IR 05000341/1991009
| ML20217D134 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 07/10/1991 |
| From: | Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20217D124 | List: |
| References | |
| 50-341-91-09, 50-341-91-9, NUDOCS 9107220064 | |
| Download: ML20217D134 (40) | |
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U. S. NUCLEAR REGULATORY COMISSION REGION 111 Report No. 50-341/91009(DRP)
Docket No. 50-341 Operating License No. NpF-43 Licensee: Detroit Edison Company 2000 Second Avenue Detroit, MI 48226 Facility Name: -Fermi 2 Inspection At: Fermi Site, Newport, flichigan Inspection Conducted: April 15,,,1991 to June 14, 1991 Inspectors: W. G, Rogers S. Stasek F. Brush K. Reimer T. Tella B. Drouin S. Gocker-
. JUL 101991-
-Approved By: 'R. eFay 'tteNief -
Reactor Projects Section 2B Date inspection Su_mma Inspection on April 15 to June 14, 1991 (Report No. 50-341/91009DRP))
Areas Inspected: Action _on previous inspection- f indings; operationa l safety; maintenance; surveillance; followup of events; LER followup; refueling activities; regional requests; and Generic Letter followu Results: Overall, onshift operator performance continued to be adequate during the inspection period. Administrative controls were generally implemented appropriatel However, in one case, operations personnel failed to recognize that a draindown evolution conducted on the Division 11 residual heat removal service water (RHRSW) piping could have impacted secondary containment integrity. When the inspector brought this concern to operations management's attention, actions were initiated to correct the problem. Refuel floor activities observed were conducted in a conservative and controlled manner. - Housekeeping in general was at an acceptable level. All. maintenance and surveillance activities reviewed during the inspection period were conducted in an acceptable manner and-in conformance with applicable procedures, work instructions and within radiation protection requirement Only one of four ESF actuations that occurred during the inspection period was-attributable to personnel error. However, an additional event did occur where an uninterruptible power supply (UPS) A was inadvertently deenergized due to a personnel error. -Implementation of the maximum extended operating domain 9107220064 910710 PDR ADOCK 05000341 0 PDR _ _ _ _ _ _ _ _ _ _ _ _ _
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(ME0D) and APRM/ RPM Technical Specification improvenients (ARTS) appeared adequate. No violations were identified. Five open items were identified (Paragraphs .c, 7.a, 7.1 and 7.m). Two unresolved items were identified (Paragraphs 2.a and 3.c).
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DETAILS
, Persons Contacted Detroit Edison Company
- R. Anderson, a perintendent Radiation Protection
- P. Anthony, Engineer, Compliance
- J. Bragg, Supervisor, Procurement QA
- S. Bruner, Director, Nuclear Training
- S. Catola, Vice President, Nuclear Engineering and Services
- L. Collins, Supervisor, Electrical, Nuclear Engineering 0 W. Colonnello, Supervisor, Plant Evaluations
- J. Contoni, Supervisor, Mechanical, Nuclear Engineering R. Eberhardt, Outage Manager
- P. Fessler, Superintendent, Technical Engineering D. Gipson, Assistant Vice President, Nuclear Operations 0 * L. Gooda.an, Director, Licensing f J. Green, Supervisor, 180, Nuclear Engineering
- K. Howard, Supervisor, Plant Systems
- A. Klcrrptner, Senior Engineer, Nuclear Engineering J. Korte, Acting Director, Nuclear Security
- A. Kowalczuk, Superintendent, Maintenance and Modifications R. Nay, Director, Nuclear Materials Management
- R. Mr.Keon, Plant Manager
- W. Miller, Director, NQA
- B. Myers, Reactor Engineer
- R. Newkirk, General Director, Regulatory Affairs G. Ohlemacher, Principal Engineer, Licensing 0 W. Orser, Senior Vice President, Nuclear Operations
- J. Pendergast, Compliance Engineer J. Plona, Superintendent, Operations Of* T. Riley, Supervisor, Compliance
- * T. -Schehr, Operations Engineer 0- L. Schuerman, General Supervisor, Plant Engineering B. Sheffel, Nuclear Production, Technical Engineering ISI
- J. Sklerczyk, Senior Engineer, PCC
- R. Slowinski, Senior Engineer, Technical F. Svetkovich, Operations Support Engineer-0 R. Stafford, General Director, Nuclear Assurance W. Tucker, General Supervisor, Engineering Design and Services 0 * J. Walker, General Director, Nuclear Engineering 0 S.-Zoma,' Engineer, Electrical, Nuclear Engineering U. S. Nuclear Regulatory Commission 0 * W. Rogers, Senior Resident Inspector 0# S. Stasek, Resident Inspector B. Drouin, Project Inspector, Rill 0 T. Tella, Reactor Inspector
- F. Brush, Reactor Inspector, Clinton 0 K. Riemer, Reactor Engineer
- S. Gocker, Coop, RIII
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- Denotes those attending the exit meeting on June 14, 1991.
- Denotes those attending the interim exit meeting on May 17, 1991.
0 Denotes those attending the interim exit meeting on June 6, 1991.
The inspectors also interviewed others of the licensee's staff during i this inspectio . Action on Previous inspection Findings (92701) (closed)OpenItem(341/84020-05(DRP)): Color banding for indicators. As discussed in the Attachment to this inspection report all Priority I & 11 human factors deficiencies associat(d with instrument / recorder color banding were resolved. The priority III color bandings remain and will be completed in Refule Outage 03 (RF03) via the engineering design packages (EDPs) associated with human engineering discreaancy (NED) 775. Therefore, implementation of corrective action to iED 775 is considered an open item (341/91009-01 (DRP)).
Also, as discussed in the Attachment, resolution to HED 719/726 appeared to have been done improperly and not per EDP 9814. This matter is considered unresolved (341/91009-02(DRP)) pending additional personnel interviews and document review (Closed)OpenItem(341/84020-07(DRP)): Identification of alarm points on recorder scales. See discussion in paragraph 2.a. abov (Closed)OpenItem(341/84020-08(DRP)): Recorder scale cnlor banding. See discussion in paragraph 2.a. abov (Closed)-Violation (341/86032-02(DRP)): Inadequate corrective actions to indication perturbations on reactor pressure vessel level indication system. During the current outage the licensee comaleted the final corrective action associated with this violation. T11s was the installation of new reactor pressure level indication racks. The inspector verified through visual observation that the new racks were installe (Closed) Violation (341/87002-01B(DRP)): Exceeding Technical Specification action statement while performing surveillance procedure NPP-44.030.154, "ECCS-HPCI Condensate Storage Tank Level Ca!ibration." The inspector reviewed the current revision of this surveillance procedure and concluded that the actions prescribed met the intent of the original violation response. Therefore, this matter is considered close Failure to nicca a drywell (Closed) Violation.(341/87048-01(DRP)):
pressure channel in the tripped condition within the allotted Technical Specification time frame. The licensee established impact statements'for instrumentation surveillance procedures to provide added direction-to on-shift operations personnel as to when to trip instrument channel __
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9 (0 pen) Violation (341/88014-01(DRP)): Failure to meet limiting condition for operation actions when a division of the noninterruptible air system (NIAS) was taken out-of-service. The licensee provided interim guidance to operators as to what actions to toke when a division of N1AS is out-of-service, lhe licensee issued a procedure for use in making operability determination The only outstanding corrective action is associated with gaining a Technical Specification explicit to the NIAS. The licensee submitted on Technical Specification request to the NRC which was withdrawn when it was apparent that the submittal did not adequately address the air supply ramifications for the torus-to-reactor building vacuum breakers. Presently, the licensee is preparing another submittal on the matter. Clos' ire of this violation is contingent upon issuance of an explicit NIAS Technical Specificatio (Closed) Violation (341/80017-02(DRP)): Inadequate channel check procedure for reactor protection drywell pressure instruments. The inspector confirmed that the instruments were added to the channel check procedur (Closed) Unresolved Item (341/88031-02(DRP)): Apparent repeated failurcs to perform adequate QC inspections. The licensee conducted training for electrical inspectors regarding the subject incident, Procedure NQP-IN1-01, " Inspections," was revised to incorporate a PQA checklist that directs inspectors to verify that all unresolved items are adequately addresse Inspector review of PQA inspector activities indicates findings are adequately resolved. This item is considered close (Closcd) Unresolved Item (341/88035-01(DRP)): Temporary modification control. The inspector reviewed the results of the temporary modification reduction program and noted a significant reduction in out. standing temporary modifications that were greater than one year old. The reduction program appears to have been effective and this matter is considered close (Closed) Open Item (341/88035-02(DRP)): Battery labelling and cell leakage. The licensee completed labelling all of the 24/48 voit
' attery banks; initiated periodic battery cell-to-cell resistance checks; and replaced a broken flash arreste Licensee inspections of the battery identified seven leaking battery cells. These leaking cells were repaired; however, minor leaks still exist on three cells. The licensee purchased eight new cells, and will replace the leaky cells during the next available system outag Based on these corrective actions, the item is considered close . (Closed) Unresolved item (341/88037-12(DRP)): As-built drawing discrepancies. The licensee revised the drawing change procedure to improve the drawing change verification process. This matter is considered close (Closed) Open Iten (341/88037-14(DRP)): Integrated planning and scheduling improvements. The licensee established a five-year plan for the integration of high impact program changes or significant facility chariges. A more structured use of engineering resources
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was established. Finally, changes to the planning and scheduling department provided more realistic outage and non-outage work schedules. This matter is considered closed, (0 pen)Openitem(341/88037-17(DRP)): Spare parts availabilit The inspector ascertained from interview and partial record review that the licensee completed surveillance and preventative-maintenance reviews for spare parts. An obsolete parts task force is in place and meeting periodically to resolve parts difficultie The only action remaining to be completed-is a' technical engineering-group review for spare parts by limiting condition for operatio This review is targeted for completion by December 199 Upon completion of-that review this matter may be closed, Post-maintenance testin (Closed)OpenItem(341/88037-19(DRP)):TheDiagnosticEvaluationTeam(DE (PMT) may not be adequat developed a concern regarding the effectiveness of existing plant PMT. The team believed that performing technical specification (TS)
required surveillances after maintenance activities may not have been sufficientAn functio to NRC ensure affected team equip inspection (ment would inspection reportperform its safety 341/91002)
performed a comprehensive review of design change acceptance tests (DCAT) and associated PMT 3erformed on modifications implemented during Refuel Outage 2. T1e team concluded that the licensee's modification testing program was adequate to ensure modified equipment would perform its decign function with no adverse impact on interfacing plant systems. Procedure NPP-CTI-06, Revision 5,
" Post-maintenance Testing Guidelines," is the controlling document for both post modification testing and PMT and was the controlling document for DCAT-and PMT performed during RF 2. Therefore, the licensee's PMT program is considered adequate to ensure equipment operability af ter maintenance has been performed. This item is-considered close (Closed) Violation (341/89008-01(DRP)): Failure to change procedures. The violation enteilco three parts.- The first two parts were closed at time of issuance of the Notice of Violatio The final part dealt with specifying operation of the drywell vent and purge system that differed from procedural constraints. The licensee evaluated the unapproved ~ negative pressure differential between the drywell and torus which can occur during containment venting-operations. The evaluation indicated that there was sufficient design margin in the containment structures and piping to accommodate the load increase resulting from the negative differential pressure. The inspector verified that System Operating Procedure NPP-23.406, " Primary Containment Nitrogen Inerting and Purge System," was revised to incorporate the engineering f the department's analysis and recommendations. The adequacy o engineering analysis is under review by NRR and is being tracked as unresolved item 341/89002-03. This violation is considered close (0 pen) Open Item (341/89008-16(DRP)): Licensee actions to improve safety relief valve (SRV) performance. During RF02, the licensee conducted setpoint testing of all 15 valves with the 5 _ _ _ . . _ _ . , . ._ _. . ___ _ _ .. _ . _ . _ , _ _ , . . _ . , .
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l resultant data indicating that 12 of the 15 valves experienced i setpoint drift greater than the 1 percent allowed in Technical Specifications during the last operating cycle. This matter continues to be an industry issue for those utilities with this type SRVs. All 15 SRVs were replaced during-RF02. The replacement SRVs contained stellite seats. The licensee continues with its evaluatio and potential corrective action (Closed) Violation (341/89011-02A(DRP)): Improper design of sixteen containment nitrogen inerting and purging system isolation valve Training was provided to operators on how to close these valves from the relay room until a design change for the 16 valves could be implemented. The design change was scheduled to be implemented during the current refuel outage and the inspector confirmed partial implementation. Implementation and final review of this modification is also necessary to fulfill closure requirements for violation 341/89011-01A. Therefore, this violation is considered closed and the design change review will be performed under the associated violation followup,
. .(0 pen) Violation (341/89011-028(DRP)): Improper design and testing of the control room ventilation syste The inspector confirmed that'the licensee performed a stress analysis of the control room and standby gas treatment auctwork using emergency and faulted conditions, that a safety evaluation of the ductwork designs were performed and a com)liance matrix to Regulatory Guide 1.52, Revision 2, was complete rom these reviews the licensee concluded that the design and testing performed on these systems were equivalent to the original requirements and no unreviewed safety questions existed. However, the inspector could not substantiate these conclusions. Subsequently, the inspector identified the discrepancies with control room design to NRC Region III management for review by the-Office of Nuclear Reactor Regulation (NRR). This information is presently with NRR for disposition. Upon completion of the NRR review, the inspector will determine the adequacy of the licensee's response to this violation. Also, a discrepancy from the original testing requirements for the standby gas treatment system was identified by-the licensee. This matter will also be forwarded to NRR_for: dispositio (Closed) Violation (341/89011-03A(DRP)): Failure to complete a limiting condition for operation when connecting a temporary battery to normal 24/48VDC battery R3200S001. Further review by the inspector of the maintenance procedures concluded them to be adequate. This completed the final action for this violation followu (Closed)OpenItem(341/89014-01(DRSS)): Condenser bay valve leakage. The valve leakage has been significantly reduced as evidenced by the reduction of leakage to the turbine building sumps. This matter is close .
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- (Closed) Violation (341/89017-02(DRP)): Inadequacies while implementing the internal corrective action system. The inspector 1 confirmed through record / procedure review and interview that a system engineer's tracking system was subsequently developed, a stricter extension request process was established, and edditional guidance was provided to DER reviewers for rejecting DER responses, (Closed)OpenItem(341/89030-06(DRP)): Drywell lighting improvements for outages. Under EDP-11044 the licensee provided a permanent maintenance power distribution system for the drywell and the reactor buildin (Closed) Violation (341/89034-03(DRP)): Failure to perform hydrogen monitoring surveillance prior to placing the of fgas system into service. The inspector completed review of the last facet of the accout.T.bility action plan through record review. This matter is Cons %kceJ closed,
- (0 pen) Ope., item (341/89200-07(DRS))
- Revision of emergency diesel generator-(EDu,' foad calculations. The licensee established the basis and methodology for the EDG load calculations. Walkdowns were completed for collecting nameplate ratings of motors and other loads. The licensee's vendor completed the Electrical Load Monitoring System (ELMS) calculation, which is currently under review prior to final approval. The licensee stated that Design Calculation (DC)-5003 will be revised by the end of July 1991, based on the ELMS. The licensee also stated that another calculation, DC-2116, would be replaced by the ELMS, which would also be issued by the end of July 1991. Pending the issuance of these EDG load calculations by the licensee, this item remains ope (Closed)OpenItem(341/89200-08(DRP)): ResidualHeatRemoval(RHR)
motor overcurrent relays. The licensee evaluated the overcurrent relay settings for the RHR and core spray aump motors and other loads on the safety buses, and concluded t1at the overcurrent relays were properly set. The licensee issued design calculation DC-5111 incorporating pump acceleration times, motor starting currents and relay response curves for RHR and core spray pump motor The relay settings were verified and appeared to be adequat a (Closed)OpenItem(341/89201-05(DRP)): Primary Contairaent Water Level Indicator. As a result of the Detailed Control Room Design Review (DCRDR) program, HED 462 and its associated HED 509, the licensee identified that primary containment water level indication to a depth of 108 feet (elevation 650 feet) was not available in the control room. Engineering Design Package (EDP)-10714, in conjunction with EDP-8483, was to add instrumentation which would moniter drywell pressure viapenetration pressure via penetrationX-206f X-27f (elevetion (elevation545 650feet feet))and torusto provide input to determine primary containment water level up to an elevation of 650 feet-(equal to a depth of 108 fect in primary containment) for display on the scfety parameter display system (SPDS) as well as on a recorder mounted on main control room panel H11-P602. The licensee subsequently completed EDPs 10714 and 8483 allowing closure of HEDs 462 and 509,
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The inspector reviewed the associated modifications, installed instrumentation, and performed a walk-through with the systems 1 engineer and concluded that the primary containment water level ;
indicator was functional. This item is considered closed, j
b (Closed)OpenItem(341/90002-05(DRP)): Corrective actions from RF01 critique. The inspector noted improvements in a number of key areas identified as weaknesses frort the first refuel outage. These included better work preparation, moic design changes ready or !
-schedule, better scaffold coordination, better outage scheduli 1 l better equipment tagout coordination. The most significant continuing weakness was in documentation ouality which is discussed 1 in more depth in inspection report 341/g1002. However, based upon overall improvements, this matter is considered close c (Closed)OpenItem-(341/90005-05(DRP)): Residual heat removal (RHR)
piping vibratio Following a number of instrument tap failures associated with RHR piping the licensee performed a stress analysis of-the subject piping, conducted walkdowns of the RHR piping for unacceptable vent and drain connection arrangements and conducted drawing reviews of the large bore reactor coolant pressure boundary piping for potential unacceptable vent and drain connecti>n arrangement The inspector ascertain . that:
Inside containment three modifications were made. Under E0P-11502 the Division 11 low pressure injectior system high point vent was changed and two feedwater connections were modifie *
For the RHR piping located both outside containment and outside the RHR pump room compartments, 75 of 75 connections were visually inspected with satisfactory results. Eleven potential high vibration candidate sites were identified. Vibration data was taken on all the connections with three being subsequently modifie ' Within the RHR pump rooms, all connections were found to be satisfactory, althongh one required additional strain data during a RHR pump run in February 199 No further inspections are planned by the license d (Closed)Open Item (341/90013-05(DRP)); TIP indexer leakage. The licensee identified the source of the TIP indexer leakage and corrected the problem. Temporary Change (T-07443) was issued to procedure 45.606.006, "TIP Guide Tube fabrication," to include testing for leakages in the TIP indexer. A leak test was performed and found to be acceptable. This item is considered closed, e (0 pen) 0)en Item (341/90013-06(DRP)): CR120A relay replacemen During tie most recent refuel outage (RF02) approximately one half of the subject relays located in the relay room were replaced. The other half is scheduled far replacement during RF03. This item will remain open until the other relays in the relay room are replace __ .-
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f (Closed)0)enItem(341/90013-07(DRP)): The licensee completed the scram disc 1arge volume ($DV) vent modification via EDP-1156 This modification included the installation of a drain line from the ventilation duct to the reactor building drain to direct condensation from the scran discharge volume vent. Pased on inspector review of the work package and walkdown of the modification site, this item is considered close (Closed) Open Item (341/90013 10(DRP)): Termination wiring discrepancies. The licensee issued as-built notices to correct the discrepancies observed. This matter is considered closed, h (0 pen) Open Item (341/90013-11(DRP)): Use of uncontrolled labels on cabinets. The licensee stated that a labelling program would be developed by August 31, 1991 to cover plant equipment such as valves, instruments switchgear, etc. The licensee stated that the icbellingisscheduledtobecompletedbyMay1994. .The licensee also stated that the operator aids attached to the cabinets will either be removed or properly identified by May 1994. Pending the above corrective actions by the licensee, this item remains ope i (Closed) Unresolved Item (341/90020-01(DRP)): Deviation from requirements of administrative procedure for maintenance taggin Night orders were subsequently issued to the Nuclear Shif t Supervisors providing better guidance for when deviation from administrative procedures is appropriate. This item is closed, j (0 pen)OpenItem(341/90020-04(DRP)): Additional licensee review for control room modifications following operator inadvertent closure of a main steam isolation valve, further evaluation for control room modifications beyond installation of protective covers on the main steam isolation valve "close" pushbuttons has not been performed by engineering. Presently, the licensee has established a date of August 14, 1991 to complete the review / evaluation. Closure of this open item is contingent upon completion of the evaluation and implementation of the management approved recommendations, k (Closed) Open Item (341/91007-05(DRP)): Implementation of a design change to the control room ventilation logic. The licensee provided the inspector with documentation supporting implementation of EDP-11889 which roodificd the ventilation logic. This matter is considered close . (Closed)OpenIten(341/88037-15(DRP)): Development of a performance evaluation program. A performance evaluation program (PEP) has been institutionalized in the licensee's written program via FMD REl, " Balance-of-Plant Reliability Improvement," NPP-REl-01,
" Performance Evaluation" and the PEP series procedures. The licensee has evaluated twenty-eight safety and non-safety related systems for inclusion in the PEP. Presently all PEP related activities are being implemente . __
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mm. (Closed)OpenItem(341/90020-03(DRP)): Tineliness of review for a GE Service Information Letter on equipment protection assembly (EPA)
breakers. The inspector determined tiat the review was timely but the impl uentation of corrective actions was slow. Subsequently, correctiveactions(EPAbreakerreplacement)wereaccomplished, nn. (Closed)OpenItem(341/89200-03(DRS)): Post modification testing (PMT) did not receive adequate engineering group review. The SSOMI team concluded that PMT was not adequately described in the applicable portion of engineering design packages (E!')) and the engineering group was not adequately involved in the development and review of PMT instructions. A NRC team inspection (inspection report 341/91002) of the licensee's modification implementation program performed April 29 through June 3, 1991, determined that the licensee modification testing program was adequate. The team determined through interview, dacument review, and field observations, that the Nuclear Engineering Department EDP " owner" was thoroughly involved in the development of EDP testing requirements, reviewing PMT testing instructions developed by the maintenance planner and in the review of Phi results. Testing requirements were generally well described in the EDP although there were a few examples of inadequate test instructions resulting from the use of a PHT procedure, NPP-CTI-06, Revision 5, " Post Maintenance Testing Guidelines." This item is close . (Closed)Openitem(341/86039-01(DRP)): Valve accessibility and serviceability for safety-related (SR) valves. The licensee had a commitment to improve the accessibility of certain SR valves for serviceability and for manual operation during abnor.nal condition The licensee had agreed to improve accessibility to 75 percent of the valves by RF02 and 100 percent of the valves by RF0 By the end of RF01, the licensee had improved the accessibility to 82.5 percent of the valves. In August 1990, the licensee reevaluated the need to improve accessibility to the remaining 17.5 percent of the SR valve The licensee's staff informally requested relief from the conmiitment to _ install the remaining valve serviceability platforn,s. The licensee determined that the valves were either easily accessible for manual operation cr that the valves were not required to be manually operated to achieve a plant safe shutdown condition. The inspector reviewed the licensee's evaluation, and discussed the methodology which determined that the valves (not readily accessible) were not necessary for safe shutdown. The Operations Support Engineer stated that the Abnormal Operating Procedures and Emergency Operating Procedures were_ reviewed to ensure that not-readily-accessible valves were not required for safe shutdown. The inspector determined through document review and discussion with the General Supervisor, Podifications (GSM), that all valves, for which conmiitment relief was-being requesteo,.could be serviced with the use load distribution points (temporary rigging) identified by Nuclear Engineering. The GSM further stated that the Operations and Maintenance Department would reevaluate the need for serviceability platforms for the subject valves if conditions changed. The inspector also reviewed maintenance histories on the subject valves. The histories did not
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appear to warrant valve serviceability platforms. The inspector ,
considered the licensee evaluation in support of the valve serviceability platform relief to be adequate. This item is close p (0 pen)OpenItem(341/88014-02(DRP)): Noninterruptible air system design document reconciliation. The licensee has one more document yet to revise. This is the part of the design specification discussing the interdivisional crosstie valves. per discussion with licensee personnel this section will not be revised until a Technical Specification change addressing noninterruptible air is issue . Operational Safety Verification (71707)(37700)(72701)
The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room o)erators throughout the inspection period. The inspectors verified tie operability of selected safety-related systems, reviewed tagout records, and verified proper return to service of affacted components. The inspectors observed a number of control room shift turnovers. The turnovers were conducted in a professional manner and included log reviews, panel walkdowns, discussions of maintenance and surveillance activities in progress or
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planned, and associated LC0 time restraints, as applicabl The inspectors conducted tours of the reactor, auxiliary and turbine buildings and the RHR complex. During these tours, observations were made regarding plant eaufpment conditions, fire hazards, fire protections, adherence to procedures, radiological controls and conditions, h_ousekeeping, tagging of equipment, ongoir.g maintenance and surveillance activities, containment integrity, and availability of safety-related equipment. Walkdowns of the accessible portions of the following systems were conducted to verify operability by comparing system lineups with plant drawings, as-built configuration or present valve lineup lists; observing equipment conditions that could degrade performance; and verifying that instrumentation was properly valved, functioning and calibrate * Emergency Diesel Generator No.11
Emergency Diesel Generator No.12 Standby Liquid Control System
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Core Spray System - Divisions I and II
' Residual Heat Removal System (LPCI mode) - Divisions I and II
Residual Heat' Removal Service Water - Division I
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Thermal Recombiners - Division I and II Additionally, the inspectors observed implementation of portions of the licensee's security program during the inspection period including:
badging of personnel; access control; security walkdowns; security response (compensatory actions); visitor control; security staff attentiveness; and operation of security equipmen Significant observations and reviews made during the period included the following:
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a. The inspectors reviewed the Engineering Design Packages (EDP) for the-Rod Worth Minimizer (RWM) re Extended Operating Domain (ME00) placement and APRM/RBM Technical(EDP-6641), the Maximu; Specification improvements (ARTS) installation (EDP-9739), and the )
3-D Monicore installation (EDP-11068). The inspectors also walked 1 down the field work for the RWM and 3-D packages and held I discussions with various personnel concerning implementation of the i Design Change Acceptance Test (DCAT) requirement The licensee used various existing plant surveillance procedures to meet the test requirements of some of the criteria delineated in the DCA Copies of these procedures were not always required to be included with the other test documents. The quality level of the PWM and ARTS /ME0D EDPs was 0-1. The quality level of the 3-D Monicore EDP was Non-Q. The inspectors q"estier.ed licensee personnel as to the criteria used in determining quality level. It was stated that one criteria was the effect the EDP had on Technical Specification (TS) requirements. The 3-D Monicore EDP was a major modification to the method used to determine fuel thermal limit These limits are in TS sc-+1on 3/4.2- The difference between the quality levels.of these EDPs s dicate inconsistencies in the licensee's program. The licensee int .ated it would further evaluate the inspectors' observations described abov b. On May 3,1991, during a review of planned work activities for the day shift, the inspector noted that to support inspection / testing of a check valve in the Division 11 Residual Heat Removal System, a draind a of the piping between the RHR heat exchanger and the RHR complex was being done. The inspector recognized that when completed, a direct air pathway from within secondary containment to the cutside environs would exist (since the isolation valves associated with the mechanical draft cooling towers had not been closed). This observation was communicated to the Nuclear Shift Supervisor who agreed that a pathway would exist following completion of the draindown and subsequently directed the subject isolation valves be close c. On June 10, 1991, the inspector obsarwd LCO 91-0098 associated with fire detection instrumentation. Attached to the LCO was DER 91-0075. This DER described the lack of a fire detector in a small cable vault adjacent to the Division I electrical switchgear roo The DER fndicated that this condition was in contradiction with the NRC SER for fire protection. The inspector noted that appropriate remedial action was being accomplished in the form of u one hour fire watch and a detector was being installed in the vault are However, this may have been required to br reported to the NRC via the ENS network under the reportability requirements associated with License Condition 2.F. This is considered unresolved (341/91009-03(DRP)) pending further revie d. During the inspection period the licensee informed the resident staff that a new containment analysis (reference DER 91-0159)
revealed that the suppression pool temperature would rise to 198 F in a hypothetical design bases accident. The original FSAR stated a
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maximum temperature of 191*f. The licensee indicated that such a )
temperature increase would not cause catastrophic equipment failure but could potentially reduce the environmental qualification (EQ) !
life. The licensee evaluated the potential EQ concerns and i
determined that-no equipment was rendered inoperable and preventative maintenance replacement schedules were modified as necessar . Monthly Maintenance Observation (62703)
Station maintenance activities on safety-related systems and components listed below were observed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specification The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemented.
Work requests were reviewed to d M ermine the status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performanc The following maintenance activities were observed / reviewed:
- WR 0002910501 Check of Division 1 Testability Cabinet Trip Unit WR 0070900725 Reactor Pressure Vessel Reassen61 WR 0120900625 SRV Valve / Pilot Removel/ Installation For Testin * WR 000Z910885 Repair of D Inboard MSIV Body to Bonnet Lea No violations or deviations were identified in this are . . Bimonthly Surveillance Observation (01726)
The inspectors observed / reviewed the following Technical Specification required-surveillance testin *
24.137.18 Main Steam Line Drain and Drain Isolation Valve Operability Test 24.307.03 Emergency Diesel Generator No.13 - ECCS Start With Loss of Offsite Power Test
- 24.307.32 Emergency Diesel Generator No.13 - 24 Hour Run Followed by Loss of Offsite Power Test 43.000.09 Reactor Pressure Vessel System Leakage Test
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The following items were considered during the inspection: the testing was performed in accordance with approved procedures; that test instrumentation was calibrated; that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test; and that any deficiencies identified during the testing were reviewed and resolved by appropriate management personne The inspectors also performed a record review of the completed surveillance tests listed below. The review was to determine that the test was accomplisted within the required time interval, procedural steps were properly initialled, the procedure acceptance cr'teria were met, independent verifications were accorrplished by individuals other than those performing the test, and that the test was signed in and out of the control room surveillance log boo .000.05 Attch 5, Pressure / Temperature Monitoring During Heatup and Cooldown 24.000.01 Attch 1, Situation Surveillances/LC0 Tracking
24.206.C03 RCIC Discharge Piping Venting and Valve Verification Test 43.401.201 Local Leakage Rate Testing for Electrical Penetrations
44.020.007 Partial, NSSSS-Reactor Vessel low Water Level (Levels I and 2) Division 1, Channel A Calibration / Functional 44.020.229 NSSSS-Main Steam Line Pressure, Division 1, Channel C Calibration / Functional No violations or deviations were identified in this are . Followuo of Events (93702)
During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR 50.72. The inspectors pursued the events onsite with licensee and/or other NRC officials. In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the licensee was taking prompt and appropriate actions, that activities were conducted within regulatory requirements and that corrective actions would prevent future recurrence. The specific events are as follows: ESF actuation due to master trip unit logic switch out of positio On April 24, during perturmance of a surveillance to calibrate the Division II Channel B Levels 1 and 2 NSSS initiating setpoints, reactor building HVAC isolated, Standby Gas Treatment System autostarted and control center HVAC transferred to the recirculation mode. The licensee subsequently determined the cause to be personnel error that resulted in a logic switch in a master trip i _ - _ _ _ - _ _ - _ - _ ___-___ ___ _ __
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unit being mis)ositioned during installation of the associated card rinto:the: testa >ility! cabinet. Similar switches in other trip units were then checked to ensure they were properly positioned.- !
b.- ' oss~ of- power to ESF clectrical bus due to out -of position cutoff
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switch.- On April 29, during calibration and logic system testing on
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j 4160v ESF bus 65F, Bus 65F lost power. The licensee subsequently a
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found an' associated cutoff switch mispositioned, flo reason for the mispositioning was found. However, after completion of the evaluation, no indication of intentional manipulation of the switch was foun i Loss of Uninterruptible Power Supply (UPS) Event. On May 3, UPS Bus A was inadvertently deenergized due to a personnel error by a DECO technician A jumper installed previously was not removed-prior to charging'the bus. The resulting short circuit damaged six fuses and-a-fuse holder which splattered metal particles damaging a static-transfer switch logic card. A further charging of the bus -
after-the fuse replacement caused loss of power to several UPS load
' areas including full core: display, half of ERIS, rod worth minimizer, feedwater control, turbine electrical governor and CRD 9 accumulator monitoring panel Deviation Event Report (DER) 91-0306 was subsequently initiated to address the event and track corrective actions. The licensee's evaluationreport-(ilAPS-91-0041) on this event identified three deficient work practices as the root causes of the event i.e.,
failure to follow procedures, inadequate.self checking and an
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inadequate-procedure. The recommended corrective actions included establishing ~a preventative maintenance procedure for the UPS, assessment of. damage and repair of_ the load circuits on UPS Bus 'A',
. redesign of fuse protection within the UPS cabinet, and establishing Jbetter control of jumper installatio LThe' inspector'noted that.while the licensce's investigation report-suggested that load circuits be verified for damage,.it did not include evaluation of the. sources of power for the short circuits-
-such as the UPS battery and the charger circuits for possible
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damage. The licensee conducted an inspection /PM on the UPS battery ~
'on May 27, 1991- for acid levels, specific gravity, yo_ltage 1evels, ~
and. cell leakag The inspector also reviewed the licensee actions on the. loss of
- power to-the CRD accumulator trouble alarm panel. At the time of this' event, the core was unloaded and all the CRD accumulators were-
- depressurized and: operator action was not needed to verify individual accumulator pressures locally when-the loss of power to the accumulator trouble panel occurred.- The inspector noted that the licensee's annunciator response procedure ARP 3010, "CR0 Accumulator Trouble," Revision-7, did .it include any operator actions in case of loss of power to the panel, such as verification of' accumulator pressures locall ,
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Since DER 91-0306 had not been dispositioned and the proposed corrective actions were not completed at the conclusion of the inspection, this matter is censidered an open item (341/91009-04(ORP))pandinginspectorreviewoflicenseeaction Control center emergency filtration makeup air charcoal filter failure to pass methyl iodide penetration test. On May 19, during performance of charcoal efficiency testTng the licenses found the charcoal to have less (98.5 percent) than the 99 parcent efficiency as required by Technical Specificrtions. 'this was apparently due in part to implementing a change to the testin the charcoal at a more realistic teraerature (g 30 methodology to check degrees C versus 80 degrees C). _ Both divisions of CCiVAC were Jeclared inoperable and the charcoal subsequently replaced, ESF actuation due to momentary spike of contro.' center makeup air radiation monitor. On May 23, CCHVAC automatically shif ted to the recirculation mode due to a spike on the subject monitor. No cause for the spike was found. The licensee performed functional testing of the circuit with no problems identified. The monitor was subsequently returned to service with no further problems note ESF/RPS actuation due to improperly installed intermediate range supply fuse cap. On May 25, a full scram signal was received but all control rods were already inserted in the reactor core. One channel of IRMs was being tested when a spuricus "INOP" signal from the companion channel was received. The most probable cause of the spurious signal was from a short in the fuse holder due to an improperly installed spring loaded fuse ca ESF actuation due to blown fuse in the Division I fuel pool ventilation exhaust radiation monitor circuitry. Un June 1, during surveillance testing of the radiation monitor, the instrument drawer
_ was required to.be withdrawn and installation of a temporary jumper installed. When the technician opened the drawer the subject fuse
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blew causing a spuricus spike on the monitor resulting in a trip of RBHVAC, autostart of Standby Gas Treatment System Division I and a shift of CCHVAC to recirculation mode. The fuse was replaced and all systems returned to servic No violations or deviations were identified in this are . Licensee Events Reports Followup (92700)
Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications.
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l I (Closed)LER89006,TurbineTrip/ReactorScramDuetoDesign Deficiency in Turbine Overspeed Rese Corrective actions to date ( have included improved operator training of circuit design and j 16
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issuance of night orders to the operators addressing the proble Implementation / scheduling of a design change is currently under review. This is considered an oaen item (341/91009-05(DRP)) pending completion of that review and suasequent implementation action (Closed)LER90012Rev.1,HPCISteamLineFlowTransmitter failure. The failed circuit boards were sent to the vendor, Rosemount, Inc. for f ailure analysis and root cause determination 1 This analysis indicated that a capacitor on the amplifier circuit board failed cue to a short circuit across the capacitor plate The vendor also concluded that this was a random failure. ~his item is considered close (Closed) LER 90013 Rev. 1 Inadequate Control During the Primary ContainmentAirGrabSampIingProcess. The-licensee identified a concern with the original methodology for obtaining containment grab samples in that administrative controls were lacking. Subsequently, the administrative controls were strengthened and the point at which the samples are taken was changed. Revision 0 to this LER was closed in inspection report 341/9100 (0 pen) LER 91006, Failure of a molded case circuit breaker, causing reactor buildino.flVAC isolation. The root causes for the failure of the breakers nre not yet determined. The licensee is considering inclusion of these breakers into its preventative maintenance program.' Pending further actions by the licensee to prevent-recurrence, this itcm remains ope e.- (Closed) LER 9 007, Error during performance of radiation monitor calibration results in activation of ESF. A high rad:ation trip signal was not reset during surveillance testing when the trip light burned out.- The event was discussed with the personnel involved and the procedure revised to require reset of the high radiation signal, (Closed) LER 91008, ESF actuation ave to mispositioned trip output logic switch. The licensee verified that similar Rosemount Analog-Trip Switches on circuit boards installed in the plant were in the correct position. A Human Performance Enhancement System (HPES)
analysis was performed for this event to identify actions to prevent
- recurrence of similar events. Because the mispositioning of this logic switch appeared to be an isolated incident, this item is considered close (Closed)'LfR 86022 Rev. 2, Technical Specification surveillance procedure inadequacies. The inspector reviewed the licencee's internal corrective action system documentation and performed independent checks to assure that the appropriate procedures were revised. No problems were identifie (Closed) LER 87048 Rev. 2 through 9. Technical Specification Improvecent Drogram review. The inspector reviewed the licensee's internal corrective action system documentation and performed independent checks to assure that the appropriate procedures were revised. No problems were identifie . .
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, (Closed) LER 91002, Emergency diesel generators 11 and 13 concurrently inoperable. The-licensee repaired the two diesels and returned them to service before a shutdown was required. Longterm corrective actions involved development of a maintenance procedure-
~for the-Woodward electric governor and periodic inspections of the motor operated potentiometers. -These two activities are to be completed by August 1991. Accomplishment of those two tasks is considered an open item (341/91009-06(DRP)). (Closed) LER 89010, Ventilation ductwork not designed or tested in accordance with UFSAR commitment. The inspector confirmed that the licensee completeo a stress analysis on the ventilation systems and modified the UFSAR to the actual construction and testing
. requirements used. However, as discussed in paragraph 2.s. of this inspection report, additional review of the licensee's analysis is warranted and will be tracked under violation 341/89011-02 (Closed) LER C0001: Rev.1, Blown fuse in testability cabinet
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H21-p083 caused entry .into Technical Specification 3.0.3. The licensee revised the applicable alarm response procedure to direct operators not to use the instrument trip sheet in this particular even . (Closed) LER 89013, Actuation of the Standby Gas Treatment System (SGTS) and isolation of Reactor Building Heating Ventilation and Air Conditioning (RBHVAC) due to per;onnel error. An I&C technician was installing a jumper to pravent initiation of the Reactor Building
. Exhaust Radiation Monitor during the replacement of a flow switc An electrica' arc was inadvertently drawn from a nearby terminal as the technician was installing the remaining lead of the jumper, resulting in the actuation of the SGTS and the RBHVAC. Licensee corrective action included an accountability meeting, the initiation of a-potential design changr (PDC) request-to evaluate testability improvemer.ts, the development of a form for suggested testability improvements on other systems, and LER 89013 was made required reading for I&C personne The inspectors verified through document review and interview that the following actions were accomplishe (1) An accountability meeting was conducted on July 3,198 (2) PDC 10577 was initiated to improve testability of the Reactor Building vent exhaust radiation monitors and approved on July 19, 198 (3) Engineering design package (EDP) - 10577, Installation of Banana Jacks for Rx Building Vent Exhaust Rad Monitnr Calibration, was installed and the system was returned to
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L service on April 15, 199 (4) A review of deviation event reports (DEP) through May 8, 1991,
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indicated that ther e had been no inadvertent actuations on the modified system resulting from inadequate testability.
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(5) .The required reading of LER 89013 by llc personnel was-completed in August 198 .
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(6) The required 1&C testability improvement form was develope This item is considered closed, Also, licensee internal corrective action report, PER 91-251, was reviewed. The report discussed a discrepancy between the actual horsepower for the E2150F031A's motor and the design requirement Design documents identified the motor as 5 ft-lbs. whereas 2 ft-lb was the motor's actual rating. The DER evaluation and corrective-actions centered upon specific motor operated-valve discrepancy resolutions. The corrective action did not address communication weaknesses apparent within the engineering organization or the untimeliness of corrective action to an identified proble Following discussion with engineering supervision the corrective actions associated with the DER were to be further evaluate The additional actions stemming from this DER situation are considered an open item (314/91009-07(DRP)).
No violations or-deviations were identified in this are . RefuelingActivities(60710)
With~ continuation of the unit's second refuel outage, the inspector observed / reviewed numerous activities associated with the refuel floor during the inspection period. Included were observations of the control -
rod blade changeout-evolution and fuel movernents during core reload on a number of occasions. The inspettor reviewed completed surveillences to verify Technical Specification requirements were met prior to the start of' core reload. Additionally, the inspector reviewed the preparations for both reactor vessel head and drywel-1 head installations as well as revieweo plant procedures related to core reload and vessel reassembl Following completion of core reload, the inspector reviewed the licensee's core verification videotapes and independently verified all
. fuel assemblies were properly configured in the core. This was done by comparing each assembly's serial number observed to that required by the master core loading pattern map. No discrepancies were note No violations or deviations were identified in this are . Regional ~ Requests During the inspection period, the inspector was requested to perform a-survey to determine the extent " reach-rods" are u.ced to operate .
valves not normally accessible at Fermi. The inspector determined that no reliance was made on reach rods for manipulating safety related valve: or valves re Operating Procedures (EOPs) quired to be operated in Emergency
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. . In response to a memoranduni dated March 8,1991, tro tha D"ector, Division of Reactor Projects, the inspector reviewed different aspects of the Thermal Recombiner System design and provided that information to NRC management during the inspection perio Specifically, the information related to the permanency of installation, testing conducted to verify operability, and qualification level of the equipmen In response to a menorandum dated June 5,1991, from the Chief, Projects Section 28, the inspector provided information on questions associated with decay heat removal and operator response to a loss of decay heat removal, In response to verbal direction from the Deputy Director, Division of Reactor projects, the inspector completed a survey form dealing with the licensee's connitment tracking syste . Gentric Letter Followup-The inspector confirmed that the licensee received the Generic Letter (GL), reviewed the GL thrcugh a formalized progrem, assigned the appropriate organizations for review and provided appropriate actions to be accomplished. The GLs reviewed were:
(Closed) GL 85003, Clarification of Equivalent Control Capacity for Standby Liquid Control Systems
.(Closed) GL 85013, Transmittal of NUREG-1154 Regarding the Davis Besse Loss of Main and Auxiliary feedwater Event (Closed) GL 86007, Transmittal cf NUREG-1190 Regarding the San Onofre Unit 1 Loss of Power and Water Hammer Event No violations or deviations were identified in this are . Open Items Open items are matters which have been discussed with the licensee, whict will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Five open items were disclosed during the inspection and are discussed in Paragraphs 2.a. 6,c, 7.a, 7.1 and .- 12. Unresolved Items Unresolved iteras are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations. Unresolved items were identified during the inspection and are discussed in paragraphs 2.a and _., . , ..__ _ .. _ _ _ _ . _ . ~ _ . _ , - - -
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i 1 Exit Interview 1he inspectors met with licensee representatives (denoted in paragraph 1)
on May 17, June 6, and June 14, 1991, and infornially throughcut the inspection period and sunnart:cd the scope and findings of the inspection activities. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee identified several docurents/ processes as proprietary and the inspectors indicated that any reference to those in the inspection report would be handled appropriately. The licensee acknowledged the findings of the inspectio Attachment: Sungnary of the Detailed Control Design R2 View (DCRDR) liuman Engineering Discn pancy (HED) program at fermi 2
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ATTACHMENT l J
SUMMARY Of ihl DETAILED CONTROL ROOM DESIGN REVIEW (DCRPR)
IlUMANENGlMERINGDISCREPANCY(llLD)FROGRAMA1IERMI2 for the inspection of the Dettuit Edison Company's DCRDR progrem, the l
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inspector evaluated the HED documentation for completness, accuracy, and verification of proper modification implementation in the fermi 2 control room, as well as the simulator. The inspector also evaluated the adequacy of prioritization and justification of IIEDs yet to be implen4nte Included at the end of this attachment is a detailed chart which lists the HEDs reviewed; the discrepancy (s) each one identified; the corrective actions taken by DECO; the associated EDP, PDC /MM, or meeting minutes identification number; and the inspection findings for both the control room and simulator at Approximately 50 HEDs were reviewe ;
Fermi Overall, Deco's in:pleroentation of HEDs identified from the DCRDR program was correct and efficient. The deferral of corrective actions for particular hEDs was generally well justified and acceptabi However, there were some concerns on select HEDs such as:
HED 704 - lhis HED required scale banding at less than 1,66 psig on the
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i Drywell-Narrow and Wide Range and Torus pressure Recorder (TSO-R806A).in order to determine torus pressure. The justification behind banding this recorder in the fi;'st place was that if torus sprays have been initiated, and the torus pressure drops below
.l 7 psig, the-sprays have to be manually terminated by the operator,
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lhe banding of-this recorder was dismissed based on the reasonirg that the recorder's scale range is from 0-80 psig. 1herefore, color coding the scale from 0-1.68 psig would not significantly enhance the instrument and the HED was close ,
When asked why some modification was not made to the recorder's original scale so that the banding would be a significant enhancunent, the licensee's response was that the data is di itally displayed on the Emergency Response Information System (ERIS and-t; e shut-off point. of 1,7 psig is noted in the corresponding Emergency Operating Procedure (LOP).
HED 967 also addresses the subject of the 1.7 psig setpoint and the-operator's ability to be dble to deterrnine when torus pressure is below this pnint in order to terminate core spray, Resolution involved EDP 11264 which installed a digital recorder.. This recorder provided torus' pressure indication in the narrow range in order to adequttely determine when the torus pressure reached 1.7 psi ,
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HED 704 was classified a priority 11 llED high safety consequences / low pots tial for error. HED 907 was classified as a priority ll! HED - los safety consequences /high potential for error. The licensee explained the inconsistency by noting there were different review tearas which identified, prioritized, and recommended corrective actions for the HEDs. According to the licensee, the difference in prioritization and corrective actions between HED 704 and HED 967 was the result of differing opinions.
, llEDs 719 8 726 - ThebandingfortheWideandNarrowRangeRecorder(C32-RE09)
Mfound to be incorrect for both the control room and the simulator. The scale installed for the wide range scale was banded amber f rom 760-920 psig instead of from 780-920 psig as required and specified in EDP 9814. Thi.s EDP was the corresponding design package which closed out these two HED Inspection of EDP 9814 and its related docurnents showed thet the finalAs-BuiltNotice(ABN)wassignedoff,indicatingthattheEDP was field ccmplet However, both recorders scales were banded incorrectly, further review showed that the purchase order sent out to acquire the wide range scales was yellow-lined by the Materials Engineering Grou ) (MEG). This meant that when the scales were shipped back to the ermi 2 plant, the Ouality Control (QC) Receiving Inspector was to follow a set of special inspection instructions. These special instructions stated that " identification will not be P/N (part number), but an inspection-comparision of each line item to its color designation as described in the requisition." (The requisition is the purchase rr.quisition MEG received from the engineer (s)whopreparedEDP9814.)
The QC Receiving Inspector received the scales and apparently perforrned the special inspection instructions because the QC Receiving Inspection Report was signed off, indicating that the wide range scales ordered for 032-R609 were correctly bande ,
Alsc noted was that the meter company, Ram Meter Inc., which did the scale banding, performed its own QC verification nf the scales for recorder C32-R609 and also signed off, indir>+ing that the scales were correctly bande All further work testing, and checks applicable to EDP 9814 for recorderC32-R60$weredocumentedascomplet The licensee issued a Deviation Event Report (DER) on June 6, 1991, pertaining to this matte HED 1046 - The banding of the RCIC pump discharge flow meter (E51-R614) was done incorrectly in the simulato Insteaa of a translucent, green scale tape from 635-645 GPM per PDC /MM 8473 banding was from 640-650 GFM with a green marker pe _ _ . _ _ . . _ _ _ _ . _ . _ _ _ ._._ ._.. ~. _ . _ _ _ _ _ -
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Itanding for the RCIC flow meter in the control room was correc Banding for the HPCI flow meter was correct in the control room and in the straulato HEDs 1233, 1374, & 1375 - Inspection of the Remote Shutdowa PancI (RSP) in the
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siinulator showed a lack of the modification implenxntation s secified in these three HEDs. Many dissimilarities existed between t1e
, simulator RSP and the plant RSP. 1he licensee explained that the prioritiration of modifications to the RSP were below those modifications to be made to the ni6in control par els in the simulator. Therefore, the RSP has not been updated yet to conform with the plan There is a two year " grace" period for changes to the simulator af ter they have been made in the plant per rcgulatory requirement The inspector noted that as of June 13, 1991, approxirnately 150 of the 300-400 Priority Ill HEDs were closed out. _However, each panel in the control '
room still has over 50% of the Priority Ill HEDs identified not corr.plete Sp cial attention was paid to HED 945 and HED 775. These two HEDs are " lead" HEDs which address the intent of other associated HEDs. HED 945 deals with label discrepancies in the control room and has 102 associated HEDs. HED 775, which deals with the banding of indicators / recorders in the control rocin, has 35 associated HED The commitment for coinpletion of Priority Ill HEDs is September 1992. This is a 16 month extension f rota the first consnitment date of completion which was the second refuel outage.
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c.ur, c IRSPECTION DEC PDC /MM, HED . PANEL ; PRIORITY DISCREPANCY CCRRECTIVE ACTIONS' OBSERVATIGh5 STATUS etc. f's-467 808 I Drywell Cooling Fy Norm. Only flood up Simulator OK C Meeting
& Area Temp. Div. I & b el indication is (4/17/91) Minutes:
817 II Recorders (T47- affected by high Contrri Rm.-old i NE-PJ-89 -
R803 A&B). Ranges drywell temps. All indicator 0024 &
are not adequate to other level replaced PDC 64E0 determine the highest indications remain (4/17/91)
run temperature availabl ._____.._______-__-__________________-_.___________ _ _ - ___-_-__--__---_-_-_,.-____-__ .-__-------
468 601 11 RCIC Turbine Speed _______-_---__--__--___(SimulatorOK Band indicator amber IC EDP 9810 Indicator Meter below 2100 rp (;/27/91)
(E51-R700) is banded Control Rm OK incorrectl It (4/12/91)
is banded green below 2I00 rpe;
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operation telow 2I00 rpm may I result in unstable system operation and squip. damag ;
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476 601 I Drywell & Torus Air none. Drywell ter Simulator OK C Meeting
& & Water Temp. Div. I will not exceed 400'F (3/27/91) Minutes 602 & II Recorders (T50- during accident Control Rm. OK f4E-PJ-87-R800 A&B) scale condition. Maximum (4/12/91) 0702 ranges are not temp. is 340*F PSTGs ar.d E0PS (10/30/87)
adequate to determine according to UFSA have been teep. range according changed to to PSTGs & E0P conform with Scales go from UFSA *F, should be fro 0-559* __-__ _
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477 601 I Drywell & Torus Air T50-PSCO A&S do have T50-R800 A&B C EDP 8472
& & Water Terp. Div..I pt. select capabilit recorders will 602 & il Recorders (T23- T23-R800 will be both be replaced R8GO & T50-R800 A&B) replaced with a with digital l don't have pt. select digital recorde recorder t captbilit Control rn:
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end of RF0 Simulator: vill be replaced by
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- m tDP WR, INSPECTION NRC PDC /M CORRECTIVE ACTIDMS OBSERVATIONS STATUS etc. f*s HED PANEL- FRIORITY DISCREPANCY Reference HED 47 T50-R800 A&B C Meetirg 624 601 1- Drywell &' Torus Air Minutes:
& & Water Temp. Div. I No action taken, recorders will T50-R800 A&B brth both be replaced NE-PJ-88-602 & II Recorders .
w'in iigital 0052 (T50-R800 A&B) don't do have pt. select have pt. select capabilit ,recor'er , (1/19/88)
capabilit f n,ne mi rm:
re faced at (m' of RF0 Stalator: will
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Provide averaging Simulator OK C EDP 8472 645 601 I Torus Bulk Water Tem Indicator.on Meter capabilit (3/27/91)
T23-R800 doesn't have Control re. OK averaging capabilit (4/12/91)
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Drywell Atm. Analysis ;None. Color coding Simulator OK IO Meeting 703 808 II Minutes:
Hydrogen /0xygen I & II scale from 2-20 (4/17/91)
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won. significantly Control rm. OK NE-PJ-89-817 recorders (T50-R806 0102 A&B) require scale enhance instrumen (4/1791)
Alarm exists to Noted in an (3/29/89)
banding at greater than 2% to verify announce to operator evaluation of.HED primary containment that an EOF condition : by Irpell Corp.:
hydrogen conc. is . exists. Upon receipt ~ existing scale greater than 2 . of alarm, operator . deviates from is able to trend stnds. established l
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hydrogen conc. on at Fermi 2 Control I recorder. Flydrogen . Room Design conc. is also Guideline available en ERI Question stiTT remains unanswered
'as to why scale wasn't changed.
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. _ _ so , .
INSPECTION !NEC PDC /MM, HED PANEL 1 PRIORITY DISCREPANCY CORRECTIVE ACTIONS' OBSERVATIONS STATUS etc. f's !
"
, t
'
i /04. 601 'Il Drywell Narrow & None. Scale goes Simulator OK Meeting i
, &- Wide Range & Torus froc 0-80 psi (3/27/91) Minutes:
l 602 Press.- Recorder Banding from 0-1.68 Control Rm. OK NE-PJ-89- ;
(T50-R8028) requires psig wouldn't (4/17/91) 013I i
- . scale banding at significantly There are still . (3/31/89) .!
j' less than 1.68 psi enhance some concerns !
i to determine Torus instrum,entatio that the inspector ! pressur has'about this
'
11ED. They are on pages 21 and 22 of the repor ' 602 II lilPCI Turb. Speed Meter Band indicator: Simulator OK C EDP 9810
,
(E41-R700)' requires Green: 2100-4000 rpm (3/27/91) I j scale banding at less Amber: 0-2100, 4000- Control rm. OK than 2100 rpm in order 4950 rpm (4/12/91) ;
to det. HPCI turbine Red: 4950-6000 rpm j
speed.
1 i
!. t
.
I r
'
I i
} }
- .
i r
! i
!
,
t ;
I
< ,
l L
- l
- - t I l i !
?
s' !
'
!
, . -
!
, - ,. .-- , , . - - _ . - _ - . . , . - . - -
. u,r , ...
INSPECTICN NRC PDC /MM, HED PANEL PRIORITY DISCREPANCY CORRECTIVE ACTIONS DBSERVATIONS STATUS etc. #'s
.
707 808 II SGTS Exhaust Cas Flow Band indicator: Simulator OK C EDP 9812
& Div. I & II Recorders Green: 3800-4180 (4/17/91) WU 024C 817 (T46-R800A&B) require CFM Control rm. OK 89 C519
. scale-banding at I (4/17/91)
operating r'nge to Mtg. c:inutes indicate acceptable NE PF-89-0102 flow rat (3/29/89) on HED 707 have banding range documented I incorrectl _____
708 603 II Reactor flow Feedwater None. Situation '
Simulator OK C Meeting
,
& Steam Recorder observed is an (3/27/91) Minutes:
(C32-R607) requires abnormal conditio Control rm. OK NE-PJ-89-
- scale banding at the Recorder is banded (4/12/91) 0275
- maximum flew rate of for normal orerating (4/18/89)
'
the condensate system condition & PDC 9814
- injection flo .:---------- .------------------------------- --
711 601 11 Post Accident Pon None. Banding wide & Simulator OK L C Meeting
& React. Press. & Lv narrow ange reactor (3/27/91) Minutes:
602 Recorders (521-R623A&B) water levels would Control rm. OK NE-PJ-89-require scale banding result in confusion (4/17/91) 0474
'
at less than 3 because scale ranges (3/31/89)
- inches to verify that are different between
-
RPV water level is the tw below 31.8 inche l _ _ _ _
________ __-- __
_ _ _ _ - . _ - -- _ __-
715 602 II HPCI Turbine Discharce None. Color banding Simulator OK C Meeting
- FicwMeter(E41-R613) is already provided (4/17/91) Miautes
! requires scale banding on turbine speed Control rm. OK NE-PJ-89-in operating range in indicator and (4/I7/91) l0024 (3/27/89)
l order to deterr: ire controllers.
< if HPCI injection flow & PDC 9816 rate is in cperating ra r.g =--
716 l 60 II Coolina Water Flow Band indicator: Siculator OK C EDP 9814 PeterlCII-R605) Red: N/A (3/27/91)
requires scale banding Amber: 0-35, Control rm. OK in operating range in 70-80 gpm (4/12/91)
order to determine if Green: 35-70 gpm CRD is operatin c
_
-
. .
s,. ..
)
. . . . - ,
i PDC /MM, INSPECTION NRC
- 0BSERVATIONS STATUS etc. 1's PANEL PRIORITY DISCREPANCY- CORRECTIVE ACTION HED
' Simulator OK C EDP 9814 717 II Reactor Level Recorder Band indicator:
' 603' Red: 160-173.5, (3/27/91)
(C32-R614) requires scale banding at less 214-220 i Control rm. OK than 171.9 inches in Amber: 173.5-193, (4/12/91).
order to confirm that 201.5-214 i Scale banding RPV water level is Green: 193-201.5 i ranges are belcw 171.9 inche incorrect in EDP written tex Correct banding ranges are provided in attached EDP drawings and supporting documentatio ..---------4 ------- __--_--_-_-_---. ___---------_-_----__ --_-_---__--__----__--- ----------.__--_-_---
- _---.-------
Simulator OK C EDP 9814 718 603 11 Reactor Level A, B & Band inoicators:
C Meters (C32-R606A, Red: 160-173.5, (3/27/91)
B, & C) require scale 214-220 i Control rm. OK bandin Amber: 173.5-193 (4/I2/91)
201.5-214 i Green: 193-201.5 i Wide range banding 0 EDP 9814 603 Reactor Press. Wide & Band indicator:
719 l II Marrow Range Recorder Wide Range: is incorrect for (C32-R609) requires Red: 1060-1200 psig both the simulator scale banding at more Amber: 780-920, 1040- and control roo than 1088 psig. in 1060 psig Banding goes from Green: 920-1040 psig 760-920 psi accordance with RPV press. guideline Narrow Range: instead of 780-920 Red: 1067.5-1150 psig psi Amber: 850-920, 1040- Note pages 22 and 1067.5 psig 23 of the repor Greer,: 920-1040 psig -----------
_------------------ .---- ------------------------
-- --.---_-------.------------_---_------------------ Simulator OK C EDP 9814 720 603 11 Pressure Reactor Loep Band indicators-A&B Meters (C32-R605 Red: 1065-1200 psig (3/27/91)
A&B) require scale Anber: 785-920 Centrol rm. OK banding at more than 1040-1065 psig (4/12/91)
1088 psig. according Green: 920-1040 psig
~
to RPV pressure guideline . -
- -
. , , . . . . . , _
- .
.
. .h ' ~
--
_ _
_ _ _ .
wr , ..
INSPECTION NRC PDC /161, HED PANEL PRIORITY' DISCREPANCY CORRECTIVE ACTIONS OBSERVATIONS STATUS etc. f's 721 601 11 Post Accident' Monitor Band indicators: Simulator OK C WR#013C
- & React. Press & Lv Red
- 1060-1500 psig (3/27/91) 890626 I
! 602 Recorders (B21-R523 Arber: 790-920, Control rm. OK & EDP 9315 A&B) require scal psig (4/12/91)
'
banding at more than - Green: 920_1040 psig
'
1088 psig according to RPV pressure i guideline ,
'
723 603 II APRM A&C, APRM B&D, None. Upscale alarm Simulator OK C Meeting RBM A & APRM E, RBM and trip are flow (3/27/91) Minutes: ;
'
B & APRM F recorders biased and not fixe Control rm. OK NE-PJ-89- ,
(C51-R603 A, B, C & Color banding would (4/12/91) 0275 ;
D) require high scale result in operatcr & PDC 9814 >
banding to confirm confusio ,
neutron flux upscale.
j _ _ _ _ _ _ . __________ __________ _._______________________,_____________________________________________.,____________________
i 725 804 II Main Condenser Vacuum Band indicator: Simulator OK jC EDP 9817 ,
j Press. Meter (N30-R823) Red: 2.7-15 psia (4/17/91)
i requires banding at the Amber: 2.2 2.7 psia Control rm. OK low vacuum trip to Green: 0-2.2 psia (4/17/91) !
I determine if mai *
!
cordenser low vacuum trip has occurre II Reactor Press. Wide & Band indicator: Wide range band'ig 0 EDP 9814 Narrow Range Recorder Wide dange: is incorrect for (C32-R609) requires Red: 1069-1200 psig both the simulator !
banding at less than Amber: 780-920, 1040 and control roo i the RHR interlock 1060 psig Banding goes from ;
pressur Green: 920-1040 psig 760-920 psig t Ndrrow Rance: instead of 780-920 :
Red: 1067.3-1150 psig psig. Also, the !
Amber: 850-920, 1040- discrepancy 1067.5 psig observation was Green: 920-1040 psig made for an .
emergency conditio ,
. Recorder is banded I in the operating rang t
' ~
Note page 22 :
and 23 of the j
-
repor .
_ _ - _ _ _ _ - _ _ _
- _ - _ _
tur,an, INSPECTION NRC PDC /MM, OBSERVATIONS STATUS etc. #'s PANEL PRIORITY DISCREPANr,Y CORRECTIVE ACTIONS -
HED Sirulator OK C EDP 9814 727 603 11 Press. React. Loop-A&B Band indicators:
Meters (C32_R605 A&B) Red: 1060-1200 psig (3/27/91)
require scale banding Amber: 785-920, Control rm. OK at less than the RHR 1040_1060 psig (4/12/91)
interlock press. in Green: 920-1040 psig Banding is in order to determine if normal range of reactor vessel pres opera tio is above RHR interlock The discrepancy pres observation was made in accordance for an emergency situatio =- -- --
_____________________ ________
= - - - - - - - - _ _ _ _ _ . _ _ _
-
Simulator OK C WR#013C89 728 601 II Post Accident Moni Band indicators: 0626
& React. Press. & Lv Red: 1060-1500 psig (3/27/91)
Amber: 790-920, Control rm. OK & EDP 9815 602 Recorders (B21-R623 A&B) require banding 1040-1060 psig (4/12/91) (4/17/91)
at.less than RHR Green: 920-1040 psig Banding is in norral interlock pres range of operatio The discrepancy observ. tion was made in accordance for an emergency situatio .___________.__________L________________________ _______________________=
---
_ _ _ _ _ _ .
None. Color banding Simulator OK C Meeting 729 603 II Standby Liquid Con Minutes:
Pumps A&B Discharge isn't required because (3/27/91)
Control rm. OK NE-PJ_89-Press. Meters (C41- system doesn't have a defined operating range (4/12/91) 0275 R600) require scale (4/18/80)
bar. ding within since it is pressure
& PDC 9814 operating range in dependen order to determine if Standby Liquid Control System has been initiate __ __________________ _______--
--
____
.__________ _____________ _________ _______ ___ ___
_ _ _ _ . .___________.__________
cmployed for normal (3/27/91) Minutes:
Eecorders (CSI-R602 NE-PJ-89-A&B) require banding operating condition Control om. OK (4/12/91) 0275 within their operating These recorders are (4/18/89)
range to determine if useful only during
& FDC 9814 standby liquid. control reactor startup.
- system has been During normal initiate operations, SRM detectors are witt. drawn
' ' and recorder reads
.. ..
.. ..
. .. .. .. ,
full scal ______
r or, nn, .
INSPECTION NRC PDC /MM, j HED PANEL FRIORIT1 DISCREPANCY CORRECTIVE ACTIONS OBSERi/ATIONS STATUS etc. f's
\
.
,
731 l 60 I Standby Liquid Tank Band indicator: l Simulator OK C EDP 1021 j r Level Meter (C41-R601) ' Amber: 0-60,71-130 : (4/12/91) !
1 requires banding inches Control rm. OK l within operating range Green: 60 71 inches I (4/12/91) i i in order to determine : 1 i if standby liquid l 1 * control has been j j f initiate ; i
_____ _________ -___________,________________________.._________________________ - _____________ _ _ _ _ _ _ _
_
,
i 732 601 Core Spray Loop A&B Band indicators: * Simulator OK ________l.___EDP C 9813 ;
Amber: 7900-10,000 (3/27/91)
'
& Flow Meters (E21 -
3 602 R601 A&B) require GPM Control rm. OK I j , banding within Green: 0-7900 GPM (4/12/91)
j g
'
j operating range to g show that core spray l j rate is correc , ;
j 733 602 II HPCI Turbine Pump None. Banding is I Simulator OK } C' Meeting :
t
,'
, Discharge Flow Meter provided en turbine ! (3/27/91) Minutes: l
,
(E41-R613) requires speed indicator & : Control rm. OK NE-PJ-89- l 4 banding to determine controller j (4/17/91) 0124 j i- , if HPCI flow rate ; & PDC 9816 :
- ) is in operating rang ___ ______-___________,._________________________'_________________________'__________________ , !
! _____
________3____________ ;
'
} 734 809 II EDG II Volts Meter Band indicator and i Sirulator OK C I EDP 9818
& g (R30-R804) requires also the other output (4/17/91) l i- 810 i banding to determine voltage indicators Control rm. OK '
,
,
i L if the EDG generators assoc'd with EDGs 12 (4/17/91) l 13, & 14 (EDG 13 &
'
output voltage is in Notation should i j operating rang are on panel 810) be made on HED 7 i p ,
Green: 108-132 volts '
to also reference i the change on ,
panel 81 !
r
> ! ,
.
.
i !
i* i i
! ;
e t t
I l !
! !
'
-
j j . .
- ,.
i
'
. e ^
'
-
-
____________t _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
<
tve, cn, INSPECTION NRC PDC /MM,
'
HED PANEL PRIORITY DISCREPANCY CORRECTIVE ACTIONS '0BSERVATIONS STATUS etc. #*s
.
855 603 I Recirc. MG Generator Add labels just above Simulator OK C PDC /MM
"
Speed Control & the pushbuttons to (3/27/91) 8475 Feedwater Flow Contro identify controller Control rm. OK '
Above two centrols are function for both the (4/12/91)
similar in appearance feedwater and recir and could be mistaten pump controller and operated in erro !
'
Controls'are close together and are !
operated while watching _ indicators on another panel sectio _ _ _ _ _ _ .. _ _ _ _ _ _ _ _ _ _ ..______________________________________________________________________________________________________
916 602 I Drywell Narrow & Wide None. Already Sinulator OK C Meeting Range & Torus Pres corrected to PSI (3/27/91) Minutes:
is scaled in " inches Control rm. OK NE-NS-87- '
of H,0." (4/12/91) 0191 ,
.
I (6/19/87) ;
918 601 I Post Accident Moni None. Already Simulator OK C Meeting j
& React. Press. & Lv corrected to PSI (3/27/91) Minutes: ;
602 is in " inches of Control rm. OK EE-NS-87- i H2 0." (4/12/91) 0191 (6/19/87)
933 809 11 Div i 480V ESF Bus Color padding will be Simulator OK C WR #009 i Volts; 480V Equip Bus added. Colc.r: Taupe (6/5/91) C890927 i Volts; and Div I 480V Control rm. OK EDP 10317 ;
'
EDG Bus Volts. The (6/4/91)
switched 480V Eu ;
Voltn=ters aren't t located consistently l with respect to the j buses on Panel 809 & !
810 and shculd be separated by some i Lind of demarcatio BIO Div. II 480 V EDG Eus (sare as above) Simulator OK C Volts: Div II 480 V (6/5/91) i
-
Equip Bus Volts; Div Control rm. OK !
II 480 V ESF Eus Volts; (6/4/91)
- , (same' reasons os above) ;
t
. + + - . - - - . . ____.____..4 -
_ _ _ - _ _ . _ _ _ _ _ _
.- - . - - _ _ . _
cur, en, INSPECT 10f! NFC PDC /MM, HED PANEL PRIORITY DISCREPANCY CORRECTIVE ACTIONS OBSERVATIONS STATUS etc. #'s
.
1015 603 II Reacter Water Level Split alarm window, Simulator OK C EDP 8476 High/ Low Annunciator " React. Water Level (3/27/91)
(3D156). LER 85-035: High/ Low," into 2 Control rm. OK A low water level alarm windows, one (4/12/91)
scram occurred due to for high level and
,
operator inattention one for low leve . to annunciated alarms Both alarm windows
'
during a period of are to be color coded high control room ambe activity.
>
1046 '601 I RCIC & HPCI Turbine Band indicators: This is one part l 0 PCD/MM i
& Pump Discharge Flow FCIC: Green: 635-645 of HED 104 Meters (E51-R614 & GPM Unable to verify E41-R614) require HPCI: Green: 5150-5250 the rest scale banding on the- GPR- Control rm. OK
, controller tape (4/17/91)
for both RCIC &
HPC Simulator: HPCI OK (4/17/91)
hCIC: Banded incorrectly from 640-650 GPM instead l
i of 635-645 1 P GPM. Note page
'
'23 of the report.
'.
'
1063 807 II "TWMS Water Level Torus level indication Simulator OK C PDChMk Trouble" Annunciator and annunciation are (4/I7/91) 9823 (70,5). available on panel 807 Control rm. OK LERs 78 n69 ,82-006: in main control roo (4/17/91)
Torus level ~-dicator Legend on annunciator
^
is locatec back says "TWMS Water Level panel e- Troable." This isn't
para: ter t definitive for the i annunciate parameter monitored.
1~
Chaage to " Torus Water
-
Level Trouble.'
. -
. _ ___ ___- __ _--- _-
!
' tur, en, INSPECTION _NRC PDC /MM, CORRECTIVE ACTIONS OBSERVATIONS STATUS etc. f's HED PANEL PRIORITY DISCREFANCY Provide color paddin Simulator OK C WR #09C 1092 809 I Thel 4160 V ESS Bus 890927
& Breaker Controis (EA-5, Color: light gree j(6/5/91) *
EB-5, EC-5, & ED-5) r ontrol rm. OK & EDP 10317 810 are located in an I(6/4/91)
ddjacent Position to to the Auto-Manual Operation Selector Switch for the Emergency Diesel Generator (EDG)
Output Breakers (EA-3, E2-3, EC-3,;& EC-3),
respectively. There-fore, when proceeding to operate the controls for the Energency DieselGenerator(EDG)
Output Breakers (EA-3, EB-3, EC-3, or EU-3),
it has been found that they can be easily confused with the controls for the 4160V ESS Bus Breakers (EA-5,EB-5,EC-5,&
ED-5) respectivel _=------ -__-- - ------.=- ----.....
CMC Switches on the Replace with switches Simulator OK C PDC /MM 1109 RSP II'
9826 100 Remote Shutdown Panel having white lettering (4/17/91)
(RSP) have black on the green & red RSP OK (4/17/91) '
lettering on the area green and red areas of the switch. This provides poor contrast when these areas are illuminate l
.- ,
..
.
.
.. . . . . .
.
.. ..
. . _ . _
_ _ _ _ _ _
- - .
..
- cur,en, INSPECTION NRC PDC /NM, CORRECTIVE ACTIONS OBSERVATIONS STATUS etc. #*s HED_ PANE PRIORITY' DISCREPANCY Employ color padding Simulator GK C WR #09C 1146 603 Feedwater Flow A 890927 Control; Feedwater to emphasize the (3/27/91) l grouping of feedwater Control rm. OK & EDP Flow E Control; and recirc. controls (4/12/91) 10317 Recirc MG A Ge Speed Cont.; Recirc MG B Gen. Speed Cont.; Recirc.- F low Master Centrol; "B" Recirc. Controller &
"A" Feedwater Flow t
Controller are all identical except for their labels and are I located next to each
.othe To facilitate rapid identification-and " blind" manipulation of these controls,
!
additional coding is neede t l
-
t
.-
_
_
.
.. . , _
. . . . . .
. ..
..
- . .
---_ _
-
._i.-
. _ _ _ _ _ _ _ _ _ - _
'
i
'
tur, an, INSPECTION URC PDC /M HED PANEL FRIORITY , DISCREPANCY CORRECTIVE ACTIONS OBSERVATIONS STATUS etc.'#*s l' i
- 1176 602 II Flow Controller Place a label in form Simulator OK C EDP 9827
- instrument provides of a decal on face of (3/27/91)
! ' demand as opposed to instr. beneath percent Control rm. OK
!. status info, and-is open/ closed scale (4/12/91)
not identified as suc reading """ demand.
f
. 603 'Recirc. Flow Master (same as above) Simulator OK C Control; Recirc. MG A (3/27/91)
'
Gen. Speed Controller; Control re. OK
- 4 Recirc. MG B Gen Speed (4/12/91)
i Centroller feedwater
! Flow A Control; and l J Feedwater Control B l Control-demand as i opposed to status inf is provided.
l, i 805 Header Drains Start-up (same as abovtl Simulator OK C
LCV I & II (4/17/91)
l - demand as opposed Control rm. OK l to status inf (4/17/91)
is provided.
j;
} 808 Adsorber RM Fan Coil Non Incorrect Simulator OK C 1 Unit I; Adsorber EM observatio (4/17/91)
{ fan Coil Unit 2; and Control rm. OK
{ Adsorber RN - Coil (4/17/91)
j Unit 3 - der ..J as
- opposed to status i info. is provide ,
i i
,
l
- .-
!
' -- ._ _-
-- - _ _ _ _ _ _
_ _ - _ _ _ _ _ _ - ,
i tur, u, INSPECTION LRC PDC /MM, HED PANEL PRIORITY DISCREPANCY CORRECTIVE ACTICNS OBSEPVATIONS STATUS etc. f's 123; RSP II Drywell pressure Provide an indication Sirulator - 0 PCD/m (100) indicator (C35-R003) of units on the scale says PSI not PSIG 9826 on RSP (panel 100) for the drywell (4/17/91)
i doesn't have units pressure indicator ir: RSP-0K (4/17/91)
indicated en display the form of a srC l Note pages 23 and fac gravoply label 24 of the repor dttaChed to the face w/the inscript.'on
"PSIG."
(Existing label on RPV press ' Wtrator which reads " PSI" will be removed F l replaced with a gravoply label with the inscription i i l "PSIG.") l
______t__________.____________________________________ ____________________________________________L_____________________
1290 601 II Reactor Water Level Previde scale for Simulator OK l C ' PDC /MM Indicator contains a indicator which (3/27/91) i 9E29 scale with white displays black Control rm. OK !
markings en a black narkings on a white (4/12/91) !
background instead of backgroun black carkings on a white backgroun Reactor Water Level (sate as above) Sirolator OK C Indicator (reasons (3/27/91)
Centrol rm. OK
-
sare as cbove) '
(4/12/91) ;
603 Reactor Vessel Flood (same as above) Simulator Or C Up Level Indicator (3/27/91)
(reasons same as .Centrol rm. OK dbcVe) (4/12/91)
f !
812 Gen. Voltage & Field (sa:e as above) Sinulator OK l C Current Indicator (4/17/91)
'
(reasons sane as Control re. OK
.
above) (4/17/91)
RSP Reactor Level Wide (same as above) Simulator OK C
.- (100) Range (reasons same (6/5/91)
as above) Control rc. OK
[r./z AMll
_ _ _ _ _
- - - .
tur. .n, i NRC PDC /MM, INSPECTION OBSERVATIONS STATUS etc. #'s PANEL PRIORITY DISCREPANCY CORRECTIVE ACTIONS HED II Mimic showing nitrogen Delete label & mimic Simulator - 0 PDC /MM 1374 RSP work 9826 (100) supplies to the SRVs for notive force to is incorrectly labeled be consistent with partially done NITROGEN SUPPLY FROM the SRV display in (6/5/91)
f INSERTING LINE to the main control Control rm. OK SRVs. The inserting roo (6/4/91)
line is not the Note pages 23 and 24 of the repor source of nitroge n--------- -----------
.--------- L ----------- -------------------------
- - - - - . .
1375 RSP II Component ----------for Provide labels Simulator - 0 PDC /MM representations component work not (e.g. , drywell, representations on complete l
'
suppression chamber) mimic (6/5/91) l on the mimics aren't Control rm. OK properly labele (6/4/91)
I Note pages 23 and 24 of the repor $ I Provide appropriate Sirolator OK C PDC /MM I408 601 II Mimic termination 9830
& labels are needed for !abels for RHR mimic (4/17/91)
RHR HD Spray, D',! Spray, ternination pts.: Head Control rm. OK 602 (4/I7/91)
Pool Spray, and Spray, Containment Cooling Line Spray, & LPCI. Remove white mimic which represents drywell on pane Simulator O C PDC /MM 601 II On the RHR mimics, Replace wide mimic 9830 1411 (4/17/91)
only the Injecticn and lines with narrower
&
Pool Cooling mimics lines where Controi rm. OK 602 should be replaced by appropriate to (4/17/91)
wide mimic lines. RHR represent smaller Pool Spray, etc. , size pipin should be represented by narrower minic lines to represent the smaller diar.eter piping system .
O
. _ _ _ _
.. _ .
l tur, an, j INSPECTION NRC~ PDC /MM,
<HED PANEL PRIORITY DISCREPA!!CY CORRECTIVE ACTIONS OBSERVATIONS STATUS etc. #*s 1- 1440 601 II Controls for SRys None. Error potential. , Simulator OK C Meeting F013 D, K, F, G, C, & is low because labels (4/12/91) Minutes:
B are presented in a and panel space Control rm. OK NE-PJ-87-
,
single row, without between controls help (4/17/91) 0702 i i extra spacing or to prevent accidental (10/30/87)
I demarcation to help actuatio locate components '
, within the array ,
- Guidelir.es 6.8.3.2.c-i (1) & 6.8.3.2c(2)
j
'
state. that more than !
5 in a row or a string
"
of core than 5 should
{. be broken up by spacing i
! and demarcatio i
{ 1498 603 II Recirc. MG A Ge No correction Siculator OK C ! e M ting
'
Speed Cont.; Recirc necessary. Control (4/17/91) Minutes:
MG B Gen. Speed Cont.; settings on Control om. OK riE-PJ-87- :
Feedwater Flow A controllers can be (4/12/91) 0722 l I
Control; and Feedwater easily adjusted with (11/19/87) !
Flow B Centro existir.c knob .
Dimensions of skirted i bias adjustment knobs en the above GEMAC controllers are less than those recommended i i by NUREG-07CO guideline t t
r I
i
,-
?
I i
.. ;
r
-
__ $ . ... - - - __ . . . _ , _ _ _ w--_ _ _ _ _