IR 05000293/1988037
| ML20244A643 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/30/1989 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20244A630 | List: |
| References | |
| 50-293-88-37, GL-83-08, GL-83-8, GL-88-14, NUDOCS 8904180115 | |
| Download: ML20244A643 (38) | |
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S. NVCLEAR REGULATORY. COMMISSION , -REGION I ' . Docket No'.': 50'-293 . Report'No.: 50-293/88-37 Licensee: Boston Edison Company ' ' -800 Boylston Street Boston, Massachusetts 02199 Facility: . Pilgrim Nuclear Power Station.
~ Location: -Plymouth, Massachusetts.
Dates: December 27, 1988 - February 5, 1989 . Inspectors: C.: Warren, Senior Residerit Inspector and' Restart Manager - T.- Kim, Resident Inspector, Pilgrim Station C.1 Carpenter, Resident Inspector, Pilgrim Station S., Barber, Resident Inspector, Millstone. Station R. Barkley,-Reactor Engineer,-Region I'(RI)- 'G. Bethke, NRC Contractor G.: Bryan, NRC Contractor T. Dragoun, Senior. Radiation Specialist,'RI.
. P. Drysdale, Reactor' Engineer, RI-A. Howe, Senior' Operations Engineer, RIJ S.'Juergens, Reactor' Engineer, RI M. Kohl, Reactor Engineer, RI 'J. Lyash, Project Engineer, ' RI l J. Macdonald, Resident Inspector, Vermont-Yankee D. Mcdonald, Project Manager, Office of Nuclear Reactor
- Regulation (NRR)-
F. McManus, NRC Contractor . , L. Miller, Technical Training Center (TTC), Office of Analysis and. Evaluation of Operat onal Data (AE0D) i , l D. Moy, Reactor Engineer, Ri J.'Raleigh, Project Engineer, NRR-T. Rebelowski, Senior Reactor Engineer, RI L. Rossbach, Senior Resident. Inspector, Indian Point 2 E. Trottier, Project Engineer, NRR-P. Wilson, Reactor Engineer, RI' ' Approved by: I NO - D A. Randy BlougtR' Chief Date Reactor Projects Section No. 3B l: Division of Reactor Projects
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Inspection Summary: Areas Inspected: Restart Staff inspection to assess licensee management con- .trols, conduct of operations, and startup testing activities during the initial phase of the licensee's Power Ascension Program.
A review of the licensee's preparations for startup was also performed on December 27-30, 1988.
Results: Violation: The report documents a licensee-identified violation involving failure to control locked high radiation area access as required by the Tech-nical Specifications (Section 8).
Unresolved Item: Further review of the licensee's newly generated Radiological Environmental Technical Specifications (RETS) surveillance implementing pro-cedures for technical adequacy, as well as review of the licensee's approach to event reporting is needed to determine adequacy (Section 5.0).
Strengths: i 1.
Licensee management provided active and effective oversight and assessment i of plant operations (Section 11.0); ' 2.
Operational evolutions were performed in a competent and professional ! manner (Section 3.0); 3.
Startup testing activities were well controlled (Section 4.0); , 4.
The licensee's design change which corrected the secondary containment track lock deficiency was implemented in a. timely manner and was well thought out from conception to implementation (Section 7.0).
Weaknesses:
1.
The licensee experienced difficulties with implementing the torus vacuum breaker block valve modification due in part to an overly aggressive implementation schedule set by upper management.
Further, weak organiza-
tional communications prevented upper management from recognizing the operating constraints imposed as a result of the initially implemented torus vacuum breaker block valve modifications (Section 7.0).; 2.
A lack of formal administrative controls for the scheduling and perform-ance of RETS surveillance caused failure to properly implement RETS (Section 5.0).
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_ _ _ _ _ _ _ _ _ _ _ _ - _ _ l Inspection Summary (Continued)
Observation The operations staff experienced some difficulties in transition from an extended outage to the operating mode.
In these instances, licensee staff response and management oversight provided for appropriate identification, i assessment and implementation of corrective actions (Section 3.0).
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- TABLE OF CONTENTS Page'
1.
Summary ofcFacility Activities............................ I'- > ; L 2.
Restart Preparation Activities (Modules 71707, 71710, 61726,62703,92701)....................................
' 2.1-Review'of Active Temporary Procedures'...............
,2.2 Review of Safety Evaluations.........................
2.3 Temporary and Permanent Radiological Shielding Program............................................
2.4 Engineering Service Requests.........................
2.5 Quality Assurance Discrepancy Reports................ 4~ 2.6 Failure-and Malfunction Reports......................
.2.7 10 CFR Part 21 Report by Limitorque Concerning.
. Elevated Ambient-Temperature Effects on RH-1 Insulated DC Motors................................
.2.8 Review of Maintenance Requests.......................- S 2.9 -Surveillance Program Status..........................
2.10 Safety ~ System Wa1kdowns..............................
2.11 Lifted Lead and Jumper Log...........................
-s 2.12 Operations Review Committee (ORC) Activities......... 8-3.
Operations (Modules 71707, 71710, 71711, 71714, 71715, 62703,61726,81046,81700,93702)......................
3.1 Sustained Control Room Observations..................
3.2 ' Plant Tour Observations..............................
3.3 Review of Training Reactivity Manipulations..........
3;4 Cold Weather Protection..............................
3.5 Review of Plant Events............................-... 12' 4.
Startup Testing Activities (Modules 61726, 72700, 72701, 61707)..................................................
4.1 Shutdown Margin Calculations.........................
4.2 HPCI and RCIC Surveillance Testing at 150 psig.......
5.
Surveillance (Modules 61726,37700,73753)...............
5.1 Radiological Environmental Technical Specification...
5.2 Routine' Surveillance Tests...........................
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~ Table of Contents (Continued) g.e
Page 6.
Maintenance and Modifications (Modules 37700, 37701, 62703,62700,62705).........'...........................
6.1 High Pressure Coolant Injection (HPCI)' System Gland Seal. Condenser Hotwell Pumo Replacement............
6.2 Intermediate Range Monitor (IRM) Detector Replacement........................................
7.
Review of Generic Letter 88-14 (Modules.92703, 90712, 37700).................................................. 19-7.1. Seconda ry Containment Integrity.......................
7.2 Inadequate Reactor Building to Torus Vacuum Breaker ' Isolation' Valve Design....................... .....
7.3 Conclusions..........................................
8.
Radiological Controls (Modules 83750,84750)..............
8.1 Radiation. Monitoring Systems.........................
8.2 Special Radiation Surveys............................
o 8.3 Control of Locked High Radiation Areas.............. '24 9.
Followup on Previous Inspection Findings (Modules 92701,- 35502,92702)...........................................
10.
Review of NRC Temporary Instructions (Modules 71707, 37700, 62703)...........................................
10.1 Verification of Quality Assurance - Diesel Generator Fuel Oil (TI 2515/93)..............................
10.2 Verification of BWR Recirculation Pump Trip (TI 2515/95).......................................
10.3-Verification of Mark I Containment Wetwell/Drywell Vacuum Breaker Modifications (TI 2515/96)..........
11. Review of Licensee Self-Assessment Activities (Module 40500)..........................................
12. Management Meetings (Module 30703)........................
Attachment I - Persons Contacted Attachment II - Facility Tour Findings by Regional Administrator
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i DETAILS-1.0 Summary of Facility Activities On December 21,-1988, the NRC Commissioners. voted unanimously to endorse the NRC staff's proposal to fermit the licensee to restart the Pilgrim Nuclear Power Station. On December 30,1988, - Mr. William T. Russell,, the Regional Administrator for Region I, approved the NRC ' Restart Assessment Panel's recommendation to release the licensee from the first Power Ascen-sion Program approval point (initial criticality to 5% rated ' power). The , program includes NRC Regional Administrator approval points at 5%, 25%, 50%, and 75% of full power, as well as a licensee written report and NRC' review after completion of testing at full power.
On December 30,1988,- at 9:54 p.m., the Pilgrim reactor achieved criti-cality Due to neutron monitoring system problems the licensee; began' a controlled shutdown at 10:14 p.m. the same day and placed the reactor in ' cold shutdown condition.
The licensee replaced failed intermediate. range , neutron monitoring detectors 'and the plant returned to criticality. at 5:05 p.m.
on January 2, 1989.
The licensee subsequently conducted reactivity. manipulation training in order.to. satisfy NRC requirements.for the reactor operators with conditional licenses.
Following reactor. heatup and pressurization to 150 psig, the licensee successfully completed. Reactor Core Isolation Cooling (RCIC) and High' Pressure. Core Injection (HPCI) system flow tests.
On January 10, 1989, the licensee commenced.a controlled' reactor - shutdown after determining that the ' torus to reactor building vacuum breaker block valves. may not perform their containment isolation function following a seismic event.
The reactor was brought subcritical.at 9:10 p.m. and reached cold shutdown at 2:15 a.m.'on January 11, 1989.
The licensee commenced reactor startup on January 27, 1989, following modifications to the' air supply and ' accumulators for the vacuum breaker block valves.
During a subsequent surveillance on the air supply, the , valves were again declared inoperable due to increased air leakage.
In l accordance with the Technical Specifications, the licensee commenced a reactor shutdown at 9:55 p.m. January 27, 1989, and an Unusual Event (UE) was declared.
The reactor was subcritical at 10:15 p.m. and the UE was terminated at the same time. The plant remained in cold shutdown for the remainder of this report period while the licensee performed additional modifications and repairs to the air supply system.
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-NRC inspection activities during this report period began with forming the onsite. Pilgrim Restart Staff led by Mr. Clay C. Warren, Senior Resident ' Inspector and Restart Manager.
The Pilgrim Restart Staff is composed of the Pilgrim resident inspectors, resident ' inspectors from other plants, NRC' regional-based and headquarters-based inspectors and an NRC contrac-tor.
During' the. week of December 27, 1988, the inspectors performed ' startup inspections.
On December 29, 1988, the Pilgrim Restart Staff f implemented 24-hour shift coverage. This coverage was reduced to extended day shift coverage at times, consistent with reduced testing activity and . plant shutdown.. A representative from the Commonwealth of Massachusetts a .was"onsite on December 29 and 30, 1988, and on January 2,1989 to observe l the NRC Restart Staff activities, q 2.0 Restart Preparation Activities Th'e Restart Staff monitored the licensee's preparations for startup activ-ities on December 27-30, 1988.
Emphasis was placed on recent status changes (i.e.,. additions, deletions, priority changes, ' revisions) since the last NRC review, documented in Inspection Report 50-293/88-33.
The purpose was to verify 'that new items and changes had been appropriately i dispositioned and that the overall <tatus was acceptable to support safe i restart'of the facility. The review included safety system valve lineups, -outstanding quality assurance discrepancy reports, maintenance requests, safety evaluations and -engineering service requests.
The status 'of licensee actions on outstanding NRC Bulletins and 10 CFR Part 21 reports H was also assessed.
. l . 2.1 Review of Active Temporary Procedures l Internal licensee memorandum (PM 88-229), " Final Review of Generic
Issues from Restart Checklist #6," Item 6.B.02.723, directed licensee ' division managers to review outstanding Temporary Procedures (TP) to determine their potential impact on startup.
Included in the review was the evaluation of potential adverse operational consequences of-l l installed jumpers, lifted leads and off-normal system alignments '{ . l resulting from partially accomplished TP's.
This_ item has'been up- > l.
dated on a continual basis by individual divisions and periodically ) i by the Operations Review Committee (0RC) for several months.
The { ' inspector reviewed the TP index with each responsible division man- .i ager and verified completed TP's were being retired, or as appropri- ' ate converted to permanent station procedures. No partially completed y TP's were identified. The inspector also noted that active TPs were i properly tracked. Licensee actions with respect to TP status review
were timely and thorough.
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2.2 Review of Safety Evaluations The inspector reviewed open safety evaluations which were projected to be active during the restart.
The open safety evaluations were of sufficient technical detail and adequately addressed 10 CFR 50.59 evaluation' criteria.
The inspector had no further questions with regard to active safety evaluations.
However, the inspector expressed concern to the licensee with respect to the use of " conditional" safety evaluations.
The licensee had routinely placed conditions or limitations on the applicability of safety evaluations.
The conditions were typically conservatism directed by the ORC in the form of operational limitations, compensatory personnel actions or increased surveillance testing of affected components.
The inspector informed the licensee that a safety evaluation should be a definitive analysis of a plant condition with respect to its design basis as described in Technical Specifications (TS) and the Final Safety Analysis Report (FSAR).
If the existing condition is evalu-ated not to be within the design basis for all applicable modes of operations then appropriate regulatory relief such as proposed TS changes, FSAR revision, justification for continued operation or temporary waiver of compliance must be initiated. The licensee con-curred with this position and committed to not invoke " conditional" safety evaluations and to revise procedures as appropriate.
It should be noted that no active safety evaluations had conditional limitations.
The licensee issued an engineering department memo to reinforce this commitment. An additional followup will be conducted in this area under an existing outstanding item (87-45-04) which addresses the licensee's use of FSAR Appendix G for the determination of conditional system operability.
2.3 Temporary and Permanent Radiological Shielding Program The inspector reviewed the licensee's program for the evaluation, installation, periodic inspection and material control of temporary and permanent radiological shielding.
The shielding program is
implemented by the Radiological Technical Support Division (RTSD) in accordance with PNPS Procedure 6.10-008, " Installation and Removal i of Shielding." A recent Quality Assurance Department (QAD) surveil-lance (88-2.1-39) of the shielding program revealed a procedural q deficiency, in that the permanent shielding request form and the I permanent shielding log were not included as attachments to Revision 1 of Procedure 6.10-008. The inspector performed a plant walkdown of . permanent and temporary shielding installations and reviewed shield- ! ing records and determined all installed shielding had been properly requested, logged and implemented irrespective of the procedural l
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deficiency. _ Revision 2 to Procedure 6.10-008,- which. incorporated corrective actions to the QAD deficiency, was approved by ORC and was being prepared for' distribution at the. conclusion of the inspection period.
The revised procedure-effectively developed.the necessary direction and accountability to ensure continued ' positive ' control of - L the shielding program. The RTSD personnel responsible for the imple-I' mentation of the program were well versed in the procedural enhance-ments of Revision 2 to Procedure 6.10-008 and were in the process of upgrading shielding logs to facilitate a smooth transition to the L revised procedure upon issuance.
l l 2.4 Engineering Service Requests l Engineering Service Requests (ESR) are used by any person within the nuclear organization and sent to the Nuclear Engineering Department to request engineering or technical support.
The inspector reviewed a listing of ESR's opened. since the last NRC review. The inspector discussed various ESR's with appropriate man-agement personnel to determine their potential' impact on the plant restart.. Two levels of management review are utilized to identify those ESR's which could affect restart of - the plant.
These two levels of management review appear to be effective.
Engineering section managers were knowledgeable of the contents of the ESRs and . the determinations made with respect to their potential affect. on
plant restart. The inspector had no further questions.
2.5 Quality Assurance Discrepancy Reports The inspector reviewed selected outstanding quality assurance (QA) audit and surveillance. reports.
These included def.iciency reports (DR), non-conformance reports (NCR) and potential condition adverse to quality (PCAQ) reports.
These reports were reviewed to determine if the licensee had appropriately identified those items requiring licensee attention and action prior to restart.
The Potential Condition Adverse to Quality (PCAQ) report is issued to resolve suspected or actual conditions adverse to quality identi-fied by the departments not using other Quality Assurance discrepancy reports, and to identify actual or suspected failures to comply with ' NRC rules and regulations or the facility license.
The inspector reviewed PCAQ's opened since the NRC's last review and identified no restart concerns.
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] Deficiency repor.ts are issued by the Quality Assurance (QA) Depart-ment during. the' conduct of. audits and surveillance for conditions , ~ contrary to management policies and: procedures, regulatory require- ] ments or licensee commitments.
The deficiency status report as of - December 26, 1989, listed eight open DR's, two 'of which were con-- sidered to be required to be dispositioned.
These two ' DR's were properly' dispositioned prior to startup.
No restart. concerns were identified with respect ' to the remaining. open DR's. - Review of a sampling ' of recently. closed DR's. indicated appropriate corrective'- - actions-were taken to correct the problem and prevent recurrence.
Good followup was performed by the person originating the DR to, ensure the corrective actions had been taken.
I L Nonconformance reports are used by operations quality control per-sonnel to document and report nonconforming materials, parts or com-ponents identified. as a result of receipt, installation and.other inspections.
Review of NCR's showed only four open NCR's.
These NCRs were written against items not installed and therefore, will not affect restart activities.
No discrepancies were identified.
~ 2.6 Failure and Malfunction Reports The failure and malfunction report (F&MR) is used to document and evaluate failures, malfunctions and abnormal operating events.
The inspector's review of F&MR's identified no additional items requiring licensee action prior to restart.
The inspector also reviewed. the licensee's methods of corrective action, root cause determin' tion. and item closeout.
The inspector a determined that the licensee had a good understanding of the root cause and appropriate actions were taken.
The inspector had no further questions in this area.
2.7 10 CFR Part 21 Report by Limitorque Concerning Elevated Ambient Temperature Effects on RH Insulated DC Motors On November 3,1988, Limitorque Corporation informed the licensee of the issuance of a 10 CFR -Part 21 notification concerning elevated ambient temperature effects on RH insulated DC motors.
Limitorque determined that in some cases SMB valve actuators with RH insulated motors may not develop full rated starting torque at elevated ambient temperatures, due to resultant DC motor resistance increases.
Limitorque recommended that the licensee review their DC motor oper-ated valves to determine if any of the listed RH insulated DC motors _ _ ____- -__-_-- -____
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Limitorque also requmea that the -- li cen see identify the order number and serial ' number from the actuator name-plate, the maximum ambient temperature and. valve requirement.
Limitorque also noted;that they have had no reported failures as 'a.
result of:this' phenomenon.
.The licensee reviewed their records on SMB valve actuators with RH { insulated DC motors installed in safety-related applications and determined that two motor-operated valves -(MOV) fit the. vendor's screening; criteria. These two valves were 2301-5 (HPCI steam line isolation. val.ve) and 2301-8 (HPCI injection line). The licensee for-warded'the information to Limitorque, and requested an analysis be performed to determine if the two DC M0V's would operate properly in the.high tempe'rature environments described in an attached. tempera- _ ture profile for each valve. Likewise, BECo ^ also requested an. evalu- . ation of operation of the valves at less than rated - voltage during the high ' temperature conditions. As of December 30, 1988, the licen-see had received verbal confirmation from Limitorque' that the-reported problem -did not apply to the referenced motor operated valves. Based.on this, the inspector concluded that the ' requirements of this Part 21 would not affect plant startup.
2.8 Review of Maintenance Requests The licensee's Work Prioritization Review Team (WPRT) meets daily to assign. priority to..each Maintenance Request (MR) and is composed of representatives of various station groups, including maintenance, operations, outage. management, construction management and fire pro-tection. The inspector attended a WPRT meeting on December 28, 1988.
Ten MR's were reviewed, two were identified as restart MR's and work was properly completed prior to startup.
The inspector also reviewed the current list of outstanding MR's to ensure that they had been properly prioritized and scheduled.
Two MR's which had been designated as non-startup items addressed defic-iencies in the emergency lighting system. The emergency lights are needed in order to facilitate operator actions to perform safe shut-down from outside the control room in the event of a loss of-station power.
Af ter questioning by the inspector, these two MR's were up-graded for completion prior to plant startup.
Repairs to the teergency lights had beer, delayed due to the lack of spare parts for the units.
Seven energency lighting units were determined by the licensee to be Inoperabla. A subsequent walkdown
of' the inoperable lighting units by the licensee demonstrated five ' ,, - _ _ _ _ - _ _ _ _ - - _ _ _ _ _ - _ _ _ _ _ _ _ -
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. y , I-of:the emergency lights were supported by adjacent or near vicinity I . lights. The inspector considered this acceptable to meet the intent of emergency lighting. The' two remaining inoperable units were used - ~ ' , to illuminate. stairways leading to the safe shutdown panels, and were determined by the licensee.to be necessary for plant startup. These
- g two emergency lighting. units were' repaired prior to ' plant startup.
After identification of the misprioritized MRs the licensee aggress-ively pursued the repair of. the. emergency lighting units.
The inspector' concluded that the required lighting units were operational r prior to startup.
With the exception of the MR's associated with emerging-lighting,.the inspector concluded, that' the : licensee had properly prioritized out-standing maintenance activities to support initial plant operations.
2.9 Surveillance Program Status The.' licensee tracks the surveillance program status as detailed in - Procedure No. 11. 8, " Master Surveillance Trackit g Program (MSTP)".
Elements ~ tracked include a listing-of all scheduled ' surveillance, windows of opportunity to perform tests and. methods to. identify late and missed ' surveillance procedures to management for increased visi-
bility and corrective actions.
The inspector reviewed the licensee's MSTP-and. evaluated a sample of technical specification surveillance requirements to determine if they.were in agreement with the MSTP, No discrepancies were identified.
2.10 Safety System Walkdowns In assessing-the plant's readiness for return to power operations, a review of emergency. core cooling system valve lineups. was performed.
A review of the licensee's current completed valve lineup for the low . pressure coolant injection system and high pressure coolant injection system was performed.
In addition, the inspector completed a walk-down and ver1fication of selected valves in each system.
This was performed with the aid of a nuclear plant cperator who physically checked system valves for the inspector during the walkdown. Na dis-crepancies were noted.
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2.11 Lifted Lead and Jumper Log Electrical lifted leads and jumpers at the ' station lare controlled by temporary modifications, maintenance requests and station procedures.
Due to the ' various. approved methods of performing lifted leads and _ jumpers, a central tracking system.is used to allow an operator to quickly. assess'.- status at.a given time.
The inspector reviewed the licensee's lifted lead 'and jumper. log as well as the licensee's con-- trolling procedure'1'.5.9.1, " Lifted Leads and Jumpers" to ensure com-pliance with the procedure.
Lifted leads and jumpers were noted. to be' placed' in accordance with procedural requirements and documented in the log. No discrepancies were identified.
2.12 Operations Review Committee'(ORC) Activities The. inspector reviewed recent ORC meeting minutes, interviewed the ORC Chairman _ regarding. 0RC restart readiness reviews and attended an - ORC meeting on December 28, 1988.
The committee appeared _to be functioning acceptably te support plant restart.
3.0 O_pe rati on s - - 3.1 Sustained Control Room Observations Based on over 500. hours of around-th ?-clock on-shif t observations during December 29 - February 5,1989, the inspectors determined that control room activities were conducted in a professional manner.
Communications in the control room were clear and formal. Operators typically. repeated back instructions which assured their understand-ing of the instructions. The flow of information among shift person-nel. was good, such that all members were aware of plant status and planned evolutions. Shift turnovers were conducted in a formal man-ner.
Appropriate ' information about system statu.; and work in pro-gress was conveyed to the on-coming shift through individual operator _ turnovers and pre-shift briefings.
The pre-shift briefing by the offgoing Watch Engineer covered encountered problems and upcoming _ evolutions in sufficient detail as to keep the on-coming personnel abreast of cVerall plant status.
Attendance at these briefings was consistert ar.d included representatives from Chemistry, Health Physics, and Outage Management groups'. Shift staffing level has been adequate.
The licensee has staffed I a four-shift. rotation with three seniar reactor operators (SR0s) and two cenctor operators (k0s) per shift, Addition of an extra SRO to each shift appears to have strengthened the shift organization with added experienca. Currently, only 8 40s have unrestricted licenses < ( ) , l_.________m________.___mm___ __.m._.___.___.m____..___m-__._m_ _. _ _ _. _ _ _____m.
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9 since.the 13 newly licensed R0s are' performing ' limited' licensed duties, pending completion of on-watch training and reactivity manip-ulations to be conducted - during the power ascension program.
At an appropriate point after restart, the licensee intends'to implement-a six-shift rotation of_ two SR0s Land 2 R0s per shift.
There are also sufficient non-licensed equipment operators'to staff six shifts.
~ The control room operators were attentive to their panels, alarms and indications.
Response to alarms and system parameter trends - was - ' appropriate. Operators were familiar with normal, abnormal 'and alarm , response procedures' and utilized them appropriately.
The. control. ' room staff generally exhibited a safety conscious and conservative-attitude.
The Technical Specifications (TS) were' ~ conservatively - ' applied. Administrative requirements were generally met.
The inspectors noted that the operations ' staff experienced some'dif-ficulties in transition from an extended outage to the operating mode. ' On occasions, the. control room ' operators and supervisors were slow in developing' a questioning. attitude, especially concerning-equipment' status.
-In certain cases, the on-shift personnel in the control room did not' know the reason for equipment being out of. ser-vice or.. the status of maintenance work on the equipment.
On January 4, 1989, with the reactor critical and primary containment - integrity required, an oxygen analyzer sample line containment isola-tion valve failed a surveillance test. The licensee delayed.taking the action required by TS~3.7.A.2.b for failed containment isolation , valves 'for 2 hours.until prompted by the inspector. While no time limit for initiation of action to close the redundant valve in the penetration is includM in ' the TS, this.2 hour delay was not war-ranted. Subsequent troubleshooting by the ' licensee revealed that the problem was the valve indication only.
The inspector discussed this event with licensee management and the licensee committed to empha- 'j size a conservative approach to determination of equipment operabil-i ity, and to instruct the operators that required actions should typically be taken within 30 minutes unless Technical Specifications specify a longer time.
The inspectors routinely reviewed various control room logs including ] the Limiting Cor.dition for' Operations (LCO) Log, the disabled Annun- ) ciator Alarm Leg, the Operations Supervisor Log,. the Reactor Opera-tors Log, the ' Lifted Lead and Jumper Leg, and the Component Leak Log.
, The inspectors noted that items were properly logged and tracked.
On
occasions. however, that control room operator logs were imprecise and activities such as 'those given in proceaure 1.3.34 " Conduct of Oper-ations" Section D were not always recorded. For example, 6 one hour
technical specification action was identified and it could not be ' ascertained by the shift inspector when the action was satisficd.
At the prompting of the inspector the licensee made a late entry to identify when the action was taken. License? management was informed of the noted weaknesses and has committed to review and take correc-tive actions.
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Control room operators. received good support from the shift technical advisors (STA) and administrative assistants.
The STAS were used in developing. failure and malfunction report; and maintaining various control room logs updated.
The administrative assistants do much of the administrative paperwork and help to reduce traffic in the con-trol room.
.. .0perations management, in'cluding the Chief. Operating Engineer and Operations Section Manager. provided. ef fective oversight of opera-tions. Operations management was observed touring the control room frequently and discussing plant status and evolutions with the Watch Engineer.
3.2 Plant Tour Observation: The inspectors made frequent plant tours and noted that the overall
material condition of rooms and equipment remained excellent during the report period.
Component labeling and tagging was good.
The licensee personnel interviewed during the tours (HP, security, oper- .ations, contractor) had experience in their positions and were know-ledgeable about their work and duties.
HPs were cognizant of work activities in progress. Housekeeping controls were being maintained during work in progress.
During' a tour of the reactor building 23 foot elevation, the inspec- . tor identified six reactor scram valve position switches which appeared misaligned.
The scram valve position switches illuminate scram lights on control room panel C905.
The scram lights are a backup indication to the operators of a scram and therefore do_ not serve any safety function.
The licensee reviewed this finding and generated maintenance requests to correct the switch alignments.
' On December 29, 1988, Mr. William. T. Russell, Regional Administrator, Region I, toured the Pilgrim Station.
Mr. Russell was accompanied by the Plant Manager, the Chief Operating Engineer and the resident inspector.
Attachment II of this inspection report lists the. items identified during the tour and the licensee's resolution of these items.
3.3 Review of Training Reactivity Manipulations Currently, there are 13 reactor operators (R0s) and a senior reactor operator (SRO) whose licenses are restricted to cold shutdown. condi- .tien.
To obtain unrestricted operating licenses, these individuals are required to perform five significant control manipulations which affect reactivity or power icvel per 10 CFR 55.3(a)(5). They also have to stand. training watches for at least one month at equal to or greater than 20*4 rated rower.
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Shortly after reaching criticality on January 2,1989, the licensee . l conducted training criticals for the 13. R0s in order to partially satisfy the control manipulations requirement. The licensee prepared Temporary. Procedure TP 88-89, " Reactivity Manipulations", to provide a description of indication and technique for the approach to criticality and return to subcritical operation.
The procedure required insertion of 45 in sequence control' rod drive notches between criticals which would preclude inadvertent critical-ity due to moderator temperature decrease.
The training manipula- , tions per TP 88-89 were observed by shift inspectors.
Each trainee was under direct supervision of an SRO. A shift training coordinator was also present in the control room during the training to assist the operators. The inspectors determined that the procedure was well developed and training activities were performed in a controlled manner.
> 3.4 Cold Weather protection An inspection was conducted on January 5,1989, to determine if the licensee had taken adequate measures to protect systems important to safety from extreme cold weather conditions to ensure operability.
The inspector verified the presence and operability of heat tracing space heaters, and insulation.
A walkdown of selected systems indicated that the licensee had taken adequate measures with the exception of the diesel-driven fire water pump room.
The space heater in the room was inoperable. This room J is located inside the screenhouse and contains the diesel driven fire water pump, its associated starting batteries, and portions of the main fire water header.
The licensee placed an additional space , heater in the room in response to the inspector's finding.
j On the same day, a fire sprinkler pipe froze due to the extreme cold weather and burst causing approximately one thousand gallons of water to drain within the condenser retube building.
The cendenser retube
building was used to support the main condenser tube replacement work ' during a previous outage. The licensee's radiological survev results indicated that there was ne spread of contamination within the build-ing. The condenser retuce building floor drains were collected in the miscellaneous radwaste. 'anks for processing.
There were no . releases to the env(ror nent and no personnel contaminations.
The i licensee subsecueatly replaced the damaged pcetions of the fire sprinkler system.
Genercily the licensev s prograr to protect against the effects of cold weather condit' ions was fomd to be acceptable.
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3.5 Review of Plant Events Rod Block Not' 0ccurring During Testing f-On January 9, 1989, the' licensee identified a problem concernintj . intermediate range monitor (IRM) wiring.that effectively bypassed.the source range monitor (SRM) " inoperable" and " upscale high" rod block. functions for "A" and "C" SRM's. The reactor was in the startup mode with? reactor pressure at about 140 psig at the time. -During perform-ance of a functional check of SRM's "A" and "C" per procedure-8.M.2-3.3,. " Source Range Monitor", the licensee determined that. the upscale trip and inoperative ' trip signals generated for the ' checks' did not -result in a rod block. At the time, IRM's. "A", "C" and "E" associated with SRM's "A" and "C" for the rod block function were on range 8 and IRM. "G" was on range 7.
This configuration should have generated a rod block since IRM "G" was on range 7, but it:did not.
The ' primary. purpose of the rod block is to ensure that the. correct. range of neutron instrumentation is.in service.
Subsequent investigation by the licensee. revealed that a wiring error associated with one of two rod block circuits of the reactor manual control. system (RMCS) caused the failure.
The wiring error bypassed two SRM's ('.'A" and "C") for the RMCS rod block function when IRM's "A" and "E" were on range 8 regardless of IRM's "C" and/or "G" range ' scales.
This' resulted in the licensee not fully complying with TS _ limiting condition for operation 3.2.C regarding the required degree of instrument redundancy during certain very limited-startup modes of operation since the initial plant operation.
The other rod block circuit was not affected by the wiring error. TS Table 3.2.C-1 iden-tifies the minimum number of operable SRM's as three.
Per TS the licensee placed SRM "D" mode switch to the STANDBY position thereby initiating a rod block. A Failure and Malfunction Report was initi-ated to document the problem.
The licensee determined the cause for the wiring error to be a per-sonnel error during original plant construction in that the wires were reversed during installation. Review of original plant drawings by thf. licenses showed the currently prescribed termination. points to be correct, and similar wiring associated with the.other rod bicek i circuit wa, visually inspected with no discrepancies noted.
The ' deficiently wired channel was corrected and the SRM functional test was subsequ*.ntly performed with satisfactory results. The inspector , reviewed licensee actions associated with this event and determined thac appropriate investigation and corrective actions had been taken.
' . The licensee promptly identified the reason for the failure to obtain the rod block, the cause of the error and verified that that no other discrepancies existed.
The irepector had no further questions.
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1 Secondary Containment Isolation During a b rveillance Test At 4:20 p.m. on January 15, 1980, the licensee e m rienced a second-ary containment' isolation and an inadvertent actuation of the "A" . standby gas treatment system (SBGT). The actuation occurred during , f the performance of surveillance procedure 8.M.2-1.5.8.1, "High Dry-well Pressure, Low Water Level and High Radiation Logic System A- , l Inboard Functional Test". The licensee's investigation revealed that , l during the performance of this surveillance test, the licensed oper- ' ator inadvertently turned the keylocked control switch, "Rx Bldg HVAC Iso Test Channel A" to the TEST position (to the right) i stead of
placing the switch to the TEST LOGIC position (to th. left) as ' instructed by Step 11 of the procedure.
The secondary containment isolation was reset and the "A" SBGT system was restored to normal standby status.
The licensee secured further performance of the test, conducted a critique and issued a failure and malfunction report.
The critique identified that human factors contributing to the error were the location of the control switch (height of switch is approximately seven inches above floor level), and the control switch terminology.
The surveillance test was successfully completed later that day.
This event was reported to the NRC via ENS at 5:10 p.m.
The inspector had no further questions.
Plant Shutdown and Notification of Unusual Event Due to Inoperable Vacuum Breaker Block Valves At 10:10 p.m.
on January 27, 1989, the licensee declared an Unusual Event (UE) due to the initiation of a plant shutdown required by the Technical Specifications.
After plant startup on January 27, 1989, reviews of the routine air supply surveillance data for the torus vacuum breaker block valve accumulators indicated increased leakage above the licensee-established limits.
Due to this increased leak-age, the licensee declared the vacuum breaker block valves inoperable and commenced a plant shutdown at 9:55 p.m.
The reactor was brought subcritical at 10:15 p.m. on January 27, 1989 and the UE terminated at that time.
Detail review of the problems associated with the - vacuum breaker block valves and the plant shutdown /UE are discussed in Section 7.2 of this report.
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'14' , 4.0 :Startup Testing Activities 4.1.
Shutdown Margin Calculations Technical Specification 4.3.A.1 requires that a sufficient shutdown margin be demonstrated following. a - re 'ueling outage.
To determine the shutdown margin (SDM), 'the license, used Procedure 9.16.1, "In-Sequence. Critical for Shutdown Margin Dt.nonstration".
The inspector reviewed the results of the procedure conducted during the initial reactor.. startup.
The test consistedof accurately determining' the reactor period during. initial criticality and monitoring recircula-tion suction temperature. The SDM was then calculated by using. the test data and.various reactivity values and correction factors in the General Electric Cycle Management Report.
Review of the test data - ' and calculation -indicated that the test was correctly performed and more than adequate SDM was present.
The inspector had 'no' further questions.
4.2 HPCI a'nd RCIC' Surveillance Testing at 150 PSIG Technical Specifications (TS) require that the high pressure coolant' injection (HPCI) system and the reactor core isolation cooling (RCIC) system be operable prior to exceeding 150 psig. To verify operabil-ity.of HPCI and RCIC, the licensee planned to perform manually initi-ated full flow rate tests in accordance with station procedures PNPS 8.5.4.3 and 8.5.5.3.
These procedures required that HPCI and RCIC be manually. started and reach rated flow.
.i The inspector determined that the procedures as written did not com-pletely verify the operability of these systems at 150 psig in that their initiating logic was not tested, nor was the time to reach . rated flow and pressure measured. After this concern was discussed with. the licensee, the procedures were changed to require system initiation via 'its associated logic circuit and to measure the time to reach rated flow and pressure. It should be noted that the licen-see planned to perform a simulated automatic actuation (i.e., initi-ation via logic) and cold quick start tests of HPCI and.RCIC at 1000 psig in acccrdance with the Power Ascension Test Program.
The inspector observed the performance of the 150 psig surveillance tests for both HPCI and RCIC.
All tests were performed satisfac- [ torily and the licensee declared both systems operable. The inspec- ! tor had no further questions.
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E 5.0' Surveillance 5.1 Radiological Environmental Technical Specifications A 1988 licensee sponsored contractor audit of Technical Specification (TS) surveillance implementing procedures identified that the Radio-l' logical. Environmental Technical Specifications (RETS) had not been-incorporated into the' licensee's Master Surveillance Tracking Program (MSTP), and in some, cases adequate procedures had not been. written'. ' A review of RETS was. initiated in response to the audit in October 1988.
During this review the ' licensee identified : th~at cumulative offsite dose contributions from radioactive. effluents had not been-calculated in accordance with the Offsite Dose Calculation Manual (0DCM) for the period of April through August, 1988 as ' required' by TS.
During ithis time only undocumented qualitative comparisons of ' current and past release data were made to determine if monthly doses were acceptable.
The cause was determined. to be an unfamiliarity with the requirements of the RETS by' licensee. personnel, and a lack of formal ' administrative controls for the scheduling and performance of RETS surveillance requirements.
This licensee identified viola-tion and the corrective' actions implemented were described in inspec-- tion report 50-293/88-33, Section 3.b.
Subsequently,.the licensee identified two additional instances of. failure to properly implement RETS..These two instances included: (1) The licensee failed to perform the 1988 garden census out to the required three mile radius.
The census was conducted to a radius of one mile.
The TS requirement had previously been expanded from one to three miles by a license amendment. Weak review of the amendment resulted in the failure to implement the revision. Licensee follow-up identified one additional garden which should have been evaluated; (2) The licensee failed to consider the contribution of gaseous tritium in completing monthly offsite dose calculations.
The cause was determined to be weak communication between the Radio-logical Protection and Chemistry Departments.
The licensee reviewed historical data and determined that the emission had no significant impact on the calculated doses, and that no TS limit had been exceeded.
. ' In ~ response to these problems the licensee elected to relocate the group responsible for environmental monitoring from the corporate engineering cffice to the site to provide for better communicatun with the balance of the organization.
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1 was. assigned,7 reporting. to the Deputy ~ Radiological Section. Manager, { to oversee review of -the RETS and development of more. formal admin-'-
istrative controls. The -licensee has identified each. RETS ' surveil-lance' requirement as. a line item in the MSTP.
Procedures have'been written to' implement each requirement.
The licensee's. approach to resolution' of _ this issue appears toibe appropriately focused and- . timely.
j The. inspector. noted that the above described licensee identified violations were determined ' not ' to be reportable under.10 CFR 50.73.
Review of the licensee's basis for this determination indicates that the judgement was. founded primari.ly on-three premises: ' (1) The - events would be included in the' 1988 Annual Radiological Environmental Monitoring Report; . (2).The failure 'to perform the TS surveillance did not result'in occurrence of a condition prohibited by TS (i.e., no dose limit wasexceeded); (3) The-surveillance requirement was solely administrative in nature.
No Limiting Condition for Operation (LCO) is ' rescribed p upon failure.to perform the surveillance.
No equipment was declared inoperable and therefore no LCO was exceeded.
The inspector questioned if the inclusion of the~ event in the annual report satisfies the reporting requirements of 10 CFR 50.73.
Fur-ther, the inspector expressed concern regarding the stated basis.for the deportability determination, and the underlying philosophy it~ suggests relative to the intent and application of RETS. The appar-ent intent of the RETS is to require licensees'to closely monitor the performance of the waste treatment' processes, and to take prompt' action if these treatment processes are'less.than effective in reduc-ing potential offsite exposure.
This action may include repair of existing equipment,' revision of operating practices to allow for_more , effective use'of equipment, or evaluation of the benefit of potential i hardware iraprovements' The-RETS should be seen as the framework for .I . a ~ sound offsite exposure ALARA program. The position that the sur- ' veillance requirements are scicly administrative in riature does not i ' appear consistent with the purpose of RETS. The inspector discussed the above concern with the licensee's Radiological and Compliance-S2ctiors.
This item will remain unresolved pending NRC specie. list. inspector , review of the technical adequacy of the licensee's newly generated l RETS surveillance implementing procedures, and the adequacy of the
licensee's approach to implementation of the RETS program in general j (UNR 88-37-01).
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._ _ _ -, .j ).- ' J 17-q The inspector also performed a brief review of the licensee's overall . process for implementing TS: revisions. The responsibility for ensur-ing. that TS changes. are appropriately ' reflected in -surveillance,.
q operating and maintenance' procedures has not-been clearly understood! by licensee personnel in the past. Some confusion between.the Modif-ications Management Group (MMG) and the. Compliance Group: existed.
j regarding the division of responsibility.
Recently= however, the i licensee has resolved this confusion by clearly assigning the task to the Compliance Group.. Applicable program procedures are being writ-ten or' revised. As described above, a contractor provided audit of.
all TS amendments through number 120_ was performed. in 1988.. To ensure-that additional amendments issued since completion of'.the.
audit' have been properly dispositioned, the licensee is reviewing.the intervening ' changes.
Licensee actions in this area have..been effective.
5.2 Routine Surveillance Tests The inspectors observed the following surveillance tests: 8.M.1-1A IRM Functional / Calibration; 8.M.2-3.3 SRM Functional; 8.M.1-4 'APRM Flow Biased Signal Calibration; 8.M.1-13 Main Steam Line High Radiation Calibration; . 8.M.2-2.5.6 HPCI Condensate Storage Tank Levels; 8.5.2.3 LPCI Motor Operated Valve Operability; 8.5.5.4 RCIC MOV Monthly / Quarterly Valve Operability Test; 8.7.1.5 Leak Rate Testing of Containment. Isolation Valves; It was determined that implementation of surveillance tests was generally well. planned, and controlled.
On occasions, ineffective communications between control. room operators and Instrument-and Control (I&C) technicians caused confusion during surveillance tests.
The licensee management agreed with the inspectors on the need to expand the " good communication practices" to other working groups (i.e. I&C) for formality and repeat backs. The inspectors will mon-itor this area in.a future inspection.
6.0 Maintenance and Modifications 6.1 Migh Pressure Coolant Injection (HPCI) System Gland Seal Condenser Hotweil Pump Replacement q During HPCI testing at 140 psig steam pressure en January 10. 1939, i the HPCI gland seal condenser level increased sufficiently to ficod and overlotd the gland seal exhauster motor.
Subsequent investiga-tion disclosed degradation in the gland seal condenser hotwell pump - (P-220). The impelier was worn and the casing eroded.
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\\ ! installed a new pump under Plant Design Changes (PDC) 89-05 and 89-08. The inspector observed installation of the new pump, associ-ated piping modifications, and post-work testing.
It was noted that adequate ALARA planning, and health physics coverage were provided for the job. Pre-job briefings for the maintenance personnel at each shift were detailed and thorough.
! In the past, the licensee had not predicated the HPCI system opera-
' bility on. the availability of the HPCI gland seal subsystem. However, the licensee changed that position based on results of a detailed review during this inspection.
Licensee calculations to determine peak HPCI room temperatures show that should the HPCI system be oper-ated without the gland sealing system functioning the room tempera-ture could rise above 130 F.
These calculations also assumed failure of one of the two room coolers.
The equipment in the HPCI room is only qualified to a mild environment (approximately 100 F) and there-l fore could not be assumed to remain operable at the calculated ele- ! vated temperatures.
As a result of this analysis, the licensee has determined that the gland sealing subsystem must be operable prior to declaring the HPCI system operable. To monitor performance of the gland sealing system, the licensee has added appropriate portions of the system to the inservice testing (IST) program and modified the system to enhance l testing.
Baseline data has been taken and future testing has been scheduled. The inspector had no further questions and considered the licensee's evaluations to be thorough.
) j 6.2 Intermediate Range Monitor (IRM) Detector Replacement During the initial startup on December 30, 1988, "B", "D", and "G" IRM failed to respond to the neutron flux in the core.
The licensee l placed the reactor in cold shutdown and began investigation.
Based on insulation resistance testing and voltage breakdown testing, it was determined to be detector failure. The licensee had replaced all eight IRM detectors during the last outage.
The licensee replaced the detector.s with spares from the warehouse.
The inspector revieed the associated documentation, including: -- The maintenance request package (MR 88-45-384); i Maintenance work plan; -- 9rocedure 3.M.2-5,13, "IRM and SRM Detector Changeout"; -- F l _ _ _ _ -. _.
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' -- Procedure 8.M.1-1,'"IRM Functional / Calibration check"; -- Procedure 3.M.2-5.6.11, " Checkout of SRM/IRM Retract Drive- -- Components"; Pre-job briefing documentation; --- Material' balance area transfer. form for special nuclear -- materials The inspector determined that these documents.were tech'nically ade. quate and thorough.
Post-work testing was performed '~per procedure '3.M.2-5.14 and 8.M.1-1.
Vibration test of the retract ' drive.
components was also completed. It appeared'that there was good over-sight by. health physics and quality control during the detector
replacement.
' The licensee is continuing with their root cause analysis on 'the detector. failure.
The inspector will continue to monitor licensee followup.
-7.0 Review of Generic Letter 88-14 7.1 Secondary Containment Integrity During ' evaluation of Generic Letter 88-14, " Instrument Air Supply System Problems Affecting Safety-Related Equipment," the licensee determined that the reactor building trucklock door inflatable seals were supplied by the non-seismically qualified instrument air system.
Since the-instrument air system is not seismically qualified,.it must; be assumed to fail during a design basis seismic event. This condi-tion is significant with respect to the reacter building inner truck-lock door inflatable. seal which constitutes a seismically designed-secondary containment penetration. Therefore, the condition and per-.
formance of the inflatable seal following a seismic e. vent directly impacts secondary containment integrity.
In order to determine the impact' of this scenario, on December 22, the licensee performed a secondary containment leak rate test.with the inner trucklock door ! seal deflated and the outer door open. A negative pressure of only 0.18 inchey of water was achieved using one train of the standby gas treatment system (SBGT). This failed to meet the reautred acceptance y criteria of 0.25 inches of water.
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i ! The licensee-subsequently designed and installed a passive mechanical . interference seal system in accordance with Plant Design Change-88-53.
On December 29, the secondary containment leak rate test was- .successfully conducted with the new door seal design in place and the '! inflatable seal deflated.
l The licensee's response to an-NRC initiative which lead to the iden- ! tification of this issue, and the subsequent corrective actions taken l to resolve the design deficiency were noteworthy.
The inspector had l no further questions, j 7.2 Inadequate Reactor Building to Torus Vacuum Breaker Isolation Valve Design ' Generic Letter 88-14, " Instrument Air Supply System Problems Affect-ing Safety-Related Equipment," requested that licensees evaluate the effects of a -loss of instrument air on the ability of safety-related . components to' perform their intended function. For example, during a seismic event, air operated safety-related components are assumed to
. fail due to a loss of the instrument air system. The. licensee evalu-ated the susceptibility of various safety-related systems to this type of failure and found that the reactor building to torus vacuum breaker block valves would not provide their required containment isolation function if instrument air was lost.
After determining that the air supply for the accumulators for the reactor building to torus vacuum breaker block valves was not seis-mically qualified, the licensee initiated a controlled plant shutdown at 9:00 p.m., on January 10, 1989. Licensee senior management deter-mined that it was prudent to shut down the plant -until the vacuum breaker block valve design was completely evaluated. The inspector considered this decision to be conservative and evidence of a sound operational safety perspective.- Cold shutdown was reached at 2:15 a.m., on January 11, 1989.
To prevent torus failure due to excessive external pressure, the reactor buildir;g to torus vacuum breaker block valves open to allow the in-series mechanical vacuum breaker to equalize, the pressure oetween the torus atmosphere and the reactor building atmosphere.
To isolate priinary containment during the initial phase of a loss of coolant accident each of the two reactor building to torus vacuum breaker block valves closes to prevent lcakage from the primary con-tainment.
These block valves (AO-5040 A&B) are held shut by air pressure and will fail-open on a loss of air pressure. TFus, a suf-ficient air supply is needed to ensure their containment isolation function.
Individual safety-related accumulators (4 cubic feet) provide a small volume of air to each valve if instrumeat air becoines inoperable.
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l Testing performed after plant shutdown proved that this small volume i was insufficient to cope with normal system leakage for the design basis mission time of 30 days. Minimum closing pressures were estab- , lished individually for each valve.
Licensee testing showed that ' minimum pressure to close valve A0-5040-A was 80 psig, and valve ! A0-5040-B was 62 psig. The difference in the minimum pressures was 'l due to dissimilar. valve operator orientation. Normal instrument air l pressure is 108 to 113 psig. Hence, only a 28 psig drop in supply pressure would result in the inability to close the limiting valve.
The licensee concluded that a air system modification was necessary to insure that the 30 day mission time was met.
Senior management believed the plant staff could complete their design. review and install any needed modifications within 5 days of reaching cold shutdown.
The plant staff responded quickly by developing a design change to install two 54 cubic foot low pressure accumulators each with its own high pressure makeup system, in series with tho existing 4 cubic foot accumulators.
Although the plant staff efforts were prompt, design options were not fully developed and many field changes were necessary to complete the modification.
The final design required operator action to make up for any losses from the low pressure system by adding air from newly installed, seismically supported, high pressure bottles. The high pressure por-tion of the system is located outside the reactor building and will be accessible during a post accident environment.
While operator action to recharge the low pressure accumulators is acceptable, the design compromises associated in part with schedule constraints shortened the operator response time requirements from 30 days to the final design operator response time of 5 days.
The inspectors reviewed the revised operator response time and concluded that the relatively simple actions to replenish the accumulator air system each five days would not result in any excessive operator burden in a post-accident situation.
The licensee completed the air system modifications and testing and commenced plant startup on January 27; the plant was critical at 2:12 p.m.
Subsequently, in reviewing the air supply fill data for the accumulators, the licensee noted increased air system leakage.
The licensee determined that the increase in leakage had occurred when the low pressure accumulators were recharged after leak rate testing was complete.
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i l L The. recharging evolution brought system pressure' to 125 psig which ' E was only 5 psi below the system relief valve setpoint of 130 psig'. , While there 'is no evidence 'that the relief valves lifted they both showed seat leakage when subsequently tested.
This new leakage path doubled the. total system leakage from the leakage rate previously determined and raised the leakage rate above the licensee's own con-servativcly established limit. for system operability.
The reactor building to torus vacuum breaker block valves were declared inoper- ' able and a plant shutdown was commenced at 9:55 p.m.
At 10:10 p.m., on January 27, the licensee declared an Unusual Event (UE) due to the: initiation of'a. plant shutdown required by Technical -Specifications.
The reactor was brought subcritical at 10:15 p.m. on January 27, and the UE was terminated at that time.
Had plant staff taken the time to fully develop the modification, to perform mere comprehensive testing, to thoroughly ' evaluate system operating margins,. and:.to ~ fully develop associated administrative controls, the plant shutdown due to excessive air system leakage might have been prevented.
The licensee considered the excessive leakage to be due to the relief l . valves on the two accumulators lifting and reseating.at erratic pressures.
The inspector observed the torus to ' reactor building vacuum breaker blo'ck valve leak tightness testing on January 30..The system was filled and checked for leakage at various fittings and mechanical joints by using a Helium detector. No appreciable leakage at these points was found.
At several points during the redesign effort the licensee revised the leakage acceptance criteria and operator re'sponse frequency for the accumulators.
In addition, the licensee is evaluating the possibil-ity of relocating the relief valves outside the normal system bound-ary.
Licensee efforts were ongoing and the inspector will continue to monitor licensee activities during the next inspection period.
7.3 Conclusions The licensee's review in response to NRC Generic Letter 88-14 was thorough and well conceived.
In conducting their evaluation the licensee identified the two design deficiencies documented above and. took prompt conservative action to address both items.
The design change which corrected the secondary containment truck lock deficiency was implemented in a timely manner and was well thought out from cenception to implementation.
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In the case of the torus vacuum breaker block valve modification the licensee's decision to modify the system was conservative and 'showed excellent safety perspective.
It must be noted that the licensee's initial attempt at implementation of the design change did not go smoothly, over fifty field revision notices were needed to complete the modification.
The inspector attributes these difficulties to overly aggressive goal in completion schedule set by upper manage- , ment, and weak vertical communications.
l Efforts to design the system, procure components, construct and test the system were all affected by schedule demands.
Although the initial design met all code and license requirements, operating characteristics and acceptance criteria were extremely restrictive and led directly to the shutdown and declaration of an UE on January 27, 1989.
The role that the Onsite Review Committee (ORC) took throughout the modification process met the requirements of Technical Specifica-tions.
ORC reviews of the design change were deliberate and thorough. Members of ORC recognized the operating constraints of the modification, however, weak organizational communications prevented upper management from recognizing the consequences of those constraints.
The licensee held a management critique of the vacuum breaker modifi-cation process in an attempt to identify root causes that led to the system failure.
Licensee management concluded that no single root cause resulted in the UE but that multiple technical design weak-nesses led to the system leakage and inoperability. Licensee manage-ment also reached the conclusion that the common factor in all the technical causal factors is that they were the result of aggressive demands on schedule. The licensee was extremely frank and self crit-ical throughout their self assessment process and the inspector has concluded that the assessment was very good.
Licensee management performance in the assessment was well focused ond came to a well balanced conclusion.
8.0 Radiological Controls Radiological controls were observed by the inspectors on a continuing basis throughout the reporting period.
In addition, a health physics specialist also reviewed portions of the licensee's radiological protec-E tion program during this inspection period.
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'8.1 -Radiation Monitoring Systems i
The l inspector reviewed the calibration' and operability status of the ' area radiation monitors. (ARM) and selected ventilation. system radia-tion monitors through. tours of the plant, - review of records and interviews with. calibration' personnel and systems engineers.- All ~j fifteen channels. of ARM were found-to be in, calibration and func-
-tional.
The next six month calibration will occur in~ March 1989.
-{ There are no repairs or improvements planned on the systec.~ prior to-j power operations.
Procedure No. 6.5-160,. " Calibration of the Area ' Radiation ~ Monitoring System,". appears to be. adequate.
Calibration dates are included in the Master Surveillance Tracking Plan (MSTP) computerized work schedule.
The licensee had installed special "high radiation area' monitors" in - + the. traversing incore probe (TIP) room and several radwaste -loca-t i o n s ~. These were declared not functional several years ago when spare parts became unavailable.
The licensee was requested by the-inspector to evaluate whether these monitors are needed and to evalu-ate. the personnel-exposure to effect repairs.
The. inspector noted - that these monitors are not required by the Technical ' Specifications.
~ ' The inspector will review ' this item during a future inspection.
8.2 Special Radiation Surveys Because of the extended shutdown and extensive equipment and plant modifications made during the outage, special radiation surveys will be conducted during plant startup and. initial operations to detect shielding changes. Airborne radioactivity is continuously monitored by Beta Aerosol Beacons (BAB) placed.in certain plant areas. When reactor steam is fed to an area for the first time there are several grab -samples 'taken of airborne particulate and gaseous activity to detect steam leaks from the equipment.
Special gamma and neutron dose rate surveys are conducted also to detect shielding changes.
These are repeated as the power level and general area dose rates increase.
Implementation of these special surveys is accomplished
through " standing orders" issued to the health physics (HP) tech-nicians. 'HP supervisors continue to review survey results.
The inspector concluded that the approach was adequate.
8.3 Control ~of Locked High Radiation Areas-t The inspector reviewed the licensee's control of locked nigh radia-tion areas (i.e., greater than 1000 mrem /hr general area).
The inspector reviewed the licensee program to ensure proper control of radiation areas during the power ascension program.
Because the plant has been in an extended outage status, many of the radiological - conditions associated with power operations have not been encountered in excess'of 30 months. In order to ensure readiness to survey, post and control access to high radiation areas resulting from normal l l L.
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power operations, equipment malfunctions and maintenance activities, the licensee developed a comprehensive high radiation area control plan for the power ascension program.
Initially, Radiological Oper-ations Division personnel retraining was provided to emphasize the responsibilities and regulatory requirements associated with the control of radiation areas. A historical review of past operational cycles was performed by the licensee to identify plant locations that-required high radiation area postings and to review reactive mainten-ance activities in which special radiological protective actions were invoked.
The licensee also contacted other similar BWR facilities to gain and exchange operational experience and industry initiatives in ! the radiological protection area.
Additionally, responses to past i Notices of Violation and NRC concerns regarding radiation area con-trols were reviewed to ensure corrective actions had been properly incorporated.
In conclusion, the inspector determined via plant walkdowns of exist-ing posted radiation areas, review of the power ascension high radia-tion area control plan, and interviews with Radiological Section per-sonnel, that proper professional attitudes and programmatic controls are present to provide positive access control to high radiation areas during power ascension.
The control of locked high radiation areas is generally adequate.
however, a locked high radiation area door to the radwaste building truck lock (RBTL) was found unlatched during a licensee's shiftly surveillance on February 3,1989.
The door, one of three personnel access paths to the RBTL, was not normally used.
Radiation surveys performed by the licensee showed a general area radiation level of 250 mrem /hr; the highest radiation level of 1200 mrem /hr was at the top of a sludge liner. The pocket optical dosimeter readings of all personnel logged into the process buildings from the time the RBTL became a locked high rtdiation area on January 31,1989, until the door was relocked on February 2,1989, were checked by the licensee and no substantial exposure was noted.
The inspector noted that the licensee's response was prompt and their investigation was thorough. The identified root cause was personnel error by technicians who checked only the door used, rather than all doors, each time work ceased in the RBTL.
The following proposed ccrrective actions were either taken or will be taken by the licensee: The technicians who did not check all doors upon exiting the - RBTL were counselled; - Discussion of this occurrence and the need for greater sensitiv-ity to locked high radiation area controls was conducted with ! l all radiological technicians on site; _ _ _ - _ _ _ _ _
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Checking of all doort to a locked high radiation area upon exit- -
ing the room will be explicitly incorporated into station pro - ! cedure 6.1-012, " Access to High Radiation Areas;"
} - The need for greater sensitivity to high radiation area controls will be incorporated into initial and' requalification General Employee Training.
Inadequate control of locked high radiation areas had been an area of previous NRC concern. Notices of Violation had been issued in the _ past during inspections 50-293/87-03, 50-293/87-11, 50-293/87-19 and
50-293/87-57 which addressed these concerns.
In regard to these violations, the licensee instituted extensive corrective actions which have been successful.
The Integrated Assessment Team Inspec-tion conducted in August 1988 determined that the licensee's program
in this area, including the shiftly surveillance' on locked high radiation area doors was effective.
l l Based on the above, the failure to comply with the requirements of Technical Specification 6.11 and implementing procedure 6.1-012 is considered a licensee-identified violation.
Consequently, no Notice of Violation will be issued (88-37-02).
The inspectors will rou-tinely monitor this area during the power ascension program.
) 9.0 Followup on Previous Inspection Findings { Closed) Unresolved Item 88-31-01, Discrepancies in Control Room High Efficiency Air Filtration (CRHEAF) System Procedures.
During inspection 50-293/88-31, five discrepancies in the CRHEAF system operating and sur-veillance procedures were identified.
The licensee's System Engineering Group subsequently reviewed the concerns and processed appropriate proced-ure revisions. During this report period the inspector evaluated licensee corrective actions, including the procedure changes and supporting engi-neering calculations.
Following is a summary of licensee actions in response to each of the five items: (1) The CRHEAF system operating and test procedures did not incorporate q position verification and securing of system manual dampers.
The i licensee revised procedure 8.7.2.7, " Measure Flow and Pressure Drop Across Control Room High Efficiency Air Filtration System," to adjust system manual dampers and to apply HVAC balance stickers to assure proper positioning is maintained.
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(2) Technical Specification 4.7.B.2.c requires that the CRHEAF system inlet heaters be capable of an output of at least 14 KW. Procedure m 8.7.2.8, " Perform a Functional Test of Humidity Controls and Inlet Heater Capabilities of the Control Room Air Filtration System," did not require that this output value be met.
The licensee has revised the procedure to require that measured output exceed 14 KW.
(3) Adequate acceptance criteria were not established in Procedure 8.7.2.8 for the single heater element in each train which is con-trolled by the humidistat.
Existing acceptance criteria required a minimum output of 3.8 KW.
This value however, was calculated using data obtained at the normal bus voltage of 490 to 500 VAC, without } adjustment to account for potential degraded voltage conditions.
! In response to the inspector's concern, the licensee collected data using a clamp-on ammeter, calculated the heater output, and adjusted the output to account for potential degraded voltage conditions.
Results of this initial testing indicated that the heater could not i fulfill its intended function.
On November 18, 1988, the licensee ! declared the system inopersble and notified the NRC via ENS of the deficiency.
The licensee implemented a Temporary Modification (TM) connecting a second heater element to the humidistat controlled cir-cuit, along with the original element, increasing the KW output to an acceptable level.
Subsequently, the licensee revised the test method to include use of in-line ammeters.
In addition, engineering calculations were per-formed which support establishment of higher expected voltages during degraded voltage conditions. The combination of a more accurate test method, and revision of the acceptance criteria to account for the new expected minimum voltage indicates that the original single heater element is capable of satisfying the design function.
The licensee plans to remove the TM and return the system to normal. (4) Proce:iure 7.1.30, "HEPA Filter and Charcoal Performance Test Pro-gram," did not contain any quantitative acceptance criteria.
The licensee has revised the procedure to include appropriate ecceptance criteria.
(5) Procedure 8.E.47.1, " Control Room /Radwaste Filtration System Instru-mentation Calibration / Logic Functional Test," requires that low ficw be simulated to effect a standby train auto start.
No method of i simulating the low flow was specified. Technicians indicated that it L would be simulated by lowering the setpoint or disconnecting the
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I . instrument tubing. The inspector questioned the impact of,this prac-- tice on instrument calibration. The licensee verified that the set-point had not been manipulated since the last instrument calibration.
.The procedure is. currently undergoing a major revision which will be completed before its next implementation..The licensee stated that as_ part of, this revision the_ procedure will be changed to specify that the low-flow signal and resultant ' standby train auto-start will be generated by securing the operating fan.
Corrective. actions ~ implemented by the. licensea in response to the above - -items are adequate.
The inspector had.no further questions regarding these procedure revisions.
During the course: of the licensee's review in response to item number 3 above, it was noted that: the Technical Specifications. require only that . the total heater output of 14 KW be verified.
Only one of the four ele- .m'ents is controlled by the humidistat and functions to reduce inlet air relative humidity to.an acceptable level.
The. remaining three elements serve'only for comfort control.
It appearsithat the TS should include an output - KW value for the single required heater element t'o ' ensure that it-has not degraded.
Technical ~ Specifications require periodic measu'rement of system flow and pressure but do.not require verification.that the sys-tem is capable of performing its basic design function, maintaining positive control room. pressure.
In both these insta'nces the licensee has performed adequate testing to demonstrate that the. design functions. are maintained.
The inspector however, expressed concern to licensee manage-ment that if TS are not adequate or, accurate they should be appropriately revised.
The licensee acknowledged this observation.
10.0 Review'of NRC Temporary Instructions-10.1 Verification of Quality Assurance - Diesel Generator Fuel Oil (TI 2515/93) . The objective of this inspection was to assure that diesel generator (DG) oil is included in the BECo Quality Assurance Program under 10 i .CFR Part 50, Appendix B requirements. Consumable items whose quality.
'. is necessary for functional performance of safety-related compo.nents, such as DG fuel oi.1, are classified as safety-related and are subject to the applicable provisions of 10 CFR Part 50, Appendix B.
The.
Pilgrim Nuclear Power Station Q-list, Revision 17, identifies those items that are safety-related. The Q-list includes System Number 61, " Diesel Generators and Auxiliary Systems," and identifie7, fuel and i lubricating oil as being included as part of the "Q" list.
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i l l Licensee Quality Control Instruction 7.07 requires that the commer-l-cial quality ' item evaluation (CQI) and. Material Procurement and Receiving Instruction (MPRI) number 52 be used to assure the quality of the DG fuel oil.
The CQI and MPRI No. 52 specify that the fuel oil conform to the specifications. in American Society for Testing and Materials (ASTM) D975-81, -" Standard Practice -for Manual Sampling of ' Petroleum and Petroleum Products," This is consistant with the DG fuel oil surveillance requirements of Technical Specifications 4.9. A.I.e and 4.12.B.1.i for the safety-related. DG's, and the : dies'el driven fire pumps. Based on the above, the inspector determined that ~ ' adequate measures are being taken to assure the quality of DG fuel . oil, : as-specified in the Technical Specifications. This inspection closes TI 2515/93.
10.2 Verification of BWR Recirculation Pump Trip (TI 2515/95) .The objective of this inspection was to verify the installation of the recirculation pump trip (RPT) function for low reactor vessel i water level or high reactor vessel pressure..The purpose of the_ RPT-is to-significantly limit the consequences of-an anticipated trans-ient.without scram (ATWS) event. A trip of the recirculation pumps on either of the' parameters identified above causes an increase in-l . the moderator voids in the reactor core; therefore, power and. press-ure surges which might occur during an ATWS event are substantially reduced due.to the negative reactivity resulting from the RPT.
Walkdown' of -the ATWS modifications to the RPT, including inspection of 'B' recirculation pump MG set field breaker was completed and i documented in inspection report 50-293/86-24.
l ! The NRC staff approved the licensee's proposed RPT including a change to 'the Technical Specifications, by letter dated May 12, 1980. This letter included Amendment: No. 42 to the Technical Specification and the supporting safety evaluation (SE).
The SE approved tripping the recirculating pumps on high reactor vessel pressure or low-low water level.
The in spec te:- reviewed flant Clesion Change Request (PDCR) 79-25, "ATWS Reactor Recirculation Pump T. rip System." lhe PDCR included a closecut nemo dated Jcnuary 11, 1984, which included a list of all applicable as-built drawings. Based on the previous system walkdown, issuance of revised Technical Specifications and the closecut memo, the inspector determined that the recirculation pump trip has been properly installed and is operable.
This inspection closes TI j 2515/95.
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30 j I .! 10.3' Verification of Mark I Containment Wetwell/Drywel1~ Vacuum Breaker ..' Modifications (TI 2515/96) -The object of this inspection was to verify that modifications to the ! Pilgrim Mark I (MKI) containment vacuum breakers in' response to'
i Generic Letter (GL) 83-08 had been completed and used the correct materials.
The GL' requested licensees to perform plant specific , analysis to determine the adequacy of the MKI containment vacuum ' ' breakers to withstand chugging and condensation oscillation loads-which would result during a loss of coolant accident (LOCA).
. I The licensee's response to GL 83-08 proposed modifications.to..the Pilgrim Mark I containment vacuum breakers.
The modifications included changes and new materials for the vacuum breaker pallets, hinge shafts, arms.and hinge arm studs.
The NRC staff approved the proposed modifications in a safety evaluation issued by letter dated
January 15, 1987.
' The inspector reviewed Plant Design Change (PDC) 83-19G, Rev. O, " Torus Vacuum Breakers Upgrade." The inspector reviewed the purchase order for the material used in the modifications and determined that the materials were consistent with that specified in the licensee's proposed modification package submitted to the NRC.
Based on the l above findings, the inspector has determined the modifications were completed and the proper materials were used. This inspection closes q TI 2515/96, 11.0 Review of Licensee Self Assessment Activities The inspectors routinely monitored the licensee's inplace programs to assess facility and personnel performance.
The licensee has implemented a. formal peer evaluation program of routine personnel performance monitor-ing.
The individuals selected for the peer evaluator program-are selected from the onsite organization, receive training on performance monitoring techniques and are assigned to monitor specific activities.
The peer evaluator program provided twenty-four hour opvations monitoring. during all periods when the facility was critical, as well as routine audits of i other areas of facility activities.
The peer evaluators held regular l debriefings with audited organizations to discuss identified strengths and weaknesses.
NRC inspectors who attended these debriefing sessions observed that the findings, both positive and negative were discussed in a frank, open atmosphere.
The audited organizations have generally been receptive to this process and the training, resolution and closecut of ! findings has been timely and thorough.
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The' inspector also noted greatly increased presence.of management in;the plant 'throughout this period. Routine' presence.of middle and senior level -_ management. in ' the control room and in the. plant was noted.
Management ' oversignt and control of = routine and abnorma1L activities showed' clearly
that the licensee has set high performance standards.
The licensee's quality assurance organization.hasfalso developed a special ' audit program Efor the duration. of the power' ascension plan.
The inspec- . tors noted an increased presence of quality assurance and quality control: personnel throughout:the inspection. period.
-Management efforts in assuring. high standa'rds of. facility and personnel-performance were ' evident throughout this inspection period. The. licensee-was highly self-critical in this self' assessment period and overall man-agement performance was' good.
12.0 Management-Meetings . At periodic Lintervals during the. course of the inspection period, as well-as. : af ter'.the close of the inspection, meetings were held fwith senior facility' management to discuss the~ inspection scope and preliminary find-ings of. the resident' inspectors.
No. written material was given ' to the licensee that was not previously tvailable to the public, a
- _ - _ _ - _ _ _ - _ _ _ _ _ _ ' t',s ATTACHMENT I Persons Contacted R. Bird, Senior Vice President - Nuclear
- K. Highfill, Site Director R. Anderson, Plant Manager D. Eng, Outage and Planning Manager E. Kraft, Training Department Manager D. Swanson, Nuclear Engineering Department Manager D. Long, Plant Support Department Manager J. Alexander, Operations Section Manager J. Jens, Radiological Section Manager J. Serry, Technical Section Manager R. Sherry, Maintenance Section Manager L. Olivier, Chief Operating Engineer J. Neal, Security Division Manager W. Clancy, Systems Engineering Division Manager F. Wozniak, Fire Protection Division Manager
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ATTACHMENT II-Facility Tour Findings by Regional Administrator - December 29', 1988 ' ' Hose l connections on the scram. discharge instrument. volume' (SDIV) drain -- lines were' unconnected at the other end; The licensee removed the hose and. capped the drain lines.
, Valve H0 301-1000 on SDIV appeared partially open; --- [ The licensee verified the valve was closed.
~ A: sight glass. was on the SDIV and it appeared that the tank was vented to -- l !- the are' through the sight glass; a The sight glass had been isolated.- -- Valve lineup on the.SDIV was last performed in February 1988; l The resident inspector independently checked a. valve lineup on December 30, 1988, and noted no discrepancies.
-- Indication of small leak on threaded fitting into RCIC pump bearing oil reservoir was noted; The licensee submitted Maintenance Request 88-13-90.
A pressure gage on HPCI, suction piping was broken; -- The licensee completed Maintenance Request 88-23-133.
Limit switches on nitrogen purge valves on HPCI were not connected; -- Indication from these valves are not used.
. Scaffolding materials, i.e., nails and wood chips, laying on floors which -- could migrate to drain systems and c3use pump or valve damage; A walkdown was performed covering all staging and scaffolds. Nails and wood chips were removed.
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