IR 05000461/1988014
| ML20151N830 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 07/29/1988 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20151N823 | List: |
| References | |
| 50-461-88-14, IEIN-87-024, IEIN-87-041, IEIN-87-042, IEIN-87-050, IEIN-87-056, IEIN-87-24, IEIN-87-41, IEIN-87-42, IEIN-87-50, IEIN-87-56, IEIN-88-005, IEIN-88-009, IEIN-88-5, IEIN-88-9, NUDOCS 8808090103 | |
| Download: ML20151N830 (25) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-461/88014(ORP)
Docket No. 50-461 License No. NPF-62 Licensee:
Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name:
Clinton Power Station Inspection At:
Clinton Site, Clinton, Illinois Inspection Conducted:
May 18 through July 6, 1988 Inspectors:.P. Hiland S. Ray D. Butler dC 'Ww a Approved By:
R. C. Knop, Chief E
Reactor Projects Braach 3 Date Inspection Summary Inspection on May 18 through July 6, 1988 (Report No. 50-461/88014(DRP))
Areas Inspected:
Routine, unannounced safety inspection by the resident inspectors of licensee action on previous inspection findings; 10 CFR 21 report followup; NRC information nntice followup; NRC compliance bulletin followup; onsite followup of written reports of nonroutine events at power reactor facilities; SIMS Item No. MPA-A-15; operational safety verification; monthly maintenance observation; monthly surveillance observation; training effectiveness; and onsite followup of events at operating reactors.
Results:
Of the 11 areas inspected, one violation was identified in the area of licensee action on previous inspection findings.
Additionally, one violation was identified in the area of operational safety verification; however, in accordane.e with 10 CFR 2, Appendix C, Section V.G.1, a Notice of Violation was not issued.
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DETAILS 1.
Personnel Contacted Illinois Power Company (IPCo)
W. Kelley, President W. Gerstner, Executive Vice President
- D. Hall, Vice President, Nuclear
- K. Baker, Supervisor, I&E Interface J. Brownell, Project Engineer / Specialist R. Campbell, Manager, Quality Assurance J. Ccok, Manager, Nuclear Planning and Support E. Corrigan, Director, Quality Engineering and Verification
- R. Freeman, Manager, Nuclear Station Engineering Department
- K. Graf, Director, Operations Monitoring Program
- S. Hall, Director, Emergency Response
- D. Holesinger, Assistant Manager, Clinton Power Station E. Kant, Director, Design and Analysis Engineering
- A. Mcdonald, Director, Nuclear Program Assessment
- D. Morris, Director, Outage Maintenance Programs
- J. Miller, Manager, Scheduling & Outage Management
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- J. Perry, Manager, Nuclear Program Coordination
- R. Schaller, Assistant Manager, Plant Operations
- F. Spangenberg, Manager, Licensing & Safety
- J. Weaver, Director, Licensing J. Wilson, Manager, Clinton Power Station Soyland/WIPCO J. Greenwood, Manager, Power Supply U.S. Nuclear Regulatory Commission (USNRC)
- B. Davis, Regional Administrator, Region III D. Butler, Inspector, Plant Systems Section, Region III
- R. Cooper, Chief, Section 3B, Region III
- E. Greenman, Director, DRP, Region III
- P. Hiland, Senior Resident Inspector, Clinton
- R. Knop, Chief, Branch 3, Region III
- C. Paperiello, Deputy Regional Administrator, Region III
- S. Ray, Resident Inspector, Clinton
- W. Snell, Chief EP Section, Region III
- Denotes those attending the management meetina on June 10, 1988.
- Denotes those attending the monthly exit meeting on July 6, 1988.
The inspectors also contacted rad interviewed other licensee and contractor personnel.
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2.
Previously Identified Items (92701) (92702)
a.
(Closed) Open Item 461/84025-01:
CPS System Descriptions Not Complete.
This item was discussed in Inspection Report No. 50-461/86048, Paragraph 2.o. at which time it was left open pending review of completed system descriptions.
The licensee completed the approval and implementation of all system descriptions in June 1988.
The inspectors reviewed a sampling of the material and fcund them to be accurate and up to date.
The licensee had programs in place to maintain the system descriptions up to date with changes in plant design and to provide current revisions of the system descriptions for use in operator training and other applications.
This item is closed.
b.
(Closed) Open Item 461/87031-06:
Control of Gas Cylinder Bottles in Containment.
This item was discussed in Inspection Report No. 50-461/87031, Paragraph 8.g. at which time the item remained open pending the licensee's action to either analyze the possible missile hazards due to unsecured gas bottles or provide a procedure to control them.
The licensee completed Modification A-128 to provide seismic restraints for six nitrogen bottles to be stored in containment.
The Modification was released for operations on October 8, 1987, with a commitment (CCT No. 046778) to revise CPS Operating Procedure No. 3304.01, "Control Rod Hydraulic and Control (RD)," to include information in the detailed impact assessments of the modification.
The original due date of the CCT was February 1, 1988.
Procedural changes requested by the impact assessments included the following:
All spare nitrogen bottles inside containment were to be s ured
in the racks including empty bottles.
No bottles were to be stored in the charging cart except when in actual use.
- Chains werit to be pulled tight around the bottles when stored.
The charging cart and tool box were to be chained to their
assigned eye bolts when not in use.
Caps were to be attached to the bottles at all times when not
in use and chains with "S" hooks were to be used on the caps to prevent the possibility of them becoming missiles.
On June 15, 1988, the inspectors toured containment to determine the status of this open item.
The following conditions were noted:
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Seven nitrogen bottles were located in containment in the
vicinity of the hydraulic control units.
Only one of the bottles was in the designated storage rack.
Four of the bottles were empty and were loosely chained
to a handrail.
One of the bottles was in the charging cart which was also
loosely chained to the handrail.
One of the bottles was on the opposite side of containment from
the rack and was chained to a handrail and equipment support.
The bottle that was in the storage rack had only one of its
two chains secured.
None of the bottles had their caps secured with a chain with
"S" hooks.
The chains designed for the bottle caps were being
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used to hold the bottles to the handrails.
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Two of the bettles had their caps removed and regulators
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installed.
The caps from those bottles were not secured in any way.
Neither the charging cart nor the tool box was chained to its
assigned eye bolt.
The inspectors determined that CPS No. 3304.01 still had not been
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revised 4 1/2 months after the due date of CCT No. 046778.
No. 1050.01, "Control of Transient Equipment / Materials," was approved on February 23, 1988, for the control of tempora *y material in seismic structures.
That procedure contained the requirement that gas bottles be securely tied off at or above their center per Site Safety Standards No. 5 and No. 14.
The Site Safety Standards were not intended to be used to restrain gas bottles under seismic conditions and a single l
tie-off would probably not be sufficient in a seismic event.
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No. 1050.01 also required that gas bottles be removed from seismic
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areas when not in use.
The four empty bottles discussed above had I
been empty for various lengths of time up to six weeks, but had not been removed from containment.
An Illinois Power Safety Bulletin written on May 25, 1988, also emphasized the requirements to remove the regulators from bottles and install the caps when the bottles were not being used and to remove empty bottles from the plant expeditiously.
This Safety Bulletin, although widely distributed on site, was apparently not successful in preventing the misuse of gar bottles in containment.
The inspectors discussed the problems with 9as cylinders in containment with the operating staff several times during this inspection period.
Each time some improvements were made, but the bottles were never secured properly in accordance with the
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detailed impact assessments of Modification A-128 because the operators had no procedural guidance or training for the implementation of the modification.
10 CFR 50, Appendix B, Criterion V as implemented by IP Operation Quality Assurance Manual, Chapter 5, states that activities affecting
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quality shall be prescribed by documented procedures of a type appropriate to the circumstances.
Failure of the licensee to provide procedures to implement Modification A-128 for control of gas cylinder bottles in containment is a violation of 10 CFR 50, Appendix B, Criterion V and IP Operation Quality Assurance Manual, Chapter 5 (50-461/88014-01(ORP)).
c.
,(Closed) Violation 461/88003-02:
Inoperable Division I Primary Containment Hydrogen Recombiner While Division II Diesel Generator Was Inoperable.
This event was discussed in Inspection Report No. 50-461/88003, Paragraph 7.b.(2).
The inoperability was caused by an operator improperly positioning a temperature controller after a surveillance and inadequate independent verification of the lineup.
The licensee responded to the Notice of Violation via IP Letter U-601154, datM % rch 15, 1988, in a timely manner.
Immediate corrective action consisted of conducting a partial electrical lineup of the Division I and II flydrogen Recombiner control panels.
This lineup restored the OPERABILITY of the Division I Recombiner and verified the OPERABILITY of the Division II Recombiner, A critique was held and LER 88-004-00 was issued.
The LER is closed in Paragraph 6.e. below.
Actions '.u prevent rect.ence included counselling the operators who improperly positioned toe centroller and performed the independent verification.
All on-shift operators have been notified of the event and the lessons learned.
Additional actions taken to prevent procedural problems which were noted by the inspectors during the investigation of this event are discussed in Paragraph 2.d. below.
Based on the inspectors' verification that corrective actions as stated in the licensee's response to this violation were completed, this item is closed, d.
JClosed) Unresolved Item 50-461/88003-03:
Control of Procedure Revisions.
This item concerned approved procedural changes incorporated in one revision of a procedure but dropped from the next revision.
This item was discussed in Inspection Report No. 50-461/88003, Paragraph 7.d.
An investigation by the licensee staff and review by the inspectors determined that this item was an isolated case of apparent administrative error.
It is believed that word processing personnel used an improper data disk containing an old version of
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the procedure when typing the later revision.
Thus, the changes made in the previous revision were lost.
Since the two revisions were to different areas of the procedure, the mistake was not noticed in the review / approval cycle of the later revision.
The appropriate word processing staff was trained on the proper methods of controlling computer disks and the procedure revision review cover shcet was changed to provide a block for the proofreader to initial.
Instructions were promulgated to the proofreaders to check the entire document and not just the parts that were expected to chan0e.
Based on the corrective actions discussed above and the fact that no other cases of similar mistakes were found, this item is closed.
e.
(Closed) Unresolved. Item (461/88009-05):
Review Similarity Analysis for Environmental Qualification of Weed Thermocouples.
During a licensee inspection of electrical equipment in the main steam tunnel, it was identified that required sealant was not installed on Weed Thermocouples.
As previously discussed in Inspection Report
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No'. 50-461/88009, Paragraph 11.b., the licensee concluded that the as-installed configuration of the Weed Thermocouples was acceptable based on a similarity evaluation to Pyco temperature measuring devices.
At the conclusior: of that inspection, this item remained unresolved pending review by a Region III specialist of the licensee's similarity analysis.
During this report period, the Region III specialist (Mr. M. Kopp)
completed a review of the licensee's similarity analysis Calculation No. CQ0-039466, dated April 25, 1988.
That review indicated that the licensee's conclusion that Weed Thermocouples were qualified in the as-installed configuration was reasonable.
This item is closed.
f.
(Closed) Open Item _(461/88009-07):
Review Proposed Changes to
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Environmental Qualifications Program Scope.
This item was previously discussed in Inspection Report No. 50-461/88009, Paragraph 11.d.
At the conclusion of that inspection, this item remained open pending completion of the inspectors' review of changes to the licensee's Environmental Qualifications (EQ) maintenance program.
During this report period, the inspector reviewed the EQ maintenance program changes with the support of a Region III specialist (Mr. M. Kopp).
The inspectors noted that changes to the EQ maintenance program had been made with the appropriate level of written justification and that those changes appeared reasonable.
This item is closed.
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3.
10 CFR Part 21 Report Followup (36100)
a.
(Closed) Part 21 Item (461/87003-PP):
High Pressure Core Spray (HPCS) System Emergency Diesel Generator (EDG) exciter leads that were damaged where they exited the generator shaft (NRR Part 21 Log Nos. 87167 and 87198).
The nature of the defect involved generator exciter leads which ran through the generator shaft from the field rectifier assembly to the main field.
Diesel generator exciter lead damage was confirmed at the point where the exciter leads exited the generator shaft.
Chafing of the outer sleeve (cotton braid) was visually observed.
The root cause of chafing was friction / rubbing by the movement of the leads with the generator shaft.
This damage resulted from the absence of Room Temperature Vulcanizing (RTV) material to secure the leads in place inside the generator shaft and from sharp edges within the shaft.
During replacement of the leads, additional damage to the exciter leads was identified.
At the interfacing surfeces of the generator fan and the pole piece, the exciter lead was crushed ana the protective tape and insulation damaged.
This condition was identified at three fan-to pole piece locations and appears to have occurred as a result of mislocating the lead during machine assembly (NRR Part 21 Log No. 87218).
The licensee replaced the exciter leads during the December 1987 maintenance outage.
The sharp edges found inside the shaft were filed down and the exciter leads were properly positioned at the interfacing surfaces of the generator fan.
RTV Type 106 (Q approved for this type of application) was used to secure the exciter leads within the generator shaft.
The EDG was satisfactorily tested and returned to service.
This item is closed.
b.
(Clored) Part 21 Item (461/87005-PP):
HPCS system EDG supplied by Beloit Power Systems was found to have rear bearing and rotor shaft d2:aga (NRR Part 21 Log No. 88006).
The damage was discovered during the exciter lead replacement discussed in Paragraph 3.a.
above.
The nicks /cnips were located on the exciter end of the oil seal portien of the shaft.
These defects were very small and not uncommon or unexpected, and could have been caused by an object inadvertently striking or hitting the lip of the shaf t during machine assembly.
The only concern with respect to the nicks / chips was th? oil sealing ability in the area of the defects.
Since no oil leak path was visible, and the friction oil seal was more than one and one-half inches from the defects, there was no reason to believe that a leak would occur at any time.
The gouged area was located approximately one quarter inch away from the friction oil seal and not in a sealing surface or bearing area.
The defect was less than one sixty-fourth of an inch deep and was
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caused by the bearing housing cap rubbing the shaft as a result of the bearing housing cap being installed incorrectly in a cocked position during machine assembly prior to delivery to Illinois Power.
Visual inspection of the bearing indicated that the damage was caused by material ground off during the shaft gouging event.
This material abraded the bearing cage and allowed the cage to drop onto the inner race of the bearing.
This condition produced more debris which allowed the Saaring damage to progress to spalling of the rolling elements.
The licensee replaced the bearing, and General Electric determined that the shaft integrity was not affected by the nicks / chips and the gouging; therefore, the shaft did not require replacement.
The nicks / chips were repaired by use of a light emery cloth to dress the affected area.
Since the gouged area was not a sealing surface, a light emery cloth was used to smooth the gouged area.
The inspectors consider the licensee's corrective action to be adequate and the EDG was satisfactorily tested and returned to service. This item is closed.
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4.
Information Notices (92701)
For the Information Notices discussed below, the inspectors verified that the licensee had received the Information Notice, had distributed the Notice to appropriate personnel, and had completed appropriate actions.
a.
(Closed) Information Notice 87-24 (461/87024-NN):
Operational Experience Involving Losses of Electrical Inverters.
This Information Notice was rece1veo oy the licensee on June e, 1987.
[ojjowingrecsipt, the jicens'e'e ass 1gnea, review rdsponsioisity in accoraance witn Licensing ana gar,ety troceaure L.}, '(ee0,oaC,K
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[rogram;.
!t Keview gneet y,fug}uj aatea June 33, 3yg/, assignea 06partment(NTO).'guepartmint(ASED),'ind'Nubida responsioliity for reviews to Liinton riant start Nuclear station "tngineefin fri niHg *
IP Memorandum Y-206384, F. A. Spangenberg to File, dated November 5, 1987, documented the results of the reviews performed.
The memorandum stated that an extensive. tachnical evaluation of Clinton's inverters
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had already been performed.
Lorrective actions in response to that evaluation included:
vendor training of CPS personnel on inverter operation and
maintenance development of an ongoing maintenance course
revisions of inverter operation and maintenance procedures to
improve inverter performance
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trending of inverter-related maintenance to identify failure
mechanisms extensive evaluations of inverter input and output voltages
and modifications to improve the voltago stability evaluations of inverter ambient conditions, area cooling
capacities and inverter loading.
Based on the actions which had already been taken and were still j
ongoing, the inspectors determined that response to this Information Notice was adequate.
This item is closed.
b.
(Closed) Information Notice 87-41 (461/87041-NN):
Failures of Certain Brown Boveri Electric Circuit Breakers.
This Information Notice was received by the licensee on August 31, 1987.
Following receipt, the licensee assigned review responsibility in accordance with Licensing and Safety Procedure L 1, "Feedback Program".
IP Review Sheet Y-205823, dated September 2, 1987, assigned responsibility for review to the Nuclear Station Engineering Department (NSED).
IP Memorandum Y-85990, E. W. Kant to F. A. Spangenberg, dated October 1, 1987, documented the review performed by NSED which determined that the Information Notice was not applicable to CPS.
All Class 1E, Division I and Division II 4160 KV breakers at CPS were provided by General Electric.
In a separate review of River Bend LER 87-004 which was one of the examples cited in the Information Notice, IP Memorandum Y-85537, E. W. Kant to F. A. Spangenberg, dated August 24, 1987, stated that no failures due to loose mounting bolts on closing spring charging motors have been reported for the Westinghouse or General Electric breakers.
As a preventative measure, NSE0 committed to revise Maintenance Procedure CPS No. 8410.01, "6900, 4100 Volt Power Circuit Breaker Generic Procedure For" to include checking and torquing, as necessary, of the mounting bolts for the closing spring charging motors for all Class 1E breakers.
That revision was incorporated on December 7, 1987.
This item is closed.
c.
(Closed) Information Notice 87-42 (461/87042-NN):
Diesel Generator Fuse Contacts.
This Information Notice was received by the licensee on September 10, 1987.
Following receipt, the licensee assigned review responsibility in accordance with Licensing and Safety Procedure L 1, "Feedback Program".
IP Review Sheet Y-205901, assigned responsibility for review to the Nuclear Station Engineering Department (NSED).
IP Memorandum Y-86417, E. W. Kant to F. A. Spangenberg dated November 9, 1987, documented the review performed by NSED which determined that the Information Notice was not applicable to CPS.
The fuse contacts discussed in the Information Notice were in a
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control cabinet supplied by General Moturs and were designed to disconnect whenever the cabinet door was opened.
CPS emergency diesel generator power and control panels were manufactured by Stewart and Stevenson and were of a completely different design than the subject panels.
They did not have mechanisms that would disconnect the fuses when the doors were opened.
The inspectors verified by direct field observation that the emergency diesel generator power and control panels were designed such that they would not be susceptible to the same type of failure as discussed in the Information Notice.
This item is closed.
d, (Closed) Information Notice 87-50 (461/87050-NN):
Potential LOCA at High-and Low-Pressure Interfaces From Fire Damage.
This Information Notice was received by the licensee on October 16, 1987.
Following receipt, the licensee assigned review responsibility in accordance with Licensing and Safety Procedure L.1, "Feedback Program."
IP Review Sheet Y-206267, dated October 23, 1987, assigned responsibility for review to the Nuclear Station i
Engineering Department (NSED).
IP Memorandum Y-86901, R. D. Freeman to F. A. Spangenberg, dated December 30, 1987, documented the review performed by NSED.
The review concluded that CPS has a piping configuration on four ECCS injection lines that is similar to that discussed in the Information Notice.
The lines are the Low Pressure Core Spray and the three Low Pressure Coolant Injection lines.
Each line has a discharge check valve with a 3/4" bypass line containing an air operated isolation valve and a 1/4" orifice.
NSED determined that the potential LOCA due to the bypass line was not a safety concern because calculated flows through the 1/4" restricting orifices would not exceed the capacity of the relief valves in the low pressure portion of the systems, thereby preventing overpressurization of the low pressure sides.
NSED issued Commitment Tracking Form No. 047290 to revise
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the Safe Shutdown Analysis as a result of this review.
This item is closed.
e.
(Closed) Information Notice 87-56 (461/87056-NN):
Improper Hydraulic Control Unit Installation at BWR Plants.
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This Information Notice was received by the licensee on November 16, 1987.
Following receipt, the licensee assigned review responsibility in accordance with Licensing and Safety Procedure L.1, "Feedback Program."
IP Review Sheet Y-206498, dated November 20, 1987, assigned responsibility for review to the Nuclear Station Engineering Department (NSED).
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IP Hemorandum Y-86734, E. W. Kant to F. A. Spangenberg, dated December 9, 1987, documented the review performed by NSED.
The review reported an NSED evaluation of Perry LER 86-14-00 which was the subject of the Information Notice.
The evaluation determined that Clinton's Hydraulic Control Units (HCUs) were properly mounted
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and met seismic requirements.
Subsequent to the Perry incident, General Electric (GE) issued a design change (FDDR LH1-5708, R/0, dated June 25, 1986) to retorque the HCU lower mounting hold down bolts.
Plant Modification RD-13 was initiated to retorque the hold down bolts from 18-28 ft-lbs to 48-52 ft-lbs to match the value used in the Wyle Labcratory Seismic Qualification Test.
Upon field implementation of the modification, the tubular HCU frames under the bolt heads became deformed and the modification was discontinued.
GE and Sargent and Lundy engireering evaluations
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detercined, by analysis, that the HCUs met seismic qualification requirements at the existing 18-28 ft-lbs torque values.
Therefore, NSED determined that the HCUs were qualified in the "use-as-is" configuration.
In addition, NSED determined that Branch Junction Modules (BJMs)
were not attached to the HCUs at Clinton.
BJMs were mounted to the containment liner consistent with the configuration used during seismic testing.
The inspectors verified by direct field observation that the BJMs were not mounted to the HCUs and did not observe any missing or obviously loose HCU mounting hardware.
This item is closed.
f.
(Closed) Information Notice 88-09 (461/88009-NN):
Reduced Reliability of Steam-Driven Auxiliary Feedwater Pumps Caused by Instability of Woodward PG-PL Type Governors.
This Inforrration Notice was received by the licensee on March 22, 1988.
Following receipt, the licensee assigned review responsibility in accordance with Licensing and Safety Procedure L.1, "Feedback Program." IP Review Sheet Y-207357, dated March 28, 1988, assigned responsibility for review to the Nuclear Station Engineering Department (NSED).
IP Memorandum Y-88075, R. D. Freeman to F. A. Spangenberg, dated
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April 25, 1988, documented the review performed by NSED.
The review determined that the only steam driven pump similar to the subject pumps at Clinton was the Reactor Core Isolation Cooling (RCIC) pump which was driven by a Terry Type GS-2 turbine.
The turbine used a Woodward EGR type governor.
There were no Woodward PG-PL type governors listed on the Clinton Mechanical Equipment List.
I The review also determined that the RCIC turbine is periodically tested using a quick start from cold conditions.
The inspectors i
reviewed records of all completed quarterly Surveillance Procedure CPS No. 9054.01, "RCIC System Operability Check" performed since the completion of the Startup Test Program.
The inspectors verified that the surveillance used a quick start from cold conditions and duplicated, as far as practical, service conditions that would exist i
if the equipment was called upon to operate.
Tb:re were no documented
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turbine governor system problems in the surveillances.
This item is closed.
No violations or deviations were identified.
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5.
Bulletins and Circulars (92701)
For the Bulletin discussed below, the inspectors verified that the licensee had received the Bulletin, had distributed the Bulletin to appropriate personnel, and had completed appropriate corrective actions.
(0 pen) NRC Compliance Bulletin No. 88-05 and 88-05 Supplement 1 (461/88005-88):
Nonconforming Materials Supplied by Piping Supplies, Inc. at Folsom, New Jersey and West Jersey Manufacturing Company at Williamstown, New Jersey.
During this report period, the licensee notified the NRC Operations Center of the identification of inaccessible flanges manufactured by West Jersey Manufacturing (WJM) installed in various systems at Clinton Power Station.
Subsequent to the conclusion of the inspection period, the licensee identified additional inaccessible flanges to the NRC Operations Center.
As of July 14, 1988, the following status was provided to the inspector concerning the licensee's review required by Bulletin 88-05:
= 614 Total WJM Flanges Purchased WJM Flanges in warehouse
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WJM Flanges installed in plant =(112]
WJM Flange Application Remaining 425 to be identified
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The inspector noted that for the identified accessible flanges the licensee was conducting "in-situ" testing.
For the inaccessible flange material, the licensee was performing evaluations to justify continued plant operation.
The licensee was continuing their review to identify the in plant application of the 425 remaining WJM flanges.
This item will remain open pending staff review of the licensee's completed investigation.
No violations or deviations were identified.
6.
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities (92700)
For the Licensee Event Reports (LERs) listed below, the inspector performed an onsite followup inspection 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 LERs.
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(Closed) LER 86-021-00 (461/86021-LL):
Reactor Water Cleanup Pump P.oom A High Temperature Trip Oue to Personnel Error.
This event was previously documented in Inspection Report No. 50-461/86072, Paragraph 11.b.(2), and inspected in Inspection Reports No. 50-461/86073, Paragraph 3.e., No. 50-461/37002, Paragraph 5.b.(2), and No. 50-461/87019, Paragraphs 6.d. and 6.f.
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At the conclusion of those inspections, Corrective Actions 7, 8, and 9 in the LER remained to be inspected, as well as a review of the licensee's plan for addressing the Lifted Leads Task Force recommendations, The inspectors reviewed training records and revised lesson plans to verify that the lessons learned from this LER had been incorporated.
The inspectors also reviewed the licensee's disposition of the Lifted Leads Task Force recommendations.
An extensive program was implemented to identify and evaluate the impact of surveillance and maintenance procedures which called for lifting leads, installing jumpers, or performing other modifications.
Corrective actions to reduce the frequency of lifted leads in procedures were being pursued on a case-by-case basis depending on impact.
In some cases procedures were changed to eliminate the need to lif t leads and in other cases equipment modifications were required.
In most cases, eliminating the lifting of leads to perform a surveillance was not practical, but additional awareness, experience, and training has reduced the frequency of errors made while lifting leads.
Based on the inspectors' verification of corrective actions as discussed in the LER, this item is closea.
b.
(Closed) LER 87-043-00 (461/87043-LL):
Automatic Actuation of the Reactor Protection System Due to Equipment / Component Malfunction, This event was previously discussed in Inspection Report No. 50-461/87031, Paragraph 10.c.(12).
The event was a reactor trip from about 60% power due to a turbine trip.
The turbine trip was caused by a high vibration signal due to a loose ccanector in the No. I turbine bearing vibration circuit.
The inspectors reviewed completed Maintenance Work Request (MWR) C-50346 which corrected loose connectors on turbine bearing vibration circuits.
The inspectors also reviewed documentation of other corrective actions including installation of a three second time delay relay in the turbine vibration trip circuit and later replacement with new circuit cards with improved noise filtering characteristics including a three second time constant.
The inspectors also reviewed the post trip review report to verify that the root cause of the trip was determined and understocd prior to the subsequent startup.
Based on the inspectors' verification that corrective actions in
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the LER have been completed, this item is closed.
c.
(Closed) LER 87-067-00 (461/87067-LL):
Overly Restrictive Design
Setpoint Trip Tolerance for Temperature Switch Results in Broken Lead Wire and Reactor Core Isolation Cooling Isolation.
This event was previously discussed in Inspection Report No. 50-461/87036, Paragraph 11.L.(9).
The event was caused by a thermocouple wire breaking af ter being moved aside to gain access
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to a terminal board during a surveillance.
The surveillance being conducted was part of a monthly surveillance requirement which required the lifting of 128 leads in order to calibrate 64 temperature modules.
The licensee's investigation of this event determined that the Sargent and Lundy (S&L) Setpoint Log for the temperature modules had been misinterpreted resulting in almost routine failures of the monthly channel functional checks.
These failures resulted in monthly calibrations of the modules.
The licensee revised the functional check procedures, which did not require the lifting of leads, to use the proper setpoint tolerances.
This allowed the channel calibrations to return to their required 18 month frequency, and much fewer lead lif ting evolutions.
The inspectors also revieved completed Maintenance Work Request C-48772 which repaired the broken thermocouple lead and records for training of appropriate personnel on the lessons learned from this event.
Based on the inspectors' verification of corrective actions, this item is closed.
d.
(Closed) LER 88-003-00 (461/88003-LL):
Malfunction of Process Radiation Monitor During Check Source Function Results in Premature Re-Landing Lead Wires and Isolation of Hydrogen /0xygen Monitor.
This event was previously discussed in Inspection Report No. 50-461/88003, Paragraph 11.b.
The cause of the event was believed to be a malfunction in the monitor's check source cycle which prolonged the cycle such that leads were relanded during
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the cycle causing the trip logic to actuate at the end of the cycle.
The inspectors reviewed CPS No. 9920.72, "Channel Functional Testing of Safety Related Process Radiation H.itors," Revision 24, which included a change to require confirmacion of check source cycle compietion prior to landing alarm relay leads.
The inspectors also reviewed completed Plant Modification AR-23 which included changes to remove the momentary activation of the output trip relay that occurred at the end of the check source cycle The inspectors also reviewed completed Maintenance Work Request (MWR) C-46680 which was to investigate and correct the cause of the prolonged check source cycle.
The problem could not be duplicated but was believed to have been caused by a defective or misaligned Geiger-Muller (GM) tube.
Between the time of the LER event and the investigation under MWR C-46680, the GM tube had been replaced, which apparently corrected the check source cycle problem.
Based on the inspectors' verification that corrective actions in the LER have been completed, this item is closed.
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e.
(Closed) LER 88-004-00 (461/88004-LL):
Manipulation of Incorrect Temperature Controller Results in Inoperable Division I Primary Containment Hydrogen Recombiner While Division II Diesel Generator Was Inoperable.
This event was discussed in Inspection Report No. 50-461/88003, Paragraph 7.b.(2).
The event was considered a violation (461/88003-02)
which was closed in Paragraph 2.c. above.
Corrective actions discussed in the LER are the same as those for the violation.
The LER discussed one time period of about 61 hours7.060185e-4 days <br />0.0169 hours <br />1.008598e-4 weeks <br />2.32105e-5 months <br /> during which the licensee was not in compliance with their Technical Specifications during this event.
The inspectors noted an additional time period of about three hours which was discussed in the licensee's response to the violation.
Since this additional time period of non-mmpliance would not have affected the safety consequences discussed IR, the inspectors did not ask the licensee to submit a revisea #A.
However, the inspectors did discuss with licensee personr.M the need to ensure the accuracy of submittals to the NRC.
Based on the inspectors' verification that all corrective actions discussed in the LER have been completed, this item is closed.
No violations or deviations were identified.
7.
Temporary Instruction 2515/93 - Inspection for Verification of Quality Assurance Request Regarding Diesel Generator Fuel Oil - SIMS Item MPA-A-15 (25593)
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The subject temporary instruction required that the inspectors verify that the licensee had included diesel generator fuel oil in its quality assurance program.
The inspectors reviewed Clinton's Safety Evaluation Report (NUREG-0853) Section 9.6.3.2 in which the NRC staff found that fuel oil quality and tests would conform to the guidelines of Regulatory Guide 1.137, "Fuel Oil Systems for Standby Diesei Generators", Positions C.2.9 through C.2.h.
The inspectors reviewed Technical Spec.fication Surveillance Requirement 4.8.1.1.2.d, CPS No. 9981.00, "Diesel Generator Replenishment Fuel Oil Sampling and Analysis", and CPS No. 9981.01,
"Diesel Fuel Oil Sampling and Analysis" to verify that they referenced and met the requirements of Regulatory Guide 1.137.
Clinton Power Station purchased commercial fuel oil from a vendor who was not on the Qualified Suppliers List; however, the oil was analyzed as required by Regulatory Guide 1.137 Position C.2.b. before being added to the storage tank.
Other required analysis were conducted by an offsite vendor laboratory (Enright) which was on the Qualified Suppliers List.
The inspector reviewed CPS No. 3506.01, "Diesel Generatcr and Support Systems" to verify that the procedure contained steps to trigger the required fuel oil samples before fuel oil was added to the storage tanks.
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The fuel oil was a so periodically sampled and analyzed while in the storage tanks.
Based on the inspectors' verification that diesel generator fuel oil was treated as a quality consumable in accordance with Regulatory Guide 1.137, this item is closed.
No violations or deviations were identified.
8.
Operational Safety Verification (71707)
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 tracking of LCOs.
Routine tours of the auxiliary, fuel, containment, control, diesel generator, turbine buildings and the screenhouse were conducted to observe plant equipment conditions including potential for fire hazards, fluid leaks, and operating conditions (i.e., vibration, process parameters, operating temperatures).
The inspectors verified that maintenance requests had been initiated for discrepant conditions observed.
The inspectors verified by direct observation and discussion with plant personnel that security procedures and radiation protection (RP) controls were being properly implemented.
Inspections were routinely performed to ensure that the licensee conducted activities at the facility safely and in cor.formance with regulatory requiremeats.
The inspections focused on the implementation and overall effectiveness of licensee's control of operating activities, and the performance of licensed and nenlicensed operators and shift technical advisors.
The following items were considered during these inspections:
Adequacy of plant staffing crd supericision.
- Control room professionalis.n, including procedure adherence,
operator attentiveness and response to alarms, events, and off normal conditions.
Operability of selected safety related systems including attendant
alarms, instrumentation, and controls.
Maintenance of quality records and reports.
- a.
On May 9,1988, while operating at 91% power, the licensee discovered that a motor-operated flush valve on Offgas Pretreatment Radiation Monitor 1RIX-PR034 was partially open.
The partially open valve caused the monitor's sample to be diluted with instrument air such that the condenser effluent gas sample 5asults were not representative of actual effluent.
In addition, during part of the time period that the flush valve was inadvertently left open, Offgas Hydrogen Monitors were declared inoperable due to inadequate gas flow.
Because of this condition, the licensee was required to take periodic grab samples of
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hydrogen concentration in the Main Condenser Offgas Treatment System.
- These grab samples were taken downstream of the partially open flush valve and therefore were invalid due to dilution.
An investigation by the licensee determined that the flush valve had most likely been open since April 7, 1988.
On that date, technicians conducted a channel functional test on the monitor as post maintenance testing.
During this test the flush valve was cycled for performance of the flush sequence.
The motor-operated flush valve was designed such that, once the flush sequence began, if the flush push button was pushed again before the valve sequence was complete, the valve would freeze in its original position.
The surveillance procedure (CPS No. 9537.66) contained a warning against pushing the buttoa again while the valve was sequencing.
In addition, since direct valve position indication was not provided, the procedure required checking local flow indication to verify that the flush valve was open during flushing, but it aid not require a similar check to verify the flush valve reclosed.
The licensee reported this event as LER 88-015-00 dated June 8, 1988.
In the LER they noted that between April 7, 1988, when the flush valve was most likely left open, and May 2, 1988, when the plant was started up from a maintenance outage, the monitor was not required to be operable.
Thus from May 2-9, 1988, several Technical Specification violations cccurred depending on the lineups of the Steam Jet Air Ejectors (SJAE) Offgas Treatment Systems, and Offgas Hydrogen Monitors during that time period.
The Technical Specifications whic.,were violated were as follows:
Technical Specification 3.3.7.1 (Table 3.3.7.1-1) which required
that the Offgas Pretreatment Radioactivity Monitor be operable during operation of the main condenser SJAE.
Technical Specification 4.11.2.7.1 which required that the
radioactivity rate of noble gases at the Offgas Recombiner be continuously monitored.
Technical Specification 3.3.7.12 (Table 3.3.7.12-1), Item 3.a.,
ACTION 1; % which required that hydrogen in the Main Condenser Offgas Treatment System be monitored.
The violations were considered to be licensee identified and met the criteria of 10 CFR 2, Appendix C, Section V.G.I.
Therefore no Notice of Violation was issued and this matter is closed (50-461/88014-02).
The incompleted corrective actions will be reviewed separately when the LER is closed.
b.
On June 16, 1988, at approximately 8:00 a.m., the inspectors noted that the boot seal on an electrical conduit penetration through the secondary containment air lock in the southeast auxiliary building was not attached to the wall.
A'r was felt leaking through the
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The inspect:rs brought the condition to the attention of the Staff Assistant Shift Supervisor.
The Shift Supervisor declared the secondary containment airlock inoperable and locked the door on the opposite side of the airlock until repairs could be made.
The plant was operating at 100% power at the time.
Approximately two hours later, while walking down the penetration to plan the repair work, the licensee noted several other secondary containment boot seal type electrical and pipe penetrations that were not installed in accordance with design drawings.
Nine penetrations, including the original one identified by the inspectors, did not have backing rings installed to help hold the boot to the wall, and four, including the one identified by the inspectors, did not have hose clamps installed to attach the boot to the conduit.
Some o' the seals without hose clamps were noticeably leaking.
The Shift Supervisor entered a four hour action statement for Secondary Containment in accordance with Technical Specification 3.6.6.".a.
The seals were caulked and hose clamps were installed before che end of four hours, but an analysis for using the seals witho..t backing rings installed was not completed, thus the licensee was required to be in at least HOT SHUTDOWN within the next 12 nours.
The licensee notified the NRC Operations Center of the event via the ENS as discussed in Paragraph 12.b.(1) below.
Approximately 35 minutes later, the Nuclear Station Engineering Department (NSED) provided IP Memorandum Y-88487, E. W. Kant to D. L. Holesinger which provided for continued operation without backing rings provided that the hose clamps were installed and the boots were firmly attached to the wall with Dow Corning caulk.
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The licensee exited the shutdown action statement before any actual power reduction.
The NSED memorandum recommended installing the backing plates by the end of the first refueling outage to ensure seal integrity for the life of the plant.
The inspectors found this approach to be reasonable.
The NSED recommendation will remain open until final disp:sition by the licensee (0 pen Item 461/88014-03).
On June 17, 1988, a critique of the event was held to review the sequence of events and determine the root cause of the installation deficiencies.
Insuffic'ent information was available at that time to determine the roet cause.
The possible unauthorized deviation from design specifications for secondary containment penetration seals will rernain unresolved until the completion of the root cause determination by the licensee (Unros91ved Item 461/88014-04).
During the critique, it was brought to light that the Dow Corning caulk used to repair the boot seal originally identified by the inspector was drawn from shop material left over from a previous job and it had exceeded its shelf life certification.
The penetration was again declared inoperable while the material certification was determined.
Within the naxt four hours, the Dow Corning caulk from the lot used in the repairs was recertified based on discussions with the vendor and a test on the materials ability to flow smoothly and set up within the specified time after application.
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In addition, it was determined in the critique that all of the areas
of the secondary containment gas control boundary which may have had boot seal type penetrations had not been inspected.
Inspections of all remaining areas were ordered.
On June 17, 1988, the Quality Assurance Department found one additional penetration that was missing its hose clamp.
The boot was immediately repaired.
c.
For the reactor trip discussed in Paragraph 12.b.(3) below, the licensee determined that the cause was due to oxidation of the contacts of an Agastat Type GPI relay.
In the course of repairing the problem, the licensee determined that all 54 of the Type GPI relays in stores also exhibited oxidation on the contacts resulting in higher than normal contact resistances.
The relays were approximately nine years old.
Preliminary analysis based on testing and discussions with the manufacturer determined that the failures may have been due to misapplications of the relays.
The relays were designed for relatively high current (1-12 amp) applications.
Some of the relays, including the one which caused the June 24 reactor trip, were being used in milliamp current applications.
In high
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current applications, it was believed that arcing across the contacts during normal opening and closing would prevent oxide buildups.
In low current applications, oxides built up on the contacts causing eventual high resistance and potential failures.
The licensee developed an action plan to address the generic issues involved.
The main points of the action plan involved determining all of the Type GPI relay applications in the plant, determining by further testing and discussions with the vendor the appropriate application parameters, and prioritizing replacement requirements The action plan was being implemented at the time of this report.
Initial results indicated that the relays would not hcve a significant effect on reactor safety. This item will remain open pending completion of the action plan for dealing with the generic issues of the Agastat GPI relays.
(461/88014-05)
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d.
The inspectors noted continuing improvements in the number of discrepant conditions in the main control room.
Reductions in the number of reduced service recorders was especially evident.
This was in part due to increased awareness and training on the damage that was being done to recorder gear trains due to improper operation.
Listed below is the status of main control room problems that were noted on July 6, 1988, while at 100% power.
These were compared to the status under similar conditions in the last report.
All main control room annunciator / instrument, and recorder problems which existed at the time of this status were scheduled for correction by the end of the first refueling outage.
THIS REPORT LAST REPORT PERIOD PERIOD
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38 Total OOS/ Disabled Annunciators
17 Total 00S Instrument / Recorders
2 Total Reduced Serv. Instr / Recorders
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Two open items and one unresolved item were identified.
One violation was identified for which a Notice of Violation was not issued in accordance with 10 CFR 2, Appendix C, Section V.G.1.
9.
Monthly MaintenTnce Observation (62703)
Selected portioi.e of the plant mair.tenance activities on safety-related systems and components were observec or reviewed to ascertain that the activities were performed in accordance with approved procedures, regulatory guides, industry codes and standards, and that the performance of the activities conformed to the Technical Specifications.
The inspection included activities associated with preventive or corrective maintenance of electrical, instrumentation and control, mechanical equipment, and systems.
The following items were considered during these inspections:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibration was performed prior to returning the components or systems to service; parts and materials that were used were properly certified; and appropriate fire prevention, radiological, and housekeeping conditions were maintained.
The inspectors observed / reviewed the following work activities:
Maintenance Work Procedure No.
Activity
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PMMOGM013 Division 1 Diesel Generator Preventive Maintenance C43485 Division III Diesel Generator Corrective Maintenance PMMVDS002 Preventive Maintenance on 1VD02CC No violations or deviations were identified.
10.
Monthly Surveillance Observation (61726)
An inspection of inservice and testing activities was performed to ascertain that the activities were accomplished in accordance with applicable regulatory guides, industry codes and standards, and in conformance with regulatory requirements.
Items which were considered during the inspection included whether adequate procedures were used to perform the testing, test instrumentation was calibrated, test results conformed with Technical Specifications and procedural requirements, and tests were performed within the required time limits.
The inspectors determined that the test results were reviewed by someone other than the personnel involved with the performance of the test, and that any deficiencies identified during the testing were reviewed and resolved by appropriate management personnel.
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The inspectors observed / reviewed the following activities.
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Surveillance / Test Procedure No.
Activity CPS No. 9443.03 Leak Detection System Drywell Air Particulate, Iodine, and Gas Radiation Monitor Calibration.
CPS No. 9861.020003 Local Leak Rate Test CPS No. 9861.020004 Local Leak Rate Test No violations or deviations were identified.
11.
Training and Qualification Effectiveness (41400 and 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 months of May and June 1988.
Personnel appeared to be knowledgeable of the tasks being performed.
No violations or deviations were identified.
12.
Onsite Followup of Events at Operating Reactors (93702)
a.
General The inspectors 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 inst, ctors reviewed cae or more of the following:
the sequence of actions; the i
functioning of safety systems required by plant conditions; licensee actions to verify consistency with plant procedures and license
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conditions; and attempted to verify the nature of the event.
Additionally, in some cases, the inspectors 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
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actions noted during the inspectors' followup are provided in l
Paragraph b. below, b.
Details l
l (1) Entering Technical Specification Action Statement Requiring Plant Shutdown Due to Loss of Secondary Containment Integrity
[ ENS No. 12572]
On June 16, 1988, the licensee reported to the NRC Operations Center via the ENS that they had entered an ACTION statement requiring them to be in HOT SHUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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The ACTION in Technical Specification 3.6.6.1.a. was entered at 2:25 p.m. due to questions about the adequacy of secondary containment penetration boot seals.
At 3:08 p.m. on June 16, 1988, an analysis was provided by the Nuclear Station Engineering Department to allow use of the seals "as-is" and the shutdown ACTION statement was exited.
The plant was operating at 100%
power during the event and no actual power reduction had taken place during the short time the plant was in the shutdown ACTION statement.
This event is discussed in more detail in Paragraph 8.b. above.
(2) Contaminated Injured Man Transported Off-site
[ ENS No. 12625]
On June 23, 1988, the licensee reported via the ENS that a contaminated injured man had been transported to a hospital off-site.
The injuries were the result of burns the individual suffered while disassembling a sight glass on the Unit 2 Radwaste Floor Drain Evaporator Vapor Body.
The evaporator was in hot standby at the time of the event with its recirculation pump running.
Maintenance Work Request (MWR) C44828 which was being worked, called for the replacement of the sight glass on the Unit 1 evaporator in Room R4-10, but the Radiation Work Permit (RWP) for the job listed the locatfor. of the work as Room R4-12.
Room R4-12 is located next to Room R4-10 and is the location of the Unit 2 evaporator.
The event happened at about 1:05 p.m. on June 23, 1988.
At 1:30 p.m, the licensee declared an Unusual Event when the injured individual was transported by ambulance for off-site treatment.
The ambulance left the site at approximately 1:32 p.m.
and the ENS call was made at 1:38 p.m.
The injured individual had fixed skin contamination levels of about 200 cpm above background on the burned areas when he left the site.
Initial treatment and final decontamination was accomplished at the John Warner Hospital in Clinton and then the individual was transferred to the burn cer.ter in Springfield, Illinois.
The individual suffered first degree burns on his upper body and second degree burns on his thighs.
Two other individuals in the room at the time of the event received minor burns and low levels of contaminat'on.
They were treated and decontaminated onsite and thtn sent to the hospital for observation.
A critique of the event was held the following ray with the Resident Inspector in attendance.
The information discussed in the critique included the following.
The maintenance planner who wrote the RWP did not recall
why he wrote the wrong room number on the RWP.
He had written the RWP about two months earlier.
The maintenance workers involved in the job had asked the
radwaste operator to verify that the Unit 1 Floor Drain
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Evaporator was tagged out of service and drained down.
The maintenance workers did not accompany the radwaste operator when he made the verification.
The radiation protection technicians who surveyed the
room and set up the step-off pads did not notice that the equipment in Room R4-12 was not the same as that listed on the RWP.
The door to the Unit 1 and Unit 2 evaporator rooms were
clearly stenciled with the proper equipment names.
The evaporators themselves were stenciled with their proper Equipment Identification Numbers.
The MWR called for working on 1WF080 but the workers removed the sight glass from 2WF08D.
The worker who was injured had carefully loosened and then
removed the bolts holding the sight glass in place but had noted no leakage until he moved the sight glass from its sealing gasket.
The workers were wearing face masks and anti-contamination
clothing as a precaution while opening a potentially contaminated system.
This fact prevented the injuries and personnel contamination from being much worse.
The coordination between the site and the hospital was
good.
There were no delays in getting the ambulance into and out of the protected area.
Site radiation protection personnel went to the hospital ahead of the ambulance to brief them on what to expect and other radiation protection personnel accompanied the ambulance crew.
As one corrective action, the Manager-Clinton Power Station issued an instruction which required that for maintenance which required a breech of a system or out-of-service tagout, an operator shall accompany the maintenance personnel before the job to confirm that the work will be done on the proper equipment and that it is safe to work.
This item will remain unresolved pending further review in the next inspection by regional radiation protection specialists.
(50-461/88014-06)
(3) Reactor Scram Due to Feedwater Transiert [ ENS No.12646]
On June 24, 1988, at 11:10 p.m., the licensee experienced a reactor scram due to low reactor vessel level (Level 3).
The reactor was operating at about 60% power at the time of the scram.
The licensee notified the NRC Operations Center of the trip via the ENS at 11:57 p.m. CST on June 24, 1988.
An investigation of the trip determined that the cause was due to failure of a relay in the startup level controller for the
"B" Turbine Driven Reactor Feed Pump (TDRFP).
The operator
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was in the process of securing the "A" TDRFP for routine maintenance.
As part of the procedure, the "B" TDRFP was switched from three elements to single element control (startup level controller).
The startup level controller failed to latch into automatic and failed into manual such that feedwater flow went to a very low value.
The operator was unable to control the decreasing reactor water level and the reactor automatically tripped on low level.
Reactor water level was recovered before reaching Level 2 and the actuation of any ESF systems.
The safety relief valves did not lif t.
All systems functioned normally after the trip except that when the Reactor Recirculation (RR) pumps automatically shifted to slow speed on the low feedwater flow cavitation interlock, the "B" RR pump did not pick up in slow speed but went to "off".
This problem was determined to be a consequence of the pump start logic and coastdown characteristics and was not a safety issue.
The problem with the startup level control relay was determined to be a potential generic problem with oxidization of Agastat type GPI relay contacts when used in low current applications.
The licensee's actions in response to the generic issues are discussed in Paragraph 8.c. above.
The licensee repaired the relay problem with a spare relay on which they had burnished the contacts.
The reactor was restarted at 4:20 a.m. on June 26, 1988.
No violations or deviations were identified.
13.
Management Meeting (30702)
On June 10, 1988, NRC management met with IP management at the Region III Office in Glen Ellyn, Illinois to discuss the licensee's performance over the preceding two months.
In addition, the management meeting discussed observations of the licensee's emergency preparedness exercise that were documented in Inspection Report No. 50-461/88012.
Key personnel attending this meeting are identified by (#) in Paragraph 1 of this report.
NRC management discussed the observations of NRC staff participants in the April 26, 1988, exercise of the Clinton Power Station Emergency Plan.
The staff critique of that exercise identified weaknesses in the communication between the licensee's participants and the staff participants.
In particular, the staff was of the opinion that additional dialogue should have occurred during the exercise when protective actions were recommended and when the recovery was initiated.
The licensee acknowledged the staff's comments.
Licensee management then discussed events over the preceding two months.
Issues discussed included Environmental Qualification of electrical equipment; failure of containment air locks; contaminated water spill; and secondary containment boundary seal failure.
The licensee discussed corrective actions taken on the issues discussed.
The licensee then discussed the results of their Spring 1988 outage.
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14.
Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspectors, and which will involve some action on the part of the NRC or the licensee or both.
Two open items disclosed during the inspection were discussed in Paragraphs 8.b. and 8.c.
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 items disclosed during this inspection were discussed in Paragraphs 8.b. and 12.b.(2).
16.
Violations For Which A "Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requirement.
However, because the NRC wants te encourage and support licensees' initiatives for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.G.I.
These tests are:
(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and (5) it was not a violation that could reasonably be expected to havt been prevented by the licensee's corrective action for a previous violation.
A violation of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued is discussed in Paragraph 8.a.
17.
Exit Meeti,ngs (30703)
The inspectors met with licensee representatives (denoted in paragraph 1)
throughout the inspection and at the conclusion of the inspection on July 6, 1988.
The inspectors summarized the scope and findings of the inspection activities.
The licensee acknowledged the inspection findings.
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 documents / processes as proprietary.
The inspectors attended an exit meeting held between a regional based inspector and the licensee as follows:
Inspector Date A. Januska June 3, 1988 25