IR 05000277/1998007
| ML20151Z247 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151Z240 | List: |
| References | |
| 50-277-98-07, 50-277-98-7, 50-278-98-07, 50-278-98-7, NUDOCS 9809210225 | |
| Download: ML20151Z247 (37) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
License No DPR-44 DPR-56 ,
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Report No '
98-07 98-07 )
i Docket No l 50-278 i
Licensee:
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PECO Energy Company Correspondence Control Desk P.O. Box 195 Wayne, PA 19087-0195 Facility: Peach Bottom Atomic Power Station Units 2 and 3 Inspection Period: June 23,1998 through August 10,1998 Inspectors: A. McMurtray, Senior Resident inspector l M. Buckley, Resident inspector '
B. Welling, Resident inspector ,
R. Nimitz, Senior Radiation Specialist ;
J. Carrasco, Reactor Engineer '
Approved by: Clifford J. Anderson, Chief Projects Branch 4 i Division of Reactor Projects i
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9909210225 900915 '
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! EXECUTIVE SUMMARY Peach Bottom Atomic Power Station ~
!. NRC Inspection Report 50-277/98-07,50-278/98-07 This inspection report included aspects of licensee operations; surveillances and maintenance; engineering and technical support; and plant support areas. The report covers a seven-week period of resident inspection and inspections by a regional radiation protection specialist and a regional engineering specialis . Operations:
e On June 8,1988, the 3 start-up transformer became inoperable following a
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severe electrical storm, but this was not recognized by operators until June 22, 1998. On June 15, the inoperable 3 start-up transformer was aligned to the 2 start-up and emergency source for over nine hours to support off-site maintenance work. Technical specification 3.8.1 checks on the correct breaker l
alignment and indicated power availability were not performed on June 15. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV), consistent with Section Vll.8.1 of the NRC Enforcement Polic Corrective actions from Licensee Event Report 96-005 were inadvertently removed from the round sheets when the sheets were converted from manual to electronic entry, which contributed to the violation of Technical Specification 3.8.1. (Section 02.1)
L e A reactor water level excursion on July 13,1998, during transfer between feedwater control system computers revealed that instrument and control i personnel did not have sufficiently specific written guidance or criteria on !
computer signal differences for performing the computer transfer. Instrument I
' and control personnel relied on inappropriate assumptions on acceptable computer signal differences. Corrective actions for this issue were good. A subsequent transfer evolution, after tuning of the control systems, resulted m no !
reactor level change. (Section 02.3) !
'o On' July 17,1998, the 2A condensate pump had to be shutdown quickly due to rapidly climbing temperatures on the thrust bearing. These high temperatures were the result of flow restriction in the thrust bearing cooling system due to ;
stem / disc separation on the turbine building closed cooling water outlet valv j High temperatures previously observed on the thrust bearing, on July 13, were
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treated as routine work rather than as priority work and were not recorded in any I of the operator logs. Equipment operators were not fully aware of the design of l the turbine building closed cooling water outlet valve and therefore did not realize ,
that overtorquing the valve on its backseat could result in separation of tt;. dise l from the stem. (Section O2.4) l I
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Executive Summary (cont'd)
e The recently issued procedure, OM-P-11.4, Revision 1, " Operations Peer Checking" provided enhanced guidance to operations personnel regarding responsibilities and criteria for performing peer checks. This procedure provided guidance that can help reduce human performance errors if fully implemente (Section 03.1)
- Nuclear Review Board members provided critical reviews in the areas of operations, maintenance, and engineering during the meeting held on August 6, 1998. (Section 07.1)
Maintenance:
e Most maintenance work and testing performed during this inspection period was professional and thorough. Technicians were experienced and knowledgeable of their assigned tasks. The work and testing procedures were present at the
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jobsite and actively used by the technicians and operators for activities observed. '
Good pre-job briefs were observed prior to the performance of the surveillances observed. Applicable procedures were present in the control room and at the job sites during surveillance testing and were appropriately use During observations of the 2A core spray pump maintenance, a quality verification inspector recorded a foreign material check of the pump casing internals as unsatisfactory. The quality verification check was performed after a worker verification of cleanliness. The quality verification inspector found some small debris and paint chips, which were subsequently removed by the worker Maintenance supervisors were taking steps to improve worker sensitivity to foreign material exclusion. In addition, the maintenance manager initiated an l investigation into a trend of less than adequate worker verifications during l maintenance, including welding verification steps and cleanliness inspection (Section M1.1)
e PECO properly completed a pressure test that verified the integrity of the unit 2 j high pressure coolant injection piping. The system was adequately returned to ;
operable status. However, the insulation removal from high pressure coolant injection piping and components was poorly contro"ed and executed during this l test. Also, operations personnel did not monitor the high pressure service water pump room temperature during the test, until after the room high temperature I alarm was received. (Section M1.2)
Enaineerina:
e The independent spent fuel storage installation (ISFSI) geotechnical report and associated regulatory analyses were acceptable. The results and recommendations of this geotechnical report had been appropriately used in the design of the storage pad and provided a satisfactory basis for planned ISFSI construction activities. (Section E1.1)
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Executive Summary (cont'd)
e The failure to perform an inservice testing surveillance requirement for main steam isolation valve stroke timing during cold shutdown conditions revealed performance weaknesses among engineering personnelin the procedure revision and review processes. Written communications for, and reviews of, a surveillance test procedure revision were poor and failed to identify an error in the test frequency. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV), consistent with i Section Vll.B.1 of the NRC Enforcement Policy. (Section E2.2)
e On June 22,1998, a reactor building equipment operator discovered during l routine operator rounds that the Unit 3 reactor core isolation cooling system I mechanical overspeed trip tappet was not fully reset. Station personnel l determined that the reactor core isolation cooling system had been inoperable j since May 4,1998 which was the last time the overspeed trip function was j manipulated and successfully tested. This condition resulted in a violation of !
technical specification 3.5.3 since the reactor core isolation cooling system was !
inoperable for greater than 14 days while Unit 3 was operating. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-l Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement Polic Engineering personnel fully investigated all of the causes of the failure of the trip l tappet to reset and adequately diagnosed and repaired each of the problems identified. (Section E4.1)
Plant Sucoort:
o PECO implemented effective internal and external exposure control programs with respect to personnel monitoring and dose assessment, personnel dosimetry use and application, and radiation and high radiation area monitoring and control.
- No significant unplanned exposures were evident during the period under review.
I No violations or safety concerns were noted. (Section R1.2)
e PECO implemented a generally effective contamination control program with respect to tracking and trending personnel contamination events, and initiating corrective measures to improve personnel work performance and limit the potential for inadvertent contact with contaminated areas and material Program cvaluations and enhancements were underway to improve contamination control and work practices and reduce instances of minor personnel contaminations. (Section R1.3)
e PECO continued to implement an overall effective ALARA program with re ,.act to work planning and control, use of dose reduction initiatives such as remote monitoring equipment, application of shielding, and work monitoring via closed-circuit television. Outage work planning and control efforts to effect improved ALARA performance are continuing. (Section R1.4)
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. Executive Summary (cont'd) !
- .PECO implemented an effective program int self-identifying and correcting self- ,
identified issues and concerns with respect to initiation of comprehensive self- !
- evaluations, usually timely corrective actions for self-identified issues, and the ;
conduct of audits with sufficient scope and depth. No violations or safety 2 concerns were identified. (Section R7.1)
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l TABLE OF CONTENTS EX E C UTIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii l
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TA8 LE O F CO NTE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi Summ ary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 I. Operations ....................................................1 O1 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Ge neral Comme nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 02 Operational Status of Facilities and Equipment ................... 2 O2.1 (Closed) LER 98-004 Failure to Meet Technical Specification Requirements for One Inoperable Off-site Power Source . . . . . . . . 2 l 02.3 Reactor Level Excursion Following Transfer of Feedwater Level l Control Computer (Unit 3) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 O2.4 Rapid Shutdown of the 2A Condensate Pump Due to High Thrust Bearing Temperature ................................6 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . 8 03.1 Review of OM-P-11.4, Operations Peer Check Program" . . . . . . . 8 l 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 07.1 Nuclear Review Board Meeting .........................8 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 08.1 (Closed) LER 97-003 High Pressure Coolant injection (HPCI)
Inope rable (Unit 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 08.2 (Closed) Violation (VIO) 50-277/98-03-02 Failure to Submit LER for TS Non-Compliance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 08.3 (Closed) Unresolved item 50-277(278)/97-07-03 Unit 2 Cooldown i Monitoring Following the Reactor Scram on November 9,1997. . 10 08.4 Institute of Nuclear Pcwer Operations (INPO) Evaluation Review (71707) ........................................10 11. M a i nt e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 M 1.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 M1.2 High Pressure Coolant Injection System Piping Pressure Test inspection (Unit 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 M8 Miscellaneous Maintenance Activities ........................12 M8.1 (Closed) IFl 50-277(278)/97-04-01 Review Maintenance Rule Program Application of 13 KV Breaker Switches . . . . . . . . . . . . 12 111. E n g i n e e ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E Independent Spent Fuel Storage Installation Geotechnical Review
.............................................13 E (Closed) LER 98-003 Missed Surveillance Requirement for Main Steam Isolation Valve Stroke Timing (Unit 3) . . . . . . . . . . . . . . 14 vi
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P E4 Engineering Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 15 ,
E (Closed) LER 98-002 Failure to Meet TS Requirements for the i Unit 3 RCIC While the Mechanical Overspeed Trip Tappet Was !
Not Fully Reset . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 !
E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 :
E (Closed) Inspection Followup ltem (IFI) 50-277(278)/97-02-03 l Station Blackout (SBO) Line Load Testing . . . . . . . . . . . . . . . . . 16
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l E8.2 (Closed) P2193-22 and 94-83 Emergency Diesel Generator Heat !
Exchanger Crossflow 10 CFR Part 21 Reports . . . . . . . . . . . . . . 17 {
E8.3 ' (Closed) IFl 50-277(278)/97-07-07 Emergency Diesel Generator Lube Oil Piping - Potential 10 CFR 21 issue . . . . . . . . . . . . . . . . . 8 ;
E8.4 (Closed) IFl 50-277(278)/97-06-04 Diesel Driven Fire Pump l
Battery Explosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 i
IV. Plant Support ................................................19 l R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 19 -
R1.1 Radiological Controls (Program Changes) ................. 19 R1.2 Internal and External Exposure Controls . . . . . . . . . . . . . . . . . . 20 t R1.3 Control of Radioactive Materials and Contamination . . . . . . . . . . 21 R1.4 ALARA Program and Unit 2 Refueling Outage Planning, Preparation, Emergent Work Control ....................23
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R7 Quality Assurance in Radiological Protection and Chemistry Activities . . 24 R7.1 Radiological Self-Assessment .........................24 ,
R8 Miscellaneous RP&C Activities .............................25 '
R8.1 injection of Noble Metals ............................25 R8.2 Housekeeping (71750) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
V. Management Meetings ..........................................26 X1 Exit Meeting Sum m ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 X2 Review of Updated Final Safety Analysis Report (UFSAR) Commitments . 26
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ATTACHMENTS a ATTACHMENT 1 - List of Acronyms Used
- Inspection Procedures Used
- Items Opened, Closed, and Discussed ATTACHMENT 2 - Maintenance and Surveillance Observations vn
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Report Details
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Summary of Plant Status i l
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a PECO operated both units safely over the period of this repor l
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Unit 2 began this inspection penod at 92% power, in end-of-cycle coast down. Power '
was reduced to approximately 60% on July 10 - 11,1998, for condenser waterbox l cleaning. Unit 2 power was at 74% at the end of the inspection perio i
- r l Unit 3 began.this inspection period at 100% power. On July 11, unit load was reduced to l L 74% for main steam isolation valve testing. Unit power level remained at 100% for the l rest of the perio l. Operations j
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01 Conduct of Operations'
l 01.1 General Comments I l
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l The inspectors observed that communications among operators in the control room
, and with others in the field were usually very good. Inspectors also attended a !
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number of control room briefings, which were typically thorough and professionally l conducte Improvements were observed in the use of peer checking. ' A new procedure, Operations Manual (OM)-P-11.4, " Operations Peer Check Program," provided j
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enhanced guidance to operators regarding responsibilities and criteria for performing i
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peer checks (See Section O3.2). .,
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On June 22,1998, a reactor building equipment operator discovered the Unit 3 ;
l reactor core isolation cooling (RCIC) mechanical overspeed trip tappet not fully l l
reset. This condition had the potential, due to vibration, to trip the RCIC turbine ]
during normal startup or upon an engineered safeguards feature (ESF) actuatio i The inspectors noted that this finding showed good questioning attitude of
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conditions observed in the field by the equipment operato .On July 22,1998, hydrogen water chemistry ir' -tion into the unit 2 feedwater system unexpectedly isolated during application of a clearance for the 2A reactor i feedwater pump. The inspectors concluded that the lack of knowledge by the l clearance writer and lack of detailin the clearance, of an interlock between the control switch for valve, AO-8366A, and the 'B' injection line, were the main cause of this event.
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. ' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report
' outline. Individual reports are not expected to address all outline topics.
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The inspectors reviewed the control room Equipment Deficiency Log, action requests designated as Main Control Room (MCR) deficiencies, and the non-outage .
maintenance backlog. The inspectors noted that although there were a few long !
standing deficiencies that posed a burden to the operators, in general the MCR deficiencies were tracked, scheduled, and planned in a manner that allowed reliable <
and safe operation of the plan )
02 Operational Status of Facilities and Equipment !
O2.1 (Ciosed) LER 98-004 Failure to Meet Technical Specification Reauirements for One inocerable Off-site Power Source ! Insoection Scoce (71707)
l On June 8,1998, the 3 start-up (SU) tap changer had become inoperable and was '
not recognized by operations personnel. On June 15, the 3 SU transformer was aligned to the 2 start-up and emergency (SUE) source to support off-site !
maintenance work. Since the 3 SU tap changer was inoperable, the 2 SUE off-site !
source was inoperable and checks required in technical specification (TS) 3. !
were not performe !
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The inspectors discussed this event with operations and engineering personnel and reviewed applicable documentatio Observations and Findinas Station personnel noted that the 3 SU tap changer had been successfully tested following maintenance on June 3,1998. Station personnel determined that there was a high probability that the 3 SU tap changer became inoperable when varistors in the tap changer shorted to ground due to lightning from severe electrical storms in the Peach Bottom area on June 6 and LER 96-005 documented a similar event w May 7,1996, when the 3 SU load tap !
changer was inoperable due to the actuaum of the tap changer surge protection circuitry. Part of the corrective actions for this event included specifying an acceptance criteria associated with the reading taken during rounds of the number of automatic load changes for the tap changer. This acceptance criteria was listed as greater than zero load changes on the manual-entry, round sheets used at the tim Operations personnel noted that the current electronic-entry, round sheets still required the equipment operators to verify that taps were occurring during the (' normal system alignment for 3 SU. However, the acceptance criteria for load i changes had been changed to zero when the logs were converted from manual to electronic entry. This acceptance criteria change occurred because the electronic
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The 3 SU tap changer circuitry was repaired, tested and returned to servic Corrective actions included reviewing the associated electronic round sheets and revising the sheets as appropriate to ensure positive verification of tap changes were documented during rounds. Also, other electronic round sheets were reviewed to ensure no similar problems exist. In addition, the system operating procedures for transferring start-up sources were revised to functionally check the manual tap changer prior to having the transformer carry an emergency auxiliary transformer and 4 kV loads. The licensee initiated a performance enhancement program (PEP) document number 10008610to investigate and correct the various problems associated with this even The inspectors performed an on-site review of this LER and noted that operations l personnel did not recognize that the 3 SU load tap changer was inoperable following the electrical storms on June 6 and 7,1998 until June 22,1998. The inspectors determined that equipment operators did not have a heightened awareness of the effects that these storms had on plant eouipment. Operations personnel were working on a storm response procedure that would check various plant equipment that could be affected by severe weather to improve awareness.
The inspectors also determined that the corrective actions in response to LER 96-005 were inadvertently removed from the round sheets, used by the equipment operators, when the sheets were converted by manual to electronic entry. The inspectors determined that this oversight contributed to the violation of TS 3. Technical specification (TS) 3.8.1 required that if one offsite circuit was inoperable, then TS surveillance requirement 3.8.1.1 must be performed within one hour and 1 once every eight hours thereafter. Technical specification surveillance requirement l 3.8.1.1 required verification of correct breaker alignment and indicated power availability for each offsite circuit. Contrary to the above, on June 15,1998 from 05:15 a.m. to 2:36 p.m., the correct breaker alignment and indicated power availability for each offsite circuit was not checked when one offsite circuit was inoperable. The offsite circuit was inoperable when the 3 SU transformer was aligned to the 2 SUE source because the 3 S'!'oad tap changer was inoperabl Station personnel determined that there w. no positive verification that the tap changer was operational prior to alignroer.t to the preferred offsite sourc Although this event would not have occurred if the corrective actions from LER 96-005 had been adequately noted in the new electronic entry logs, LER 96-005 occurred beyond the past two years of this event. Therefore, this event was considered non-repetitive per the "NRC Enforcement Policy." This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-277(278)/98-07-01)
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4 Conclusions On June 8,1988, the 3 start-up transformer became inoperabie following a severe electrical storm, but this was not recognized by operators until June 22,1998. On June 15, the inoperable 3 start-up transformer was aligned to the 2 start-up and
, emergency source for over nine hours to support off-site maintenance work.
l Technical specification 3.8.1 checks on the correct breaker alignment and indicated I power availability were not performed on June 15. This non-repetitive, licensee-l.
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identified and corrected violation is being treated as a Non-Cited Violation (NCV),
consistent with Section Vll.B.1 of the NRC Enforcement Polic Corrective actions from Licensee Event Report 96-005 were inadvertently removed from the round sheets when the sheets were converted from manual to electronic entry, which contributed to the violation of Technical Specification 3.8.1.
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O2.2 Review of Station Preoaredness for Loss of Conowinao Pond Event Insoection Scope (71707. 37551)
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The inspectors reviewed selected station facilities, equipment, and procedures that would be used in the event of a loss of the ultimate heat sink (Conowingo Pond). Observations and Findinas i
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Station Special Event Procedure SE-3, " Loss of Conowingo Pond," included instructions to lineup the emergency cooling tower (ECT) and associated systems to ensure that both units can be safely shutdown in the event of such scenarios as an unusually low river level or failure of the Conowingo Da The inspectors reviewed the surveillance procedures for the ECT and found that the l procedures fulfilled the surveillance requirements specified in TS 3.7.3.
, The inspectors noted that the actions in SE-3 were consistent with the descriptions in the Updated Final Safety Analysis Report (UFSAR). The inspectors also performed a walkdown of selected equipment and observed that the equipment
, necessary to accomplish contingency actions according to the SE-3 procedure were l in place. Specifically, the inspectors found that temporary pumps and hoses for l makeup water were present in sufficient quantitie Routine testing of the temporary pumps in early August, however, revealed some degraded conditions. Operations personnel found that one of three pumps did not i have a required suction fitting, and another pump had engine problems and did not L pump water. Action requests were drafted to correct the deficiencies. The i
operators concluded that two pumps remained operable. Thia met the requirement for at least one operable pump as described in SE-3 and technical specifications.
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Operators noted that pump performance problems and other material condition deficiencies were repeat problems and had been documented in 1994 and 1995. . . .
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Further review by the inspectors noted that five deficiencies on the pumps were l identified in 1997, but only one of these was actually resolved. The incorrect l suction fitting had been previously documented as a deficiency. Although these deficiencies were of minor significance, the inspectors concluded that the corrective actions for previous pump performance problems and deficiencies were poor. Conclusions l The systems, equipment,'and procedures that operators would use in the event of a i
loss of the ultimate' heat sink (Conowingo Pond) were adequately controlle . Repetitive material condition deficiencies on temporary pumps used for emergency l cooling tower replenishment were identified by operators through routine testing, and repairs were requeste .3 Beacter Level Excursion Followina Transfer of Feedwater Level Control Comouter (Unit 3)
! Insoection Scope (71707 & 62707)
i: .The inspectors examined an operational event which occurred during a transfer of
.the on-line feedwater level control computer and led to a reactor level excursion and low level alarm. The inspectors discussed the event with operations and instrument and control (l&C) personnel and the system manager.
l l Observations and Findinas On July 13,1998, instrumentation and controls (l&C) personnel were preparing to perform corrective maintenance on the 'X' digital feedwater control system computer. The work order required swapping the in-control feedwater control system computer from the 'X' to the backup, "Y' computer. During the transfer, e - operators experienced a drop in reactor water level from its normal value of 23" to approximately 16" (low level alarm !s 17"). Operators entered the appropriate operational transient procedure for reactor low level, and the feed pumps responded and level was restored without operator actio Prior to the computer transfer, operators and l&C personnel conducted a pre-job brief, which addressed the possibility of reactor level transient. Operators discussed contingency actions in the event of a significant level drop, but they expected a l maximum level change of only a few inches. Operators stated that level !
l perturbations during past computer transfers were minor; they were unaware of any I L level changes below 20".
l&C technicians stated that, during the evolution, they monitored the output signals L of the 'X' and 'Y' computers and attempted to perform the transfer when the difference between the two outputs were at a minimum. They indicated that the signals differed by about 8 - 9% during the transfer, which was the minimum attainabl . - . . _ - . - _ _ -. _ -. . . - - -. - - - ...-_. - .- . .
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The inspectors noted that the work order contained no specific written guidance relative to the signal difference, nor any acceptance criteria or maximum allowed difference. The l&C technicians stated that they did not have any historical data / trending information on the signal differences. They also acknowledged that i they made assumptions on the magnitude of the level change without such dat The system manager and the l&C technicians stated that they were aware that they could tune selected parameters in the feedwater control system computers to
! minimize the level perturbations. However, there was no procedure or guideline for tuning the parameters.
I The inspectors reviewed the corrective actions taken for this event and then observed a feedwater control system computer transfer on July 31,1998. The inspectors attended the pre-job briefing and reviewed the corrective maintenance work order. The inspectors considered the briefing to be very good, and they noted that the work order specified that the control system be tuned to allow a maximum difference between signals of only 2%. The subsequent transfer was uneventful, l
resulting in no discernable reactor water level perturbation. Based on these results, I the system manager was developing additional procedural guidance on tuning the l l feedwater control systems. The inspectors concluded that these corrective actions l were effectiv I Conclusions A reactor water level excursion on July 13,1998, during transfer between feedwater control system computers revealed that instrument and control personnel did not have sufficiently specific written guidance or criteria on computer signal differences for performing the computer transfer. Instrument and control personnel relied on inappropriate assumptions on acceptable computer signal difference Corrective actions for this issue were good. A subsequent transfer evolution, after tuning of the control systems, resulted in no reactor level chang i 02.4 Raoid Shutdown of the .2A Condensate Pumo Due to Hioh Thrust Beanna Temperature Inspection Scope (71707)
On July 17,1998, the 2A condensate pump had to be shutdown quickly due to rapidly climbing temperatures on the thrust bearing. Since the unit was at 82%
power at the time, no power reduction was required to secure this condensate pum The inspectors reviewed the documentation for this event and discussed the issue with operations personnel.
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$ Qbservations and Findinas
I The cause of the high thrust bearing temperature on the 2A condensate pump was determined to be restricted flow due to stem disc separation on the TBCCW outlet ,
valve. The licensee determined that the disc had separated from the stem due to -
excessive manual force when the valve was back seated as some time prior to !
July 13,199 '
Higher than normal thrust bearing temperatures had been identified on this pump and an Action Request written on July 13,1998. When the temperatures increased c on July 13, an equipment operator mechanically agitated the turbine building closed cooling water (TBCCW) outlet valve which caused movement of the valve disc, a :
flow increase, and the thrust bearing temperature to return to normal. On July 17, 1998 the valve disc relocated to a position which caused the low flow conditio Operations personnel initiated a PEP to document the 2A condensate pump -
shutdown issue. The PEP documented that the loss of thrust bearing cooling on July 13 was not captured in any of the operators logs and that this issue had been scheduled as routine work by station personnel, including Operations, rather than an issue requiring increased priorit The inspectors determined that equipment operators were not fully aware of the design of the TBCCW outlet valve and therefore did not realize that over torquing l
the valve on its backseat could result in separation of the disc from the stem. The inspectors were also concerned that operations personnel did not place a higher priority on evaluating the July 13 temperature increase, since a rapid shutdown of J the condensate pump was require Conclusions On July 17,1998, the 2A condensate pump had to be shutdown quickly due to rapidly climbing temperatures on the thrust bearing. These high temperatures were the result of flow restriction in the thrust bearing cooling system due to stem / disc
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separation on the turbine building closed cooling water outlet valve. High
! temperatures previously observed on the thrust bearing, on July 13, were treated as routine work rather than as priority work and were not recorded in any of the operator logs. The inspectors concluded that equipment operators were not fully aware of the design of the turbine building closed cooling water outlet valve and therefore did not realize that overtorquing the valve on its backseat could result in separation of the disc from the stem.
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03 Operations Procedures and Documentation I
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l 03.1 Review of OM-P-11.4. "Ooerations Peer Check Proaram" ,
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. Inspection Scooe (71707)
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l' The inspectors reviewed the OM-P-11.4, Revision 1, "Opeiations Peer Checking" procedure that was recently issued to address the human performance deficiencies, j by operations personnel, that have been noted during several events that occurred over the past ten month ;
l lJ Observations and Findinas i
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~ Procedure OM-P-11.4 described the responsibilities and qualifications for peer checkers and criteria for determining and specific situations requiring peer checkin Peer checkers were to verify that the intended action was correct and the correct
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component was selected. Peer checking was to be performed for first time 1 equipment manipulations by the performer, infrequently performed tasks, tasks that j
[ : are complex from a human factors standpoint, tasks that may change core l- reactivity, and tasks that have the potential to initiate a plant transient.
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l The inspectors determined that the new procedure for peer checking provided - l
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enhanced guidance to operations personnel regarding responsibilities and criteria for !
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performing peer checks. The inspectors noted that this procedure provided
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guidance that can contribute to a reduction in human performance errors if the l procedural instructions were fully implemented by operations personne Conclugign.g _n I
. The recently issued procedure, OM-P-11.4, Revision 1, " Operations Peer Checking,"
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' provided enhanced guidance to operations personnel regarding responsibilities and criteria for performing peer checks. This procedure provided guidance that can help reduce human performance errors if fully implemente ' 0 Quality Assurance in Operations L
l 07.1 Nuclear Review Board Meetina
! Insoection Scoce (71707) l p i The. inspectors attended portions of Nuclear Review Board (NRB) Meeting #350,
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i The inspectors observed detailed questioning by NRB members in a number'of key areas of plant performance. For example, they inquired as to the status of l corrective actions for some operations-related violations identified in recent NRC I l inspection reports, and they raised questions about the procedure revision process.
l The inspectors noted good analysis of some initiatives that are intended to improve operations department, such as the Event Free Operations and Back to Basics program. Good discussions were c.lso observed in the areas of Maintenance and Plant Engineerin Conclusions
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l Nuclear Review Board members provided critical reviews in the areas of operations, maintenance, and engineering during the meeting held on August 6,199 Miscellaneous Operations issues 1 08.1, (Closed) LER 97-003 Hioh Pressure Coolant Iniection (HPCI) Inocerable (Unit 2)
The licensee determined that the momentary lifting of the common power wire lug to support jumper installation, caused a portion of the HPCI logic power circuitry to
! deenergize which resulted in the alarms and automatic opening of the HPCI valve Corrective actions for this event included providing the lessons learned to appropriate operations personnel. The lessons learned focused on the potential effects associated with installation of jumpers and lifting of wire lugs in control circuit The inspectors performed an on-site review of this LER. The inspectors reviewed E General Procedure (GP)-25 Appendix 10, " Installation of Trips /Isolations to Satisfy TS Requirements for inoperable Instrumentation" and OM-P-7.7, Revision 0, " Boots and Jumpers" and noted that no changes had been made to either of these procedures since this event. The inspectors learned from Operations management that the lessons and corrective actions from this event were discussed in I operational training sessions given to all operations personnel. Based on this ;
information, the inspectors determined that no NRC requirements were violated during this event. The inspectors had no additional concerns with this issu .2 (Closed) Violation (VIO) 50-277/98-03-02 Failure to Submit LER for TS Non- l Compliance PECO failed to submit a Licensee Event Report (LER) within 30 days, as required by 4 10 CFR 50.73 for a deviation or condition prohibited by plant Technical Specifications (TS). The LER for the event was included in a subsequent LER and
.therefore no additional submittal of an LER was required.
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The inspectors reviewed the corrective actions for this issue. On July 17 and 22,.
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1998, PECO conducted counseling on the usage of NUREG-1022," Event Reporting Guidelines 10 CFR 50.72 and 50.73" for personnel who routinely determine reportability and draft LERs. The inspectors discussed the counseling sessions with PECO personnel and had no additional concerns with this violatio .3 (Closed) Unresolved item 50-277(2781/97-07-03 Unit 2 Cooldown Monitorina Followina the Reactor Scram on November 9.1997 PECO performed an engineering review to evaluate the cooldown data and the subsequent NRC questions regarding cooldown and heatup after the November 9, 1997 scram. This evaluation included temperature change of the bottom head after the scram, after recirculation pump start, and the possibility of stratification in the reactor vessel after the reactor scra The inspectors questioned if stratification existed within the reactor vessel for some period of time prior to recirculation pump start following the reactor scram on November 9,1997. The recorded heatup rate at the bottom head drain after the start of the first recirculation pump, although greater than 100 F/Hr., was withi expected values for the conditions at the time. The steam dome temperature changed +2 *F dunng that hour. The inspectors noted that operations personnel started the first recirculation pump with a 145 F difference between steam dome and bottom head drain which was the maximum temperature difference allowed by TS The licensee's analysis concluded that the rapid increase in temperature of the bottom head drain line represented a potential for contribution to the thermal fatigue usage of the reactor vessel bottom head. Also, engineering personnel determined that this event should be counted as a partial sudden start of a recirculation pump in a cold recirculation loop for the thermal cycle counting for the bottom head of the reactor pressure vessel. The thermal cycle account for this event will be included
. by performance of ST-J-080-940-2(3),Rev 5, " Reactor Pressure Vessel Transients Cycles Record." Also, the counting of this event as a partial cycle for fatigue usage reconciled the impact of this transient on the reactor vesse The inspectors concluded that the recorded heatup rate in the bottom head drain temperature after the start of the first recirculation pump was within expected values for the conditions at the time. Inclusion of this event as a partial thermal ,
cycle was considered appropriate, but resulted from questions raised by the NRC j inspectors. The inspectors had no further concerns with this issu .4 Institute of Nuclear Power Operations (INPO) Evaluation Review (71707)
i The inspectors reviewed the INPO assessment report for the evaluation performed at Peach Bottom in February 1998. The report was reviewed to determine if there were any safety issues which were previously unknown to the NRC. The inspectors noted that the report documented findings of similar programmatic problems to
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those previously identified by the NRC and the licensee, especially in the areas of Operations and Engineering.
i 11. Maintenance t
L M1 Conduct of Maintenance !
. M 1.1 General Observations l
l' NRC Inspection Procedures 62707 and 61726 were used in the inspection of plant t maintenance and surveillance activities. The inspectors observed and reviewed l
selected portions of the maintenance and test activities listed in Attachment 2.
I i t The work and testing performed during these activities was professional and thorough. Technicians were experienced and knowledgeable of their assigned i tasks. The work and testing procedures were present at the jobsite and actively used by the technicians and operators for activities observed. Good pre-job briefs were observed prior to the performance of the surveillances observed. Applicable procedures were present in the control room and at the job sites during surveillance '
testing and were appropriately used. Some exceptions were observed and are described below and in Section M1.2.
l During observations of the 2A core spray pump maintenance, the inspectors observed that a quality verification (OV) inspector recorded a foreign material check i of the pump : casing internals as unsatisfactory. The QV check was performed after !
a worker ve'ification of cleanliness. The QV inspector found some small debris and paint chips, which were subsequently removed by the worker .
t in later discussions, the maintenance manager stated that maintenance supervisors were taking steps to improve worker sensitivity to foreign material exclusion. In addition, the inspectors noted that the maintenance manager initiated an investigation into a trend of less than adequate worker verifications during i l maintenance, including welding verification steps and cleanliness inspection '
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l M1.2 Hiah Pressure Coolant Iniection System Pioina Pressure Test Insoection (Unit 21 Insoection Scope (61726. 62707)
The inspectors observed portions of a high pressure coolant injection (HPCI) system piping pressure test from the control room and HPCI roo *
l Observations and Findinas On July 21, operators commenced ST-O-023-611-2,"HPCI System Piping Pressure !
Test inspection" after a pre-activity brief of operations and maintenance personne The HPCI system was declared inoperable at the beginning of the test and remained so until properly returned to service on completion of the test. The licensee verified the integrity of the HPCI piping as satisfactory with no leaks.
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During performance of the test, engineering personnel determined that the some of the piping required to have insulation removed was still insulated. This information was passed to maintenance personnel for immediate action. The inspectors also questioned the engineer about the insulation remaining on the drain line from the HPCI turbine steam admission (MO 14) valve. The engineer clarified the scope of the insulation removal to include the NRC-identified drain lin Work Order RO755303 activity 4 described the insulation to be removed. Although the job scope description required the removal of the appropriate insulation, it gave I in some cases, a very general description of the insulation to be removed, which caused confusion for the maintenance personnel. The insulation removal was not verified to be correct before start of the test. The test was halted and resumed after the insulation was properly remove During the HPCI system test, all four unit 2 high pressure service water (HPSW)
pumps and the residual heat removal system were operated to maintain the torus temperature in the required band. Only two HPSW pumps are operated during a design basis event. The heat load of all four HPSW pumps caused the indicated HPSW room temperature to exceed the alarm setpoint of 115*F. The highest l indicated temperature in the room was 120*F, which engineering personnel I determined would not cause damage to the HPSW pump or other e ;uipment in the l room. The inspectors noted that operations personnel did not monitor the HPSW !
room temperature during the test, until after the room high temperature alarm was !
received even though all the pumps were running and the outside air temperature I was high during the HPCI test. Subsequently, the room temperature was monitored l through the remainder of the HPCI test and torus cooling was maintained using only two HPSW pump I 1 Conclusions PECO properly completed a pressure test that verified the integrity of the unit 2 high pressure coolant injection piping. The system was adequately returned to operable j status. However, the irtsulation removal from high pressure coolant injection piping l and components was poorly controlled and executed during this test. Also, operations personnel did not monitor the high pressure service water pump room temperature during the test, until after the room high temperature alarm was receive M8 Miscellaneous Maintenance Activities M8.1 (Closed) IFl 50-277(278)/97-04-01 Review Maintenance Rule Proaram Acolication of 13 KV Breaker Switche After a manual power transfer from the main generator to the offsite power supply on March 9,1997, an auxiliary breaker position contact in a 13 KV breaker failed to change positio .
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The inspectors reviewed this issue and determined that initially PECO considered the 13 kV breaker a functional failure (FF). The failure had not been considered a ,
maintenance preventable functional failure (MPFF) based on vendor instructions not l
requiring any preventative maintenance on the auxiliary contact and that no l
previous failures of the contact had occurred. That determination was subsequently I updated to a MPFF based on industry operating experience documentatio The 13 kV system, although within the scope of the maintenance rule program, was considered by PECO to be a low risk system. Therefore the expert panel established by the maintenance rule program was not required to review nor did the panel consider it prudent to evaluate the 13 kV system for a change in status based on a single MPFF. The inspector's review of the maintenance rule performance criteria for the 13 kV system determined that no performance criteria had been exceede The inspectors determined that PECO took generally effective actions to address the 13 kV auxiliary switch failure and initiated proper followup actions. The application of the maintenance rule to this issue was appropriate and was within the requirements of the licensee's program. No further issues were identified by the inspector l lll. Enaineerina E1 Conduct of Engineering E1.1 Indeoendent Spent Fuel Storaae Installation Geotechnical Review a. Insoection Scoce (60851)
This inspection focused on the geotechnical investigation for the independent spent fuel storage installation (ISFSI) The inspectors reviewed the pad design and report No. SDOC-NE-266-1, Revision 0, entitled " Report for Geotechnical Investigation for j Independent Spent Fuel Storage Installation." This report summarized the results of the subsurface investigation to determine the actual soil condition for the ISFSI, including storage pad, retaining walls, anc ccess road including the proposed bridge over Rock Run Creek. The inspechs evaluated the results and recommendations of the geotechnical investigation and its use in the design of the pad and retaining walls. Also, the inspectors reviewed the 10 CFR 50.59 evaluation for the pad / wall installation, and the preliminary 10 CFR 72.212 analysi b. Observations and Findinas The inspectors determined that the gaotechnical report was acceptable. The inspectors noted that some of the cores were not drilled all the way to the sound bedrock due to a core refusal (e.g., drill can not go any further). This approach was acceptable to the American Society for Testing and Materials (ASTM) D 1566-84 Reapproved 1992. The inspectors verified that the soil borings showed that ground
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, water would not be a concern for the pad and that ground water had been properly l addressed in the design of the bridg The inspectors walked down the physicallayout and evaluated the location of the borings as described in the geotechnical report. The inspectors verified that the 19 borings had been satisfactorily locate i The inspectors noted an inconsistency between the 10 CFR 50.59 evaluation and l l the preliminary 10 CFR 72.212 analyses. The 10 CFR 50.59 evaluation defined the
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design basis for the east retaining wall as erosion mitigation of the bank, while the preliminary 10 CFR 72.212 analysis defined the design basis as flood mitigation l during maximum probable flood condition. PECO stated that these documents were
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not in the final approval status because the TN-68 Safety Analysis Report (for the I
storage cask general license) was under review by the NRC. The inspectors noted i that the wall design appeared to be capable of meeting both design objective PECO stated that this inconsistency would be resolved in the final analyse The inspectors noted that PECO has a quality assurance (QA) program in place for the ISFSI and that an early QA action had been to hold the release for the first j shipment of rebar until acceptable material traceability could be establishe : Conclusiord The independent spent fuel storage installation (ISFSI) geotechnical report and associated regulatory analyses were determined to be acceptable. The results and recommendations of this geotechnical report had been appropriately used in the design of the storage pad and provided a satisfactory basis for planned ISFSI construction activitie E2.1 (Closed) LER 98-003 Missed Surveillance Reauirement for Main Steam Isolation Valve Stroke Timino (Unit 3)
On July 10,1998, engineering personnel identified that technical specification surveillance requirement 3.6.1.3.9 was missed during the 3J12 maintenance outage in March 1998.
, The inspectors reviewed the causes of the missed surveillance requirement with the
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IST engineer, the surveillance test coordinator, and other station personnel. The inspectors also attended a plant operations review committee (PORC) meeting at which the investigation and the draft Licensee Event Report (LER) were discusse The inspectors also completed an in-plant review of the final LE Inspectors reviewed the completed and planned corrective actions for this event and considered them adequate. Engineering personnel corrected the surveillance test i
frequency and reviewed operating history to verify that MSIV testing was not missed on Unit 2. Planned corrective actions included providing training to
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communications and procedure reviews. Station personnel also stated that they will evaluate the need for improving guidance for procedure revision The failure to perform surveillance requirement 3.6.1.3.9 while in cold shutdown during the 3J12 outage in March 1998 was a violation of technical specification l This non-repetitive, licensee-identified and corrected violation is being treated as a 1 Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement 1 Policy. (NCV 50-278/98 07-02)
The inspectors concluded that the failure to perform en inservice testing surveillance requirement for main steam isolation valve stroke timing during cold shutdown conditions revealed performance weaknesses among engineering personnel in the procedure revision and review processes. Written communications for, and reviews of, a surveillance test procedure revision were poor and failed to identify an error in the test frequenc E4 Engineering Staff Knowledge and Performance l E (Closed) LER 98-002 Failure to Meet TS Reauirements for the Unit 3 RCIC While the Mechanical Oversneed Trio Taonet Was Not Fully Reset Insoection Scone (37551)
The inspectors reviewed documentation associated with the discovery, repair, and testing of the RCIC mechanical overspeed trip tappet that was found not fully rese The inspectors discussed this issue with operations and engineering personne Observations and Findinas ;
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On June 22,1998, a reactor building equipment operator discovered during routine j operator rounds that the Unit 3 RCIC mechanical overspeed trip tappet was not fully l reset. Station personnel determined that the RCIC system had been inoperable l since May 4,1998 which was the last time the overspeed trip function was i manipulated and successfully tested. Technical specification (TS) 3.5.3 required that if RCIC was inoperable, then RCIC must be restored to an operable condition in 14 days. The failure to restore RCIC to an operable status or be in Hot Shutdown within 14 days was a violation of TS 3.5.3. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV), consistent l
with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-278/98-07-03)
The inspectors reviewed the licensee's corrective actions and found them acceptable. As part of the on-site review of this LER, the inspectors noted that engineering personnel used administrative procedure, AC-CG-50, Revision 0,
" Equipment Investigation and Troubleshooting Guideline" during troubleshooting i activities. Several problems were identified with the overspeed trip mechanism i during these activities. The inspectors also noted that engineering personnel
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investigated all of the causes of the failure of the trip tappet to fully reset and I adequately diagnosed and repaired each of these problems. Based on these i _ . . , _ _ _. . . _ ,
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observations, the inspectors determined that troubleshooting activities were thorough and comprehensive. The inspectors reviewed the mechanical and overspeed trip testing and determined that the trip tappet was adequately tested l following troubleshooting and repairs.
l Conclusiqn_g l
On June 22,1998, a reactor building equipment operator discovered during routine operator rounds that the Unit 3 reactor core isolation cooling system mechanical overspeed trip tappet was not fully reset. Station personnel determined that the reactor core isolation cooling system had been inoperable since May 4,1998 which l was the last time the overspeed trip function was manipulated and successfully tested. This condition resulted in a violation of technical specification 3.5.3 since the reactor core isolation cooling system was inoperable for greater than 14 days while Unit 3 was operating. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement Policy.
t Engineering personnel fully investigated all of the causes of the failure of the trip l l
tappet to reset and adequately diagnosed and repaired each of the problems identified.
l E8 Miscellaneous Engineedng issues E8.1 (Closed) Insoection Followuo item (IFI) 50-277(278)/97-02-03 Station Blackout (SBO) Line Load Testina During the week of April 28,1997, the inspectors reviewed the testing conducted on the SBO line. The inspectors questioned whether the SBO line test acceptance
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criteria of greater than 7000 KW was adequate during testing. An engineering L calculation at the time determined that 7420 kW was required through the SBO line ]
to adequately mest the load requirements to shutdown both units during a blackou l Following questioning by the inspectors, engineering personnel reanalyzed the load calculation and determined that 6940 kW was sufficient to meet the load requirements to shutdown both units during a blackou The inspectors reviewed the new calculation and the Peach Bottom SBO requirements and commitments. The inspectors also reviewed station special event (SE)-11 procedures, including SE-11.1, Revision 0," Operating Station Blackout Line During a Loss of Offsite Power Event."
l The inspectors wm concerned with the lack of engineering rigor in the initial evaluation performed by the system manager to address the concerns raised dealing I with the differences between the SBO line load testing and the required loads calculation. However, the inspectors had no concerns with the reanalysis of the engineering calculation and determined that the SBO line load testing met the plant shutdown load requirements. Also, no discrepancies existed between the SBO
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loading information submitted to the NRC and the plant shutdown load requirements l'
noted in the revised engineering calculatio E8.2 LClosed) P2193-22 and 94-83 Emeraency Diesel Generator Heat Exchanaer Crossflow 10 CFR Part 21 Reports Insoection Scope (92903)
The inspectors reviewed the status of licensee actions for a 10 CFR Part 21 Report j on the potential for crossflow conditions in the emergency diesel generator (EDG)
jacket water and air cooler heat exchangers. This issue was being tracked by two inspection Follow-up System Part 21 items,93-022 and 94-083, for Units 3 and 2, respectivel Observations and Findinas l
The inspectors reviewed a 10 CFR Part 21 notification issued on September 16, i 1991, by EDG manufacturer Fairbanks Morse, which identified potential concerns with the EDGs at Arkansas Nuclear One (ANO) Unit 2. Specifically, the notice stated that the EDGs may not be able to carry the required load due to air cooler and jacket water heat exchanger limitations. The inspectors noted that PECO received a letter from Fairbanks Muse on July 8,1992, which provided additional information on the potential for crossflow conditions, which could impact heat exchanger performance at Peach Botta Engineering personnel initiated action requests that documented initial evaluations of the applicability of this issue at Peach Bottom and the plans for detailed heat exchanger testing. The inspectors noted the following:
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- Engineering personnel determined that the potential significance of any crossflow conditions would be much less at Peach Bottom than at ANO due to a lower design basis emergency service water temperature, j l
- Maintenance personnel used a thermography camera in an attempt to detect ,
crossflow conditions during routine testing, but the results were int onclusive.
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- Engineering personnel made plans for final resolution of the crossflow issue through comprehensive heat exchanger performance testing.
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system manager and learn 9d that the testing was deferred until the completion of an EDG modification and the completion of testing at Limerick. Currently the heat exchanger performance testing at Peach Bottom is scheduled for completion in March 199 The inspectors concluded that both the interim actions taken to evaluate the crossflow conditions and the long-term plans for comprehensive heat exchanger
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performance testing were adequate.' Inspection Follow-up System Part 21 items '93-022 and 94-083 are closed. The results of the heat exchanger performance
. testing will be reviewed through routine resident inspection activitie E8.3 IClosed) IFl 50-277(278)/97-07-07Emeraency Diesel Generator Lube Oil Pioina - I Potential 10 CFR 21 Issue Insoection Scooe (37551. 92903)
-The inspectors reviewed licensee actions following the identification of cracked welds on emergency diesel generator (EDG) lube oil piping at other nuclear power l station Observations and Findinas in September 1997, Coltec Industries - Fairbanks Morse Engine Division submitted a ;
potential 10 CFR 21 letter to the NRC regarding failed welds on EDG lube oil piping l at Millstone Unit 2. This letter listed Peach Bottom as being affected by a potential quality issue with associated piping welds. The affected welds were performed during original constructio Since September 1997, PECO has removed sections of tube oil piping from two EDGs in order to analyze them for the weld quality issues identified in the Coltec letter. All of the analyzed welds were found to be satisfactory. Engineering l personnel were continuing to follow the issue and were awaiting addit!onal i information from Colte ,
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The inspectors discussed the completed and planned actions with the EDG system manager. The inspectors concluded that these actions provided reasonable
- assurance that similar weld quality issues were not present on Peach Bottom EDG E8.4 (Closed) IFl 50-277(2781/97-06-04 Diesel Driven Fire Pomo Batterv Explosion l
l Inspectior Scoce (92902 & 92903)
The inspectors reviewed the finalinvestigation and corrective actions for an explosion of the diesel driven fire pump battery that occurred in August 199 Observations and Findinas On August 14,1997, during . veillance testing, the diesel driven fire pump starting battery exploded shortly fter the start of the pump. There was no personnel injury or significant impact on the plant or other equipmen Industrial Risk Management (IRM) personnel completed a Performance Enhancement i Program (PEP) full root cause investigation for this event. They found that the explosion occurred due to the ignition of hydrogen within the battery. Contributing to this occurrence were degraded battery cables and a severely aged battery. The i
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19 PEP report also noted that predictive maintenance personnel had identified uneven battery electrolyte heating, and a separate action request had identified higher than normal current on the battery charger. The PEP report noted that the root cause for this event was lack of maintenance on the battery. The battery was not in a l preventive maintenance program and several corrective maintenance action requests had not been promptly worke The inspectors reviewed the PEP report and discusse.d the issue with IRM personnel. The inspectors noted that the investigation adequately addressed the causes of the event. Corrective actions for the event appropriately included acceptance criteria for battery charging current and the development of preventive maintenance for battery replacement. However, the inspectors noted that the PEP did not explore why equipment operators had not noted the degraded condition of the battery and cables during round Inspection of the diesel driven fire pump room, which was normally luu, o was the responsibility of the equipment operators. The inspectors concluded that the equipment operators missed opportunities to identify fully the degraded condition of the diesel driven fire pump battery and cables during their rounds. The inspectors were not concerned with undetected degradation of other station batteries since they are in unlocked rcoms and were routinely inspected by other personnel including the resident inspector In addition, the inspectors determined that work management and maintenance personnel did not recognize the potential hazard associated with the combination of uneven electrolyte heating and a high charging current. Therefore, the priority for battery and cable replacement was not raise The inspectors identified no violations of NRC requirements and had no further concerns with this ite IV. Plant SuDDort i R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Radioloaical Controls (Proaram Chanaes) inspection Scope (8375_0)Q The inspectors reviewed selected radiological contiols program changes. Areas reviewed included organization and effing, facilities and equipment, and procedure changes. The inspectors evaluated , erformance in this area through interviews of cognizant station personnel, re@uv of documents, and observations during station tour . . -. . . .- - - -. _ _ . - -
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20 . Observations and Findinas PECO hired a new radiation protection manager (RPM)in early 1998. The RPM was in a PECO developed training program to familiarize the individual with current industry radiological control practices, regulatory requirements and :
l recommendations, and reactor operations. An appropriately qualified individual was I acting as RPM pending completion of the new RPM's training program. Two new supervisors were also selected in radiation protection operations. The individuals met applicable Technical Specification qualification requirement PECO changed its vendor for personnel dosimetry services to improve performance in this area. The new whole body dosimetry met the requirements of 10 CFR 20.150 Conclusions
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No program changes were identified that reduced the effectiveness of the radiological controls program. PECO took effective actions in improving personnel dosimetry equipment and performance. No safety concerns or violations were identifie R1.2 Internal and External Exoosure Controls Insoectlon Scope (83750)
The inspectors selectively examined the internal and external exposure control programs. The inspectors reviewed the radiological controls performance associated with the October 1997, Unit 3 refueling outage, calendar year 1997, and for calendar year 1998 up to July 1998. The following aspects were reviewed:
airborne radioactivity sample results, whole body count results, personnel contamination reports, radioactive materialint'ake assessments, and calculations of effective doses. Additionally, the effectiveness of airborne radioactivo material ,
controls, including use of respiratory protective equipment was examine The inspectors reviewed records and discussed the program with cognizant personnel. The inspectors also observed exposure control practices during tours of the RCA and observation of work activities. The inspectors reviewed high radiation area controls and general radiological posting, implementation of the radiation work permit program, and implementation of the dosimetry progra Observations and Findinas There were no significant recorded internal exposures identified during the time period reviewed including the 1997 Unit 3 refueling outage. Overall, PECO effectively controlled airborne radioactivity, conducted appropriate bioassays when personnel contamination monitors indicated potential internal contamination, and performed appropriate internal dose calculations and calculated DAC-hours, as necessary, for indications of intake of radioactive materia .
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PECO continued to implement and maintain effective real time personnel exposure control by use of an electronic dosimetry / access control system. There were no unplanned deep dose equivalent exposures of workers during the period reviewed and electronic dosimeters results compared favorably with vendor processed thermoluminescent dosimeters. Anomalous TLD/ electronic dosimeter results were evaluated, as appropriate. PECO readily detected apparent processing errors in its TLDs and corrected doses and results accordingly. There were no accumulated individual exposures greater than 3 rem in 1997 and six individuals sustained aggregate exposure between 2 and 3 re PECO provided workers with neutron monitoring, in accordance with guidance in NRC Regulator / Guide 8.14, whenever personnel were required to work in neutron radiation areas. No significant personnel exposures to neutrons were evident during the period reviewe Radiological areas (e.g., high radiation areas, radiation areas) were properly posted and locked (as appropriate).
With the exception of one worker observed to not be adhering to posted protective clothing use requirements (discussed in Section R1.3), workers were implementing good radiological work practices including radiation work permit requirement Conclusions PECO implemented effective internal and external exposure control programs with respect to personnel monitoring and dose assessment, personnel dosimetry use and application, and radiation and high radiation area monitoring and control. No significant unplanned exposures were evident during the period under review. No violations or safety concerns were note R1.3 Control of Radioactive Materials and Contamination Inspection Scone (83750)
The inspectors selectively reviewed radioactive material and contamination control practices including calibration and performance checks of survey and monitoring instruments and the use of personal contamination monitors and friskers. The inspectors also evaluated personnel skin contaminations and skin dose assessment rnethodolog The inspectors reviewed the current in-plant radiological source term and specific radionuclides present within the station and discussed contamination monitoring practices, Observation,3 and Findinas
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PECO implemented generally effective contamination control work techniques and l prompt correction and cleanup of contamination. At the time of this inspection, the
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station exhibited approximately 4% of accessible floor areas as contamination areas !
(excluding the drywall). PECO tracked and tended personnel contaminations for '
programmatic corrective action purposes, identified increases in low level contaminations of personnel in late 1997 and early 1998, and initiated root cause evaluations and corrective actions. PECO developed and implemented corrective actions to reduce the instances of low level contaminations including enhanced cleaning of the radiological controlled portions of the station, and establishment of corrective measures for deficient work and worker practices which contributed to the increasing trend in personnel contamination PECO performed a comprehensive evaluation of plant contamination conditions in early 1998 and developed and implemented recommendations to reduce personnel contamination events. PECO implemented appropriate dose assessments for personnel contaminations and calcul . id shallow dose to the skin (SDE), as appropriate. Skin contamination resulted in generally low skin dose (well within i applicable NRC limits).
PECO continued to implement room-specific control of those areas of the station which exhibited electron capture decay nuclides (e.g., Zinc-65) to provide enhanced monitoring of material removed from these rooms and work performed therei PECO was also evaluating new state-of-the-art equipment for monitoring of such radionuclides. PECO was also controlling injection of hydrogen into the reactor primary systems to preclude any adverse effects on survey and monitoring practices due to increase in general area radiation field PECO conducted a work stand-down May 1998, to convey management expectations with respect to adherence to station procedures, including radiation safety procedures. Such action was taken due to a licensee-identified trend involving procedural compliance. Notwithstanding, on July 13,1998, the inspectors identified an individual, within a posted contamination area in the Unit 2 reactor building (established on a tractor trailer bed), who was not wearing all the protective clothing prescribed by clearly posted instructions at the entrance to the area. The individual did not wear the top portion of his personnel protective clothing (scrubs) but rather wore a personal tee-shirt. The worker was not in compliance with requirements stated in procedure HP-C-310, Revision 3, Section 5.6, which required that all plant personnel comply with established posting Upon notification, a radiation protection supervisor requested the individual to leave the area. Subsequently, the individual was surveyed for contamination and counseled with respect to complying with station procedures. The observed performance deficiency was entered into the corrective action tracking system, the individual's access to the radiologically controlled area was suspended pending
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retraining in general employee skills and expectations. The observation was l discussed at station management meetings and generic corrective actions are being l considered. Station radiation protection personnel were informed to monitor use of
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protective clothing in the RCA. The worker was not contaminated and overall
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l l A similar issue was identified by the inspectors on August 7,1998. In this l l instance, an individual was working in a posted contamination area with proper j protective clothing, but the sleeves of the scrub top were partially rolled up. This !
l was contrary to station radiation protection practices. Immediate corrective actions were accomplishe The above failures constituted a violation of minor significance and is not subject to formal enforcement action, Conclusions PECO implemented a generally effective contamination control program with respect to tracking and trending personnel contamination events, and initiating corrective measures to improve personnel work performance and limit the potential for inadvertent contact with contaminated areas and materials. Program evaluations and enhancements were underway to improve contamination control and work i practices and reduce instances of minor personnel contamination R1.4 ALARA Prooram and Unit 2 Refuelino Outaae Plannina. Preoaration Emeroent Work Control Insoection Scope (83750) '
The inspectors selectively reviewed various ALARA program elements and reviewed the planning and preparation for the Unit 2 refueling outage, including control and review of emergent work. The inspectors reviewed records, discussed outage planning, and observed activities to verify necessary planning, preparations, and l management support for the implementation of radiological controls. The inspectors reviewed lessons learr:ed from previous outages to determine if they were incorporated into planning and preparations for future outages. In addition, the inspectors reviewed the licensee's conformance with self-imposed ALARA occupational exposure goals for 1997 including the 1997 Unit 3 outag Observations and Findinas PECO implemented numerous initiatives and standard practices to reduce overall l occupational radiation exposure. These included such items as shielding, remote l reading teledosimetry, cameras, and selection and use of experienced personne PECO had also implemented chemistry initiatives such as injection of " depleted zinc" in October 1996 to reduce drywall radiation dose rates. PECO's post outage report identified the initiatives and also identified self-assessment items and areas
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for improvemen PECO exceeded its 1997 Unit 3 outage exposure goal (282 person-rem versus 225 person-rem based on electronic dosimeter). The outage report discussed the l
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exposure reduction for the outage. PECO concluded that the goal was exceeded l primarily due to an increase in work hours to complete a modification within the
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torus and an increase in valve repairs due to failed leak checks, within the drywal i l
PECO was actively performing reviews of proposed work activities for the October I 1998 Unit 2 refueling outage. Work scope additions were reviewed by station management to ensure the appropriateness and necessity of accommodating additional work that emerged since the original work scope was determined. PECO implemented enhanced review of radiological work to improve evaluation and I coordination of work between the health physics planner, work supervisor, health physics supervisor and cognizant radiological engineer )
The stations's ALARA committee was actively reviewing and evaluating the station's occupational exposure accumulation and considering dose reduction initiative PECO has initiated efforts to more closely track, evaluate, and control the exposures of its workers who move between its various stations as a mobile work force. The effort is to ensure that the work patterns of those workers approaching administrative limits, do not result in the worker inadvertently exceeding the administrative limit Overall, records indicated that workers, who routinely worked in radiation and contaminated areas (e.g., radwaste technicians and radwaste engineers) over the past 15 years, generally exhibited declines in their accrued personnel exposur Selective review of records for these individuals did not indicate any significant external or internal exposures. Where anomalous exposure information was noted (e.g., TLD and electronic or pocket dosimeter differences) records reflected conservative evaluations and dose assessments and no exposures in excess of regulatory limit Conclusions PECO continued to implement an overall effective ALARA program with respect to work planning and control, use of dose reduction initiatives such as remote monitoring equipment, application of shielding, and work monitoring via closed-circuit television. Outage work planning and control efforts to efhet improved ALARA performance are continuin R7 Quality Assurance in Radiological Protection and Chemistry Activities R 7.1 Radioloaical Self-Assessment 1 Inspection Scoce (83750)
The inspectors selectively reviewed oversight activities for radiological controls. In pprticular, the inspectors reviewed PECO's evaluations and actions associated with
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self-identified issues and concerns documented in its se*t identification programs i
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(e.g., personnel contamination reports, radiological occurrence reports, performance enhancement issues, quality assurs .a audits and surveillances, and industry audits). The inspectors reviewed so., aed licensee self-identified issues covering calendar year 1997 and 1998.
I Observations and Findinas PECO implemented generally good audit and self-assessment activities for its radiological controls program. The 1997 audit of the radiation protection program was of good depth and scope to identify problems in the areas audite Appropriately qualified personnel were used as audit team members. An audit program was established and implemented in accordance with 10 CFR 20.110 PECO took action on identified issues. PECO implemented generally timely corrective action on self-identified concern l PECO initiated a comprehensive self-evaluation of its radiation protection program l processes to improve its efforts to improve performance. Color coded charts were used to identify areas of strength, satisfactory performance, and areas for improvement and weaknesse Conclusions PECO implemented an effective program for self-identifying and correcting self-identified issues and concerns with respect to initiation of comprehensive self-evaluations, usually timely corrective actions for self-identified issues, and the conduct of audits with sufficient scope and depth. No violations or safety concerns were identifie R8 Miscellaneous RP&C Activities l
R8.1 iniection of Noble Metals Insoection Scope (83750)
The inspectors selectively reviewed the radiological implications associated with PECO's plans to inject noble metals into the reactor coolant system. The purpose i
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of the injection of the metals in to the reactor coolant system was to reduce certain types of corrosio The inspectors discussed the plans with cognizant licensee staff including its ;
implications for the stations radiological controls program, b. Observations and Findinas PECO was in the process of developing a 10 CFR 50.59 safety evaluation to support injection of the noble metals. PECO had reviewed current industry
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experience on this matter and concluded that no radiological controls impact was associate with the injection of the metal PECO was evaluating potential impets on certain types of reactor coolant and containment atmosphere sampling systems, Conclusions PECO was appropriately reviewing the potential impact of injection of noble metals into the reactor coolant system at Unit 2. No apparent radiologicalimpact was note R8.2 Housakeeoina (71750)
The inspectors toured the facility and noted overall very good plant conditions including areas outside the station. Areas were generally neat and no leaking equipment was noted. PECO was actively cleaning and painting the facility and had taken action to reduce nuisance personnel contamination V. Manaaement Meetinas X1 Exit Meeting Summary .
The inspectors presented the results of the inspection to members of licensee management on August 14,1998. The licensee acknowledged the findings presented. No proprietary information was identified by the license X2 Review of Updated Final Safety Analysis Report (UFSAR) Commitments A discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practices, procedures and/or parameters.
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ATTACHMENT 1
- LIST OF ACRONYMS USED AO abnormal operating AR action request !
ASTM American Society for Testing & Materials l
ALARA as low as reasonably achievable ;
CAD containment atmosphere dilution '
DAC derived air concentration DBD design basis document DDE deep dose equivalent . ,
ECT emergency cooling tower i ECCS emergency core cooling system EDG emergency diesel generator FME foreign material exclusion FF functional failure GP' general procedure HCU hydraulic control unit HYDRO hydrostatic '
HP health physics HPCI high pressure coolant injection HPSW high pressure service water ISFSI independent spent fuel storage installation l lRM industrial risk management ISI inservice inspector IFl inspector followup item I&C instrumentation & controls LCO limiting condition for operation LER- licensee event report LOCA loss of coolant accident MCR main control room MSIV main steam isolation valve MPFF maintenance preventable functional failure l MOV . motor operated valve NCV non-cited violation NCR non-conformance report NOTICE notice of violation NVLAP national voluntary laboratory accreditation program NRB nuclear review board OSC Operations Support Facility PA protected area PECO Peco Energy PEP performance enhancement program PORC Plant Operation Review Committee PDR public document room QV quality verification
- RP radiation protection
[ RPM ' radiation protection manager
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Attachment l
- RCA'- . radiological controlled area -
.RP& radiological protection and chemistry {
i RPM radiation protection manager .
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RW radiation work permit RCICL J reactor core isolation cooling -
RFPT' . reactor feed pump turbine
' RWCU - reactor water cleanup .
RP reactor protection syste l
~RPV reactor pressure vessel '
.SDE : shallow dose equivalent
' SO . . system operating SSPV scram solenoid pilot valve SRV safety relief valve SR surveillance requirement ST surveillance test
?TLD thermoluminescent dosimeter TS ' technical specification TSA technical specification action
' URI
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unresolved item UFSAR . updated final safety analysis report
' VOTES valve operation test and evaluation system l
INSPECTION PROCEDURES USED )
IP 37551 Onsite Engineering Observations I IP 60851 - Design Control of IFSFI Components 1 IP 61726 - Surveillance Observations ,
. IP 62707 Maintenance Observations .j iP 71707 - Plant Operations i IP 71750 : Plant Support Observations IP 83750 Occupational Radiation Exposure i
. IP 92901 Followup - Operations IP 9290 Followup - Maintenance IP 92903 Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
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- 50-277/98-07-01 NCV ' Failure to Meet Technical Specification Requirements for L One Inoperable Off-site Power Source
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50-278/98-07-01 NCV Failure to Meet Technical Specification Requirements for
, One Inoperable Off-site Power Source i
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50-278/98-07-02 NCV Missed Surveillance Requirement for Main Steam !
Isolation Valve Stroke Timing !
-50-278/98-07-03 NCV Failure to Meet Technical Specification Requirements for !
l Reactor Core isolation Cooling While the Mechanical ,
Overspeed Trip Tappet Was Not Fully Reset l
l 50-277/2-98-04 'LER Failure to Meet Technical Specification Requirements for l One Inoperable Off-site Power Source
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50-278/2-98-04 LER Failure to Meet Technical Specification Requirements for One Inoperable Off-site Power Source
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50-277/2-97-03 LER High Pressure Coolant injection inoperable (Unit 2)
50-277/2-97-07 LER Potential for Bypass of Pressure Suppression Pool 50-278/2-97-07 LER Potential for Bypass of Pressure Suppression Pool 50-278/3-98-03 LER . Missed Surveillance Requirement for Main Steam
! Isolation Valve Stroke Timing 50-278/3-98-02 LER Failure to Meet Technical Specification Requirements for Reactor Core Isolation Cooling While the Mechanical !
Overspeed Trip Tappet Was Not Fully Reset l l 50-277/97-07-03 URI Unit 2 Cooldown Monitoring Following the November 9 l Reactor Scram 50-278/97-07-03 URI Unit 2 Cooldown Monitoring Following the November 9 Reactor Scram 50-277/98-07-01 NCV Failure to Meet Technical Specification Requirements for One Inoperable Off-site Power Source 50-278/98-07-01 NCV Failure to Meet Technical Specification Requirements for !
l One Inoperable Off-site Power Source 50-278/98-07-02 NCV Missed Surveillance Requirement for Main Steam Isolation Valve Stroke Timing 50-278/98-07-03 NCV Failure to Meet Technical Specification Requirements for Reactor Core Isolation Cooling While the Mechanical i Overspeed Trip Tappet Was Not Fully Reset 50 277/98-03-02 VIO Failure to Submit LER for TS Non-Compliance 50-277/97-02-03 IFl Station Blackout Line Load Testing 50-278/97-02-03 IFl Station Blackout Line Load Testing 50-277/97-06-04 IFl Diesel Driven Fire Pump Battery Explosion 50-278/97-06-04 IFl Diesel Driven Fire Pump Battery Explosion 50-277/97-04-01 IFl Review Maintenance Rule Program Application of 13 KV Breaker Switches 50-278/97-04-01 IFl Review Maintenance Rule Program Application of 13 KV Breaker Switches 50-277/97-07-07 IFl Emergency Diese! Generator Lube Oil Piping-Potential
, 10 CFR 21 issue
- 50-278/97-07-07 IFl Emergency Diesel Generator Lube Oil Piping-Potential
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10 CFR 21 issue
- - 50-278/93-22 P21 Emergency Diesel Generator Heat Exchanger Crossflow l 10 CFR Part 21 Report
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-50-277/94-83 P21 Emergency Diesel Generator Heat Exchanger Crossflow 10 CFR Part 21 Report
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ATTACHMENT 2 !
I Maintenance Observations: Observed On:
M-056-001 . Testing and Control of 600 Volt Class - July 21,1998 Molded Case Circuit Breakers and Setpoints M-C-700-232 480 Volt Motor Control Center Circuit July 21,1998 Breaker Assembly and Cubicle Terminal Maintenance M1145903 Unit 3 high pressure coolant injection July 7,1998 system valve AO-3-23-53 solenoid replacement C0182895 Unit 3 rod select push button July 8,1998 !
replacement I C0181694 Floor boring for conduit entry July 13,1998 C0182933- . Main control room emergency ventilation July 14,1998 !
filter train replacement !
M1159282 2B core spray loop flow detector July 15,1998 calibration M1147807 Standby gas treatment system solenoid July 6,1998 valve SV-0-368-00011 replacement C0182966 2A core spray pump impeller July 28,1998 replacement C0183147 Unit 3 digital feedwater control computer July 31,1998 Replace digital output board M1152386 Unit 2 scram discharge volume inboard July 29,1998 vent valve corrective maintenance Surveillance Obsorvations: Qhserved On: l ST-M-030-475-2 Sluice Gate Functional and Remote June 29,1998 l Position Indication Verification Test i
ST-O-023-301 -3 HPCI Pump, Valve, Flow and Unit July 21,1998 l Cooler Functional in-Service Test l ST-O-023-611 -2 HPCI System Piping Pressure Test July 21,1998 ,
inspection ST-O-052-701-2 E1 Diesel Generator 24 Hour Endurance July 24,1998 l Test ST-O-052-703-2 E3 Diesel Generator 24 Hour Endurance August 5,1998 Test j ST-M-037-311-3 Detailed VisualInspection of Penetration July 13,1998 Seals and Difficult to View Fire Barriers ST-O-14-301-2 2A Pump, Valve, Flow and Cooler July 30,1998 Functional and in-Service Test e ST-O-14-301 -3 3A Pump, Valve, Flow and Cooler Functional and in-Service Test July 14,1998
- RT-O-052-202-2 E2 Diesel Generator Load Run July 15,1998
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