IR 05000293/1998005
| ML20236N925 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/09/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236N923 | List: |
| References | |
| 50-293-98-05, 50-293-98-5, NUDOCS 9807160063 | |
| Download: ML20236N925 (23) | |
Text
__
__ __ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - __ _ -__ -
. .
Enclosure U.S. NUCLEAR REGULATORY COMMISSION
REGION I
License No.: DPR-35 l
l Report No.: 98-05 Docket No.: 50-293 Licensee: BEC Energy 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Inspe.ction Period: April 21,1998 through June 8,1998 Inspectors: R. Laura, Senior Resident inspector R. Arrighi, Resident inspector R. Summers, Project Engineer S. Dennis, Operations Engineer R. Ragland, Radiation Specialist Approved by: Curtis J. Cowgill, Ill, Chief Reactor Projects Branch No. 5 Division of Reactor Projects phbDO O
[
G
.
. .
__ ____ _ _ - _ - _ _ _ _ - _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ - - - - - - - - . - - - - - _ - _ - - - - - - - - - -
. .
EXECUTIVE SUMMARY Pilgrim Nuclear Power Station NRC Inspection Report 50-293/98-05 This integrated inspection included aspects of licensee operations, engineering,
,
maintenance, and plant support. The report covers resident inspection for the period of l April 21,1998, through June 8,1998;in addition, it includes the results of announced inspections by a regional radiation specialist and by an operations enginee Operations
The conduct of operations was professional and safety-conscious. A May 28, 1998, plant down power was well controlled and executed. (Section 01.1)
The recently revised process for tracking equipment and system degraded conditions and performing and reviewing operability determinations was well defined, clearly documented, and consistently applied. The new tracking program provides operations personnel with sufficient information to understand the impact of the degraded equipment on a component and system basis. (Section O2.1)
Licensed operator performance in training scenarios was generally adequate. The Pilgrim training staff identified a weakness in an operator's use of an emergency operating procedure (EOP) and provided appropriate remedial activitie Additionally, training evaluations were acceptable. Plant experience was incorporated in the scenarios. (Section 05.1)
Maintenance
- The inspector observed portions of selected maintenance activities and determined that activities were performed using approved procedures and completed with satisfactory results. Communications among work and support groups were good and supervisor oversight was acccptable. The inspector also verified through document review that technical specifications were satisfied and maintenance was performed by qualified personnel. (Section M1.1)
NRC identified a potentially degraded solenoid valve (SV) 302-21C which supplies air to the west scram discharge instrument volume outboard vent and drain valv A detailed operability evaluation was completed including performance based testing to determine that an immediate operability problem did not exist. A detailed test procedure and work package were written and implemented which replaced the potentially degraded solenoid valve. A root cause investigation was planned to identify any potentialinternal defects or failure mechanism. (Section M1.2)
Enaineerina
An implemented increase in the response time for the scram discharge volume high reactor water level scram instrumentation reduced spurious half scrams during post ii
.. - _ _
- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ . _ _ _ _ _ _
. .
f l
scram conditions and met the safety related setpoint change methodology criteri A calculation and minor design change carefully evaluated the effect of increased response time. (Section E8.1)
l Plant Sucoort l
- Reasonable As Low As Is Reasonably Achievable (ALARA) dose goals were established for 1998, the ALARA Committee was actively investigating dose reduction measures, and several significant dose reduction initiatives were planned including installation of permanent shielding in the drywell and reactor building, and a chemical decontamination of portions of the residual heat removal (RHR) system.
,
(Section R1.1)
l l
- Contamination controls in major plant work areas were generally good as evidenced by a spacious radiological controlled area (RCA) access control facility, a separate
'
RCA tool room, and maintenance of clean areas in the high pressure coolant injection (HPCI) room, reactor core isolation cooling (RCIC) room, and areas around the hydraulic control units (HCUs). Rain water intrusion into several enclosed radioactive material storage areas located outside of the plant had the potential to spread contamination. (Section R1.2)
- Housekeeping deficiencies in the torus room and retube building detracted from otherwise good conditions in major plant work areas. (Section R2.1)
- Quality assurance oversight and the problem report system were effective in the identification, evaluation, and resolution of radiological control program deficiencie (Section R7.1)
- The NRC identified several potential deficiencies in the fire protection program. One issue involved the identification of a major defect in an Appendix R raceway enclosure which protects safe shutdown cables. Further NRC review will be required to determine the safety significance and proper enforcement action for these deficiencies. (Section F1.1)
l l
iii ( _ . _ _ _ __. - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ -
. _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
. .
l TABLE OF CONTENTS EX EC UTI V E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii Summary of Plant Status ............................................1 1. O P E R AT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 O1 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 General Comments (71707) ...........................1 02 Operational Status of Facilities and Equipment ...................2 02.1 Operability Determinations and Equipment Status Control . . . . . . . 2 05 Operator Training and Qualification ...........................3 05.1 Operator Training and Qualification (71707) . . . . . . . . . . . . . . . . 3 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 08.1 (Closed) URI 50-293/97-10-01: Shortage of Operations Support Center (OSC) Personnel Qualified to Wear Self-Contained Breathing Apparatus (SCBA) During EP Exercise. . . . . . . . . . . . . . . . . . . . . 3 08.2 (Closed) VIO 50-293/97-11-01:Tagout Problems ............ 4 11. M AI N T E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M1 Conduct of Mair.tenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M 1.1 General Maintenance (62704,62707) . . . . . . . . . . . . . . . . . . . . 4 M1.2 On-line Replacement of Scram Discharge instrument Volume (SOLV)
Sole n oid Valve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Ill . EN G I N E E R I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 E3 Engineering Procedures and Documentation ..................... 6 E (Closed) IFl 50-293/96-08-01: Safety Evaluation Process . . . . . . . 6 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 E Response Time Change of Level Setpoint associated with Scram Discharge Volume (SDIV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 IV. PLANT S U PPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................ 9 R1 Radiological Protection and Chemistry (RP&C) Controls ............. 9 R initiatives to Maintain Radiation Exposures As Low As is Reasonably Achievable ( ALAR A) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 R1.2 Contamination Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 R2 Status of RP&C Facilities and Equipment ......................11 R2.1 H o u s e k e e pin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 R7 Quality Assurance in Radiological Protection and Chemistry Activities . . 12 R7.1 Identification and Resolution of Radiological Control Deficiencies . 12 P8 Miscellaneous Security and Safeguards Issues ..................13 P8.1 (Closed) IFl 50-293/96-04-01: Security Assessment Aid Weaknesses
..............................................13 ;
F1 Control Of Fire Protection Activities . . . . . . . . . . . . . . . . . . . . . . . . . . 13 l F (Open) URI 50-293/98-05-01: Major Defect in Appendix R Fire i Enclosure and inadequate Corrective Actions From 1996 Audit . . 13 iv i
!
!
l l
- _ _ _ . -. _- -. . _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ - - - _ _ _ _ _ _ _ _ _ _ _ - - _
. . Management Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
.
X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 X4 Review of UFSAR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 ITEMS OPENED, CLOSED, AND UPDATED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 LIST O F AC RONYM S U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1 V
i
_- _ _ _ _ . _ _ - _ - _ _ _ _ _ _
. .
t REPORT DETAILS Summarv of Plant Status Pilgrim Nuclear Power Station (PNPS) entered the repart period at 100 percent reactor power. Power was reduced to approximately 50 percent on two occasions to perform a thermal backwash of the main condenser. On May 28,1998, power was reduced to approximately 35 percent to repair a leak of the "A" feedwater flow instrument lin Power was restored to 100 percent on May 31,1998, where it remained for the remainder of the inspection perio l. OPERATIONS l
01 Conduct of Operations 2
'
01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspector conducted frequent reviews of ongoing plant operations including the May 28,1998, plant down power. In general, the conduct of operations was professional and safety-conscious. Shift turnover briefings were detailed and provided effective continuity. The plant down power was determined to be well cortrolled and executed. Anomalies noted during .
plant tours were brought to the attention of the licensee. These issues were immediately resolved and/or entered into the licensee's corrective action proces During the inspection period, operators identified a number of plant material deficiencies including the leak of the "A" feedwater flow sensing line that required a plant down power to affect repairs. The Operations Department Manager
- conducted meetings with operations department personnel re-emphasizing management's expectations regarding the identification and initiation of corrective actions for material deficiencies identified during operator tours. As a result of this focus, the number of licensee identified deficiencies significantly increased. Several teams, composed of personnel from the various organizations, also conducted tours of selected plant areas. Senior management initiated the teams to inspect all areas of the plant to identify material deficiencies. Plant inspection guidelines were issued to station personnel to identify selected attributes to be focused on during the plant inspections / tour .
~ ' Topical headings such as 01, M8, etc., are used in accordance with the NRC l- standardized reactor inspection report outline. Individual reports are not expected to address all outline topics. _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ -
_ - __ __ _ _ - - _ _ _ _ - _ _ _ _ _ - _ _ - _ _ . J '
02 . Operational Status of Facilities and Equipment O2.1 Operability Determinations and Eauioment Status Control Insoection Scone (71707 and 40500)
- Throughout the report period, the inspectors reviewed the implementation of the l
'
~ revised operations department procedure for documenting and tracking the operability status of safety related and important to safety equipment. The inspectors observed use of new guidance provided in Revision O of Pilgrim Nuclear Power Station Procedure No.1.3.34.5," Operability Evaluations," and Revision 3 of Nuclear Engineering Services Group Procedure No.16.04," Preparing Engineering Evaluations " both dated May 18,199 The inspectors examined a sample of the active Operability Evaluations (OEs) and the associated plans for corrective action. The purpose of this examination was to conduct a preliminary assessment of the licensee's controls to ensure timely I
' resolution of the degraded condition ' Observations and Findinas
. . . . I The inspectors noted that the new process for tracking the status of important to safety equipment was well defined, and accurately reflected existing regulatory information and guidance, primarily Revision 1 to Generic Letter 91-18. Generally -
consistent application of the new program was noted, in part as the result of the coordinated effort between engineering and operations in developing the new
'
guidance and a prescriptive methodology for using the process. The status of equipment degraded conditions was clearly documented, improving the ability of operators to readily evaluate the overall impact of system discrepancies on plant safet Of particular significance, the inspectors learned that the new program requires quarterly that the engineering and operations department heads review the status of the current operability evaluations to ensure timely corrective actions are progressing towards resolution of the degraded conditions and whether the aggregate of the degraded conditions affects plant safet The list of active OEs totaled 48, as of May 29,1998. In preparation for developing the new OE program, the Pilgrim staff had reviewed all of the identified degraded conditions being tracked through various engineering, maintenance and operating department level programs to ensure that the current list was comprehensive. The licensee's summary listing, " Operability Evaluation Log" reflected a significant re-assessment and resulted in a much improved understanding of the degraded conditions and impact on a system basis than was previously availabl i a
l
_ _ _ - - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ -_
, _ _ _ - _ - _ _ _
. .
3 Conclusions The recently revised process for tracking equipment and system degraded conditions and performing and reviewing operability determinations was well defined, clearly documented, and consistently applied. The new tracking program l
'
provides operations personnel with sufficient information to understand the impact of the degraded equipment on a component and system basi '
05 Operator Training and Qualification l
05.1 Operator Trainina and Qualification (71707) ! Insoection Scone j l
The inspector observed portions of three simulator training scenarios for licensed {
senior and reactor operators to assess operator performance, training evaluations, 1 and incorporation of plant experience in trainin l
! Observations & Findinas The inspector observed that operator overall performance was adequate. The inspector noted that one licensed operator demonstrated a weakness in identifying )
and performing the correct portion of an emergency operating procedure (EOP). The i inspector also verified that the training evaluators correctly assessed the weakness and provided remedial activities as required by the Pilgrim training progra Additionally, the inspector noted that the operators, facilitated by the training staff, actively participated and were self critical in post-scenario critiques, and plant experience was incorporated in the scenario Conclusions Licensed operator performance in training scenarios was adequate. The Pilgrim training staff identified the weakness and provided appropriate remedial activitie Additionally, training evaluations were acceptable. Plant experience was incorporated in the scenario Miscellaneous Operations issues (92700,92901)
08.1 (Closed) URI 50-293/97-10-01:Shortaae of Ooerations Suooort Center (OSC)
Personnel Qualified to Wear Self-Contained Breathina Anoaratus (SCBA) Durina EP Exercis During a November 1997 emergency preparedness (EP) exercise, the licensee identified that only four of the twenty responders who were available for repair teams (no maintenance personnel) were qualified to wear SCBA. Although other members of the response team were trained, they did not have a valid face mask
" fit" for SCB in response to this concern, a memo was issued documenting management's expectation that it iu the individual's responsibility to ensure all of the qualifications
.
to wear SCBA and other protective equipment be kept up to date. The inspector l was informed by the maintenance manager that the maintenance deoartment
'
administrative assistant had also been assigned the responsibility of tracking and scheduling SCBA " fit" tests for the maintenance organization and that all maintenance personnel are scheduled to be SCBA qualified by September 199 The inspector reviewed the qualification database and noted that over half of the maintenance personnel are presently qualified to wear SCBA. The other emergency response team members are SCBA qualified or are scheduled to complete all qualification element Procedure 6.7.1-104," Issue and Return of Respiratory Protection Equipment,"
requires that radiation protection personnel verify an individual's qualification prior to the issuance of respirators. Discussions with licensee personnel and inspector review of the past six months of problem reports revealed no instances where an unqualified individual was issued a respirator. The inspector was also informed by the licensee that the means to perform a " fit" test exists in the OSC, which is accessible post accident. The irspector concluded that the licensee's actions to address this issue were adequate. No violations of regulatory requirements were identified; this item is close .2 (Closed) VIO 50-293/97-11-01:Taaout Problems Corrective actions included retraining plant operators in human error prevention techniques, and revising procedure 1.4.5, "PNPS Tagging Procedure," to require that an independent technical review of the tagout adequacy be performed. The inspector reviewed the adequacy of tagouts for selected maintenance requests and verified the proper placement and positioning of tags on plant components. In addition, a review of problem reports generated since January 1998 was performed to determine if similar tagout problems had recurred, of which none were identifie The inspector concluded that the licensee's corrective actions were adequate, this item is close . MAINTENANCE M1 Conduct of Maintenance M 1.1 General Maintenance (62704,62707)
The inspector observed portions of selected maintenance and surveillance activities to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to limiting conditio.'s of operation, and correct system restoration following maintenance and/or testin Portions of the following activities were observed:
- S9804023 RWCU high flow functional test
. _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ -_
. .
- S9801698 RWCU high flow functional test
- S9804273 HPCI steam line high flow functional test
- 19800103 IRM "D" pre-amp - remount in cabinet
- 19602611 EPA's 5 + 6 card replacement and breaker calibration
- S9712638 "B" Suppression chamber water temperature calibration
- 19800887 Installation of test equipment for diesel validation test procedure TP98-007
- 19801098 Replace west scram discharge instrument volume (SDIV) air supply solenoid valve The inspector determined the activities were performed using approved procedures and completed with satisfactory results. Additionally, communications between the various work and support groups were good and supervisor oversight was acceptable. The inspector also verified through document review that technical i specifications were satisfied and maintenance was performed by qualified I personnel, i
M1.2 On-line Reo!acement of Scram Discharae Instrument Volume (SDIV) Solenoid Valve l Insoection Scope (62703)
During a plant inspection tour, the inspector identified a loud rattling sound originating from solenoid valve (SV) 302-21C which supplies air to the outboard vent and drain valve on the west scram discharge instrument volume (SDIV). The inspector was concerned that the solenoid valve may be defective and possibly prevent a safety related function. The inspector informed control room operators of i the apparent deficiency and observed the licensee's evaluation and corrective action Observations and Findinas The licensee generated problem report 98.9257 to document, evaluate and correct the condition. Testing of the valve and review of prior inservice tests did not show any adverse trends of valve response times; however, the licensee elected to replace the solenoid to eliminate any potential failure mode. The replacement activity required the SDIV to be isolated. The licensee wrote a temporary procedure (i.e., TP98-017)to measure the leakage into the SDIV to ensure enough time existed to perform the maintenance. Performance of the TP determined that enough time existed to conduct the wor Prior to the performance of the maintenance activity, a mockup of the solenoid change out was performed by the Instrumentation and Control (l&C) departmen These evolutions were performed to ensure that the activity could be performed on-line without adversely affecting the plant. The pre-evolution brief was well attended by plant management and all of the salient points, including the potential affect on the plant and abort criteria were covered in detail. The inspector verified that the applicable technical specification was entered during the performance of the work l
- ___ _------- .____ _ ___ _______ ____ _ _ _______ _ _ _
.
. .
and that approved procedures were properly being followed. The inspector also noted that the work was closely monitored by l&C management. The work to replace the solenoid valve was well planned and executed. Also, the licensee initiated a direct cause analysis.
1 Conclusions NRC identified a potentially degraded solenoid valve (SV) 302-21C which supplies air to the west SDIV outboard vent and drain valve. A detailed operability evaluation was completed including performance based testing to determine that an immediate operability problem did not exist. A detailed test procedure and work package were written and implemented which replaced the potentially degraded ;
solenoid valve. A root cause investigation was planned to identify any potential internal defects or failure mechanis Ill. ENGINEERING E3 Engineering Procedures and Documentation l
E3.1 (Closed) IFl 50-293/96-08-01:Safetv Evaluation Process j
,
A potential inconsistency was noted between engineering work instruction, NEDWI 395 revision 4, and engineering program procedure NOP83E5 (i.e., Safety Reviews)
in the area of performing preliminary 10 CFR 50.59 safety evaluations for degraded conditions. NEDWI 395 specified the use of preliminary 10 CFR 50.59 safety evaluations for degraded conditions while NOP 83E5 was silent on that matte The NRC issued revision 1 to NRC Generic Letter (GL) 91-18, on October 8,1997, to clarify the use of 10 CFR 50.59 for degraded conditions. Preliminary 10 CFR 50.59 evaluations were not required to address degraded conditions. A 10 CFR ;
50.59 safety evaluation is required if the final resolution will be a licensing basis )
change to accept the condition as-is or if interim corrective action requires a l procedure change or temporary modification. A 10 CFR 50.59 safety evaluation is not required for a degraded condition provided corrective actions are scheduled at l the first available opportunity such as the next refueling outag !
During this assessment period, the inspector confirmed that the licensee made the ,
appropriate procedure changes to implement the guidance of revision 1 to NRC l Generic Letter 91-1 Procedure 16.04 replaced NEDWI 395 revision 4 and NOP 83E5 was revised. The inspector identified no concerns and determined that the licensee procedures complied with the intent of NRC GL91-18, revision 1. IFl 50-293/96-08-011s close . _ . _ . - _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ _
. .
-7
.
E6 Miscellaneous Engirnering issues
'E Resoonse Time Chanae of Level Setooint associated with Scram Discharae Volume (SDIV) Insoection Scone (37551)
The inspector reviewed a revised setpoint calculation and associated documents involving a change in the response time of the scram discharge level switches that was implemented during refueling outage no.11-(RF011) conducted February 199 to April 1997. This review assessed whether the change was within the existing design basis requirements and to confirm its implementation in the plant, and was consistent with the licensee's established procedure and the NRC guideline Observations and Findinas
- Two level elements, also referred to as FCI probes, with corresponding switches (LS-302-83 A&B - West SDIV and 302-82C&D - east SDIV) and two Rosemount level transmitters are provided for each scram discharge instrument volume (SDIV).
The logic for each scram discharge volume (SDV) high water level scram signal is one out of two taken twice. This scram signal ensures sufficient volume remains in the SDV to accommodate discharged water to ensure the scram goes to completion and proper control rod scram times resul During RFO11 an instrumentation and control (l&C) surveillance (i.e.,8.M.1-20)
found that the response times for the FCI probes exceeded the response time criteria of less than or equal to 1 second. The maximum time recorded was 1.15 seconds. Problem report (PR) 97.9217 was issued on March 20,1997, to document, evaluate and correct the conditio An l&C design engineer determined that the response time could be increased from 1 second to 2+/ .2 seconds based on a calculation and field revision notice (FRN).
This change was beneficial to preclude spurious half scram signals during post scram conditions when the scram conditions reset. Additionally, better human factors existed to adjust the response time in the range of 2 seconds. The I&C engineer was also aware, based on previous engineering work for the SDV high water level scram function, that the setpoint analysis remained valid with a response time of up to 4 seconds. l&C technicians increased the FCI probe response time to 2 +/ .2 seconds and the system was declared operable. Pilgrim technical specifications (TS) and updated final sdfety analysis report (UFSAR) do not contain response time requirements. Hence, the inspector confirmed that the original 1 second response time surveillance requirement was self imposed by tha license A second PR (i.e, 97.2798)was initiated by the licensee on September 18,1997, which questioned the adequacy of implementing the aforementioned response time change via a calculation and FRN without a complete 10 CFR 50.59 safety
_ _ _ _ - _ _ _ _ ____ . _ _ - _ - -
.
evaluation. As a conservative measure, the licensee completed Safety Evaluation 3121, dated October 17,1997, which concluded that no unreviewed safety question (USQ) was involved with the response time change. Inspector review of SE 3121 during this inspection period identified that it had not been approved by the Operations Review Committee (ORC). l&C engineering management informed the inspector that SE 3121 was previously presented to the ORC approximately 6 months ago, but the ORC needed more time to review the details. During this inspection period, the ORC approved SE 3121 with no changes required. The licensee initiated a PR to document this administrative oversight of not obtaining the ORC approval. This failure constitutes a violation of minor significance and is not subject to formal enforcement actio The inspector reviewed the evaluation and corrective actions associated with PR 97.2798 on the need for a detailed 10 CFR 50.59 safety evaluation. The licensee concluded that the response time change was properly evaluated for 10 CFR 50.59 requirements by the FRN and calculation. The response time change was performed under Plant Design Change (PDC) 97-04, " Standing PDC For Instrument and Control Modifications - 1997." Administrative limitations in PDC 97-04 allowed minor modifications to instrument setpoints provided the change does not involve a change as described in the Pilgrim UFSAR or TS. The setpoint change is further limited for changes that do not degrade the system function or operability. The licensee had adequate setpoint calculation controls in place per NESG Procedure 3.05 and NEDWI 39 The inspector reviewed the aforementioned response time change which was documented and evaluated by FRN 97-04-35 dated April 9,1997, and setpoint calculation l-NI-106 revision 1, dated April 4,1997. The setpoint calculation was performed in accordance with NRC Regulatory Guide 1.105, " Instrument Setpoints For Safety Related Systems," and industry standard document, ANSI /ISA-S67.04,
"Setpoints For Nuclear Safety Related instrumentation." The increase in response time allowed more water to enter the SDIV prior to the scram initiation; however, the TS allowable value of 38 gallons and the design basis analyticallimit of 62.835 documented in UFSAR Table 7.2.1 remained unchanged. The increase in response time corresponded to approximately 6 additional gallons of water prior to scram initiation. The additional 6 gallons was within the margin for the total instrument loop uncertainty which was calculated as part of the setpoint change methodolog Based on this, the licensee concluded that the as found condition of greater than 1 second response time was not reportable pursuant to 10 CFR 50.72/73 criteri The inspector identified no technical or reporting concerns with these issue I Preliminary evaluation checklists (PEC) were completac' for both FRN 97-04-35 and i calculation I-NI-106 revision 1. Tne PECs concluded that a detailed 10 CFR 50.59 safety evaluation was not required. This was based on remaining within the limitations of PDC 97-04 and a related pending UFSAR change request (i.e.,2396)
which deleted the SDIV accuracy data. Inspector review of the PECs identified a weakness in the documentation reference between the PECs and the FRN, calculation and pending UFSAR change request. The Engineering Department
]
!
i
,
,
___._._..___.___________ . _ - - - - - __
- - _ - _ _ - _ ._ ._
.
Manager agreed with the inspector's concern and indicated that management
. expectations in this area on completing PECs with proper references would be reviewed and clarified. Conclusions l
l
' An increase in the response time for the scram discharge volume high reactor water level scram instrumentation reduced spurious half scrams during post scram l- conditions and met the safety related setpoint change methodology criteria. A
- calculation and minor design change carefully evaluated the effect of increased response tim IV. PLANT SUPPORT
-
-
. 'R1 Radiological Protection and Chemistry (RP&C) Controls 1 R1.1 initiatives to Maintain Radiation Exoosures As Low As Is Reasonably Achievable .
(ALARA) . Insoection Scooe (83713.)
A review was performed to evaluate current radiation exposure goals and planned initiatives to maintain radiation exposures ALARA. Information was gathered by a review of the 1998 dose goal, radiation dose tracking records, selected ALARA committee meeting minutes, a document entitled " Boston Edison - Pilgrim Nuclear Power Station Dose Reduction Strategy," Rev.1, and through interviews with cognizant personne ]
I Observations and Findinas !
l The station dose goal for 1998 had been established at 110 person-rem. The goal !
included input from all major work groups and specific goals for special projects such motor operated valve work, chemical decontamination of portions of the residual heat removal (RHR) system, and installation of permanent scram volume header shieldin ..
.
l A review of selected ALARA committee meeting minutes indicated that the station was actively involved in the investigation, prioritization, and implementation of dose reduction measure The Pilgrim Nuclear Power Station Dose Reduction Strategy, approved by the ALARA Oversight Committee, readily acknowledged that performance improvement L' was warranted and outlined a strategy fx reducing occupational dose. The report l indicated that the two-year rolling dose average for the best quartile of boiling water I reactors in the United States was approximately 150 person-rem, and Pilgrim's best
[ . two-year rolling average since 1990 was 299 person-rem. Significant dose L
.. . - - _ _ - _ _ _ _ - _ _ _ _ _ - - _ - - - - - - . __ _ -- - - . .
.
reduction initiatives included plans for the installation of permanent shielding for recirculation and RHR piping in the drywell, installation of permanent shielding on the scram volume header in the reactor building, and plans for a chemical decontamination of RHR piping.
l Conclusions Reasonable ALARA dose goals were established for 1998, the ALARA Committee was actively investigating dose reduction measures, and several significant dose reduction initiatives were planned including installation of permanent shielding in the drywell and reactor building, and a chemical decontamination of portions of the RHR systei R1.2 Contamination Control insoection Scope (83750)
A review was performed of contamination control boundaries, information was gathered through tours of the facility and discussions with cognizant personne Observations and Findinas The major radiological controlled area (RCA) access was spacious, well staffed, and well equipped with contamination monitoring equipment. A tool control facility was located within the RCA and served to minimize the number of tools brought into the RCA. Many areas had painted surfaces to allow for better surface decontaminatio The high pressure core injection (HPCI) room, the reactor core isolation cooling (RCIC) room, and areas around the hydraulic control units (HCUs) were maintained as clean areas and were readily accessibl Contamination Control Weaknesses Several weaknesses in contamination control were identified during tours of enclosed radioactive material storage areas located outside of the plant including several temporary storage (" Kelly") buildings and a building referred to as the
"retube" buildin Kelly Buildina No. 3:
Water was observed on the floor of several temporary storage buildings used to store radioactive material outside of the plant. It was not clear if the water had entered the building from a leak in the roof or from around and beneath the rollup door. A piece of equipment labeled as an RHR test fixture with 20,000 dpm/100 cm fixed contamination was observed in Kelly building number 3, and was partially unwrapped and sitting on a wooden palate in a puddle of water. The inspector noted that water leakage onto surfaces with fixed contamination could potentially spread contamination to floor surfaces and to outside areas. A health physics (HP)
_ _ _ _ _ _ _ - - _ ___
_ _ _ _ _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ - _ _ - _
,
!
! 11 t
l supervisor stated that the condition was unacceptable, and the radiation protection
!
manager (RPM) indicated that the following corrective actions would be taken: 1) a walkdown of the Kelly buildings would be performed to identify materials / equipment with inadequate coverings and a waste control technician (WCT) would be assigned
'
to properly wrap / cover the identified items; 2) inspections would be performed during a rain storm to identify water entry points and repairs would be initiated as necessary; and 3) absorbent tubes would be installed on both sides of the rollup door to minimize water in-leakage, j
! Retube Buildina:
Rain water had leaked into the lower elevation of the retube building traveling i through a posted contamination area into a clean area walkway; creating the I potential to spread of contamination. A HP supervisor stated that the water and the affected " clean" area had been sampled and were not contaminated. The RPM acknowledged that the condition was inadequate and stated that a cleanup of the area would be initiated as soon as maintenance could support the movement of boxes and drums to allow for a proper decontaminatio l l Conclusion i
Contamination controls in major plant work areas were generally good as evidenced !
by a spacious RCA access control facility, a separate RCA tool room, and maintenance of clean areas in the HPCl room, RCIC room, and areas around the HCU Rain water intrusion into several enclosed radioactive material storage areas located outside of the plant had the potential to spread contaminatio R2 Status of RP&C Facilities and Equipment
. R2.1 Housekeeping Insoection Scope (83750)
Plant tours were conducted to evaluate housekeeping and cleanliness, material conditions, and maintenance of radiological boundaries. Areas inspected included selected areas of the reactor, radwaste, and turbine buildings; the trash compactor facility; and various enclosed radioactive material storage areas located outside of the plant, Observations and Findinas Overall, housekeeping in major plant work areas was good in that aisles and l walkways were clear and free of debris, tools and equipment were stored in
- designated areas, and radiological boundaries were clearly delineated. The I following exceptions to good housekeeping practices were identified
- 1)
!
l
- _ _ _ _ _ _ - - _ _ _ _
_ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ _ _ ____________
. .
miscellaneous debris including a hard hat, gloves, paper and plastic, and flashlights were observed in the torus room; 2) tools, equipment, and trash were stored in a j haphazard manner in the re-tube building; and 3) poor lighting was observed in the reactor building torus room requiring use of flashlights. Based on these observations, licensee staff dispatched a crew to pick up miscellaneous debris in the l torus room; stated that a cleanup of the "retube" building was planned prior to the next refueling outage; and stated that a plant design change had been initiated to improve lighting in the torus roo Conclusion Housekeming deficiencies in the torus room and retube building detracted from otherwise good conditions in major plant work area R7 - Quality Assurance in Radiological Protection and Chemistry Activities R7.1 Identification and Resolution of Radiological Control Deficiencies Insoection Scoce (83750)
A review was performed to determine if licensee audits and problem reporting system were effective in the identification, evaluation, and resolution of radiological control deficiencies. Information was gathered by a review of selected quality assurance surveillance and an " Oversight Program Review of the Radiation Protection Program (No.97-04)," procedure 1.3.121, " Problem Report Program,"
selected Problem Reports, and interviews with cognizant personne Observations and Findinas l
The quality assurance group (OAG) had replaced the traditional audit of the radiation protection program with a series of surveillance and a year end program revie This change was made to provide more timely feedback and to allow for a better evaluation of adverse trends. The review was broad in scope, trends were discussed, and findings were placed into the problem reporting syste A review of radiological control issues placed into the problem rcport system
,
revealed a high volume-low threshold system that the staff readily used to address l program deficiencies. Approximately 164 problem reports related to radiological controls were initiated between the period of September 1,1997 and April 1,199 '
Ten problem reports were selected to evaluate the effectiveness of the system for-resolving problems. Problem evaluations including identification of cause and corrective actions taken were reasonable and commensurate with the significance i of identified issue !
l 1 I
l L
- _ _ _ _ _ _ _ - _ - - - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _
. .
13 Conclusions Quality assurance oversight and the problem report system were effective in the identification, evaluation, and resolution of radiological control program deficiencie P8- Miscellaneous Security and Safeguards issues P (Closed) IFl 50 293/96 04-01: Security Assessment Aid Weaknesses I
Weaknesses in the capability of security assessment aids were identified in NRC i inspection Report No. 50-293/96-04, dated April 17,1996. This weakness was followed up by security specialist inspectors from NRC Region I as documented in
'
NRC Inspection Report No. 50-293/98-03, dated April 3,1998. The NRC identified multiple examples (eel 50-293/98-03-01)of equipment failures associated with the protected area assessment system. An enforcement conference was held on April 27,1998, and the NRC subsequently issued a Severity Level lil violation with a
$55,000 civil penalty. The inspector noted that eel 98-03-01 replaced IFl 96-04-01, and as such IFl 96-04-01 is closed. The issues of eel 98-03-01 (NOV ID#
01013) remain open pending licensee response to the Notice of Violatio F1 Control Of Fire Protection Activities F (Open) URI 50-293/98-05-01:Maior Defect in Anoendix R Fire Enclosure and
- Inadeauate Corrective Actions From 1996 Audit- l Insoection Scope (71750. 37551)
l
= The inspector reviewed the status of compensatory measures established for degraded fire barriers and visually inspected several fire barriers to ensure proper fire barrier configuration. The fire barrier acceptance criteria of procedure 8.B.29,
" Inspection of Fire Barriers," was used during the inspection. Additionally, the
.
inspector reviewed and followed-up on the 1996 triennial quality assurance (QA)
audit performed for the fire protection progra Observations and Findinas The inspector inspected the following three Appendix R cable raceway enclosures:
'
l
- ' Fire barrier 194.5018-encl.3 located in the cable spreading room (CSR).
_
i
- Fire barrier 194.504A-encl.1 located in the ceiling of the "B" train lower switchgear roo * Fire barrier 194.504C-encl. 2 located in the outer hallway of the "B" train lower switchgear roo ,
- _ . _ . . - - . ----.__-_______-_--_-__-----_-_.-_-_._-_---_-__.___w
I
. .
The inspector identified three deficiencies including a two square foot major defect on the back side of barrier 194.501B-encl. 3 and also two minor gaps (1/8 inch wide by 5 inches long) between block walls and fire barriers 194.501 B-encl. 3 and 194.504C-encl. 2. The major defect was evident by exposed metal lath without the required 1.5 inches of Pyrocrete fireproofing material. The inspector informed fire protection program personnel of the fire barrier deficiencies who initiated corrective
. actions. Fire barrier 194.501B-encl.3 was immediately declared inoperable and a fire watch was established per UFSAR requirement 10.8.4.6.1, " Fire Barrier System Technical Requirements." Fire barrier 194.504C encl. 2 was already inoperable due to another small gap in the protective coating previously identified by the licensee i during a 1996 fire protection program audit. The licensee initiated three problem reports to address the fire barrier deficiencies identified by the inspector. The licensee initiated an evaluation and inspection of the full problem scope related to other possible degraded fire barriers at Pilgri The inspector interviewed several fire protection program personnel and reviewed relevant design documents including calculations that were related to fire barrier 194.5018-encl.3. These three raceway enclosures were constructed with a 3/4 inch channel iron framework spaced every 16 inches. The frame was then covered with sheets of metallath with a 1.5 inch layer of Pyrocrete fireproofing materia The enclosures house safe shutdown electrical cables and were rated as '3-hour fire barriers to maintain the cold (unexposed) side less than 325 degrees Fahrenheit during a postulated fire. The enclosures were built in 1980 per Plant Design Change (PDC)79-03C.2 to comply with Section Ill.G.2 of Appendix R, " Fire Protection Programs For Nuclear Facilities Operating Prior To January 1,1979."
The cables inside of barrier 194.5018-encl.3 were power cables for the "A" safe shutdown train including the "A" core spray pump, the "A" and "C" residual heat removal pumps and the power feed cable from the "A" emergency diesel generator to the A5 emergency bus. The inspector noted that safe shutdown could be achieved during a postulated fire by using the unaffected "B" safe shutdown train equipment and alternate safe shutdown control stations located outside of the control room. The inoperable fire barrier was a concern due to the NRC identification of a longstanding deficiency which was missed during routine licenseo I surveillance inspection ~ The longer term corrective action to' correct the major defect was under evaluation at the end of this inspection period. The licensee performed a detailed inspection of
<
the major defect and noted other deficiencies with the subject enclosure including varying thicknesses of the Pyrocrete fireproofing material. Potential corrective ;
actions included the installation of a newly designed Appendix R raceway enclosure, )
rerouting safe shutdown cables and/or further detailed analysi '
The inspector reviewed quality assurance audit 96-04, " Fire Protection Program Triennial Audit," dated May 1996. The audit documented one finding and two potential findings. QA personnel initiated several problem reports to evaluate and .
- - _ _ _ _ _ - _ _ - _ _ -
.
correct each of the audit findings. The inspector independently reviewed the close-out of the audit findings to evaluate the effectiveness of the corrective action PR 96.9323 was issued to address potential finding (PF) 1 which raired 9 specific concerns on the design adequacy of the three aforementioned Appendix R raceway enclosures. One concern was the lack of a configuration specific test of the three
\ fire snclorares which utilized a 1.5 inch thick layer of Pyrocrete fireproofing. Only a
" qualitative" calculation (FP-37) existed which was performed in 1986 that utilized vendor test data from a flat panel fire test. The audit report identified that using a queWative calculation in lieu of a specific configuration test was not consistent with NRL deneric Letter 86-10," Implementation of Fire Protection Requirements."
in 1996, engineering personnel completed an operability evaluation for PR 96.9323
\ which concluded the enclosures were operable based solely on the 1986 calculation FP-37. During this inspection period, the inspector determined that 9 specific concerns listed as PF1 in audit report 96-04 were not evaluated in detail and were improperly closed out. The inspector determined that PR 96.9323 was closed out with the evaluation and corrective actions transferred to PR 96.9324. PR 96.9324 was initiated to address PF2 of audit 96-04 viiiich related to the acceptability of several fire barriers (i.e., FPEE 4,101,61 on block walls, structural steel and joint compound). The licensee determined that these FPEE issues were technically justified as 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> barriers, but may require further evduation including a 10 CFR 50.59 safety evaluatio Inspector review of the close-out to PR 96.9324 determined that no detailed evaluation and corrective action was taken to address either PF1 or PF2. The basis for closure of PR 96.9324,which was prepared and approved by Regulatory Affairs Department personnel, was inadequate. The 1986 calculations (FP-36 and 37) for the Appendix R raceway enclosures were used as the basis for closing out a different concern on block walls. Hence, the inspector determined that both potential findings raised in the 1996 audit were improperly closed-out. The quality assurance staff initiated PR 98.1053 to resolve these issue The inspector also reviewed the QA Fire Protection Audit 97-05. This report identified a broad degradation in the fire protection program that led to the initiation of a significant conation adverse to quality (SCAG) problem report. The licensee determined that the root cause was a number of organizational changes, staff and resource reductions made withcut understanding the impact of the changes on the fire program integrit The inspector discussed the potential programmatic nature of the aforementioned NRC findings with senior site management. The inspector concluded that the organizational changes resulted in no clear ownership of the fire Frogram; it had been spread between several departments. Additional resources and the appointment of a fire protection program team leader were assigned during this inspection period. The inspector identified three concerns including the fire enclosure defects, untested configuration and ineffective corrective actions from the i l
_ _ _ _ _ _ _ _ _ _ _ _
__ . _ _ _ _ _ _ -_____________ _ _ _-_ _
.
1996 audit findings. These issues will remain unresolved (50-293/98-05-01)
pending further NRC inspection to assess the licensee's corrective actions and response to prior audit findings, as well as to determine the safety significance of the identified degraded condition Conclusions l The NRC identified potential programmatic deficiencies in the fire protection program. One issue involved the identification of a major defect in an Appendix R raceway enclosure which protects safe shutdown cables. Further NRC review will be required to determine the safety significance for these deficiencie V. Management Meetings X1 Exit Meeting Summary The inspector presented the inspection findings to members of the licensee management after conclusion of the inspection on June 19,1998. The licensee acknowledged the findings presente I X4 Review of UFSAR Commitments A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR description highlighted the need for a special focused review that compares plant practices, procedure and/or parameter to the UFSAR description While performing the inspections discussed in this report, the inspector reviewed the applicable portions of the UFSAR that related to the areas inspected. One problem was !
identified with compliance of an Appendix R fire barrier which is discussed in Section F I of this repor j
i
!
,
i l
_ __ _ ._ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _
i .
.)
L l
-lNSPECTION PROCEDLMES USED l i
IP 37551: Onsite Engineenng '
IP 40500: Effectiveness of Licensee Controls in Identifying', Resolving, and Preventing Problems -
IP 61726:
..{
Surveillance Observation IP 62707: Maintenance Observation ]
IP 71707. . Plant Operations l IP 71750: Plant Support Activities IP 82301: Evaluation of Exercises for Power Reactors
. lP 83728:-- Maintaining Occupational Fxposures ALARA iP 83750:. Occupational Radiation Exposure IP 92700: Onsite Followup of Written Reports of Nonroutine Eveas at Power Reactor
- Facilities IP 92901:- Followup - Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering IP 92904: . Followup - Plant Support IP 93702: Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED, CLOSED, AND UPDATED .
Opened '
URI 50-293/98-05-01 Fire protection program deficiencies Closed IFl 50-293/96-04-01 Security assessment aid weaknesses IFl 50-293/96-08-01 Safety evaluation process URI 50-293/97-10-01 Shortage of self-contained breathing apparatus (SCBA)
qualified personnel during emergency planning exercise
- VIO 50-293/97-11-01 Inadequate tagout implementation .
.
(
LIST OF ACRONYMS USED ALARA' As Low As is Reasonably Achievable
~ APRMs ' Average Power Range Monitors BECo Boston Edison Company CFR- Code of Federal Regulations CRD Control Rod Drive CS Core Spray E Emergency Preparedness EPIC ' Emergency and Plant Information Computer ESF Engineered Safety Feature -
. _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
___ _ _ _ - __-____-___ _ _ _- - _ -___ _ _ - ____ _ __ . _ _ _ _ _ _ _ _
. e
,
18 gpm gallons per minute HP Health Physics -
l&C . Instrumentation and Contrnis-IFl inspection Follow-Up ' tem
~ lR - Inspection Report
. LEit Licensee Event Report MG Motor Generator
- MR Maintenance Request NCV Non-Cited Violation NOV Notice of Violation NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NWE Nuclear Watch Engineer PDR Public Docket Room PNPS Pilgrim Nuclear Power Station PR- Problem Report QA Quality Assurance QAG Quality Assurance Group RHR Residual Heat Removal RP Radiological Protection RPM Radiation Protection Manager SALP Systematic Assessment of Licensee Performance SRO Senior Reactor Operatcr
- T . Temporary Modification
'TS Technical Specification UCSAR' Updated Final Safety Analysis Report
- VIO Violation
- WWM Work Week Manager l
l
!
.I l
i e
e