ML17335A390
| ML17335A390 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/02/1998 |
| From: | Grobe J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Powers R AMERICAN ELECTRIC POWER CO., INC. |
| References | |
| NUDOCS 9812160047 | |
| Download: ML17335A390 (50) | |
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UNITED STATES NUCLEAR REGULATORYCOMMISSION REGION III 801 WARRENVILLEROAD LISLE, ILLINOIS605324351 December 2,
1998 Mr. R. P. Powers Senior Vice President Nuclear Generation Group American Electric Power Company 500 Circle Drive Buchanan, Ml 49107-1395
SUBJECT:
MID-YEARINSPECTION RESOURCE PLANNINGMEETING - D. C. COOK
Dear Mr. Powers:
On November 4, 1998, the NRC staff held an inspection resource planning meeting (IRPM).
The IRPM provided a coordinated mechanism for Region IIIto adjust inspection schedules, as
. needed, prior to the conclusion of the Plant Performance Review cycle in April 1999.
In November 1998, you informed me that you have chartered a short-term independent Engineering Issues Review Group to evaluate the effectiveness of the engineering restart readiness process as described in your restart plan. You expected this effort to be completed.
in mid-December and indicated that this assessment would likely result in additional actions that would impact planned NRC engineering and restart inspections.
Based on this discussion, we have placed engineering and restart inspections on hold, pending review of your revised schedule that incorporates the actions from your short term assessment.
Enclosure 1 to this letter, advises you of our planned inspection effort for the next 6 months at D. C. Cook.
This attached information is provided to minimize the resource impact on your staff and to allow for scheduling conflicts and personnel availability to be resolved in advance of inspector arrival onsite. The rationale or basis for each inspection outside the core inspection program is provided so that you are aware of the reason for emphasis in these program areas.
Resident inspections are not listed due to their ongoing and continuous nature. contains a historical listing of plant issues, referred to as the Plant Issues Matrix (PIM), that was considered during the IRPM. The PIM includes only items from inspection reports or other docketed correspondence between the NRC and American Electric Power Company.
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R. Powers We willinform you of any changes to the inspection plan.
Ifyou have any questions, please contact Anton Vegel at (630) 829-9620.
Sincerely, Steven A. Reynolds for Docket Nos.: 50-315; 50-316 License Nos.:
DPR-58; DPR-74 John A. Grobe, Director Division of Reactor Safety
Enclosures:
1.
2.
Inspection Plan Plant Issues Matrix cc w/encls:
J. Sampson, Site Vice President R. Eckstein, Chief Nuclear Engineer D. Cooper, Plant Manager R. Whale, Michigan Public Service Commission Michigan Department of Environmental Quality Emergency Management Division Ml Department of State Police D. Lochbaum, Union of Concerned Scientists Distribution:
RRB1 (E-Mail)
RPC (E-Mail)
Project Mgr., NRR w/encls J. Caldwell, RIII w/encls C. Pederson, Rill w/encls B. Clayton, Rill w/encls SRI DC Cook w/encls DRP w/encls TSS w/encls
~
DRS (2) w/encls Rill PRR w/engls PUBLIC IE-gt w/encls ~
Docket Fileiw/encls GREENS IEO (E-Mail)
DOCDESK (E-Mail) p Q Q'3 DOCUMENT NAME: G:iCOOKttINSPPLN7.DCC To receive ~ co of this document, Indicate In the box M w w/o ettechtenct P i wlattachtenct tt ~ No OFFICE Rill NAME Schweibinz/co DATE 12/2/98 Rill
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Ve el 12/
/98 Rill Grobe 12/&98 OFFICIAL RECORD COPY
DONALDC COOK INSPECTION I ACTIVITYPLAN IP - Inspection Procedure Tl - Temporary Instruction Core - Minimum NRC Inspection Program (mandatory all plants)
Regional Initiative - Discretionary Inspections INSPECTION I
ACTIVITY IP40501
- IP40500, 71001 IP86750 IP84750 IP81700 IP81701 IP71 001 IP37550 TITLEI PROGRAM AREA SSFI Oversight Inspection Readiness Restart Assessment Phase 1
Corrective Action Team Inspection Readiness Restart Assessment Phase 2 Transportation and Solid Radwaste Effluents and Chemistry Security EP Program Licensed Operator Requal Engineering Team Inspection NUMBER OF NRC INSPECTORS I
INDIVIDUALS PLANNED DATES November 16, 1998-February 5, 1999 November 16-December 11, 1998 January 11 - 29, 1999 TBD March 15 - 19, 1999 May 17-21, 1999 April5-9, 1999 April 12 -16, 1999 February 1 - 5, 1999 TBD TYPE OF INSPECTION/
ACTIVITY-COMMENTS Regional Initiative Regional Initiative Regional Initiative Regional Initiative Core Core Core Core Core
, Regional Initiative
PLANT ISSUES MATRIX'C Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Cotumn = 'Date';
Beginning Date = '10/1/97';
Ending Date = '9/30/98 11/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 9/3/1998 Positive IR98019 NRC Plant 1C Support Effluent radiation monitors were well maintained, and monitor alarm setpoints were conservatively set. Workers responsible for these monitors were aware of procedural requirements and of contingency actions for monitor inoperability (Section R2.2).
2 9/3/1998 Positive IR 98019 NRC Plant 1C Support The radiolcgical environmental monitoring program was well conduct and associated results were documented as required. An increase in amount of tritium activity discharged to the lake had not resulted in doses exceeding regulatory limits and was adequately addressed by the licensee (Section R1.1).
3 9/3/1998 Positive 4
8/27/1998 Negative 5
8/27/1998 Positive IR 98019 IR 98016 IR 98016
-NRC Plant 5A Support NRC Operations 1A NRC Operations 1A Self-assessments performed by the licensee were thorough but continued to identify problems with radiation worker practices.
These problems were confirmed by inspector observations, and the licensee was planning additional corrective actions to address this issue (Section R7.1).
During control room observations the inspectors determined that the licensee's contingency planning for the loss of the remaining train of Residual Heat Removal (RHR) did not clearly address the least risk significant course of action. Because the contingency action was not clear, uncertainty existed among operators as to which action should be taken in the event the remaining train of decay heat removal was lost.
The remaining train of RHR was not lost and the licensee restored operabi1ity to both trains of RHR. (Section 01.2)
A shift Technical Advisor demonstrated a good questioning attitude wh he challenged the adequacy of compensatory measures contained in a surveillance procedure during a surveillance test. The compensatory measures provided guidance to prevent plant operators from removing both trains of the component cooling water system from service.
(Section 01.1a)
Page1of 35
PLANT 1SSUES MATR1X
.DG Gook Search Sortedby Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = 't0lt/91';
Ending Date =
'9130I98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 6
8/27/1998 Negative 7
8/27/1998 Negative IR 98016 IR 98016 NRC Operations 1B NRC Operations 2A 2B A review of the cause for a higher than expected value for calculated Pressurizer delta-T determined that, following the removal from service of the west Component Cooling Water (CCW) pump and the placement of the east CCW pump into operation, CCW system flowincreased, resulting in lower Volume Control Tank temperature.
The review also determined that control room operators did not monitor plant paramet closely enough to detect the change in Pressurizer delta-t, resulting in exceeding procedural limits. Technical Specification limits were not exceeded.
(Section 01.1b)
During a routine review of the control room caution tag logs, the inspectors identified a number of tags that were still hanging even though the issues listed on the tag had been resolved.
A quality assurance audit had previously identified a discrepancy between two procedures, one which required monthly reviews of the caution tag logs and one which excluded the review during outages.
The inspectors identified that the licensee's corrective action to eliminate the discrepancy was not implemented in a timely manner.
(Section 01.3) 8 8/27/1998 Negative IR 98016 NRC Maintenance 2A 3A Overall, maintenance work was performed using approved work procedures and reflected good maintenance practices.
However, the licensee identified two maintenance errors: a wrong valve was cut out on the Unit 2 main condenser and a maintenance test resulted in an electrical arc on the 2CD auxiliary transformer supplemental cooling system circuitry. These events challenged the adequacy of the correc actions in progress to improve the work control program and indicated that room for improvement remains.
(Section M1.1) 9 8/27/1998 Negative IR 98016 NRC Maintenance 2B The inspectors concluded that the work control process weaknesses identified in 1996 still exist. Past observations, recent observations, and licensee self-assessments indicated that the weaknesses have not been corrected.
The inspectors did not identify any specific examples where these weaknesses resulted in a risk significant issue.
However, the inspectors noted that the licensee's operations personnel were repeatedly challenged to ensure that safety margins were maintained.
(Section M1.3)
Page 2 of 35
PLANT ISSUES MATRIX DG Gook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 10 8/27/1 998 NCV 11 8/27/1998 Positive IR 98016 IR 98016 NRC Mainteriance 28 5A 5C NRC Maintenance 28 While lining up the fire water tanks in preparation for a test, the suction source for all three fire pumps was isolated, resulting in the automatic start of all of the fire pumps. The inspectors concluded that the auxiliary equipment operator was directed to use the engineering test procedure to align the fire water tanks, an activity for which the test procedure was not intended.
(Section M1.2)
Ice project personnel identified shortcomings in the control of ice bagged for off-site storage, and prompt corrective actions were implemented.
Foreign material exclusion practices were improved to provide assurance that the ice would be of the required quality when loaded into the Unit 1 ice condenser.
The physical security of the ice during transport and storage was improved to maintain control over the ice during all phases of the work. (Section M1.4) 12 8/27/1998 Positive 13 8/27/1998 Positive IR 98016 IR 98016 NRC Maintenance NRC Maintenance 28 5A 5C 28 5C 2A Repair crews showed good attention to detail during follow up ice basket inspections which identified additional damage to the ice baskets.
The licensee took prompt action to determine and correct the causes of the original inspections which had missed some ice basket sections needing repair. (Section M1.5)
Inspection and repair work to restore damaged and out of specification components for the Unit 1 ice condenser appeared affective and comprehensive.
The decision to melt the ice in both ice condensers facilitated effective inspection and repair, which demonstrated the licensee's commitment to a quality repair effort for the ice condensers.
However, as a consequence of engineering and maintenance staff inexperience with this non-routine evolution (ice condenser melt out),
water was entrained within the concrete ice bed subfloor of Unit 1.
Lessons learned from Unit 1, enabled the licensee to implement actions that prevented substantive water intrusion into the Unit 2 floor. (Section M2.1) 14 8/27/1998
- Positive, IR 98016 NRC Engineering 48 Licensee engineering personnel were involved in the assessments, repairs, and modifications to the ice condenser.
Extensive involvement of engineering was noted by the inspectors during their assessment of the ice condenser corrective actions.
(Section E1)
Page3of 35
Search Sorted by Date (Descending) and SMM Code PLANT ISSUES MATRIX DC Cook s(Ascending):
SearchColumn
= 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 15 8/27/1998 EEI 16 7/16/1998 Negative 17 7/16/1998 Negative 18 7/16/1998 Negative 19 7/16/1998 Negative IR 98016 IR 98015 IR 98015 IR 98015 IR 98015 NRC Operations 5A 5C NRC Operations 1C NRC Maintenance 2A NRC Maintenance 2B 2A NRC Maintenance 2B 3C The licensee had previously identified that their Corrective Action Program was ineffective. Reviews of current Condition Reports by NRC inspectors determined that additional changes to correct the ineffective program were required and were being performed by the licensee.
(Section 07.1) s The inspectors identified that the licensee was inappropriately enterin and exiting an administrative Limiting Condition for Operation (LCO) whenever the outside air temperature exceeded 88oF. The licensed operators had questioned whether this was a conservative practice but had not taken actions to resolve their questions or to ask for management assistance.
The inspectors'eview of other licensee entries into TS LCOs determined there appeared to be an appropriate use of LCO time clocks (Section 01.2).
The inspectors concluded that the licensee staff responded appropriately to mitigate the consequences of an oil leak in the Unit 2 main turbine lubricating oil cooler. However, the inspectors concluded that this leak was caused by equipment material condition problems (Section M2.2).
CO2 was inadvertently discharged into the auxiliary building crane bay.
Use of a procedure intended for operability testing of the CO2 system as the post maintenance testing was identified by the licensee as a significant contributor to the incident. An investigation was promptly initiated and interim preventive actions taken. A formal root cause investigation is being conducted to evatuate this event (Section Ms.tt.
Steady progress was being made in repairs to the ice condensers in both units. Some instances of foreign material intrusion into the ice making system were quickly identified and corrected.
Initial lapses in command and control which resulted in minor scheduled impacts were part of the reason the licensee assigned additional project management (Section M2.1).
20 7/16/1998 Negative IR 98015 IR 98015 NRC Plant Support 3A 3B The inspectors noted several minor occurrences of lack of attention to detail concerning anti-contamination personnel protective clothing dress requirements.
NRC inspection activities willcontinue to monitor worker compliance with radiation work permit requirements (Section R4.1).
Page 4 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 21 6/11/1998 Positive 22 6/11/1998 Negative 23 6/11/1998 Negative 24 6/11/1998 Positive 25 6/11/1998
- Negative IR 98012 IR 98012 IR 98012 IR 98012 IR 98012 NRC Operations 1A NRC Operations 1A NRC Operations 1A NRC Operations 1A 5A NRC Operations 1C During routine evolutions, the inspectors noted that the operators were attentive to their panels and to annunciators.
During special evolutions, such as the reactor coolant system partial drain down for maintenance, the inspectors observed that good quality pre-job briefings were held (Section 01.1).
'everal days prior to a RCS drain down, the inspectors determined th licensee had planned to drain down the RCS to approximately 1 to 2 feet above the reactor vessel flange. The inspectors'eview of the operating procedure determined the procedure did not address draining down to 1 to 2 feet above the flange. Prior to draining the RCS, the licensee changed its plans for the RCS draindown to be consistent with the drain down procedure (Section 01.2).
The inspectors determined that command and control of the operating shifts during Modes 5 and 6 was not well defined with the Unit Supervisor out of the control room.
In addition the inspectors determined that the procedure addressing absence of the Unit Supervisor was weak.
In response to the inspector's questions, the licensee provided guidance for limits on the duration of absence and whereabouts of the Unit Supervisor and who was in charge during the Unit Supervisor's absence from the control room (Section 01.3).
During a drain down of the reactor coolant system (RCS), the Inspecto~
observed the operators stop the drain down upon observing a discrepancy between RCS level indicators.
Even though the procedure allowed the drain down to continue for several more feet, the operating shift decided to stop the evolution until instrumentation and control personnel could identify the cause.
This was indicative of a conservative operating philosophy (Section 01.2).
The inspectors determined that prompt operability determinations for action requests and condition reports were weak and contained inconsistent documentation.
The licensee had recently identified the need to provide additional guidance and training and was revising their process for performing operability determinations (Section 03.2).
Page5of 35
P.LANT ISSUES MATRIX DC Cook earch Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 26 6/11/1998 Negative 27 6/11/1998 VIO/SL-IV 28 6/11/1998 Negative 29 6/11/1998 Negative 30 6/11/1998 Negative IR 98012 IR 98012 IR 98012 IR 98012 IR 98012 NRC Operations 1C NRC Maintenance 2B 5A 3A NRC Maintenance 3A 2B NRC Maintenance 3A 5A NRC Maintenance 3B 3A The licensee's procedural guidance for the required position of ventilation system hand switches to support the operability of certain safety-related components was weak. The procedures for the normal lineup of safety-related components had ventilation guidance that was inconsistent between systems.
The procedure addressing ventilation contained erroneous guidance allowing hand switches to be placed in off withou addressing equipment operability (Section 03.1).
The inspectors identified that the job order used for fillingice bags did not contain detailed instructions or assign crew responsibilities and was not appropriate to the circumstances.
A violation of 10 CFR Part 50, Appendix B, Instructions, Procedures, and Drawings was issued (Section M1.2).
On May 5, 1998, while lining up the fire system suction source in preparation for a routine surveillance, an auxiliary equipment operator inadvertently isolated both fire water tanks, resulting in the automatic start of all three fire pumps.
A root cause investigation team assigned by licensee management to investigate this incident had not yet issued a final report. An inspection followup item was opened pending the inspectors'eview of the team's report (Section M1.3).
Licensee contractor personnel identified that they had a wrong unit error and inadvertently worked on a Unit 1 valve instead of a Unit 2 valve. T~
wrong unit error occurred on the non-safety-related portion of the feedwater system while the system was cooled down and depressurized.
Licensee management met with the contractor senior management and the contractor management informed the licensee of the seriousness in which they were taking the error (Section M1.1).
During a contractor performed freeze seal on a non-safety-related portion of the non-essential service water system (NESW), licensee employees failed to take the temperature data required by the job order. The licensee employees assumed that they could transfer the contractor's data after the work had been performed.
However, when the licensee maintenance workers attempted to copy the contractor's data after the freeze seal had been thawed, they determined that the contractor's procedure did not require recording temperature data at every location that the licensee's job order required (Section M1.1).
Page6of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date; Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 31 6/11/1998 Negative 32 6/11/1998 Negative 33 6/1 1/1 998 VIO/SL-IV 34 6/2/1998 Negative IR 98012 IR 98012 IR 98012 IR 98013 NRC Engineering 4B NRC Engiriee ring 4C 5A NRC Plant Support 5C NRC Maintenance SC 1C The inspectors determined that the procedural guidance which controlled the ventilation requirements for safety-related equipment was weak. The guidance within the procedure was provided by Engineering Department personnel.
Engineering Department personnel did not adequately consider the consequences of inoperable ventilation equipment when providing guidance to a procedure utilized by the Operations Departm (Section E1.1).
The inspectors determined that the System Engineering Review Board (SERB) and the Restart Oversight Committee (ROC) appeared to be doing an effective job of reviewing the items identified by the system engineers; however, there was a failure to comply with the SERB charter.
This resulted in the initial failure of the Restart List to reflect the shared concerns of Operations, Maintenance and Engineering.
The inspectors discussed observations of this failure with licensee management several times before effective corrective actions were taken (Section E7.1).
Due to a failure to adequately correct a previous occurrence, the licensee performed an expected plant cooldown and failed to comply with surveillance requirements designed to monitor the formation of a gas bubble in the reactor vessel head.
Inconsistent procedures, inattention to detail, and a large number of items in the open items log contributed to the repeat failure to followthe procedural requirements.
A violation for inadequate corrective action was issued (Section M7.1).
The licensee's investigation of this event as documented in a letter to the NRC dated March 23, 1998 was reviewed.
Several discrepancies in the sequence of events as well as the events themselves were noted between the licensee's findings and the workers recollection of events.
Upon review during the inspection, the inspectors concluded that the licensee's investigation was thorough, however, the information had not been fullyincorporated into the letter sent to the NRC.
35 6/2/1998 VIO/SL-IV IR 98013 NRC Plant Support 5C The inspection concluded that on January 4, 1998, a contract painter having an open wound entered containment in violation of station procedures.
This violation was a result of poor communications between work groups which led to the radiation protection group not being notified of the open wound prior to entry, as required by procedures.
Page7of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Cotumn = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 I
DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 37 5/21/1998 Positive IR 98010 36 5/21/1998 Positive IR 98010 NRC Maintenance 4B 5B 5C NRC Maintenance 5A 4B The independent determination of the ultrasonic examination capability for the part length control rod housings was considered a sound engineering decision.
Since other sites had previously performed comparable inspections without this determination, the licensee demonstrated a positive commitment to safety.
Examinations of the Unit 2 part length control rod housings were accomplished in accordance with the Electric Power Research Institute demonstrated procedure, using the same equipment and essential variables.
38 5/14/1998 Positive 39 5/14/1998 Positive 40 5/14/1998 Positive 41 5/14/1998 Positive IR 98305 IR 98305 IR 98305 IR 98305 NRC Operations 1C NRC Operations 1C 3B NRC Operations 1C 3B NRC Operations 1C 3B With the exception of several of the operating JPM followup questions, the facilityexamination development team provided a balanced, acceptable JPM examination tool for evaluating applicant competency.
Several of the submitted follow up questions were not useful for determining applicant competency.
(Section 05.3)
With the exception of the one administrative JPM rejected by examiners, facilityinstructors provided a balanced, acceptable examination tool to evaluate applicant competency.
The applicants appeared well prepared for this portion of the test.
(Section 05.2)
Allapplicants passed their respective retake examinations and were issued a Senior Reactor Operator's or Reactor Operator's license.
The facility's examination development team provided NRC examiner with a comprehensive, balanced dynamic simulator examination tool for evaluating applicant competency.
The applicants appeared well prepared for the examination and exhibited strengths in several areas identified as weaknesses during the July 1997 examination the applicants failed.
(Section 05.4) 42 5/8/1998 Positive IR 98011 NRC Plant Support 1A In general, the conduct of security operations was professional with-marked improvement in the level of management support for the physical security program.
This was evidenced by an increase in the management oversight of the program, increase staffing in security operations and training areas, and improvement in the staff communications and.the initiation of employee incentives.
Security force performance was effective. (Section S6.1)
Page8of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date ~ '10/1/97';
Ending Date =
'9/30/98'ATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 44 5/8/1998 Negative IR 98011 Licensee, Plant 2A 5A Support 43 5/8/1998 Negative NRC Plant 1C Support The approved security plan did not provide a description of the duties and responsibilities/delegation of authority of the security management oversight position of the Security and Emergency Operations Manager established in November 1997. This matter is considered an Inspection Follow Up Item (IFI). (Section S3)
The licensee identified that access control to a vital area was lost for approximately two and a half hours on October 9, 1997, because a portion of the physical boundary was removed without establishing compensatory measures.
(Section S8.1) 45 5/8/1998 Misc 'R 98011 NRC Plant 5A 5B Support Self-assessments conducted by the Plant Protection Department were effective in determining procedural compliance but lacked the depth and scope to determine overall program effectiveness.
(Section S7) 46 5/8/1998 Negative 47 5/7/1998 Negative 48 5/7/1998 Negative 49 5/7/1998 Negative IR 98011 IR 98004 IR 98004 IR 98004 IR 98004 NRC Plant 5C Support NRC Engineering 4B NRC Engineering 4B 3B 4C NRC Engineering 4A 4B Management's corrective actions were not totally effective in addressing the loss of temporary lighting which caused the licensee to go below the required.2 foot candles on four occasions subsequent to the last security inspection. This matter is considered an Inspection Follow Up Item (IFI).
(Section S6.2)
The containment spray heat exchanger room heat gain calculation used design input values which were not consistent with the Updated Final Safety Analysis Report (UFSAR). This could result in the containment spray heat exchanger room temperature limits being exceeded during certain accident scenarios (Section S2.2.b.1.10).
Several calculations appeared to be obsolete but were still identified as valid calculations in the calculation index. A condition report was initiated to address calculation control issues (Section S2.2).
The licensee did not recognize that the refueling water storage tank (RWST) low level alarm setpoint change was a change to the plant. As a result, procedure specific safety evaluations were not performed (Section S5.2).
Page9of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = 10/1/97; Ending Date = '9/30/98' DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 50 5/7/1 998 EEI 51 5/7/1 998 EEI 52 5/7/1998 Negative 53 5/7/1998 Negative 54 5/7/1998 Misc 55 5/7/1998 Positive IR 98004 IR 98004 IR 98004 IR 98004 IR 98004 IR 98004 NRC Engineering 4B 4C NRC Engineering 4B 4C NRC Engineering NRC Engineering 4B 4C 3B 4B 4C 3B NRC Engineering 5A 4B NRC Engineering 4C 5C The inspectors were concerned that safety evaluations continue to have deficiencies.
Seven (7) apparent violations of 10 CFR 50.59 were identified. Two (2) of these examples were previously reviewed by AEP staff during the short term assessment reviews and were found to be acceptable.
The inspectors concluded that weaknesses still exist in the safety evatoation program (Sections C2.2, S2.4 and S5.2).
~
The inspectors concluded that, overall, the licensee successfully completed job order activities and modifications to the Unit 1 and Unit 2 recirculation sump to re-install sump roof vent holes and add foreign material exclusion devices to prevent foreign material from entering the recirculation sump. However, the inspectors identified three (3) apparent violations of 10 CFR 50.59 where 10 CFR 50.59 screenings were not completed as required due to the licensee not recognizing that the plant design was being changed (Section C2.2).
The licensee's 50.59 reviewer qualification training did not treat anticipated transients without scram and station blackout scenarios as accidents requiring the same level of review as UFSAR Chapter 14 accident scenarios.
This was considered a weakness (Section S5.1).
Initial corrective actions implemented for the installation of leak collection devices appeared to be reasonable, however, leak collection devices were recently installed under the RWST overflow pipe without followin~
the temporary modification procedure.
It appeared that other contract~
personnel were not properly informed that leak detection devices were considered'a change to the plant. As such, the previous corrective actions did not preclude repetition (Section S7.2).
The inspectors concluded that the root causes applied to the CAL items and programmatic weaknesses, such as calculation control, were appropriate.
However, AEP submittals to the NRC identified potential AEP to Westinghouse interface weaknesses (Section S2.1).
The licensee was adequately addressing other plant processes that could bypass the design control process.
The review identified several processes, such as action requests, that had implemented chang'es to the plant (Section S2.3.b.3).
Page10of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date; Beginning Date = '10/1/97';
Ending Date = '9/30/98' DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 56 5/7/1998 Negative 57 4/27/1998 VIO/SL-IV 58 4/27/1998 Negative 59 4/27/1 998 EEI 60 4/27/1998 URI IR 98004 IR 98008 IR 98008 IR 98007 IR 98007 NRC Engineering 5A 4C NRC Operations 1A 3A 3B NRC Operations 1C 2A NRC Engineering 2A NRC Operations 2A The inspectors concluded that due to quality assurance (QA) organization audit methodology weaknesses, the licensee's QA organization did not identify the extent of the problems identified by the AE design inspection team (Section S7.1).
The inspectors determined that the continuous use procedure coverin~
plant heatup for Unit 2 was not known by the operators to be in effect,~
was not readily available, and was not in use. A violation for failure to followprocedure was identified (Section 01.2).
The licensee's procedure providing guidance for ventilation equipment required to support Technical Specification (TS) equipment was weak.
The procedure addressed a complete failure of the ventilation equipment but failed to address degraded performance issues.
A review of licensee documentation failed to identify any examples of inoperable TS equipment as a result of the weak procedure (Section 01.3).
During a review of power operated relief valve (PORV) operability, the inspectors identified a 2-year period in which one PORV did.not have an operable backup air supply. following consultation with the Office of Nuclear Reactor Regulation, it was determined that the operability of the PORVs depends upon the operability of the associated backup air supply. An apparent violation was issued for a failure to comply with TS requirements upon discoverer oi an inoperable PORV (Section E8.1). ~
During routine walkdowns of selected engineered safety features systems, the inspectors identified some equipment in poor material
'ondition.
Examples of items identified included bent and dirty motor air inlet screens on the residual heat removal (RHR) pumps, missing bolts on the containment hydrogen mixing system (CEQ) and electrical junctions boxes with gaps on the CEQ system.
The inspectors concluded that these material condition issues did not render the CEQ system inoperable.
An unresolved item was opened to address the analysis of the open junction box on the operability of the CEQ system (Section 02.1).
Page11 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 61 4/27/1998 URI 62 4/27/1998 EEI 63 4/27/1998 Negative 64 4/27/1998 Negative 65 4/27/1 998 Misc IR 98007 IR 98007 IR 98008 IR 98007 IR 98007 NRC NRC Maintenance 2A 2B 5A 2A 3A 5A Plant Support NRC Maintenance 2A 5A NRC Maintenance 2B NRC Maintenance 2A 2B The inspectors identified duct tape as being used as an installation aid that did not appear to be properly controlled by procedure.
In addition, during a review of the installation procedure for the divider deck barrier seals, the inspectors identified a step which appeared to authorize a blanket bypass of the 10 CFR 50.59 process.
Additional information was required to resolve the questions and two unresolved items were issu (Section M1.4).
During a review of the Technical Specification (TS) surveillances on the hydrogen recombiner, the inspectors identified: 1) an apparent violation for declaring the recombiner operable with recorded data which exceeded the TS limits, 2) an apparent violation for an inadequate procedure which failed to measure resistance to ground immediately following the heat up test, 3) an apparent violation for a procedure which caused inconsistent performance of TS surveillances, and 4) an apparent violation for failure to correct a previously identified condition regarding preconditioning of equipment prior to a surveillance test (Section M1.2).
The inspectors identified a non-safety-related High Efficiency Particulate Absorber filterinstalled in an unapproved manner in the steam generator storage building. Licensee personnel failed to identify the improper installation even though multiple entries had been made by radiation protection personnel to perform routine surveys (Section R1).
As noted in Sections 02.1, 02.2, 02.3, 02.4, and M1.2, the inspector identified a number of material condition issues.
Most of the issues ha3 not been recognized by licensee personnel and were long-standing (Section M2).
During a review of the surveillance testing program for the distributed ignition system the inspectors determined the licensee was performing surveillance test of measuring voltage and current of the igniters. The need to conduct visual verification of igniter energization or to measure igniter temperature was identified as an inspection followup item (Section M1.3).
Page12of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date = '9/30/98' DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 66 4/27/1998 VIO/SL-IV 67 4/27/1998 URI 68 4/27/1998 URI 69 4/27/1998 Misc 70 4/27/1 998 EEI IR 98008 IR 98007 IR 98007 IR 98008 IR 98007 NRC Maintenance 3A 2A NRC Engineering 4A NRC Engineering 4A 2A 5B NRC Engineering 4B NRC Engineering 4B 4C The inspectors determined that during the installation of a minor modification (MM)in March 1997, the contract workers installing MM-438 on the 1 CD emergency diesel generator loosened and improperly reinstalled the exhaust manifold bracket bolting without the jam nuts as required by the job order. This improper bolting configuration could have led to a failure on that engine similar to the exhaust manifold bracket ~
failure which caused the 2 AB D/G to become inoperable for repairs. ~
failure to install jam nuts in accordance with the job order was a violation of TS 6.8.1 (Section M2.1).
During an assessment of the distributed ignition system (DIS), the inspectors identified an unresolved item on the whether the DIS was required for beyond design basis accidents.
The inspectors also identified inspector followup items on the initiating signals used to manually actuate DIS, possible water impingement on the DIS, and drawing discrepancies from the as-built configuration (Section 02.2).
During a routine plant tour, the inspectors identified loose hold down nuts on some of the divider deck barrier missile blocks. The licensee informed the inspectors that while these particular bolts were not known to be loose, this was a repetitive problem and had been evaluated.
As of the end of the report period, the licensee was unable to find the evaluation.
Pending the review of the evaluation, this issue remained unresolved (Section 02.3).
For those items sampled, the inspectors determined that the System Engineering Review Board (SERB) and Restart Oversight Committee (ROC) appropriately determined whether the item was required to be corrected prior to restart of the units. However, the ROC appeared to perform only a minimal review and assessment on those items the SERB did not recommend be corrected prior to restart (Section E7.1).
During followup to a licensee identified blockage of a CEQ line, the licensee identified low flow rates in other lines and trains of both units'EQ systems.
The licensee determined that the low flow rates were attributed to the system design, inadequate pre-operational tests, and the failure to maintain a proper distribution of system flows. An apparent violation was identified for the failure to comply with 10 CFR 50.59 (Section E1.1)
Page 13 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 71 4/27/1 998 VIO/SL-IV 72 4/15/1998 EEI 73 4/15/1998 EEI IR 98008 IR 98009 IR 98009 NRC Engineering 5B NRC Engineering 4A 4B NRC Engineering 4A 4B The licensee's investigation report into the exhaust manifold bracket failure on the 2 AB emergency diesel generator (2 AB D/G) did not document the root cause for the missing jam nuts on the 1 CD and 2 AB D/Gs. The inspectors concluded that appropriate corrective actions for this significant condition adverse to quality could not be implemented without an adequate root cause determination.
A violation of 10 CFR 50, Appendix B, Criterion XVI,was identified (Section 07.1).
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to correctly translate RWST Appendix R inventory requirements into specifications, drawings, procedures, and instructions.
Specifically, calculation No. TH-90-02, 'RCS Volume Make-up Required AfterAppendix R Fire," RWST volume requirements were not incorporated into procedure No. PMP-4100, "Plant Shutdown Safety and Risk Management.
(Section E8.18)
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to verify or check the adequacy of Engineering Control Procedure (ECP) calculation Nos. 1-RCP-09 and 2-RCP-09, "RWST Level." Specifically, the suction pipe entrance head losses and Bernoulli velocity head losses were not included in the uncertainty analysis.
(Section E8.1) 74 4/15/1998 EEI IR 98009 75 4/15/1998 EEI IR 98009 NRC Engineering 4A 4B NRC Engineering 4A 4B An apparent violation of 10 CFR 50, Appendix B, Criterion III, was identified pertaining to the failure to verify or check the adequacy of E Nos. 1-CG-39 and 2-CG-39, 'Refueling Water Storage Tank Level.'pecifically, vortexing (air entrainment) was not addressed when the RWST low-Iow level setpoint was developed.
(Section E8.3)
An apparent violation of 10 CFR 50, Appendix B, Criterion III, was identified pertaining to the failure to verify or check the adequacy of ECP Nos. 1-2-N3-01, "CNTMTSump Water Level Indication, 1-RPC-14 and 2-RPC-14, 'Containment/Containment Sump Level.
Specifically, post-accident containment environment effects were not incorporated in the uncertainty analysis.
(Section E8.5)
Page 14 of 35
J
PLANT iSSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Cotumn = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 76 4/15/1998 EEI 77 4/15/1998 EEI 78 4/15/1998 EEI 79 4/15/1998 EEI 80 4/15/1998 EEI 81 4/15/1998 EEI IR 98009 IR 98009 IR 98009 IR 98009 IR 98009 IR 98009 NRC Engineering 4A 4B NRC Engineering 4A 4B NRC Engineering 4A 4B NRC Engineering 4A 4B NRC Engineering 4B 4C NRC Engineering 4B 4C An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure'to correctly translate containment water
inventory requirements into specifications, drawings, procedures, and instructions.
Specifically, it was not demonstrated that sufficient water could be recovered during a design basis accident to prevent pump vortextng. (Section E8.6)
~
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to correctly translate CCW heat exchanger design flow into specifications, drawings, procedures, and instructions.
Specifically, the cooldown analysis and operating procedures used a CCW flowthat exceeded the UFSAR design value.
(Section E8.14)
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to correctly translate the % inch recirculation sump roof vent hole design into specifications, drawings, procedures, and instructions.
Specifically, the vent holes were plugged without verifying their design basis.
(Section E8.31)
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to correctly translate
~/4 inch containment sump particulate retention requirements into specifications, drawings, procedures, and instructions.
Specifically, the containment sump sere~
sections contained
~/~ inch gaps and the % inch sump roof vent holes ~
were not covered with screening material. (Section E8.8)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not fullyanalyzing unit operation above UFSAR Tables 6.3-2 and 9.5-3 ESW 76'F ultimate heat sink (lake) design temperature.
Specifically, the units were operated in 1988 for 22 continuous days at an average lake temperature of 81'F. (Section E8.28)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments,'as identified for not considering the loss of spent fuel pool (SFP) cooling during a design basis accident.
Specifically, the sdfety evaluations for the Unit 2 dual train CCW/ESW outage did not address the reduction in SFP time-to-boil if the Unit 1 CCW flow isolated due to a Unit 1 design basis accident.
(Section E8.29)
Page15of 35
PLANT ISSUES MATRIX DC Cook Search Sortedby Date (Descending) and SMM Codes(Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 82 4/15/1998 EEI 83 4/15/1998 EEI 84 4/1 5/1 998
~
EEI 85 4/15/1998 EEI 86 4/1 5/1 998 EEI 87 3/19/1998 Negative IR 98009 IR 98009 IR 98009 IR 98009 IR 98009 IR 98005 NRC Engineering 4B 4C NRC Engineering 4B 4C NRC Engineering 4B 4C NRC Engineering 4B 4C
~
NRC Engineering 5C 4B NRC Maintenance 2A 2B 3A An apparent violation of 10 CFR 50.59, 'Changes, Tests, and Experiments," was identified for creating a single failure vulnerability in a procedure revision to ES-1.3, 'Transfer to Cold Leg Recirculation.
Specifically, Revision 2 to ES-1.3 piggy-backed all high head injection pumps onto one residual heat removal pump. (Section E8.30)
An apparent violation of 10 CFR 50.59, 'Changes, Tests, and Experiments," was identified for.not performing a safety evaluation for unit operation with CCW temperatures in excess of the 95'F UFSAR Table 9.5-3 design value. (Section E8.32)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not performing a safety evaluation for unit operation with reactor coolant pump thermal barrier flow less than the 35 gpm UFSAR Table 9.5-2 design value. (Section E8.33)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments,'as identified for not performing a safety evaluation for residual heat removal operation without automatic overpressure protection as described in UFSAR Section 9.3, "Residual Heat Removal System."
(Section E8.34)
An apparent violation of 10 CFR 50, Appendix B, Criterion XVI,was identified pertaining to not promptly correcting an identified condition adverse to quality. Specifically, calculation No. DCCHV12CR11N,
'Control Room Temperature Evaluation," identified in 1990 that control room equipment/component life could be reduced to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ifthe ESW temperature reached 87.5'F. Adequate documentation to demonstrate control room equipment shutdown capability at elevated temperatures could not be located.
(Section E8.12)
The ice condenser was degraded to a poor state of materiel condition such that the operability of the ice condenser was in question.
(Section M2.1)
Page16of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date =
10/1197';
Ending Date =
'9i30198'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 88 3/19/1998 EEI 89 3/19/1998 EEI 90 3/19/1998 EEI 91 3/19/1998 EEI 92 3/19/1998 EEI 93 3/19/1998 EEI 94 3/1 9/1 998 EEI IR 98005 IR 98005 IR 98005 IR 98005 IR 98005 IR 98005 IR 98005 NRC Maintenance 2B 1A NRC Maintenance 2B 2A NRC Maintenance 2B 2A NRC Engineering NRC Engineering
- NRC Engineering 4A 4B 4C 4A 4B 4C 4C 4A 4B NRC Maintenance 2B - 2A Eight apparent violations of 10 CFR 50 Appendix B (three Criterion V, four Criterion XI, and one Criterion Vll)and two violations of technical specifications (TS) were identified pertaining to inadequate surveillance testing of the ice condenser.
Specifically, these violations pertained to inadequate instructions, inadequate acceptance limits, inadequate control of contractors, failure to implement TS requirements and entry into an~
unanalyzed condition for ice condenser surveillance testing. (Section ~
M1.1)
Two examples of an apparent violation of 10 CFR 50 Appendix B, Criterion V were identified for the licensee's failure to followthe procedure change process for changes made to completed surveillance tests.
(Section M1.1)
Collectively, the apparent violations associated with surveillance testing activities represented a breakdown in the surveillance testing program for the ice condenser.
(Section M1.1)
Three apparent violations of 10 CFR 50 Appendix B, Criterion XVIwere identified for the licensee's failure to identify conditions adverse to quality.
Conditions not previously identified by the licensee in the ice condenser included: blocked flow passages, missing ice segments, dented/buckled basket webbing, unweighable ice baskets, and nonencapsulated insulation. (Section M2.1)
Collectively, the apparent violations identified in Section E7.1 represent programmatic breakdown in the maintenance of the design basis for the ice condenser.
(Section E7.1)
Seven apparent violations of 10 CFR 50.71(e) were identified pertaining to the licensee's failure to update the Final Safety Analysis Report (FSAR)
Appendices J and M, which contained the detailed description and design basis for the ice condenser.
(Section E7.1)
Four apparent violations of 10 CFR 50 Appendix B, Criterion IIIwere identified pertaining to the licensee's failure to followthe established design control process for ice basket modifications. (Section E7.1)
Page17of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = Date';
Beginning Date =
10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 95 3/19/1998 EEI 96 3/19/1998 EEI 97 2/27/1998 Positive 98 2/27/1 998 EEI 99 2/27/1 998 EEI IR 98005 IR 98005 IR 97017 IR 97017 IR 97017 NRC Engineering 5C 2A 2B NRC Engineering 5C 2A 2B NRC Maintenance 2A 4B NRC Operations 2B NRC Engineering 4A Page18of 35 Four apparent violations of 10 CFR 50 Appendix B, Criterion XVIwere identified for the licensee's failure to identify and correct conditions adverse to quality on ice condenser components.
Specifically, these violations pertained to the licensee's failure: to implement prompt corrective actions for missing ice basket sheet metal screws, to implement effective corrective actions for preventing the recurrence of~
loose U-bolt nuts and separated ice baskets, and to take appropriate ~
corrective actions for the ice baskets with defective hold down bar welds.
(Section E2.1)
Collectively, the apparent violations identified in Sections M2.1 and E2.1 represent a breakdown in the licensee's corrective action program for the ice condenser.
(Section E2.1)
The corrective action implemented subsequent to the final evaluations of material condition problems in containment wet'e extensive.
The licensee removed over 13,000 pounds of various materials from the containments, including insulation, coatings (paint), labels, rust, tape, high efficiency particulate air (HEPA) filters, granular charcoal, and other foreign material.
In Unit 1, paint was removed from the lower containment floor down to the base concrete and new coatings were applied.
Equipment such as welding machines, vacuums, and man lifts used during outages were also removed (Section E1.1.b.3).
The containment inspection tours procedure which defined how to perform containment inspections, was inappropriate to the circumstance'n that it made no reference to inspection for fibrous material or insulation that could clog the recirculation sump. This is an apparent violation and is being considered for escalated enforcement (Section E1.1.b.7).
A lack of design control led to the installation of Fiberfrax, a fibrous damming material found inside both containments that could potentially cause blockage of the containment recirculation sump screens during the recirculation phase of a loss of coolant accident (LOCA) resulting in insufficient emergency core cooling water flowto the reactor vessel.
The lack of sufficient measures to assure that the design basis was correctly translated into specifications to control the installation of material that could potentially have an adverse effect on the safety-related functions of the containment system is an apparent violation and is being considered for escalated enforcement (Section E1.1.b.7).
PLANT ISSUES MATRIX
.DC Cook Search Sorted by Date (Descending) and SMM Codes(Ascending):
Search Column = 'Date; Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 100 2/27/1998 EEI 101 2/27/1998 Negative IR 97017 IR 97017 NRC Engineering 4A 4B 4C NRC Engineering 4B 2A The installation, replacement, and repair procedure for silicone fire barrier penetration seals, was inappropriate to the circumstances in that it did not require that fibrous damming material be removed or encapsulated following sealing operations.
This is an apparent violation and is being considered for escalated enforcement (Section E1.1.b.7).
The initial engineering evaluations performed to determined whether not the recirculation sump was operable in the as-found condition were not well supported and did not fullyaddress the amount of debris found in containment.
Subsequent engineering evaluations and support to restore the containment material condition were thorough.
Engineering personnel also did an extensive evaluation of the coatings on equipment that could have an adverse effect on the performance of the recirculation sump screens following a LOCA (Section E1.1.b.6).
102 2/27/1998 EEI 103 2/27/1 998 URI 104 2/20/1 998 VIO/SL-IV IR 97017 Licensee Engineering 4C IR 97025 NRC Maintenance 2A IR 97017 URI NRC Engineering 4C 97017-05 The installation of unjacketed fibrous insulation inside of both containments could potentially cause blockage of the containment recirculation sump screens during the recirculation phase of a loss of coolant accident resulting in insufficient emergency core cooling water flowto the reactor vessel.
The lack of sufficient measures to assure that the design basis was correctly translated into instructions which would be changed in a controlled manner is an apparent violation and is being considered for escalated enforcement (Section E1.1.b.7).
One unresolved item relative to TS operability of the as-found conditio the containment recirculation sump was also identified.
The unresolved item willbe dispositioned subsequent to the completion of a detailed analysis by the engineering department (Section E1.1.b.7).
The questionable material condition of the SSPS master relay dust covers resulted in both trains of SSPS being declared inoperable for operability under seismic conditions. A violation was identified when the licensee failed to make a timely report to the NRC concerning an unanalyzed condition that significantly compromised plant safety. An Unresolved Item was opened to track the review of additional data to determine when licensee personnel were aware of the degraded condition of the SSPS relays (Section M2.1).
Page 19 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 105 2/20/1998 Positive 106 2/20/1998 Positive 107 2/20/1998 Negative 108 2/20/1998 URI IR 97025 IR 97025 IR 97025 IR 97025 NRC Operations 2A 5A 4A NRC Maintenance 3A NRC Maintenance 3A NRC Engineering 4B 5C The inspectors concluded that cold weather preparations had been properly implemented.
The inspectors also concluded that it was prudent to procure a stand-by boiler in order to ensure that safety-related equipment and other plant spaces were not adversely affected by cold weather in the event the plant heating boiler was unable to operate (Section 01.2).
The inspectors concluded that the work activities observed were performed in a quality manner with procedures present and in use.
The high quality of the Instrumentation and Control (l8 C) technician work on the Solid State Protection System (SSPS) relay-repairs was especially noteworthy (Section M1.1).
The Instrumentation and Control (18 C) technician work on the Solid State Protection System (SSPS) relay repairs was noteworthy. An exception to this good performance was a personnel error by an l&Ctechnician on another activity which resulted in an invalid reactor trip signal.
In addition, the inspectors determined that the required report regarding the inadvertent trip signal was not going to be made to the.NRC until after the inspectors questioned the lack of a report (Section M1.1).
Performance testing of the containment hydrogen skimming system continued during this inspection report period. Computer modeling and engineering assessments were being performed in an effort to determi~
the as-found operability and to ensure the system would be returned t~
operable condition. An unresolved item on the as-found condition of the hydrogen skimming system remained open. The inspectors concluded that the Licensee Event Report (LER) issued contained inappropriate statements (Section E1.1).
109 2/20/1 998 VIO/SL-IV IR 97025 NRC Engineering 4C 5C 110 2/17/1998 LER LER 315-98009 NRC Maintenance 2B During a review of selected LERs, the inspectors identified one event that was improperly retracted and another event that was not reported in a timely manner.
A third LER had an insufficient basis for retraction until the licensee performed additional calculations in response to inspector questions.
One violation for failure to submit a timely LER was issued (Section E8.1).
Hydrogen recombiner surveillance requirement not being met results in unanalyzed condition Page 20 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Coiumn = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 111 2/12/1998 LER LER 315-98008 Licensee Maintenance 2B 2A 112 2/11/1998 LER LER 315-98007 NRC Maintenance 4C 113 1/29/1998 LER LER 315-98006 NRC Maintenance 4B 114 1/22/1998 LER LER 315-98004 NRC Maintenance 2A 4A 115 1/22/1998 LER LER 315-98005 Licensee Maintenance 4A 2A Failureof ice basket to withstand simulated accident loadings during testing results in unanalyzed condition.
ice Condenser Weights Used to Determine Technical Specification Compliance Not Representative Procedural Option for Weighing of Ice Baskets in Modes 3 and 4 Determined to be a Potentially Unanalyzed Condition Restricted lce Condenser Flow Passages Found to Constitute an Unanalyzed Condition Lack of Adequate Number of Screws in Ice Basket Coupling Rings Determined to Constitute Unanalyzed Condition 116 1/16/1998 Positive 117 1/16/1998 Positive 118 1/1 6/1 998 NCV 119 1/16/1998 Positive 120 1/16/1998 Positive 121 1/16/1998 Positive IR 98003 IR 98003 IR 98003 IR 98003 IR 98003 IR 98003 NRC Operations 1A NRC Operations 3B NRC Engineering 3B NRC Operations 3B NRC Operations 3B NRC Operations 3B Control room conduct of operations and decorum were effective at maintaining appropriate focus on. plant and system evolutions.
(Section 01.1)
LOR program adequately revised to administer comprehensive examinations (Section 05.3.1)
The engineering support personnel training program failed to maintain a systems approach to training (NCV 50-315/31 6-98003-01).
(Section E5.1)
Overall examination material improved since the previous annual LO program NRC inspection, although some job performance measure (JPM) and written examination deficiencies continued to exist. (Section 05.2)
In general, the requalification examination material bank contained the minimum attributes to provide an adequate evaluation of operator skills.
(Section 05.2)
In general, the annual requalification examination met the minimum criteria in accordance with the guidance given in NUREG 1021, "Operator Licensing Examination Standards for Power Reactors, Interim Revision
- 8. (Sections 05.2, 05.3)
Page21 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 CZ
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 122 1/16/1998 Negative IR 98003 NRC Operations 3B Some operator performance weaknesses for communications and crew briefs were observed during the requalification operating examination.
(Section 05.3) 123 1/16/1998 Positive 124 1/16/1998 Positive 125 1/16/1998 Positive 126 1/16/1998 Negative 127 1/16/1998 Positive IR 98003 IR 98003 IR 98003 IR 98003 IR 98003 NRC Operations 3B NRC Operations 3B NRC Engineering 3B NRC Maintenance 3B NRC Operations 5A Licensee documentation and evaluation of operator performance showed improvement since the previous LOR program inspection.
(Section 05.3)
Appropriate security measures were implemented throughout the annual requalification examination.
(Section 05.3)
Licensee corrective actions for the engineering support personnel training program appeared adequate and were on schedule.
(Section E5.1)
Implementation of a systems approach to training for the maintenance and technical training programs was adequate with known weaknesses in the area of continuing technical training and program evaluation.
(Section M5.1)
The inspectors determined that the licensee's Training Department feedback process was satisfactorily implemented.
(Section 5.4) 128 1/13/1998 EEI 129 1/13/1998 Positive 130 1/13/1998 Positive IR 98006 IR 98006 IR 98006 NRC NRC NRC Plant 1C 3A Support
- Plant, 2A Support Plant 2A Support The licensee failed to maintain positive control over a radioactive material shipment, resulting in its leaving the site without the driver having the required shipping paperwork or emergency response instructions. Tw~
apparent violations of regulatory requirements were identified. (Secti~
R1.1)
The overall calibration and maintenance system for the portable instrument program was effectively implemented with the exception of the inspector identified RM-14/RM-20 response check weakness.
(Section R2.1) 3 Material condition of the Aptec PMW-3, PCM-1B whole body friskers and portal monitors was good, and workers used this equipment appropriately.
Calibrations and tests were performed at the required frequencies and response test results were good.
Independent reviews indicated that the monitoring instrumentation was capable of detecting the established external and internal personal radioactive contamination limits. (Section R2.2)
Page22of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sortedby Date (Descending) and SMM Codes(Ascending):
Search Cotumn = 'Date';
Beginning Date = '10/1/97';
Ending Date = '9/30/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 131 1/13/1998 Positive IR 98006 NRC Plant 2A 1C 5A Support Rugs with low levels of contamination build-up on them were being cleaned outside of the protected area.
The contamination to the soil was identified and remediated, and corrective actions were appropriate.
(Section R1.2) 132 1/13/1998 VIO/SL-IV 133 1/13/1998 VIO/SL-IV IR 98006 IR 98006 NRC NRC Plant 3A Support Plant 3A 3B Support A procedural violation was identified for workers who did not check th~
electronic dosimetry prior to entering the radiologically controlled area~
ensure that they were tumed on. The licensee's review of this matter had not sufficiently considered the procedural requirement that workers check their electronic dosimetry.
(Section R4.1)
A procedural violation was identified when workers in the lower ice condenser were found not wearing two pairs of gloves as required by the radiation work permit. This violation, as well as the violations cited in sections R4.1 and R4.2, in the aggregate, indicate a decline in radiation worker performance.
Of further concern was the fact that workers had been disregarding the RWP dress requirements for two to three weeks and RP had failed to identify this practice.
(Section R4.3) 134 1/13/1998 VIO/SL-IV IR 98006 NRC Plant 5A Support Aworker's failure to contact radiation protection after alarming the security gate house portal monitors was a procedural violation.
Additionally, of concern, was the licensee's conclusion regarding where the hot particle had come from. The condition report concluded that it had come from outside the radiologically controlled area and even possibly outside the buildings but did not address other more likely scenarios.
Further, the inspectors were concerned that the condition report addressing the procedural violation had not been responded to at the time of the inspection.
(Section R4.2) 135 1/13/1998 Negative 136 1/9/1998 Positive IR 98006 IR 98006 IR 98002 NRC NRC Plant Support Plant Support 5C 5A 1C 3A The licensee was effective at identifying problems, but was not as effective in resolving and/or documenting them.
(Section R7.1)
The self evaluations performed by the plant protection department were good. Effective corrective actions were taken to resolve identified problems.
(Section F7) 137 1/9/1998 Positive IR 98002 NRC -
Plant 2A Support Material conditions of the fire protection equipment appeared to be good.
Minimal amounts of combustible and impairments were noted in the plant. (Section F2)
Page 23of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Coturnn = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 138 1/9/1998 Positive IR 98002 NRC Plant Support 2A 4B The licensee was proactive in performing 100 percent destructive testings of gap seals.
(Section F3) 139 1/9/1998 Positive 140 1/9/1998 Positive IR 98002 IR 98002 NRC NRC Plant Support Plant-Support 3B 5A 3A The training provided to the fire brigade appeared to be adequate.
The annual physical examinations were kept up-to-date.
(Section F5)
The performance of the observed fire drill was good with two weaknesses.
The area to the equipment lockers of the brigade members was partially obstructed and the brigade leader did not dispatch another.
person as a communication link when encountering radio problems (Section F4).
141 1/9/1998 Positive IR 98002 NRC Plant 5C 5A Support 143 1/7/1998 Negative (50.72 event Licensee Maintenance 3A notification 33505 (retracted))
142 1/8/1998 LER LER 31 5-98002 Licensee Maintenance 2A 4A The corrective actions taken to resolve problems noted during welding, burning and grinding activities appeared to be adequate and were driven by self evaluations (Sections F1 and F7).
Degraded Solid State Protection System Master Relays Result in Condition Outside the Design Basis During a surveillance procedure an la C technician missed a step to reset the solid state protection system memory resulting in a reactor trip while shutdown and the reposition of the'main steam isolation valve dump valves.
144 1/7/1 998 145 1/4/1 998 LER LER 315-98003 Self-Maintenance 3A Revealed LER LER 315-98001 Licensee Engineering 2A 4A Missed Procedure Step Results in Engineered Safety Features and Reactor Protection System Actuation Containment AirRecirculation System Flow Testing Results Indicate Condition Outside the Design Basis 146 12/27/1997 VIO/SL-IV IR 97024 VIO NRC Engineering 1A The inspectors identified a violation in which the licensee failed to treat the manual backwashing of the ESW strainers in accordance with quality standards commensurate with the importance of the safety functions to be performed.
The licerisee committed to complete the corrective actions necessary prior to placing either unit in Mode 4, Hot Shutdown (Section E1.1).
147 12/27/1997 NCV IR 97024 NCV Licensee Maintenance 2A Page24of 35 A non-cited violation was issued when licensee personnel identified that non-safety-related parts were used during maintenance activities on the Unit 2 AB diesel generator.
The licensee looked for other instances where non-safety-related parts could have been used and identified one other example (Section M4.1).
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 I
DATE TYPE SOURCE ID BY
- SALP SMM CODES DESCRIPTION 149 12/27/1997 URI IR 97024 URI Licensee Maintenance 2A 4B 148 12/27/1997 URI IR 97024 URI Licensee Maintenance 2A 4B During periodic maintenance on the Unit 2 upper containment airlock, licensee mechanics found that the "0" ring seal material installed on the inner bulkhead interlock shaft was Teflon packing rather than the specified EPDM elastomer.
This is considered an unresolved item pending the results of the root cause assessment and a determination of the safety sfgntgcance of the use of Teflon seals (Section M1.3).
Licensee personnel identified a blocked hydrogen skimmer connection to a steam generator enclosure.
The blockage of this line, coupled with failure of the opposite train skimmer, could have resulted in an excessive buildup of hydrogen gas in the steam generator enclosure following a postulated loss of coolant accident.
This issue is considered an unresolved item pending the results of testing the individual flow connections from each containment enclosure (Section M1.2).
150 12/27/1997 Negative 151 12/10/1997 LER IR 97024 Self-Maintenance Revealed 4B 5B LER 316-97010 Licensee Maintenance 3A 2A The licensee's efforts to modify the Unit 2 AB diesel generator (D/G) reflected a weakness in design control in that multiple changes were made to engine components without a thorough understanding of the interrelations of proposed modifications. The engine timing change, performed to reduce cylinder pressure, in conjunction with the fuel line changes, resulted in several other engine parameter changes which were not anticipated by the licensee.
The troubleshooting and analyses required to correct the engine parameters resulted in a significant delay in the licensee's restoration of the 2 AB D/G to service (Section M2.3)..~
Use of Teflon Packing on Containment AirlockDoor Interlock Shaft Results in Potentially Degraded Condition 152 11/26/1997 Strength IR 97022 NRC Plant Support 1B The seal for the inner bulkhead interlock shaft was found to consist of Teflon packing rings rather than the specified EPDM elastomer.
The emergency preparedness program was in a commendable state of operational readiness and emergency response facilities were well-maintained, as evidenced by their material condition. A new Emergency Operations Facility, located in a portion of the Buchanan Office Building should be a program enhancement when completed.
Plant personnel performed effectively during two actual activations of the Emergency Plan. Quality assurance oversight of the program was very competent.
Page25of 35
PLANT ISSUES MATRIX DC Cook SearchSortedbyDate(Descending)andSMMCodes(Ascending):
Search'Column
= 'Date';
Beginning Date = 10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 153 11/26/1997 Positive IR 97022 NRC Plant Support 1B Licensee personnel performed appropriately during two actual activations (Unusual Events) of the Emergency Plan. A procedure had been developed for review of actual events.
Review of the events was very good.
(Section P1.1) 154 11/26/1 997 LER LER 316-97009 Licensee Maintenance 2A Blockage of Containment AirRecirculation Inlet Line Results in Condi
'utside Design Bases One of two Train B inlet lines for the hydrogen removal and air recirculation from the Unit 2 number 2 and 3 Steam Generator enclosures was found to have been blocked by concrete. The blockage occurred during the Unit 2 Steam Generator replacement in 1988.
155 11/26/1997 Positive 156 11/26/1997 Positive 157 11/26/1997 Strength 158 11/26/1997 Positive IR 97022 IR 97022 IR 97022 IR 97022 NRC NRC NRC NRC Plant Support Plant Support Plant Support Plant Support 2A 3A 1C 3B 3C 1C Each emergency response facilitywas well maintained and in an excellent operational state of readiness.
The proposed EOF had adequate space, lighting, telephone and electrical outlets. The proposed layout of the facilitywas acceptable.
An appropriate plan for turnover from the old facilityto the new.
(Section P2.1)
Procedure reviews indicated that EP procedures provided detailed guidance on responding to declared emergencies.
No procedural problems were identified. Condition Report documentation indicated that plant staff made good use of an Operating Experience report to identify a problem, and had taken aggressive actions to correct the lack ofcurre~
respirator qualifications for members of the Emergency Response Organization (ERO). (Section P3)
The EP training program was very effective. Drills were used as an efficient and performance-based requalification method. Training records were complete, indicated that ERO members were trained within the required frequency. The quality of drill cntiques improved during 1997.
The ERO's staffing levels were excellent for all key and support positions.
(Section P5)
The overall effectiveness of the licensee's emergency preparedness facilities, equipment, training, and organization was effective. (Sections P2.1, P3, P5, and P6)
Page 26of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date =
10/1/97';
Ending Date = '9/30/98' DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 159 11/21/1997 Positive 160 11/21/1997 Weakness 161 11/21/1997 Positive IR 97023 IR 97023 IR 97023 NRC Operations 1C 5A NRC Operations 5A NRC Operations 5A 3A The inspectors determined that the reviewed procedures adequately addressed the licensee's goals and expectations for effectively implementing the corrective action process and the audit and self-assessment programs.
The ability of the licensee to track and trend conditions identified in conditions reports was marginal. The initial input of text data had problems due to the limited data field and the reliance of clerical staff to summarize the technical data into the small field. The causal codes failed to provide a meaningful sort capability which left the staff with a need to manually sort through numerous pages of data printouts and conditions reports.
Finally the root cause investigations were not part of the KTP and had no automated method of tracking and trending the findings.
The corrective action program was in transition and progress has been made in improving the quality of CR content.
The licensee identified the problem with the training department not writing CRs, and not conforming to the new program and management expectations.
The staff was identifying good items as they performed their normal work.
162 11/21/1997 Weakness 163 11/21/1997 Negative IR 97023 NRC Operations 5C 5A IR 97023 Licensee Engineering 5A 3B During the latter part of this inspection period. the inspectors became aware that in two recent self-initiated audits, you identified a riumber of issues involving inadequate training and qualifications of engineering personnel.
The audit/surveillance program covered the required areas and was identifying problems and concerns.
Audit findings were documented in condition reports, which were used for tracking and to obtain corrective actions. The inspectors concluded that the licensee's surveillances and audits were effectively being conducted to identify problems and concerns.
However, as noted in previous licensee findings and independently confirmed by the NRC, the followup (resolution) of the identified problems and concerns was not being resolved in a timely fashion.
164 11/21/1997 Negative IR 97023 NRC Operations 5C 5A The licensee was effective in identifying problems through audits and through the staff observing problems during their daily activities. "
However, the ability of the licensee to followthrough to completion on corrective or preventative action was questionable.
Page27of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 165 11/16/1997 VIO/SL-IV IR 97024 VIO NRC Operations 2A 4B The inspectors identified a violation in which the licensee had been removing control room annunciators from service without maintaining records which contained a written safety evaluation as required by 10 CFR 50.59 (Section 03.1).
166 11/7/1997 Negative 167 11/7/1997 NCV 168 11/7/1 997 URI 169 11/7/1997 Positive 170 11/7/1997 NCV NRC IR 97018 Plant 2A Support IR 97018 NCV NRC Maintenance 2A IR 97018 NRC Maintenance 2B IR 97018 NCV Licensee Operations 3A IR 97018 URI Self-Maintenance 2A 4B Revealed The inspectors identified lights under a temporary trailer that were inoperable.
The specific root cause was not identified (Section S2).
The inspectors identified unsecured foreign material near the recirculation sump in the Unit 2 lower containment.
The sump was not required by Technical Specifications to be operable, and the amount of material would not have significantly degraded the performance of the sump. This was a violation of minor significance (Section M4.1).
Following a failure of the 2 AB D/G flywheel end exhaust manifold bracket, the licensee discovered that required jam nuts on the bracket bolts were missing from two emergency diesel generators, 1CD D/G and 2 AB D/G. The licensee speculated that the missing jam nuts may have allowed the bracket bolt to come loose, resulting in a fatigue failure of the bracket; however, the minor modification package paperwork indicated that the jam nuts had been installed. An unresolved item was opened pending a review of the licensee's investigation into the root cause of the bracket failure (Section M2.2).
The control air system safety valves appeared to be properly installed dedicated as safety grade components.
The inspectors questioned the use of work procedures annotated for non-safety-related work to install safety-related valves; however, no violations of NRC requirements were identified (Section M3.1).
The licensee identified that during a 4 day period six human performance errors by licensed and non-licensed operators occurred.
None of the errors resulted in personnel injuries, equipment damage or an engineered safety features actuation. A non-cited violation for a failure to follow procedures was issued (Section 01.2).
Page 28 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 171 11/7/1997
~
LER LER 97028 NRC Plant 4A Support 10 CFR 50, Appendix R, Section III 6.2.(b) requires a twenty foot separation between trains with no intervening combustible materials with fire detection and suppression installed. At Cook Nuclear Plant, fire stops were being used to prevent the possible spread of fire across the twenty foot separation space at two locations in the AuxiliaryBuilding (el.
587'nd 609') where there are intervening combustibles (open cable trays~
inside the twenty foot separation space.
However, an exemption to 10~
CFR 50, Appendix R, Section 6.2.(b) had not been requested.
172 11/7/1997 Positive IR 97018 NRC Engineering 4B Engineering personnel were involved in several of the issues discussed in this report (refer to Section 03.1, Procedures forCross-Tying 250 Vdc Buses During Maintenance Activities (Unit 2), and Section 03.2, Emergency Operating Procedures Containing Incorrect Set points).
Engineering support to the rest of the licensee organization appeared to be good, but the support was supplied in response to NRC questions (Section E1).
173 11/7/1997 VIO/SL-IV IR 97018 VIO NRC Operations 4B 174 11/7/1997 VIO/SL-IV IR 97018 VIO NRC Operations 4B The inspectors identified a discrepancy between the pressurizer pressure low safety injection set point as referenced in emergency operating procedure E-0 and as listed on a control board operator aid. While evaluating the inspectors'uestions concerning this discrepancy, the licensee identified discrepancies between the plant set-point document and reactor trip set points as listed E-0. The inaccurate procedure was a third example oi a violation oi NRC reqoirementa (Section 03.2).
The inspectors concluded that the licensee's initial plans and procedures to cross-tie safety-related electrical buses lacked adequate analysis and controls to support plant operation in the proposed configuration. Two examples of a violation for procedural inadequacy were identified. The inspectors were concerned that the licensee did not conduct an adequate evaluation of cross-tying 250 Vdc buses until questioned by the inspectors (Section 03.1).
175 10/31/1997 Positive IR 97020 NRC Plant 3A Support The reactor head set was well controlled. The inspector observed good communications between the radiation protection technicians (RPTs) and other workers regarding ALARAconcerns at the pre-job briefing.
Although communications between the two RPTs went well, there was some difficultyin communications between the other workers during the work evolution (Section R1.2).
Page29of 35
PLANT ISSUES MATRIX
.DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 176 10/31/1997 Negative IR 97020 NRC Plant 3A Support As-low-as-is-reasonably-achievable (ALARA)controls were effective in maintaining the station dose totals below the goal. However, the inspector was concerned by the lack of conservative decision making that allowed work under a radiation work permit to continue until the dose estimate had been exceeded by 33 percent (Section R1.1).
177 10/28/1997 LER LER 315-97027 Other Engineering 4A Westinghouse Integral Fuel Burnable Absorber (IFBA) Fuel Rods The licensee was informed by Westinghouse that in the process of developing new fuel rod cladding corrosion and rod internal pressure models it was determined that these new models showed a decrease in rod internal pressure margin. This decrease in margin has the potential to place plants in a condition that is outside of their design basis with respect to pellet-to-clad gap re-opening.
178 10/24/1 997 LER 179 10/17/1997 Weakness 180 10/17/1997 VIO/SL-IV LER 97008 IR 97019 IR 97019 Self-Operations 2A 1B Revealed NRC Plant 3B 1C Support NRC Plant 3C Support
'On October 24, 1997 at 0008 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with Unit 2 in Mode 6, Refueling, an unplanned ESF actuation occurred.
A high radiation alarm and high fail signal were generated by the Train A Upper Containment Area Radiation Monitor due to equipment failure. The high radiation alarm resulted in a Containment Ventialtion Isolation. The associated isolation valves closed as designed.
Staff reductions significantly affected the capability of conducting non-Appendix B training. The licensee eliminated table top tactical respons~
drills, the combat weapons course, and shift deployment exercises. Th elimination of much of the non-appendix B progam was considered a weakness in the contingency response program.
(Section S5.1)
Actual site security response drills conducted during February and June 1997 demonstrated response team members were located distant from some plant vital areas making it unlikely responders could interdict adversaries before reaching target sets.
This is a violation of 10 CFR 73.55(a) and 10 CFR 73.55(h)(4)(i)(A). (Section S4.1)
Page30of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98
¹ DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 181 10/17/1 997 Negative 182 10/15/1997 Negative 183 10/15/1997 Negative 184 10/15/1997 Positive 185 10/15/1997 Positive 186 10/15/1997 Positive 187 10/15/1997 Positive IR 97019 IR 97015 IR 97015 IR 97015 IR 97015 IR 97015 IR 97015 NRC Plant 3C 1C Support NRC Operations 1C NRC Maintenance 2B NRC Maintenance 2B 1C NRC Operations 2B 4B NRC Plant 3A Support NRC Maintenance 3A The number of required armed responders specified in the approved security plan was not met on eighteen occasions between October 26, 1996 and May 31, 1997 when the weapons of one to two guards per shift were removed and their radio frequencies switched to the fire protection radio channel, in order for these individuals to participate as members of the site fire brigade during the routine monthly tire brigade drills (Senti S4.2).
The inspectors observed that the licensee was not evaluating cascading Technical Specifications consistently in order to determine the operability of supported equipment.
The inspectors'eview-did not identify any violations of Technical Specifications (Section 02.1).
The inspectors reviewed the reference and acceptance parameter ranges for inservice testing of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 pumps.
Not all reference ranges were within ~ 2 percent of the reference parameter as required (Section M1.4).
The licensee initiated an extensive upgrade program in response to previous NRC concerns with new fuel receipt. The inspectors observed portions of the licensee's handling of new fuel in preparation for the Unit 2 refueling outage and identified no new concerns (Section M4.1).
During a followup review of an automatic start signal to the turbine driv auxiliary feedwater pump, the inspectors determined that a modification was already in progress to prevent inadvertent start signals in the future (Section 02.2).
The inspectors observed appropriate and safe laboratory techniques during a receipt analysis of emergency diesel generator fuel oil (Section R4.1).
Maintenance worker and licensed operator performance during the replacement of cell N34 to the safety related Unit 2 CD battery was good.
Excellent communications and coordination were observed by the inspectors (Section M1.3).
Page 31 of 35
PLANT ISSUES MATRIX DC Cook 11/23/98 Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date = '9/30/98' DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 188 10/15/1997 URI IR 97015 URI NRC Plant Support 5A During a review of an Operating Experience report, the licensee identified that the program for maintaining operator respirator qualifications did not include a provision for ensuring that the operators maintained corrective lenses available when necessary.
The program also failed to include provisions for tracking respirator qualification for members of the emergency response organization who were required to be respirator qualified (Section P5.1).
189 10/15/1997 Positive IR 97015 NRC Maintenance 5B 3A The licensee's troubleshooting efforts following a failure of the 2 AB emergency diesel generator voltage regulator were aggressive and thorough (Section M1.2).
.190 10/14/1997 Negative IR 97018 IFI NRC Maintenance 2A The 2 AB D/G experienced a number of electrical and mechanical failures since May 1997. Two valid run failures resulted in the 2 AB D/G being placed on an accelerated testing frequency. The inspectors were concerned that these failures were indicative of poor material condition.
An inspection followup item was opened to track resolution of the material condition of the 2 AB D/G (Section M2.1).
191 10/14/1997 LER 192 10/14/1997 Positive 193 10/14/1997 Positive IR 97020, LER 97007 IR 97016 IR 97016 Licensee Plant 3B Support NRC Maintenance 5C NRC Maintenance 5C Radiation protection (RP) personnel identified that a contractor had arrived at the facilitywith several hot particles located in his shoe.
The RP department was evaluating where and when the contractor had become contaminated with these particles.
A preliminary estimate determined that the dose was approximately 82.7 rads to the skin, whic~
would be an exposure in excess of the 10 CFR 20.1201 limits. The R~
staff was able to demonstrate that the contamination did not originate at the station based on the results of the entrance whole body count, so the inspector determined that no violation of NRC requirements had occurred at D. C. Cook. This estimate was a preliminary number and further investigation by RP personnel, as well as the final dose assigned to the contractor, willbe reviewed in future inspections (Section R1.3).
Five findings involving availability performance criteria, quarterly assessments, masking of components, structure monitoring, and audit corrective actions were closed.
Two findings remained open. The first involved the required refueling cycle evaluation which was scheduled for completion on January 31, 1998. The second item was related to the acceptability of reliability performance criteria.
Page32of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date =
'9/30/98'1/23/98 I
DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 194 10/10/1997 LER LER 316-97006 Licensee Maintenance 2A Equipment in Containment Rendered Inoperable Due to Faulted Floodup Tubes 195 10/3/1997 Negative
~
IR 97019 NRC Plant 1C Support An inspection of unit 2 floodup tubes conducted during its current refueling outage identified three tubes containing cables connected to safety related components with through wall holes caused by welding activities.
Staff reductions significantly affected the capability of conducting non-Appendix B training. The licensee eliminated table top tactical response drills, the combat weapons course, and shift deployment exercises. The elimination of much of the non-appendix B program was considered a weakness in the contingency response program.
(Section S5.1) 196 10/3/1997 Negative IR 97019 Licensee Plant 1C Support The number of required armed responders specified in the approved security plan was not met on eighteen occasions between October 26, 1996 and May 31, 1997 when the weapons of one to two guards per shift were removed and their radio frequencies switched to the fire protection radio channel, in order for these individuals to participate as members of the site fire brigade during the routine monthly fire brigade drills. (Section S4.2)
197 10/2/1997 Strength 198 10/2/1997 Positive 199 10/2/1997 Negative IR 97012 IR 97012 IR 97012 NRC Operations 1A NRC Maintenance 2B NRC Operations 2B The licensee has successfully initiated change to the nuclear organization to shift to a more operations centered focus which has resulted in improvement in prioritization of maintenance activities factoring operational consideration into the decision making process (Section 01.2).
Through observation of the planning, troubleshooting, and preparation for the change out of a solenoid valve on the air supply to a letdown isolation
, valve inside containment, the inspectors determined that the work activity was well planned and implemented (Section M1.2).
The licensee complied with Technical Specifications during maintenance on containment isolation valve 2-DCR-204, but the valve was not declared inoperable.
Afterthe inspectors discussed their concerns about the valve with the operators, the Unit Supervisor declared the valve inoperable in order to avoid a misunderstanding which could result in the valve being improperly restored to service (Section 01.3).
Page 33 of 35
PLANT ISSUES MATRIX DC Cook Search Sorted by Date (Descending) and SMM Codes (Ascending):
Search Column = 'Date';
Beginning Date = '10/1/97';
Ending Date = '9/30/98' DATE TYPE SOURCE ID BY SALP SMM CODES DESCRIPTION 11/23/98 200 10/2/1997 Positive IR 97012 NRC Plant 3A Support Thorough ALARAplanning and work preparations contributed to the successful restoration of operability for a letdown orifice block isolation valve (Section R1.1).
201 10/2/1997 Positive 202 10/2/1997 Positive IR 97012 IR 97012 NRC Plant 5A 28 Support r
NRC Engineering 58 5A The inspectors concluded that the licensee's program for sampling and storing emergency diesel generator fuel oil was in compliance with Technical Specifications, regulatory guides, and industry testing criteri (Section R3.1).
Engineering personnel were slow to initiate a condition report for steam generator level channel deviations.
However, engineering personnel were thorough in their evaluation of the causes for the steam generator level deviations (Section E2.1).
Page34of 35
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