IR 05000315/1996014
| ML17333A749 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/22/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17333A747 | List: |
| References | |
| 50-315-96-14, 50-316-96-14, NUDOCS 9701280224 | |
| Download: ML17333A749 (93) | |
Text
U.'S.
NUCLEAR REGULATORY COHHISSION
REGION III
Docket Nos:
50-315, 50-316 License Nos:
50-315/96014'0-316/96014 Licensee:
Indiana Hichigan Power Company Facility:
Donald C.
Cook Nuclear Generating Plant Location:
1 Cook Place Bridgman, HI 49106 Dates:
October
November 23, 1996 Inspectors:
B. L. Bartlett, Senior Resident Inspector B. J. Fuller, Resident Inspector J.
D. Haynen, Resident Inspector N.
D. Hilton, Resident Inspector, Byron Plant Approved by:
Lewis F. Hiller, Jr., Chief Reactor Projects Branch
9701280224 970i22 PDR ADOCK 05000315
Executive Summar D. C.
Cook Units 1 and
NRC Inspection Report 50-315/96014, 50-316/96014 This inspection included aspects of licensee operations, maintenance, engineering, and plant support.
The report covers a 6-week period of resident inspection and includes the follow-up to issues identified during previous inspection reports.
~0eratioos
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A licensed senior reactor operator directed isolation of the number
Boric Acid Transfer Pump (BATP) without first verifying that all prerequisites had been completed.
Isolation of the number
BATP with an incomplete prerequisite resulted. in an inadvertent TS LCO Action statement entry for TS 3. 1.2.2 "Reactivity Control Systems Flow Paths."
Fortuitously, the error was identified and corrected within the allowed out of service time.
(Section 01.2)
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The licensee promptly and appropriately responded to degraded oil levels in the number 23 reactor coolant pump (RCP).
As plant conditions and equipment status changed, the licensee altered the repair plan as necessary.
Good, conservative decision making was evident during all phases of this evolution.
(Section 01.3)
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The inspectors determined that the licensee had properly implemented cold weather preparations, as required by procedure, for those items selected by the inspectors for verification.
(Section 01.4)
Haintenance The inspectors identified that the licensee was placing pressurizer power operated relief valves (PORVs) into a manual mode of operation without declaring the PORVs inoperable.
The inspectors determined that this practice did not appear to be outside the licensee's authorization basis, but that the operability evaluation memo supporting this position was poor. (Section H1.2)
The inspectors identified an example of work being performed on plant equipment without prior notification of the Shift Supervisor.
Licensee procedures did not have a precise definition of when Shift Supervisor authorization was required prior to starting work on plant equipment or systems, and the guidance which was provided appeared to be narrowly focussed.
The licensee initiated a review to assess the adequacy and effectiveness of their procedures and program for this aspect of work control.
This issue will be tracked as an inspector follow-up item.
(Section H1.3)
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The inspectors identified that the licensee's safety evaluation for a fication to the Spent Fuel Pool ventilation system provided inadequate documentation of the bases for a determination that no, USg existed.
After an independent review of the source documents, the inspectors concurred with the licensee's determination that a
USg did not exist.
Examples of poorly documented safety evaluation bases had been previously identified in IR 50-315/316-95012.
The failure to adequately document the bases for a safety evaluation was considered to be a violation of 10 CFR 50.59.
(Section El. 1)
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The inspectors observed the licensee's annual emergency preparedness drill and determined that the drill was planned and executed in a deliberate and well thought out manner.
The scenario appeared to be challenging to the operators and other drill participants.
The inspectors had no substantive comments that differed from the licensee's self-critique items.
No drill weaknesses were identified.
(Section Pl. 1)
Re ort Details Summar of Plant Status Unit 1 main transformer temperature limitations forced operation of the Unit at between 88.5 percent reactor power and 92.0 percent reactor power for most of the report period.
Reactor power was briefly reduced to 78.5 percent on October 15, 1996, in order to perform maintenance on one circulating water, pump.
Power was restored to 88 percent later that same day.
Unit 2 entered this reporting period in Mode 1 at 100 percent power.
On October 26, 1996, reactor power was reduced to 96 percent in order to perform testing on the auxiliary feedwater system.
Power was restored to 100 percent later that same day.
On November 23, 1996, reactor power was reduced to approximately 30 percent in order to add oil to the number 23 reactor coolant pump lower bearing reservoir.
Power was returned to 100 percent on November
'24, 1996.
01 Conduct of Operations 01. 1 General Comments 71707 Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations.
The conduct of operational activity that was observed was generally good.
Specific events and noteworthy observations are detailed in the sections below.
01.2 Un lanned Isolation of A Boration Flow Path Unit 2 a.
Ins ection Sco e
71707 On November 8, 1996, licensee personnel identified that the boration flow paths from the boric acid tanks (BATs) to the Unit 2 charging pumps via the boric acid transfer pumps (BATPs)
had been inadvertently isolated.
Technical Specification (TS) 3. 1.2.2.a required that one of the flow paths from the BATs through the charging pumps be available.
The boration flow path from the refueling water storage tank (RWST)
through the CCPs remained fully operable as required by TS 3. 1.2.2.b.
The inspectors reviewed the, licensee's immediate corrective actions and performed routine follow-up of the licensee's root cause analysis.
b.
Observations and Findin s
On November 7, 1996, licensee personnel isolated the boric acid filter in order to perform routine maintenance.
A clearance was hung in order to safely perform work on the filter.
The clearance isolated the number
BATP from the boration flow path.
A special instruction in the clearance order specified that the number
BATP be aligned prior to hanging the clearance on the number 4 BAT Due to a personnel error, the unit supervisor (US) failed to ensure that the number
BATP was properly aligned in accordance with the special instruction prior to authorizing the non-licensed equipment operators to hang the clearance order on the number 4 BATP.
The number
BATP was being used for recirculation of a BAT, and was incapable of providing a
source 'of boric acid to the CCPs.
Thus, when the non-licensed operators isolated the number 4 BATP, they isolated the only operable BATP.
TS 3. 1.2.2 required, in part, that a flow path be operable from the boric acid tanks through a
BATP to the CCPs and into the reactor coolant system (RCS).
Mith this path not available, the TS Limiting Condition for Operation (LCO) Action statement required that the path be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or that the unit be shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The number
BATP was isolated when the clearance was hung at 2:50 pm on November 7, 1996.
The number
BATP was made operable on November 8, 1996, at 5:45 am.
The licensee was in the LCO for 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> and 55 minutes, which was less then the TS limit of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The failure to ensure that at least one BATP was properly aligned was identified when another operating shift prepared to remove a boric acid blender from service for maintenance.
Prior to removing the blender from service, the operating crew attempted to verify that the boration flow path from the BATs was available, and identified the inadvertent isolation of all flow paths from the BATs, via a BATP, to the CCPs.
The number
BATP was promptly made operable and a condition report was written (CR 96-1794).,
The US stated that he could not remember the reason that he had believed that it was proper to authorize the non-licensed operators to hang the clearance order on the number 4 BATP.
The licensee temporarily suspended the US's senior reactor operator license pending additional training, and subsequent evaluation of the individual by licensee management.
The licensee took prompt corrective actions to address this failure to follow procedures.
,Therefore, this licensee-identified and corrected violation is being treated as a Non-Cited Violation (50-316/96014-01)
of
CFR 50, Appendix B, Criterion V, consistent with Section VII.B.1 of the NRC Enforcement Polic
.
Conclusions The self-identification of this error by licensed operators showed a
good questioning attitude and conservative operational practices.
The initial failure by the unit supervisor to ensure that the special instructions of the clearance order were followed resulted in an inadvertent TS LCO Action statement entry.
Fortuitously, the error was identified and corrected within the allowed out of service tim eration with a Leakin Reactor Coolant Pum Oil S stem Unit 2 Ins ection Sco e
71707 On October 30, 1996, a low oil level annunciator was received on reactor coolant pump (RCP). number 23 (Unit 2, number 3 pump).
The licensee performed an evaluation and determined that it was acceptabl'e to continue operation of the unit with the oil level annunciator energized.
The inspectors evaluated the licensee's analysis, monitored the performance of the RCP, and assessed the licensee's performance in this area.
Observations and Findin s
Licensee personnel performed a calculation following the receipt of the annunciator to determine when RCP lower bearing damage could occur.
A straight line interpolation using a constant oil leak rate showed that damage would not occur until early January, 1997.
This was based upon an estimate of the reservoir oil level at which damage had previously occurred to RCP 14.
In late September 1993, Unit 1 operators received a
low oil level alarm on the No.
RCP motor lower radial bearing.
The low oil condition eventually led to damage to the lower radial bearing.
Inspection reports 50-315/316-94002, 94007, 95009, and 96007 discuss this issue in additional detail.
Following the low oil level alarm of October 16, 1996, the licensee monitored RCP 23 vibration and bearing temperatures.
Additionally,
~ the licensee maintained component cooling water temperature as constant as possible to preclude oil level changes caused by shrink and swell.
The licensee planned to add oil to RCP 23's lower bearing reservoir on December 7,
1996; however, on November 17, 1996, one of the three probes (the X probe)
which measure shaft vibration failed.
After reviewing the failure data, the licensee conservatively decided to reschedule the RCP 23 work to November 23, 1996.
On November 20, 1996, licensee personnel were taking data on the RCP
vibration system in an effort to troubleshoot the cause of the X probe failure.
While data was being gathered, the Y probe failed (the troubleshooting efforts were not a proximate cause for the Y probe failure).
Licensee personnel utilized the third probe, the key phase, to monitor RCP 23 for vibration until the X and Y probes could be repaired.
On November 23, 1996, the licensee reduced Unit 2 power to about
percent.
The power. reduction reduced dose rates near RCP 23 and allowed workers to add oil to the reservoir and replace vibration monitoring equipment.
Approximately 2.5 gallons of oil were added.
The licensee had estimated that a little over 2 gallons would be required.
The vibration monitoring Y probe was repaired; however, local inspection showed that the X probe's failure mechanism was different.
For personnel safety reasons, no attempt was made to repair the X probe.
Unit 2 returned to full power the next day, following a slow increase.
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Based upon a constant leak rate, licensee personnel concluded that one additional power reduction to add oil would be required prior to the next scheduled outage of Unit 2 (Fall 1997).
c.
Conclusions The licensee promptly and appropriately responded to degraded oil levels in the number
RCP.
As plant conditions and equipment status changed, the licensee altered their repair plan as necessary.
Good, conservative decision making was evident during all phases of this evolution.
01.4 Cold Weather Pre aration a.
Ins ection Sco e
71714 The inspectors verified that the licensee had implemented the procedures necessary to prepare the plant for cold weather.
The inspectors reviewed procedures, verified that work documents were performed, and conducted plant walkdowns.
b.
Observations and Findin s
The inspectors reviewed selected action requests to ensure that the licensee had properly scheduled work that could effect plant components'esponse to cold weather conditions.
Early in the report period the inspectors identified a room heater that was not scheduled for repair until January, 1997.
The inspectors identified this issue to the licensee.
The licensee increased the priority of the heater repair.
This was the only such item identified by the inspectors.
The licensee subsequently identified one similar example of an opportunity to better schedule work required to prepare the plant for cold weather.
During the previous pe}formance of this inspection module (documented in inspection report 50-315/316-95012),
the inspectors had identified a
weakness in the appropriate use of temporary modification packages for the installation of sheet metal covers.
During'his inspection, the inspectors verified that temporary modifications had been performed, when appropriate, for the installation of sheet metal covers.
c.
Conclusions The licensee had properly implemented cold weather preparations, as required by procedure, for those items selected by the inspectors for verification.
Miscellaneous 0 erations Issues 08.1 0 en Violation 50-316 96004-01:
Temporary procedure changes which resulted in a change to the intent of the procedure.
TS 6.5.3. l.a required that procedures which effected plant nuclear safety, and changes thereto, be prepared, reviewed, and approved.
TS 6.5.3. l.e also required that procedural changes be reviewed to determine if an
unreviewed safety question existed.
To facilitate implementation of changes that did not change the intent of approved procedure, TS 6.5.3. l.a described the use of a temporary change, process.
The temporary change process allowed non-intent procedure changes to be issued up to 14 days before a safety evaluation was performed.
The inspectors identified two examples of procedure changes where the intent of the procedure had been effected, but the temporary change process had been used.
In the licensee's'esponse to this notice of violation (letter AEP:NRC: 1238A), the licensee discussed one aspect of the inspectors concerns.
However, the licensee failed to address the generic issue that changes to the intent of a procedure should not be implemented without prior performance of a safety evaluation.
During the exit meeting for this inspection report (50-315/316-96014),
the licensee stated that the corrective actions for the notice of violation consisted of extensive work in the area of defining "procedure intent,"
and that an additional response to the NOV would be provided.
This violation will remain open pending the receipt of the revised response.
II. Maintenance Ml Conduct of Maintenance Ml.l General Comments
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Ins ection Sco e
62703 and 61726 Portions of the following maintenance job orders, action requests, and surveillance activities were observed by the inspectors:
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IHP 4030.SMP.113 Pressurizer Pressure Protection Set III Functional Test and Calibration
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OHP 4021. 032. 001CD Operation of Diesel Generator 2 CD, Revision
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OHP 4030.STP.027CD
~ C0035798
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IHP 4030.STP.039
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IHP 6030. IMP.373 CD Diesel Generator Operability Test (Train A), Revision
Implement Design Change Package DCP-56, Modify Unit 2 Large Bore Piping Supports Reactor Coolant Pump No.
4 Underfrequency Bus lA Surveillance Test, Revision
Reactor Coolant Pump 2 Seal Water Injection Flow Lower Bearing Seal Water and No.
1 Seal Outlet Temperature Alarm and Indication Calibration, Revision
~ C0037881
~ R0044693
~ C0036841 Adjust Packing and Clean Valve 1-IFI-320-V2 Refurbish Actuator to Valve 2-VCR-204 Repair Splice on Ice Condenser Solenoid Valve 1-XSO-12 Observations and Findin s
The inspectors found the work performed under these activities to be generally of good quality with procedures present and in use.
0 erabilit of Pressurizer Power 0 crated Relief Valves PORV Durin Protection Set Calibration Unit
Ins ection Sco e
61726 During the performance of
IHP 4030.SHP. 113, Revision 1, the licensee removed the automatic operation function of *the PORVs, but did not consider the PORVs to be inoperable.
The inspectors interviewed licensed operators and IKC technicians and performed an independent review of the UFSAR in an effort to determine the appropriateness of the licensee's operability position.
Observations and Findin s
Technical Specification 4.4. 11. 1 required that the licensee perform a
calibration check and functional test of the pressurizer pressure protection channels every 18 months.
Procedure SHP. 113, referenced above, was the procedure used to perform the functional test and calibration.
Procedure SHP. 113 required that pressurizer pressure protection set III be made inoperable for performance of the surveillance.
The licensee recognized that rendering pressurizer pressure protection set III inoperable defeated the automatic open signal to two of the three pressurizer PORVs.
SHP. 113 directed that the PORV hand switches be placed in "manual" while the pressurizer pressure protection set III was inoperable, but the licensee did not declare the PORVs to be inoperable in this configuration.
Licensee staff had written a memo on October 23, 1990, addressing the issue of whether the PORVs continued to be operable during the time the PORV hand switches were in manual.
The memo stated that the PORVs were considered to be operable because:
a)
"Technical Specification 3/4.4. 11 contains a footnote which states that
"PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this
. leakage are not considered inoperable."
A parallel can be
I8)
drawn between this footnote (the closing of three block valves is permissible when isolating leaking PORVs)
and placing the control switches in the closed position when performing the surveillances under...."
b)
The UFSAR states that the PORVs operate either automatically or by remote manual control.
There is no statement in the UFSAR which laces a limitation on the -use of remote manual control.
This combined with the fact that manual activation of the isolation valve is acce table for achievin technical s ecification com liance leads to the conclusion that a
PORV with the control switch in the closed position ma be considered o erable...."
(Original emphasis retained).
The inspectors reviewed the UFSAR and determined that automatic PORV operation did not appear to be required by the plant's authorization basis, and concurred that the licensee's position on PORV operability appeared to be acceptable.
The inspectors were concerned, however, that the memo which served as the basis of the licensee's operability determination was weak.
Specifically, the inspectors were concerned that the memo did not specifically discuss the PORV's ability to perform its design basis functions while in the manual mode.'he inspectors also considered the comparison of PORV seat leakage to controller position contained in paragraph
"a)" above to be inconsistent with the applicable TS.
The licensee issued a condition report (CR) to address the inspectors concerns.
c.
Conclusions The inspectors determined that the licensee's practice of placing the PORV control switches in the manual mode position with out declaring the PORVs inoperable appeared to be consistent with the plant's authorization basis.
The inspectors were concerned, however, by weaknesses which they identified in the licensee's written basis for the PORV operability determination.
The licensee initiated a
CR to review and address the inspector's concerns.
Ml.3 Shift Su ervisor SS Authorization of In-Plant Work Both Units a.
Ins ection Sco e
62703 While performing a routine walkdown of a plant modification, the inspectors identified that licensee procedures did not require SS authorization prior to commencing weld modifications on Component Cooling Water (CCW) pipe.
The inspectors reviewed licensee procedures, interviewed personnel, and evaluated the licensee's program for obtaining SS authorization prior to working on plant components.
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Observations and Findin s
On November 8, 1996, the inspectors were performing a routine walkdown of a plant modification.
During the walkdown of DCP-56,
"Modify Unit 2 Large Bore Piping Supports,"
the inspectors noted that SS authorization had not been obtained prior to the start of work activities.
Licensee staff informed the inspectors that the SS would authorize certain portions of the modification (such as making a fire detector inoperable during welding activities), but that the SS did not authorize the implementation of all modifications.
For DCP-56, the SS would not be aware of, or authorize, the welding-of lugs on component cooling water (CCW) piping (the focus of most of DCP-56).
The inspectors reviewed the licensee's procedures for planning and performing work and identified only two, vague, SS authorization statements.
Nuclear Plant Maintenance Manual NPM-02CM, Revision 6,
"Process Instruction Corrective Maintenance,"
stated in step 5.7. 1.D, that
"The SS/US Required (approval to start work required) field shall by "Y" if work will, affect the operability of equipment and a Clearance will not be used while performing the work or as required by procedure."
In addition, step 5. 13.3.B, stated,
"Obtain Shift Supervisor permission to start work, as applicable by signature on the JOA (job order activity)."
No definition of "as applicable," other than step 5.7. 1.0, existed in the licensee's procedures.
The design engineering work planners made the decision as to what work activities the SS should be required to authorize prior to the initiation of work based on these instructions.
The inspectors were concerned that the licensee's procedural requirements were both narrow and vague.
The inspectors discussed with licensee staff the potential that a welder could accidentally
"burn through" the CCW pipe and cause equipment to become unexpectedly inoperable in a manner which introduced a plant transient.
Operator response to such a transient could be complicated by a lack of awareness of the potential source of equipment degradation.
The licensee personnel stated that they performed ultra-sonic testing (UT) of the CCW pipe in the area to be welded to ensure that there was enough wall thickness to prevent a concern with "burn through".
Licensee staff also stated that if there were special considerations, such as a tight work space, wor k would be performed during an equipment outage when the applicable system or equipment could be taken out of service.
In response to the inspectors'oncerns regarding the effectiveness of the licensee procedures and program for control of work, the licensee initiated CR 96-1892 and placed a hold on similar support attachments until the CR was resolved.
Conclusions The inspectors identified that licensee procedures did not have a
precise definition of when SS authorization was required prior to starting work on plant equipment or systems, and that the guidance
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provided'ppeared to be narrowly focussed.
The licensee initiated a
review to assess the adequacy and effectiveness of their procedures and program.
This issue will be tracked as an inspector follow-up item (50-315/316-96014-02)
pending resident inspector review of the licensee's assessment of procedure and program effectiveness.
Miscellaneous Maintenance Issues Closed Licensee Event Re ort 50-316 94004-00:
Use of Notice of Enforcement Discretion to Extend Type B&C Leak Rate Testing Interval.
The licensee made a" timely request for a license amendment extension of Type B&C leak rate testing intervals.
Due to an unexpected delay in the publication of the Federal Register, the approved license amendment was not issued until the TS surveillance interval was exceeded.
A three day extension to the TS leak rate surveillance interval was granted by the NRC to cover the time that the TS surveillance interval was exceeded before the issuance of a license amendment.
This was an isolated case, and the inspectors had no further concerns.
This minor violation of TS 4.6. 1.2.d is being treated as a Non-Cited Violation (50-316/96014-03),
consistent with Section IV of the NRC Enforcement Poli c
.
This LER is closed.
Closed Unresolved Item 50-315 316-96007-02 Review of the possible pre-conditioning of the station batteries during surveillance testing.
In response to a vital battery cell being identified as inoperable during a surveillance test, licensee personnel performed a review of the test procedure.
The personnel recognized that the procedure could allow the pre-conditioning of the batteries during surveillance testing if a note in the front of the procedure was not strictly complied with by the electricians.
The licensee performed a review of the history of the surveillance tests and interviewed the electricians and determined that no such pre-conditioning had occurred.
The licensee's review found no examples where the procedure caused a failure to comply with TS by delaying the electricians notification of the shift supervisor.
The inspectors performed a review of the original test procedure and the licensee's revised test procedure, and no additional concerns were identified.
The inspectors consider this a closed item.
III. En ineerin Conduct of Engineering Inade uatel Documented Safet Evaluation For A Plant Modification E ui ment Common to Both Units Ins ection Sco e
37551 During a review of DCP 12-DCP-0049, Revision 0, "Spent Fuel Pool (AFX)
Filtration System Bypass Damper Replacement;"
the inspectors identified that the licensee's safety evaluation was inadequate.
The inspectors also.reviewed Revision 1 of 12-DCP-0049 which added the modification of
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the engineered safety features (AES) ventilation system (two ventilation trains per reactor for a total of four units) to the scope of work.
Observations and Findin s
The licensee's spent fuel pool exhaust (licensee system designator AFX)
and engineered safety features (licensee system designator AES)
ventilation systems contained roughing filters, high efficiency particulate air (HEPA) filters, and charcoal filters.
In order to prolong the life of the charcoal filters, bypass dampers were utilized to bypass the air flow around the charcoal filters when they were not required.
By design, the bypass dampers automatically close following an AFX or AES actuation signal, forcing the air to flow through the charcoal beds prior to discharge out the unit vent exhaust.
Excessive leakage past the bypass dampers during a postulated event could result in degradation of system effectiveness in controlling the release of radioactive material.
TS 4.7.6. 1 required the performance of periodic surveillance tests of the efficiency of the AFX and AES filtration units.
The existing bypass dampers had periodically failed to meet their bypass leakage limits due to age induced degradation.
The licensee decided to replace the bypass dampers as a result of these surveillance results.
The new dampers were not identical replacements, but would perform the same function.
As of the date of this inspection, the licensee had started the replacement of the AFX dampers, but due to operational limitations had not yet begun the replacement of the AES dampers.
As required by 10 CFR 50.59, the licensee performed a safety evaluation prior to starting the replacement of the AFX bypass dampers.
Licensee personnel performed both a safety screening and an unreviewed safety question determination (USED)
as required by their procedures.
The inspectors reviewed the safety screening and USED.
The UFSAR did not contain specific design and operation information for the dampers, but did discuss the function of charcoal filter bypass.
The licensee's documented basis for determining that the damper replacement did not represent a US( relied upon the lack of specific design and operation information for the dampers in the UFSAR, and did not adequately discuss the functional and performance characteristics of the original and proposed replacement dampers.
As a result of this inadequacy in the licensee's documentation, the inspectors were required to independently retrieve and review the dampers'ender literature in order to determine whether any significant differences in the function or operating characteristics of the original and replacement dampers existed.
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Based upon the results of their independent review of the modification package and. the damper vendors'iterature, the inspectors concluded that the performance characteristics of the replacement dampers were such that their installation would not alter the function or performance of the AFX or AES systems, and that a US( did not exist.
CFR 50.59,
"Changes, tests and experiments,"
section (b)(l),
requires, in part,
"The licensee shall maintain records of changes in the facility...
These records must include a written safety evaluation which provides the bases for the determination that the change...
does not represent an unreviewed safety question."
The licensee's failure to include in the written safety evaluation an adequate bases for the determination that the change to the AFX system did not represent a
USA is a violation (50-315/316-96014-04)
of 10 CFR 50.59.
In response to the inspectors findings the licensee began immediate corrective actions.
These actions consisted of:
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Reviewing the safety evaluations for the minor modifications, preventive maintenance, requests for change (RFC),
and DCPs that had been performed by design engineering for the upcoming unit
and unit 2 1997 refueling outages.
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Reviewing the safety evaluations for modifications being implemented while on line, prior to the start of the modifications.
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Giving training to the safety evaluation performers, reviewers, and supervisors with the training to be completed by December 31, 1996.
Inspection report (IR) 50-315/316-95012, discussed two examples of poor quality 10 CFR 50.59 safety evaluations.
The safety evaluations discussed in that report were incomplete, non-responsive to the questions, and lacked detail.
In one of the two examples in IR 50-315/316-95012, the identified concerns were nearly identical to the concerns identified in this report.
The inspectors considered the repetitive nature of this performance issue in accordance with the guidance of Section IV of the NRC Enforcement Polic
.
Conclusions The inspectors identified that the licensee's safety evaluation for a modification to the Spent Fuel Pool ventilation system provided inadequate documentation of the bases for a determination that no USA existed.
After an independent review of the source documents, the inspectors concurred with the licensee's determination that a
USA did not exist.
Examples of poorly documented safety evaluation bases had been previously identified in IR 50-315/316-95012.
The failure to adequately document the bases for a safety evaluation was considered to be a violation of 10 CFR 50.59.
Hiscellaneous En ineerin Issues Closed Violation 50-316 96002-01 Failure to have a radiation monitor that met the requirements of 10 CFR 70.24.
This violation was written when it was discovered by NRC inspectors that the licensee did not have a new fuel vault criticality monitor that complied with 10 CFR 70.24.
The licensee attributed the failure to comply with 10 CFR 70.24 to a lapse in their program for integrating regulatory requirements into the plant design. bases, and a loss of institutional expertise in criticality accident monitoring.
These deficiencies resulted in the installation of portable criticality monitoring equipment that was not captured in either the modification process or on plant drawings, and to the lack of procedures for response to this monitoring equipment.
The licensee temporarily installed a radiation detector inside the new fuel vault to comply with 10 CFR 70.24, wrote procedures and instructions to ensure the monitor setpoint complied with 10 CFR 70.24, added new fuel vault criticality actions to the emergency plan, wrote a
procedure to guide radiation protection personnel in the response to a criticality alarm, and provided training to RP personnel concerning the new procedure.
On October 28, 1996, the licensee was issued an exemption from the requirements of.10 CFR 70.24 by the NRC.
The bases for the exemption was that inadvertent or accidental criticality would be precluded through compliance with the Cook TSs, the geometric spacing of fuel assemblies in the new fuel storage facility and spent fuel storage pool, and administrative controls imposed on fuel handling procedures.
The inspectors consider this item closed based upon the licensee's corrective actions and the NRC issued exemption to
CFR 70.24.
Closed Unresolved Item 50-316 96002-03 The NRC inspectors had questioned the licensee's emergency procedures for a criticality event, including the need for drills and the need for either constant monitoring or a detector which would send signals to a remotely monitored location.
NRC resolution of the question whether the licensee, was required to comply with 10 CFR 70.24(a)(l)
or (a)(2)
was an Unresolved Item.
As stated in paragraph E8. 1 above, the licensee was granted an exemption to
CFR Part 70.24 on October 28, 1996.
Prior to that time, the licensee had revised the emergency plan to incorporate actions to take in the event of an inadvertent criticality and had modified the plant equipment to ensure compliance with 10 CFR 70.24.
This unresolved item is closed based upon the licensee's actions and the NRC issued exemption to
CFR 70.24.
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Rl Radiological Protection and Chemistry Controls (71750)
During the resident inspection activities, routine observations were conducted in the areas of radiological protection and chemistry controls using Inspection Procedure 71750.
No discrepancies were noted.
Pl
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Pl.l Emergency Preparedness Emer enc Pre aredness Exercise a.
Ins ection Sco e
71750 b.
On November 5, 1996, the licensee held the annual emergency preparedness drill required by NRC regulations.
The NRC inspectors observed selected portions of the annual drill.
Observations and Findin s
The inspectors observed portions of the drill from the simulator (used as the control room for the drill), the technical support center (TSC),
the emergency off-site facility (EOF),
and the joint public information center (JPIC).
Prior to the exercise, the inspectors reviewed the licensee's scenario and concluded that it was well written and that it did a good job of maintaining realism considering the multiple postulated failures necessary to accomplish the drill goals.
During the drill, the inspectors observed that licensee staff made and implemented conservative decisions in a timely manner.
One example of a conservative decision was the SS direction to initiate a manual reactor trip instead of a controlled shutdown.
This decision was discussed with other personnel prior to implementation, and was based upon the safety conscious desire to avoid potential fuel damage which might have resulted during a controlled shutdown under the conditions established in the drill scenario.
The NRC inspectors'bservations and concerns were all independently captured by the licensee's self-assessment process, and were properly categorized therein.
Improvement items identified by the licensee during post-drill critiques included:
The site area coordinator should have confirmed a protective action recommendation (PAR) with the State of Michigan prior to the PAR being formally sent to the State personnel.
Good communications were practiced in phones conversations, including the use of repeat-backs; however, repeat-backs were not always utilized during face-to-face communications.
A licensee manager identified the desirability of additional management guidance regarding the appropriateness of authorizing higher doses to plant workers if that dose could prevent dose to members of the general public.
The inspectors observed that the NRC office in the EOF was being utilized for the storage of empty licensee file cabinets.
Approximately one half of the NRC office was taken up in this manner.
The inspectors requested that the licensee store the cabinets elsewhere.
c.
Conclusions The inspectors observed the licensee's annual emergency preparedness drill and determined that the drill was planned and executed in a deliberate and well thought out manner.
The scenario appeared to be challenging to the operators and other drill participants.
The inspectors had no substantive comments that differed from the licensee's self-critique items.
No drill weaknesses were identified.
Sl Conduct of Security and Safeguards Activities (71750)
During normal resident inspection activities, routine observations were conducted in the areas of security and safeguards activities using Inspection Procedure 71750.
No discrepancies were noted.
Fl Control of Fire Protection Activities (71750)
During normal resident inspection activities, routine observations were conducted in the area of fire protection activities using Inspection Procedure 71750.
No discrepancies were noted.
Xl Meetings and Other Activities Xl. 1 Management Meetings On November 18, 1996, there was a management meeting between NRC and the licensee to discuss the licensee's performance of operability determinations, the operations improvement initiative, the FSAR revalidation project, and an update on equipment preconditioning prior to surveillance tests.
The Regional Administrator (RA) concluded the meeting by stressing the importance of the licensee operating within the plant's design basis envelope.
The RA also stated that success of the licensee's improvement initiatives would be measured by changes in performance, not processes.
A copy of the material used by the licensee to discuss these matters is attached.
J I
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Xl.2 Exit Neeting The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on December 2,
1996.
The licensee acknowledged the findings presented.
PARTIAL LIST OF PERSONS CONTACTED Licensee
¹J. Allard," Maintenance Superintendent
¹T. Andert,
- H. Ackerman, Manager Nuclear Licensing
- H. Barfelz, NUC Licensing
¹*A. Blind, Site Vice President
- S. Brewer, Manager Regulatory Affairs
¹H. Depuydt, Licensing
- E. Fitzpatrick, Senior Vice President Nuclear Gen
- S. Foley, Senior Engineer
- R. Gillespie, OPS Manager
- J. Kingseed, Manager Nuclear Safety and Analysis
¹*J. Kobyra, Manager Nuclear Engineering
¹D. Hafer, Manager Plant Engineering
¹S.
Hodge,
¹H. Horvath, Plant Performance Assurance
¹D. Loope,
- G. Hartin, Jr.,
Nuclear Licensing Engineer
¹D. Noble, Radiation Protection Superintendent
¹T. Postlewait, Site Engineering Support Manager
¹R. Ptacek, Licensing
¹T. guaka, Project Management 5 Inst. Services
¹P. Russell, Plant Protection
- J. Sampson, Plant Manager
¹*P. Schoepf, Manager Safety-Related Systems
¹G.
Van Bladeren, Maintenance Rule Coordinator
¹L. Van Ginhoven, Material Management Department NRC eration Group
¹*B. Bartlett, Senior Resident Inspector, RIII
- A. Beach,
- J. Caldwell, Acting Director, DRP, RIII
- B. Clayton,
¹*B. Fuller, Resident Inspector, RIII
- J. Hickman, Project Manager, NRR
¹*J. Maynen, Resident Inspector, RIII
- L. Hiller,
- H. Ring, Chief, Division of Reactor, RIII
¹Denotes those present at the December 2,
1996 exit meeting.
- Denotes those individuals attending the AEP meeting on November 18, 1996.
IP 37551 IP 61726 IP 62703 IP 71707 IP 71750 IP 71714
~0ened INSPECTION PROCEDURES USED On-site Engineering Surveillance Observations Maintenance Observation Plant Operations Plant Support Activities Cold Weather Preparation ITEMS OPENED and CLOSED 50-315/96014-01 (DRP)
50-315/316-96014-02(DRP)
50-316/96014-03(DRP)
50-315/316-96014-04(DRP)
Closed.
NCV IFI NCV VIO Unit Supervisor failure to follow procedure Licensee evaluation of SS work authorization Leakrate testing interval extension Inadequate justification for a safety evaluation 50-316/94004-00 50-316/96002-01 50-316/96002-03 50-315/316-96007-02 50-315/96014-01 50-315/316/96014-03 Discussed 50-316/96004-01 LER Failure to comply with Tss VIO Failure to have a radiation monitor that met the requirements of 10 CFR Part 70.24 URI guestions concerning the lack of drills and emergency response plans for the new fuel vault URI Review of the possible pre-conditioning of the station batteries during surveillance testing NCV Unit Supervisor failure to follow procedure NCV Leakrate testing interval extension VIO Temporary procedure changes which resulted in a change to the intent of the procedure
AES AFX AEP BAT BATP CCP CCW CFR CR DCP DRP ELO EOF FSAR HEPA I&C JPIC LER LCO NPH NRC PAR PDR PORV PWR RCP RCS RFC RWST SAR SS TS TSC US USAR USQ USQD UT List of Acronyms Engineered Safety Features Ventilation System Fuel Pool Ventilation System American Electric Power Boric Acid Tank Boric Acid Transfer Pump Centrifugal Charging Pump Component Cool'ing Water Code of Federal Regulations Condition Report Design Change Package Division of Reactor Projects Emergency Leak-off Line Emergency Operations Facility Final Safety Analysis Report High Efficiency Particulate Absorber Instrumentation and Controls Joint Public Information Center Licensee Event Report Limiting Condition For Operation Nuclear Plant Hai.ntenance Hanual Nuclear Regulatory Commission Protective Action Recommendation Public Document Room Power Operated Relief Valve Pressurized Water Reactor Reactor Coolant Pump Reactor Coolant System Request For Change Refueling Water Storage Tank Safety Analysis Report Shift Supervisor Technical Specification Technical Support Center Unit Supervisor Updated Safety Analysis Report Unreviewed Safety Question Unreviewed Safety Question Determination Ultra-Sonic Testing
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AQERfCAM EEECTRM PCVfEP COOK NUCLEAR PLA.NT Bridgrrian, Michigan AEP/NRC RIIIManagement Meeting November 18, 1996 Wh
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AEP / NRC RIIIManagement Meeting November 18, 1996 AGENDA Operability Determinations J. B. Kingseed Operations Improvement Initiative B. K. Gillespie FSAR Revalidation Project G. Martin Jr.
1 ~
~
reconaitj.omng - P G Schoepf
Operabili Determinations J. B. Kingseed Manager - Nuclear SafetylAnalysis Nuclear Engineering Department AEP / NRC RIIIManagement Meeting 11/18/96
t
~ Recognition ofneed
~ Quality
~ Improvement Initiatives AEP / NRC RIIIManagement Meeting 11/18/96
Recent Operabili Performance Failure to recognize need for formal determination Degraded Conditions Test Results Failure to appropriately document condition AEP / NRC RIIIManagement Meeting 11/18/96
(
Recent Operabili Performance
~ Determinations correct
~ Document quality not always acceptable Failure to perform a rigorous assessment Over-reliance on judgment Format not consistent with procedural requirements AEP / NRC RIIIManagement Meeting 11/18/96
GL 91-18: Degraded And Nonconforming Conditions
~ Scope ofapplicability
~ Scope ofdetermination What is degraded State safety function Assess circuxnstances Deteimi.ne requirements Determine safe Plant configuration Provide basis for Operability AEP / NRC RIIIManagement Meeting 11/18/96
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Operability Improvement Initiatives Plant Manager Standing Order PMSO)
173, Interim Guidance on Operability Determinations
~ Reinforce identification ofconditions
~ Applicabilityto test results
~ Covered Operability and condition identification in Root Cause training AEP / NRC RIIIManagement Meeting 11/18/96
Operability Improvement Initiatives
~ STAs reviewing Action Requests for Operability concerns New procedure Development underway Procedure/Training By 12/31/96
~ Heightened supervisory awareness iri interim AEP / NRC RIIIManagement Meeting 11/18/96
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Operations Improvement Initiative Bob Gillespie Operations Manager, Plant Operations Group AEP / NRC RIIIManagement Meeting 11/18/96
Operations Improvement Initiative Ops Manager Principles Ops Centered Culture No Events Focus on Fundamentals AEP / NRC RIIIManagement Meeting 11/18/96
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Operations Centered Culture Definition - Operationally based thinking, and approach to the way we perform our jobs.
Plant wide Conservative and Questioning Attitude Expected Plant response anticipated &, controlled
~ Consistently results in; Proper preparation and scheduling ofwork, Contingencies well defined, Briefings and support staff established, and Availabilityofneeded resources.
AEP / Nl<C MIIManagement Meeting 11/18/96
Operations Centered Culture Implementation Strategy
~ Work ItNow
.
~ Eliminate Barriers to Plant Performance Procedure adherence standards
"Infrequently Performed Evolutions" process When operating outside normal configuration, apply design review process.
Oper'ator workarounds PSA use
~ Strengthen Team Perfonnance
. Shift Manager Development Strengthen operational experience in support groups Management support ofconservative decision making AEP / NRC RIIIManagement Meeting 11/18/96
Conservative Decision Making Wlien faced with unexpected or uncertain conditions, personnel must promptly identify a transition point at which efforts to keep the unit on-line or on-schedule, are no longer conservative or reasonable.
Once this point is reached, actions to place the unit in a safe condition by reducing power, tripping the reactor, or suspending core alterations must be taken without hesitation.
Fundamentals OfConservative Decision Making EVhcn activities or deficiencies result in unexpected or uncertain conditions, a prompt assessment is made to dctcrmine a safe and conservative course ofaction i Thc possible adverse impact ofconservative actions on a unit's Capacity I'ac(or is irrelevant. In thc long term, conservative decisions willhave a positive effect on plant performance overall.
+ 'Conservative actions taken under transient conditions arc always appropriate and ncvcr second guessed.
+
Reactivity additions are applied only with thc utmost care and vigilance.
Supervisory personnel assigned oversight roles arc clearly Identified and focus broadly on crew pcrformancc and overall plant conditions.
Conservative Decision Makers;
+ Maintain individual integrity
+
Keep industrial and nuclear safety as the top priority
+
Prevent or minimize cquipmcnt damage
+ Know what to cxpcct prior to taking actions
+
Plan forcontingencies
+ Involve all team members in key decisions and cffcctivclycvaluatc team input
+
Seek and accept assistance
+
Insist on high quality training and high quality procedures
+ Take pride h their understanding ofsystems design and integrated phnt response and actively pursue increasing their knowlcdgc AEP / NRC RIIIManagement Meeting 11/18/96
Operations Dept.
Human Performance Triangle SigniQcant Events Human performance errors that result in reportability to any regulatory agency or notice ofviolation Reportable Events year-to-date Events Human performance errors that did cause, or were likely(as determined by the Shift Manager, Section Head, or Department Superintendent) to have caused, equipment damage, personnel injury, radioactive release or spill, or loss ofgeneration; or that significantly reduce the level ofsafety ofthe plant.
Events year-to-date Errors Any human performance error not definable in two previous categories AEP / NRC RIIIManagement Meeting 11/18/96 Errors year-to-date
I f f I
OPS Initiated vs Investigated 6 ninth rollingaterage ofCRs/thy 2.5 1.5 0.5
~ Initiate
~ Investiltte Jan - Jun Feb-Jul Apr-Sept May-Ckt AEP / NRC RIIIManagement Meeting 11/18/96
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Focus On Fundament;als
~ Make Expectations Known
~ Provide Coaching and Feedback
~ Promote and Expect Accountability AEP / NRC RIIIManagement Meeting 11/18/96
Ops Manager's Expectations Expectations are reinforced by; Operations Manager Meetings with ShiA/Crew each Week, Operations Manager Meetings with Training each requalification period, Daily Operations Department Meeting, and Ops lead correspondence.
AEP / NRC RIIIManagement Meeting 11/18/96
Coaching And Feedback Operations Manager evaluates Shift/Crew performance during Simulator Training Shift Managers evaluate Shift/Crew performance during Simulator Training Daily Visits/Oversight to Control Room Complex AEP / NRC RIIIManagement Meeting 11/18/96
Accountabili
~ Lessons Learned are expected for Human Performance errors
~ A Human Performance Scorecard is provided to each Shift/Crew and Section Manager
~ Individuals are held accountable for their actions AEP / NRC RIIIManagement Meeting 11/18/96
Operations Improvement Initiative Ops Manager Principles Ops Centered Culture No Events Focus on Fundainentals AEP / NRC RIIIManagement Meeting 11/18/96
FSAR Revalidation Project, G. Martin Jr.
Project Manager/Licensing Engineer AEP / NRC RIIIManagement Meeting 11/18/96
]II
FSAR Revalidation Project
~ Purpose
~ Reason for Project
~ Project Goals
~ Implementation FSAR Accuracy Reviews Programmatic Reviews Project Execution Project Schedule AEP / NRC RIIIManagement Meeting 11/18/96
~ ~
~
FSAR Revalidation Project U
OSC Verifythat the Cook Plant FSAR accurately represents the Plant Design and Processes.
Resolve any identified. discrepancies.
Ensure that mechanism are in place to maintain that accuracy in the future.
AEP / NRC RIIIManagement Meeting 11/18/96
FSAR Revalidation Project easons or ro ect Respond to NRC generic concerns regarding the accuracy ofthe FSARs and inconsistencies with Plant Design and Operating Procedures.
Minor discrepancies between the Cook Plant FSAR and Plant Design/Procedures, none of which are safety significant, found during the course ofpreliminary review efforts.
AEP / NRC RIIIManagement Meeting 11/18/96
FSAR Revalidation Project ro ect oa S
To demonstrate that;
~ The Cook Plant FSAR is accurate and can be used with confidence, and
~ Adequate programmatic controls exist to ensure that Cook Plant FSAR willaccurately reflect f'uture changes to the Plant Design/Processes.
AEP / NRC RIIIManagement Meeting 11/18/96
FSAR Revalidation Project m
ementation FSAR Accuracy Review Initial assessments ofFSAR accuracy will address at least six Plant Systems.
Assessments willconsist ofin-depth, line by line, reviews ofthe FSAR versus the actual Plant design and Procedures.
Reviews willaddress all aspects ofselected Systems.
AEP /NRC RIIIManagement Meeting 11/18/96
I
FSAR Revalidation Project m
ementation FSAR Accuracy Review continued System and Criteria Assessments willbe performed using information Rom;
~ Technical Specifications,
~ Plant Procedures,
~ Interviews with Plant Operations personnel,
~ Design Basis Documentation Project,
~ System Specifications,
~ 50.59 Reviews,
~ Docketed Correspondence, and
~ Independent System Walkdowns.
AEP / NRC RIIIManagement Meeting 11/18/96
)
II t
FSAR Revalidation Project Im ementation FSAR Accuracy Review continued Necessary changes to the FSAR willbe included, to the extent practicable, in the 1997 yearly FSAR Update.
Discrepancies identified during the course of the project willbe documented, tracked, and resolved systematically.
50.59 Reviews willbe performed where necessary.
AEP / NRC RIIIManagement Meeting 11/18/96
FSAR Revalidatioo Project m
ementation Programmatic Review
. Examine the Processes that result in changes to the Plant Licensing Basis, Plant Design or Operating Procedures and the manner in which such changes are documented in the FSAR.
Identify changes, ifany, that have not been accurately reflected in the FSAR.
AEP / NRC RIIIManagement Meeting I I/18/96
F ARReva i ation Project Im ementation Programmatic Review - continued Determine why inconsistencies between the FSAR and Plant Design/Procedures, ifany, exist.
Take or establish actions to ensure that mechanisms exist to ascertain that future changes to the Plant Design or Procedures are appropriately incorporated in the FSAR.
AEP / NRC RIIIManagement Meeting 11/1 8/96
FSAR Revalidation Project m
ementation Project Execution Allproject activities are to be conducted in a manner consistent with the philosophy ofNEI 96-05 Cooperative effort between AEPNGG and EPRI AEP / NRC RIIIManagement Meeting 11/18/96
F R Reva i ation Project m
ementation Project Execution - continued Project Team-consists offour individuals with appropriate technical and regulatory experience (i.e. FSAR Update, SSFI, DBD, etc.) under the direction of an AEPNGG Project Manager.
AEPNGG personnel (Engineering, Licensing, Operations, Maintenance, and Quality Assurance)
willsupport the dedicated Project Team.
AEP / NRCIIIManagement Meeting 11/18/96
FSAR Revalidation Project m
ementation Project Schedule Pilot review ofthe ECCS Section of the FSAR started in October, 1996.
Project preparations presently underway.
Formal project kick-offon January 6, 1997.
AEP / NRC RIIIManagement Meeting 11/18/96
'
le
~
~
eva i ation m
ementation 10 ect Project Schedule - continued.
Necessary changes to the FSAR willbe included, to the extent practicable, in 1997 FSAR submittal.
Open item resolution phase to continue beyond July, 1997, ifnecessary.
Follow-up (post-July 1997) project activities willbe based on the results ofthe initialproject review.
AEP / NRC RIIIManagement Meeting 11/18/96
Pleconditioning Paul G. Schoepf Manager - Safety Related Mechanical Systems Nuclear Engineering Department AEP / NRC RIIIManagement Meeting 11/18/96
~ Cook Plant History
~ Efforts to Address Issue
~ Perspectives on Issue AEP / NRC RIIIManagement Meeting 11/18/96
Preconditioning Cook Plant History February 1996 - Notice ofViolation TDAFP Turbine TATValve Tested/Operated prior to Pump Surveillance
~ EDG rolled prior to Fast Start Surveillance Lubricating/Exercising EDG Fuel Racks prior to Surveillance AEP / NRC RIIIManagement Meeting 11/18/96
l
P re oond.itioning Efforts To Address Issue
~ Marcli, 1996 - Notice ofViolation response Reason - Absence of guidance Task Team formed - Guidance document developed/issued 3-15-96 Reviewing Inage and Outage Procedures No significant procedural issues discovered to date AEP / NRC RIIIManagement Meeting 11/18/96
Preconditioning Cook Plant Definition
"Anyaction taken by the Licensee which may mask System or Component malfunction, degradation, or design deficiencies and which, ifperformed ri r t a Technical Specification (TS) Surveillance, may affect the TS Surveillance results" AEP / NRC RIIIManagement Meeting 11/18/96
T
P reconditioning Perspectives On Issue
~ Lack ofIndustry information only minor Inspection history
~ Lack ofan Industry accepted definition and/or detailed Inspection guidance Potential to over-ride good operating practices/jeopardize equipment AEP / NRC RIIIManagement Meeting 11/18/96 40