (3) The cold shutdown boron weight calculations did not correspond to values given in the Technical Specifications. These differences resulted from the fact that the E0P calculations used Revision 1 of the applicable design specification and the calculations supporting the Technical Specifications used Revision 2. The E0P calculations were revised using Revision 2 of the design specifications. Since the values in the E0Ps based on the original calculations were conservative, no immediate changes to the E0Ps were require (4) The boron injection initiation temperature was calculated using rated flow rather than minimum flow for the standby liquid control pumps as used in the Technical Specifications. The calculation was revised using minimum pump flo There was no impact on the E0P (5) The graph of " Primary Containment Pressure Limit" given in the E0Ps did not match the results of the calculation. The graph presented in the E0Ps was based on a revision submitted to the NRC as part of the Post-Accident-Containment-Venting Plan. The graph in the E0Ps was correc (6) There were two calculations labelled as "C20.0.". One gave results in terms of a graph of " Minimum Core Flooding Interval" and the other in a curve of " Maximum Core Uncovery Time". The E0Ps used the correct versio (7) The flow stagnation water level given in the P-STG was different from that given in the supporting calculatio The difference had no impact on the E0P The difference resulted from a slight difference (0.71 inches) in the zero elevation of

. the fuel zone level indicators given in different reference The calculation issues had been adequately addressed and this item is close No violations or deviations were identifie l

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. . Review of Allegations (99014)

(Closed) Allegation (RIII-86-A-0166): On October 27, 1986, Region III submitted the following concerns to IP for their review and followup. On December 11, 1986, IP notified Region III by letter U-600779 that their review and followup was completed. The inspector reviewed IP's response to the concerns as documented belo '

Concern #1 The cable installed on the containment hygrometer was not installed in accordance with the manufacturer's specification Review The inspector reviewed the licensee's letter; reviewed the Model 4000 Moisture Computer Operating and Service Manual; reviewed Plant Nodifications CM-20 and Engineering Change Notice (ECN) 7726; reviewed portions of the Clinton Master Equipment List; and perfonned an inspection of the termination of the hygrometer cable The inspector determined the following: The licensee identified that the hygrometer cables were not installed in accordance with the manufacturer's specification Specifically, the two conductors associated with the moisture probes did not have separate shields as specified by the manufacturer. In addition, both ends of the probe shields were grounded in lieu of the requirement that only one end be grounde Subsequent to receipt of the allegation, the licensee had initiated plant modification CM-020 and ECN 7726 to correct the cable installation deficiencie Rework activities to correct the hygrometer cable deficiencies were completed by the licensee for each of the hygrometer moisture probe The drywell hygrometer is identified in the Clinton Master Equipment Index as being non-safety related. The inspector discussed the hygrometer safety classification with the Plant Operations Supervisor and determined that the drywell hygrometer serves no safety function. From routine NRC observation of safety related cable installation, there is no reason to believe that these non-safety cable installation deficiencies may reflect on the adequacy of safety grade cable installations, The inspector observed the current hygrometer cable terminations and identified no deficiencie Conclusion The allegation was substantiated; however, the licensee had acted to correct the hygrometer cable deficiencies. Although substantiated, this concern does not have safety significance in that the drywell hygrometer is a non-safety related component. This allegation is close . - _ - - . . .. ._ - _ . - .

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Concern #2

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Hygrometers installed at other plants which were similar to the Clinton hygrometer were replaced due to unspecified problems. The Clinton hygrometer was not replace Review

The inspector reviewed the licensee's letter and information relative to the licensee's review of the Operating Plant Experience and Operations and Maintenance Information Exchange data bases of the Institute of Nuclear Power Operations (INP0) Nuclear Networ The inspector determined that the INP0 Nuclear Network information ,

solicited by the licensee did not identify any generic problems with i Panametrics Model 4000 hygrometer Conclusion The allegation was not substantiated in that there is no available data which identifies generic problems with the Clinton hygrometer. Also, since the hygrometer is not a safety related component, there is no bases for a safety concern. This allegation is close Concern #3 The hygrometer installed at Clinton was never calibrate i Review The inspector reviewed the licensee's letter; reviewed Special Test Procedure XTP-CM-03, Revision 0; reviewed Preoperational Test Procedure PTP-CM-01; and discussed the results of these reviews with the licensee's startup departmen The inspector determined the following: The drywell hygrometer was calibrated on September 4, 1984 and February 8,1986, as directed under XTP-CM-03, "Drywell Air Humidity

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Analyzer Calibration". An additional partial calibration of the hygrometer was completed in January 1987 during performance of Preoperational Test Procedure PTP-CM-0 Conclusion The allegation was not substantiated in that the drywell hygrometer had been calibrated by the licensee. This allegation is closed.

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No violations or deviations were identifie . Employee Concerns (99014)

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The inspector reviewed concerns expressed by site personnel from time to time throughout the inspection period. Those concerns related to regulated activities were documented by the inspector and submitted to Region III. One concern was transmitted to the regional office during this report perio No violations or deviations were identifie . Licensee Action on Three Mile Island (TMI) Action Plan Requirements (25401)

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The NRC Office of Inspection and Enforcement issued Temporary Instruction (TI)2514/01, Revision 2, dated December 15, 1980, to supplement the Inspection and Enforcement Manual. The TI provides TMI-related inspection requirements for operating license applicants during the phase between pre-licensing and licensing for full power operation. Part I lists requirements that were closed prior to fuel load. Part 2 lists requirements that must be closed prior to full power operation. Part 2 of the TI was used as the basis for inspection of the following TMI item

, found in NUREG-0737, " Clarification of TMI Action Plan Requirements".

(0 pen) Item I.G.1: Training During Low Power Testin During this report period, the inspector reviewed the licensee's program for compliance to the subject TMI Action Plan. Nuclear Training Department (NTD) procedure 2.7, "NTD Startup Initial Test Program Training", revision 0 (through ACN 1/3 dated May 5, 1987)

dated January 2,1986, described the training for licensed personnel during the Startup Test progra The inspector noted that the program described by the above procedure detailed the training requirements, required participants, and training records to be used. Responsibilities for assuring the implementation of this training program had also been included in the procedure. This item will remain open pending the inspector's review of the completed trainin No violations or deviations were identifie . IE Information Notice Followup (92701)

(Closed) Information Notice (461/87008-NN): On February 4, 1987, the

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NRC Office of Inspection and Enforcement issued Information Notice

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No. 87-08. The subject of that notice was degraded motor leads in Limitorque DC motor operators. The inspector reviewed the actions taken by the licensee in response to IE Information Notice No. 87-08.

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As described in IP memorandum Y-83548, dated February 26, 1987, the licensee identified three Class 1E valves that used the DC motors described in Information Notice 87-08. The three affected valves (1E51-F019, -F046, -F095) were in the Reactor Core Isolation Cooling (RCIC) system. The licensee initiated Condition Report (CR) 1-87-02-152 to document the potentially defective DC motors and to provide corrective actio Maintenance Work Request (MWR) C33400 was initiated to remove the DC motors and ship to an offsite vendor for repair. This action was completed prior to the licensee's initial entry into mode 2 on February 26, 1987. In addition, the licensee reviewed the spare parts inventory to assure no additional defective DC motors, as described in Information Notice 87-08, were on sit No violations or deviations were 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 inspectors performed an in-office review of each LER to determine that 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 determined appropriate, the LER was scheduled for onsite followup inspection or other necessary action by cognizant NRC personne (1) (Closed) LER 87-005-00 (461/87005-LL) [ ENS No. 07564]:

Automatic Actuation of Containment Isolation Valves 11A006 and IIA 007 Due to Control and Instrumentation Technician Erro The licensee identified the cause of this event to be personnel error. Two leads being landed on the same termination point during performance of a surveillance resulted in the closure of valves IIA 006 and IIA 007 when the trip signal was inserte The licensee revised procedure CPS No. 9030.01C016, Step 5. and Plant Manager's Standing Order (PMS0) No. 030 to provide adequate instruction to prevent recurrence of this type of event. This LER is close (2) (0 pen) LER 87-006-00 (461/87006-LL) [ ENS No. 07757]: Partial Group I Containment Isolation Due to Blown Fuse on Circuit Card in Containment Isolation Logi The licensee identified the cause of this event to be a random card failure (blown fuse) in the containment isolation logic that resulted in closure of 1821-F016 during performance of

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procedure CPS No. 9431.58, Main Steam Line High Radiation Functional Test. The blown fuse was replaced and containment isolation logic was reset and valve 1821-F016 was reopene There has not been any previous occurrence of this type of event. At the conclusion of this report period, the licensee informed the inspector that a supplemental report was to be issued concerning the event. This item will remain open pending the inspector's review of that supplemental repor (3) (Closed) LER 87-007-00 (461/87007-LL) [ ENS No. 07628]:

Automatic Actuation of Safety Relief Valve (SRV) B21-F041F Due to Electrical Ground in Control Circuitr The licensee identified the cause of this event to be a 10 kilo-ohm short on the upstrean side of actuation circuit for SRV B21F041F. After the repair of a downstream ground, the circuit fuse was inserted which resulted in the SRV actuation due to the 10 kilo-ohm short. There was no evidence of procedural or personnel error in the maintenance activity being performed. This LER is closed.

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(4) (Closed) LER 87-009-00 (461/87009-LL) [ ENS No. 07976]: Manual i Actuation of the Standby Gas Treatment System Train "A" Due to Fuel Building Ventilation System Damper Solenoid Failur The licensee identified the cause of this event to be a fuel building supply outboard isolation damper solenoid venting air while energized. To maintain the technical specification limit for secondary containment differential pressure, the licensee manually started the Standby Gas Treatment system. The

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isolation damper was repaired and returned to operation. This LER is close (5) (Closed) LER-87-010-00 (461/87010-LL) [ ENS No. 07990]: Main Control Room Ventilation System Shifted to the High Chlorine Mode Due to Trip Setpoint Being Exceeded by the Cumulative j Effects of Component Operating Parameter Investigation by the licensee determined that the root cause of the two ventilation system shifts that occurred, one each in Trains' "A" and "B" were due to a discolored chlorine sensitive tape and an electrical transient in the alarm circuitry, respectively, in combination with minor variations

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in operating characteristics such as setpoint drift, line noise, tape surface texture variations, and photocell sensitivity which probably caused the detector output to

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fluctuate near the 0.4 ppm alarm setpoint. A plant modifi-cation was initiated to increase the alarm setpoint of the chlorine monitors to 1.7 ppm to prevent these spurious

actuations. This LER is closed.

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(6) (Closed) LER-87-011-00 (461/87011-LL) [ ENS No. 08002]: Main Control Room Ventilation System = Shifted to High Chlorine Mode Due to Trip Setpoint Being Exceeded by the Cumulative Effects of Component Operating Parameter Investigation by the licensee determined that the two Train "B" ventilation system shifts that occurred were attributable to buildup of dust on the chlorine sensitive tape and a higher than specified detector air flow, respectively. These effects in combination with the variations described in Item (5) above are believed to be the root cause. Also see Item (5) for the licensee's corrective actio Since these events, the licensee has placed the main control room ventilation system in the recirculation operational mode to prevent spurious isolation of this system into the high chlorine mod In addition, in a Safety Evaluation dated April 17, 1987, issued by NRR with the Clinton full power license, the staff approved a technical specification change requested by the licensee in a letter dated March 20, 1987, that provides alternate methods for use and storage of chlorine gas on site. This technical specification change only requires automatic shifting of the control room HVAC to the high chlorine mode when onsite chlorine is stored in containers with a capacity greater than 150 pounds or at distances less than 100 meters from the nearest control room intake. This LER is close (7) (Closed) LER-87-013-00 (461/87013-LL) [ ENS No.s 08006, 08122, and 08126]: Automatic Isolation of Reactor Water Cleanup System Due to High Differential Flow Signa Automatic isolation of the reactor water cleanup system (RWCU)

due to a high differential flow signal occurred five times between March 10, 1987 and March 31, 1987. The licensee's investigation of these events determined that the cause of these isolations was the generation of a false high differential flow trip signal in the RWCU system at very low (approximately 300 gpm) flow rates when the return flow from the RWCU is directed to the condenser due to thermal consider-ations in the feedwater lin In this lineup the return line from the RWCU system to the feedwater line was isolated (no flow from the RWCU system was passing through the feedwater line); however, the flow sensing instrument was not electrically isolated from the RWCU system flow summing circuitry. This results in false RWCU system flow signals being generated when reactor vessel feedwater flowrate adjustments or changes are made in this configuration. The licensee's corrective action, to preclude recurrence of these isolation events, was to remove the RWCU system from service during startup of the feedwater system and during periods of very low feedwater flowrate This LER is close .. ..

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(8) (0 pen) LER-87-014-00 (461/87014-LL) [ ENS No. 08050]: Automatic Actuation of the High Pressure Core Spray System (HPCS) Due to Utility Control and Instrumentation Technician Erro The licensee's investigation of this event determined that the cause was due to a C&I technician not notifying appropriate management and supervisory personnel when the level instrument the technician was calibrating could not be rettored to operation utilizing the procedure (CPS No. 8801.12) being use The technican restored the level instrument ylthout procedural guidance, which resulted in a hydraulic surc,e causing actuation of the HPCS syste The C&I technician was counselled regarding procedural compliance and maintenance department training will be provided that emphasizes the lessons learned froin this event. The licensee revised the procedure to minimize the possibility of recurrence of this event. The Senior Resident Inspector (SRI) will verify this training has been provided and procedure CPS No. 8801.12 has been revised and provide the results in a subsequent report. This LER remains ope (9) (0 pen) LER-87-019-00 (461/87019-LL) [ ENS No. 08153]: Automatic Actuation of Safety Relief Valve 1821-F041C Due to Failed Load Driver Circuit Car This event and the effects of load driver card failures in other circuits on system operational and plant safety were discussed in Inspection Reports 50-461/87011 and 87013. The Region III inspectors and the NRR staff have completed a preliminary review of the analyses provided by the licensee in letters dated April 9 and April 18, 1987, and the corrective actions taken to prevent spurious actuation of the ADS valves during surveillances due to a load driver card failure and determined that the actions are acceptable. This LER will remainopenpendingfinalrepewbyNR (10) (0 pen) LER-87-020-00 (461/87020-LL) [ ENS No. 8197]: Automatic Isolation of Reactor Water Cleanup System Due to High Differential Flow Signal as a Result of Operator Erro Investigation by the licensee determined the cause of the

Reactor Water Cleanup system isolation was due to an operator i

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not properly verifying the pressure had been equalized around Main Steam Isolation Valve (MSIV) "C" before opening the valv The licensee provided training for the operators on equalizing pressure aryund and opening the MSIVs and revised procedure CPS No. 3002.01rto clarify the pressure equalization process. The

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SRI will verify the licensee's corrective actions and provide the results in a subsequent inspection report. This LER remains ope ., .- _ - - - . - .

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(11) (0 pen) LER-87-021-00 (461/87021-LL) [ ENS No. 08207]: Automatic Isolation of Reactor Core Isolation Cooling (RCIC) System Due to Utility Personnel Erro The licensee's investigation determined the cause of this event was due to a combination of the following: The procedure did not specify the RCIC would isolate when a surveillance trip signal was inserted; the surveillance procedure was previously only performed with the RCIC isolated; the surveillance Impact Matrix required by a Plant Manager's Standing Order No. 30 was not completed prior to performance of the surveillance (this resulted in a violation that was documented in Inspection Report 461/87011); and the Line Assistant Shift Supervisor did not realize that the Impact Matrix had not been completed prior to providing authorization to perform the surveillance. The licensee revised procedure CPS No. 9030.01C035 to provide for removing the RCIC system from service prior to performing the surveillance. The licensee is also conducting a review to determine if there are other ESF systems which trip or isolate on a one-out-of-one logic and will review and modify applicable surveillance procedures, if necessary, to preclude similar actuations. The corrective actions to the violation will be addressed in a subsequent inspection report. This LER remains open No violations or deviations were identifie b. Onsite Followup Of Written Reports Of Nonroutine Events At Power Reactor Facilities (92700)

For the LER listed below, the inspector performed an onsite followup inspection of the LER to determine whether response to the event was adequate and met regulatory requirements, license conditions, commitments, and to determine whether the licensee had taken corrective actions as stated in the LE (1) (Closed) LER 87-008-00 (461/87008-LL) [ ENS Nos. 08316, 08328, 08330,08332]: Violation of Technical Specification Resulting from Surveillance Procedure Error Due to Utility Personnel Erro As discussed below in paragraph 13.b.(4), the inspector verified the corrective actions as stated in this LER were completed. The violation of technical specifications resulted from an inadequate procedure review. The licensee performed a review of Technical Specification instrumentation requirements against existing surveillance procedures and confirmed that the error was limited to one surveillance procedure. 8ased on the inspector's review of this LER and the followup discussed below in paragraph 13.b.(4), this LER is close No violations or deviations were identifie t

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8. Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with technical specification The following items were considered during this review: 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; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemente The inspector observed the maintenance performed to repair containment isolation valve CM-023 in accordance with Maintenance Work Request C49771. The valve could not be opened and repairs were performed by electrical maintenance mechanics. During the work the following problems were observed: The troubleshooting mechanics identified a faulty solenoid coil by resistance readings. These readings were not effective because a rectifier in the circuit would have indicated an open circuit in any cas The solenoid coil was replaced and subsequent checks found a Conax conductor shorted to the housing and melted through. A root cause for the Conax short is being investigated by the license During reconnection of the valve electrical leads, it was discovered that mechanics had installed wrong size lugs on the Conax lead The junction box leads could not be connected because the screws would not fit through the lugs. Material controls ensure that correct part numbers are issued but no other controls govern installation. Two sizes of lugs were being used in the work area and were inadvertantly exchanged. The licensee's review found that lugs of different sizes were mixed in shop stores and Condition Report 1-87-04-178 was written to resolve thi The licensee conducted a critique of the problems encountered during the performance of the above maintenance activity. IP memorandum MJM-0032-87, dated April 27, 1987, described the problems and corrective actions that were taken. The inspector reviewed the critique writeup and concluded that problems identified during the conduct of this maintenance activity had been properly addressed by the license No violations or deviations were identifie _ _ _ . _ _ _ _ _ - - .- _ _.___ . _ _ - _ . -_ -. .. _ _ _ - _.

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. + Surveillance (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 requirement Items which were considered during the inspection included whether adequate procedures were used to perform the testing, test instrumenta-tion 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 personne The inspectors observed / reviewed the following activitie Surveillance / Test

Procedure N Activity

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CPS No. 9031.11 APRM Flow Biased Power / Flow Channel Check

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CPS No. 9432.18 RWCU Ventilation Delta Temperature Channel j Functional / Calibration CPS No. 9431.58 Main Steam Line Radiation Channel Functional

! "C" & "D" CPS No. 9432.13 RCIC Equipment Area Temperature E-31 l Channel / Calibration Functional 1 CPS No. 6902.01 Off Gas Hydrogen Concentration Determination

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CPS No. 3322.01 Traversing Incore Probe (TIP)

CPS No. 9030.01 Residual Heat Removal Minimum Flow Channel Functional CPS No. 9080.02 Diesel Generater 1C Operability l No violations or deviations were identifie . 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 inspectors verified the operability of selected emergency systems and verified tracking of LCOs. Routine tours of the auxiliary, fuel, containment,

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control, diesel generator, and turbine buildings and the screenhouse were conducted to observe plant equipment conditions including potential for

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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 implemente During the approach to power operation (Mode 1) the licensee attempted a channel check of the Average Power Range Monitor (APRM)

as required by Technical Specification (TS) 4.3.1. The surveillance was for Functional Unit 2.b. " Flow-Biased Simulated Thermal Power -

High" of TS Table 4.3.1.1-1 requiring a daily channel check to verify that measured core flow is greater than or equal to established core flow at the existing loop flow control (APRM %

flow). The channel check .'esults did not meet this requirement and further review by the licensee determined that results of this surveillance wotid not be meaningful until the " established" core flow data was determined during startup testing. To resolve the performance of this surveillance during startup testing, the licensee prepared a Licensing Document Interpretation (LDI)

documenting their intent to take data during startup and to conservatively set the APRM Flow-Biased Simulated Thermal Power -

High Power during the low power testing. (During the low power testing, flow will still be too low for meaningful data).

Additional reactor protection was provided by the high neutron flux rate trip setpoint which is set in accordance with Regulatory Guide 1.68, Section 4.d, Appendix C at a value no greater than 20% beyond the power of the next level bounding the associated startup test condition. For example, for Test Condition 1, defined between 5 and 20% power, the set point would be no greater than 40%. The LDI was discussed with NRC staff and was found acceptable, During the report period, the inspector discussed the licensee's disposition of a condition report that addressed a non-conforming condition that existed on the Division 2 Residual Heat Removal (RHR)

water leg pump motor. Clinton Power Station Nonconforming Material Report (NCMR) 2-1039 was initiated on December 23, 1986, to document an identified discrepancy on motor leads to RHR water leg pump 1E12-C003. Subsequent evaluation of the condition described in NCMR 2-1039 resulted in Condition Report (CR) 1-87-03-014 dated March 2, 198 The condition description block of CR 1-87-03-014 stated that the Class 1E motor on water leg pump 1E12-C003 no longer met its Environmental Qualification because of the damaged motor lead The original disposition of the identified condition determined that the motor should be reworked before exceeding 5% powe Nuclear Station Engineering Department revised the disposition of CR 1-87-03-014 when the original rework was not performed prior to 5% powe _ _ - _ -

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The revised disposition justified the continued operability of the RHR water leg pump beyond the original 5% power limit based on:

1) the water leg pump was operating properly; 2) the water leg pump low discharge pressure annunciator was not actuated and it would alarm indicating pump failure and; 3) technical specifications do not address operability of the water leg pum The RHR water leg pump was reworked during a scheduled maintenance outage the week of May 4,1987. However, the inspector discussed with cognizant licensee personnel apparent discrepancies with the licensee's resolution of this issu In particular, the statement in the revised disposition to CR 1-87-03-014 that " Technical Specifications do not address operability of the water leg pump" was in error. The definition of " Operability" is contained in CPS Technical Specification 1.27. The CPS Final Safety Analysis Report (FSAR), Section 6.3.2.2.5 clearly defined the support function of the RHR water leg pump was to keep this ECCS system discharge header piping filled to prevent water hammer effects during system initiatio The inspector also discussed the lack of a clear written justification for continued operation during the interval of initial discovery of the nonconforming condition (December 23, 1986) and the first disposition provided on CR 1-87-03-014 (March 2, 1987). The inspector discussed the guidelines contained in Generic Letter 86-15 relating to compliance with 10 CFR 50.49 and requested the licensee to conduct further evaluations of their corrective actions to this issue. This will remain an unresolved item pending additional information from the licensee (461/87015-01). Throughout the inspection period, the inspectors performed off shift inspections (i.e., swing shift and midnight shift observations).

During those inspections, the control room operators were observed to be alert and attentive to plant operations. No unnecessary reading material was found in the control roo One unresolved item was identifie . Modification Testing (72701)

A review of modification packages NR-2 and NR-7 was performed. Th :e packages were reviewed for proper documentation of the work performed; that work was accomplished using approved procedures; and that post modification tests were appropriately approved, performed, and evaluate The review of the package for modification NR-7 found Condition Report (CR)1-87-01-038 was initiated by the licensee documenting that a review of the post modification testing, which was an excerpt from existing test procedure CPS No. 2830.11 "CRD/RPS Investigation on IRM Spiking", was not performed in accordance with 10CFR 50.59. The modification reduced the amplification of neutron monitors in an attempt to eliminate reactor trips due to Intermediate Range Monitor spikes reported in Licensee Event Report 86017 revision 00, 01, 02, and 03. The post-modification test

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used individual control rod scramming to reproduce the condition during which spiking was originally observed. The existing test procedure (CPS No. 2830.11) was changed so that only the prerequisites and steps involving specific rod scrams would be performed. The modification and use of this test was done twice by the same individual within a short period. Technical Specification 6.8.3 required that temporary changes to procedures which affect plant nuclear safety be reviewed as required by 10 CFR 50.59 and approved by unit management staff. However, the licensee's engineer initiating the test change failed to process the temporary change form required by administrative procedures resulting in the omission of the required 50.59 review and approval The licensee identified this problem after the test was performed and correctly completed the review and approval process which found that the test change did not involve an unreviewed safety question. A memo was sent to all Technical Department engineers redirecting conformance to procedure change requirements. As part of the Condition Report, a review of reportability was performed which concluded that a Licensee Event Report (LER) was not require The inspector determined that Technical Specification 6.8.3 was violated; however, the licensee met the requirements of 10 CFR2, Appendix B, Paragraph W in that; (1) the violation was identified by the licensee; (2) the violation fits in Severity Level IV of Supplement I (failure to meet regulatory requirements that have more than minor safety significance); (3) it was reported, if required (was not required by 10 CFR 50.72); (4) the violation was corrected by performing the appropriate safety reviews; and (5) the violation could not have been prevented by the licensee's corrective action, since there had not been any previous violations of this nature. As discussed in 10 CFR 2, Appendix B, the NRC will not generally issue a notice of violation for violations that meet the above requirements to encourage and support licensee initiative for self-identification and correction of problems."

Therefore, a notice of violation will not be issue No violation or deviations were identifie . Training (41400 & 41701)

The effectiveness of training programs for licensed and nonlicensed personnel were reviewed by the inspectors during the witnessing of the licensee's performance of routine surveillance, maintenance, and operational activities and during the review of the licensee's response to events which occurred during the month of April 1987. Personnel appeared to be knowledgeable of the tasks being performed, and nothing was observed which indicated any ineffectiveness of trainin No violations or deviations were identifie . . . _ -- . . . . - - - . . - . - =_ - _ - - - . -- . - . -

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13. Onsite followup of Events at Operating Reactors (93702)

I General t

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 actions; and

, interviews of licensee personnel. For each event, the inspector

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reviewed one or more of the following: the sequence 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 even Additionally, in some cases, the inspector verified that licensee investigation had identified root causes of equipment malfunctions i

and/or personnel errors and were taking or had taken appropriate

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corrective actions. Details of the events and licensee corrective i actions noted during the inspector's followup are provided in j paragraph b. below.

) Details

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(1) Loss of Annunciators i

At approximately 7:20 p.m. CST on April 5,1987, while at 4%

power, approximately 75 annunciator windows activated. After i acknowledgement and reset, approximately 15 remained

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illuminated. The remaining 60 windows were black and not operational. The licensee determined by reviewing plant  :

i parameters that the annunciator signals were not valid. The

-circuits involved were on the ECCS and reactor console ;

l The licensee attempted to isolate the fault by positioning '

, sectionalizing switches without success. The fault was removed after a faulted ground detector card was removed. The faulted

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card was replaced. An additional fault was found in an j optical card supplying one annunciator while restoring the

sectionalizing switches. This fault was cleared by lifting a j lead to the optical isolation circui During the event reactor power level was maintained constan ,

Information was available for the affected circuits by review of SPDS, Data Process Computer, Indicators, Recorders, and

equipment status lights. The system was restored to normal

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with the exception of one annunciator by 2:00 a.m. on April 6, t

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Additional inspection of the above event was performed and i the results documented in Inspection Report 50-461/87013, paragraph i

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(2) Manual Isolation of Reactor Core Isolation Cooling System

[ ENS No. 08290]

At about 3:30 p.m. CST on April 6,1987, the licensee manually isolated the Reactor Core Isolation Cooling (RCIC) system in response to a High Differential Temperature Alarm on the RCIC room cooler. The licensee identified a packing leak on RCIC valve E51-F095 and elected to manually close the system's containment isolation valves (E51-F063 & F064). The reactor plant was in mode 2 at about 2% power. The licensee notified the NRC Operations Center of this event via the ENS at about 6:40 p.m. CST on April 6, 198 (3) ESF Actuation Due To Auto Start of High Pressure Core Spray

[ ENS No. 08296]

On April 7, 1987, the licensee experienced an ESF actuation when the Division 3 High Pressure Core Spray (HPCS) system started. During the performance of a routine surveillance, C&I technicians attempted to vent a process instrument line through a level transmitter (B21-N080D). The instrument being vented shared the process line with two Division 4 level transmitters (1821-N073H & N073D) that actuated the HPCS low 1 level - level 3 logic when they sensed a hydraulic transient during the venting process. Control room opeiItors verified that the reactor water level was within the acrmal operating band and secured the HPCS pump. Actual pump 'un time was 13 seconds. After the Division 3 diesel generatur was verified to be properly running, it was also secured and the system i

was returned to a standby status. The reactor plant was in mode 2, approximately 2% power during this event. The licensee

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notified the NRC Operations Center of this event via the ENS at 7:50 p.m. CST on April 7, 1987.

i (4) Violation of Operating License [ENSNo.08316,08328,08330, and 08332J

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On April 8, 1987, the licensee discovered during investigation of Condition Report (CR) 1-87-03-031, that the requirements for Technical Specification Table 4.3.2.1-1. Item 1.b, were not being met by Surveillance Procedure CPS No. 9000 'll, " Control Room Surveillance Log". Procedurestep8.6.10.1.$,was

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performing the technical specification required channel check for the containment high pressure isolation on Residual Heat Removal (RHR) system instruments (ATM 1E12-NH2 A, B, C & D)

instead of the containment system (CRVICS) instruments VG-145 and VG-14 The licensee investigated the occurrence and determined that the procedure CPS No. 9000.01, was in error. A procedure change request (PDR 87-0681) and CR-1-87-04-050 were issued to correct the inadequate procedure. Additionally, the licensee performed the correct channel check and verified that the instruments were operable. The inspector reviewed this

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occurrence with the licensee and determined that all technical specification required actions had been taken (technical specifications required that the affected valves were to be closed within one hour), since the affected valves had been closed and tagged prior to initial fuel load (September 29, 1986). However, the inspector determined that the technical specifications had been violated since the licensee entered Operating Mode 2 (initial criticality) on February 26, 1987, without performing the required channel check Because of the potential importance of the violation, Region III requested the Ifcensee to conduct a 100% investigation of all surveillance requirements to determine any additional or pending technical specification violations resulting from inadequate surveillance procedures. The licensee completed their investigation on April 10, 1987, and determined that additional violations existed (reference IP letter U-600905).

Listed below are the violations and the instruments that were not channel checke TECH. SPEC. TABLE VIOLATIONS FUNCTION INSTRUMENT 4.3.2.1-1 ITEM CONTAINMENT PRESSURE H!SH VG 145 & 147 4.3.2.1-1 ITEM REACTOR VESSEL WATER LEVEL 1821-N692 4.3.5.1-1 ITEM a LOW-LOW LEVEL 2 A, B, E & F 4.3.2.1-1 ITEM REACTOR VESSEL PRESSURE-HIGH 1821-N679 (RHR CUT-IN PERMISSIVE 135#) A, B, C & D 4.3.2.1-1 ITEM 3.c & d (c) EQUIPMENT AREA TEMP-flGH E31-N618 A&B (d) EQUIPMENT AREAS DELTF T E31-N626 A&B In addition, the licensee discovere d that procedure CPS N .01 was also inadequate in chasnel checking the Reactor Vessel Water Level 8 isolation of the Feedwater and Turbine Systems. This was not a violation of Technical Specification 4.3.9.1-1, Item 2.a because this i ntation feature was only required for Operational Condi; ion 1 and as such was not applicable for the plant's operatioral statu The inspector reviewed the licensee's conclusions and agreed that all of the violations perta*ned only to channel checks (channel functional and calibrations were not affected) and that all of the errors were limited only to surveillance procedure CPS No. 9000.01. The inspector also verified that all of the affected instruments were correctly channel checked satisfactorily on April 9, 1987, and that the licensee had made the required notification ._-

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As discussed above in paragraph 7.b., Licensee Event Report (LER)87-008-00 dated May 7, 1987, described the above event, detailed the root cause of this event and the corrective action taken by the licensee. The inspector reviewed CPS N .010001, " Control Room Operator Surveillance Log - Mode 1, 2, 3 Data Sheet", revision 24, dated May 13, 1987, and CPS No. 9000.02D001, " Unit Attendant Surveillance Log Data Sheet",

revision 23, dated April 30, 1987. These procedures had been revised to include the required channel check Technical Specification 4.0.4 states: " Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement (s)

associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified." Operational Condition 2 was entered on February 26, 1987 without performing all of the surveillance requirements (channel checks) associated with the following Limiting Conditions of Operations; (1) LC0 3.3.2 for CRVICS isolations on containment high pressure and reactor low-low water level (instruments VG-145, VG-147 (containment isolation)

and 1821-N692 A, B, E & F (RCIC isolation)); (2) LC0 3. Reactor Core Isolation Cooling (RCIC) system isolation (instruments 1821-N692 A, B, E & F); (3) LC0 3.3.2 vessel high pressure permissive for RHR shutdown cooling (instruments 1821-N679 A, B, C and D; and (4) LC0 3.3.2 Reactor Water Cleanup (RWCU) system isolations on area high temperature and high differential temperature instruments (E31-N618 A and 8 and E31-N626 A and B). Failure to perform the above is a violation of Technical Specification 4.0.4 (461/87015-02).

Technical Specification 6.8.1.d states: " Written procedures shall be established for surveillance activities of safety-related equipment." Written procedures were not established for Technical Specification surveillances 4.3.2.1-1 (Items 1.k. 1.b, 5.e, 3.c, and 3.d), 4.3.5.1-1 (Item a), and 4.3.9.1-1 (Item 2.a). Failure to establish required written procedures is a violation of Technical Specification 6.8. (461/87015-03).

As discussed above, the licensee conducted an investigation to determine the extent of these violations. The initial results of that investigation were documented in IP letter U-600905 dated April 10, 1987. LER 87-008-00, dated May 7, 1987, provided the licensee's sumary of this event and described the root cause and corrective action taken. The inspector's review of these correspondence and verification that corrective action had been completed indicated that no further response from the licensee was necessary. These items (461/87015-02 and 461/87015-03) are closed.

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(5) ESF Actuation [ ENS No. 08349]

At about 7:30 p.m. CST on April 11, 1987, the licensee identified that an unexpected ESF actuation had occurred earlier that day at about 1:00 p.m.. During the performance of a routine surveillance for calibration of process radiation monitors, the procedure in use required the installation of a temporary jumper. When this action was performed, the division 1 H202 containment isolation valves closed. The licensee declared the division inoperable while they evaluated the cause for the unexpected closure. At the time of this event, the ;

reactor plant was less than 2% power in mode 2 (startup). The licensee notified the NRC Operations Center of this event at 10:00 p.m. CST on April 11, 198 (6) ESF Actuation [ ENS No. 08426]

At about 8:00 p.m. CDT on April 19, 1987, the licensee experienced an automatic actuation of containment isolation valves in the division 2 H202 monitoring system. The isolation occurred when Process Radiation Monitor (PRM) 1RIX-PR001D failed low causing the associated containment isolation valves to actuate closed. The licensee discovered the failed PRM at about 1:00 a.m. on April 20, 1987. At the time of the event, the reactor plant was in mode 2 (startup) at about 4% powe The licensee had left the division 2 H202 monitor isolated while troubleshooting the PRM failure. The licensee notified the NRC Operations Center of this event via the ENS at about 2:00 a.m. on April 20, 198 (7) Minor Chlorine Gas Leak [ ENS No. 08430]

At about 7:30 a.m. CDT on April 20, 1987, the licensee identified a chlorine gas leak had occurred at their sewage treatment plant outside the protected area. While performing shiftly rounds, a radwaste operator identified and immediately isolated a leaking chlorine bottle at the sewage treatment plant. The licensee estimated that between 10 and 150 pounds of chlorine gas was released. In accordance with plant procedures, the licensee notified federal and state agencies of this release. The reactor plant was in mode 2 (startup)

at about 3% power. The licensee notified the NRC Operations Center of this event via the ENS at about 11:00 a.m. on April 20, 198 (8) ESF Actuation [ ENS No. 08542]

At about 5:15 a.m. CDT on May 1, 1987, the licensee experienced an ESF actuation when the Reactor Water Cleanup System (RT)

isolated en high differential flow. At the time of event

, occurrence, the licensee was manipulating the RT system flow path to allow letdown flow to the main condenser. The licensee

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determined the high differential flow signal resulted from the system lineup being performed. The RT system was restored to its normal lineup at about 6:00 a.m. The Senior Resident Inspector was present in the control room during this event and noted the operators responded in accordance with plant procedures. At the time of this event, the reactor plant was in mode 2 (startup) at less than 1% power. The licensee notified the NRC Operations Center of this event via the ENS at about 6:15 a.m. on May 1, 198 (9) ESF Actuation [ ENS No. 08551]

At about 8:55 p.m. CDT on May 1, 1987, the licensee experienced an ESF actuation when the Reactor Water Cleanup System (RT)

isolated. The isolation occurred due to a pressure transient sensed in the Division 1 High Delta Flow transmitter while venting the Division 2 High Delta Flow transmitter. The Division 1 and 2 transmitters share a common reference le The licensee was troubleshooting the cause for an earlier RT isolation that occurred while manipulating the systems flow paths (reference ENS No. 08542). At the time of occurrence the reactor plant was in mode 2. The licensee notified the NRC Operations Center of this event via the ENS at about 11:00 on May 1, 1987.

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(10) Manual Scram Due to Feedwater Valve Failure [ ENS No. 08584]

At 6:17 a.m. CDT on May 6, 1987, the licensee manually scrammed the reactor from 17% power due to increasing reactor vesscl water level. The increasing water level was caused by the feedwater regulating valve failing in the as-is position due to an oil leak on the operator. All systems responded normall The event was cleared at 7:09 a.m. CDT. A NRC inspector was in the control room at the time of the event and noted plant

operators responded in accordance with plant procedures.

j (11) ESF Actuation [ ENS No. 08640]

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At about 7:30 p.m. CDT on May 11, 1987, the licensee

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experienced an ESF actuation of Division III ECCS components.

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The event occurred during troubleshooting of a failed pressure transmitter (B21-N400). While attempting to vent the failed transmitter, a hydraulic transient was sensed on common instrument lines to the Division III level transmitter The result was an auto start of the HPCS pump, Division III Diesel Generator, and Division III Shutdown Service Water. HPCS did not inject into the reactor vessel due to a high water level 8 that was reflecting plant conditions. Plant operators verified an actual low water level did not exist and restored all Division III equipment to a standby status at about 8:00 The reactor plant was in mode 2 at the time of this event. The licensee notified the NRC Operations Center of this event via the ENS at about 9:15 p.m. CDT on May 11, 1987.

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(12) ESF Actuation [ENSNo.08657]

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At about 1:00 a.m. CDT on May 13, 1987, the licensee '

experienced an ESF actuation when the Reactor Water Cleanup system (RT) isolated due to a high differential temperature sensed in the RT Heat Exchanger Room Cooler. Upon l investigation, the licensee determined the cause for the high differential temperature was due to a failed temperature regulating valve to the room cooler. Cooling water to the RT Heat Exchanger Room Cooler was restored by opening a bypass valve. This action cleared the high differential temperature sensed by the RT room cooler. At the time of this event, the reactor plant was in mode 2 (startup) at less than 1% powe The licensee notified the NRC Operations Center of this event via the ENS at about 3:30 a.m. CDT on May 13, 1987, 14. Startup Test Witnessing (72302)

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The inspector witnessed the conduct of Startup Test Procedure (STP)

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25A-1, "MSIV Functional Test".

The inspector determined by direct observation that licensee operating

and test personnel were knowledgeable in their individual roles and responsibilities. Adequate communications were established and

, maintained throughout the test. Prior to, during, and subsequent to l the subject test 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 specifications regarding licensed operators.

i The proper versions of the test procedures were in use and were being followed. All referenced procedures had been reviewed and approved.

Each of the prerequisites had been satisfied, f

Changes or revisions to the test procedures were properly reviewed and approve *

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 inspectio No violations or deviations were identifie . Management Meeting (30702)

On May 15, 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 i to enhance the licensee's performance. Key personnel attending this meeting are identified by (#) in paragraph 1 of this repor ,

e + 1 The licensee discussed plant operations to date and summarized significant events; the licensee discussed the status of their Maintenance Improvement Program; the licensee's status of plant modifications was presented; the licensee discussed their corrective action and operational monitoring program; and the licensee presented the status of their INP0 training accreditation progra NRC (Region III) management acknowledged the licensee's status and plan The meeting concluded with a tentative agreement to meet again in June 1987, at the Clinton site with a similar agend . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are receptable items, violations, or deviations. One unresolved item disclosed during this inspection was discussed in paragraph 1 . Exit Meetings (30703)

The inspector met with licensee representatives (denoted in paragraph 1)

throughout the inspection and at the conclusion of the inspection on May 18, 1987. The inspector summarized the scope and findings of the inspection activities. The licensee acknowledged the inspection finding The inspector 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 <