IR 05000277/2002011
| ML022180114 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 08/05/2002 |
| From: | Doerflein L Division of Reactor Safety I |
| To: | Skolds J Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-02-011 | |
| Download: ML022180114 (22) | |
Text
August 5, 2002
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION - NRC INSPECTION REPORT 50-277/02-011, 50-278/02-011
Dear Mr. Skolds:
On June 21, 2002, the NRC completed an inspection at the Peach Bottom Atomic Power Station. The enclosed report documents the inspection findings which were discussed on June 21, 2002, with Mr. Gordon Johnston and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety system design and performance capability of the residual heat removal (RHR) and high pressure coolant injection (HPCI) systems and compliance with the Commissions rules and regulations. The inspection consisted of a selected examination of calculations, drawings, procedures and records, observations of activities and interviews with personnel.
Based on the results of this inspection, the team identified two findings of very low safety significance (Green), one of which was determined to involve a violation of NRC requirements.
However, because of its very low safety significance and because the issue has been entered into your corrective action program, the NRC is treating this issue as a non-cited violation, in accordance with Section VI.A.1 of the NRCs Enforcement Policy, issued May 1, 2000, (65FR25368). If you contest this non-cited violation, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Peach Bottom facility.
John L.Skolds
In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web-site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Lawrence T. Doerflein, Chief Systems Branch Division of Reactor Safety Docket No.
50-277,50-278 License No.
DPR-44,DPR-56
Enclosure:
Inspection Report 50-277/02-011 and 50-278/02-011
REGION I==
Docket No.
50-277, 50-278 Licensee No.
50-277/02-011, 50-278/02-011 Licensee:
Exelon Generation Company, LLC Correspondence Control Desk 200 Exelon Way, KSA 1-N-1 Kennett Square, PA 19348 Facility:
Peach Bottom Atomic Power Station Units 2 and 3 Location:
1848 Lay Road Delta, Pennsylvania Dates:
June 3 - 7 and June 17-21, 2002 Inspectors:
F. Arner, Senior Project Engineer S. Chaudhary, Reactor Inspector F. Jaxheimer, Reactor Inspector R. Moore, Reactor Inspector (Part time)
B. Norris, Senior Reactor Inspector L. Scholl, Senior Reactor Inspector P. Wagner, USNRC Contractor Approved by:
Lawrence T. Doerflein, Chief Systems Branch Division of Reactor Safety
ii SUMMARY OF FINDINGS IR 05000277-02-011, IR 05000278-02-011; Exelon Generation Company; on 06/03-06/21/2002; Peach Bottom Atomic Power Station; Units 2 and 3. Safety System Design and Performance Capability.
The inspection was conducted by five region I inspectors, one region II inspector (part time),
and one NRC contractor. Two findings of very low safety significance (Green) were identified, one of which was considered a non-cited violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609 Significance Determination Process (SDP). Findings for which the SDP does not apply may be green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
Cornerstone: Mitigating Systems Green:
The team identified a finding concerning an inadequate emergency operating procedure (EOP) for returning the suction of the high pressure coolant injection (HPCI) pump to the condensate storage tank (CST) to ensure the self cooled HPCI lube oil temperatures would remain within the analyzed limit. This issue was associated with the HPCI safety function during a postulated anticipated transient without scram.
The issue was considered to be of very low safety significance (Green) based on a Phase 1 evaluation of the SDP since there was no actual loss of the HPCI system, and was determined to be a non-cited violation (NCV) of the Peach Bottom Technical Specifications, Section 5.4.1.b., Procedures.
(Section 1R21.1)
Green:
The team identified that the HPCI and Reactor Core Isolation Cooling (RCIC)
surveillance procedures incorporated steps which cycled 12 HPCI system valves and 8 RCIC valves, some several times, before the ASME in-service timing test.
The team determined that this practice was unrecognized equipment preconditioning which had the potential to mask the as found condition of the valves.
The issue was determined to be a finding of very low safety significance (Green)
based on a Phase 1 evaluation of the SDP because there was no actual loss of a valve safety function. (Section 1R21.2)
Report Details 1.
REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1R21 Safety System Design and Performance Capability (IP 71111.21)
a.
Inspection Scope The team reviewed the design and performance capability of the residual heat removal (RHR) system and the high pressure coolant injection (HPCI) system. The significance determination process (SDP) worksheets and the individual plant examination (IPE)
were reviewed to identify initiating events where these systems were credited with the capability of performing mitigating functions. Specific core damage accident sequences were selected in order to review the success criteria, including the required mission time for both the RHR and HPCI systems. The capability of HPCI to respond to a small break loss of coolant accident was reviewed. In addition, the HPCI function relative to a loss of offsite power initiating event was reviewed to ensure injection capability against a reactor vessel pressure corresponding to the setting of the lowest safety relief valves.
Furthermore, the plant risk assessment model credited HPCI as a viable injection source during a station blackout. For this scenario, the team reviewed the capability of the system to function without room cooling, with the suction of the pump aligned to the torus with increasing water temperature, while powered by the station batteries until the alternate source of power was assumed to be restored at one hour into the event. The team also reviewed the ability of the HPCI system to function in mitigating the anticipated transient without a scram (ATWS) initiating event. With respect to the RHR system, the team focused on specific modes of operation, including low pressure coolant injection (LPCI), containment (torus) cooling, and the capability of the high pressure service water pumps to perform late injection through the RHR piping for transients without the power conversion system.
The scope of the mechanical design review included: (1) a review of the HPCI turbine and governor controls including maintenance performed and test results for lubrication and control oil parameters; (2) a review of the performance of the Unit 2 and 3 HPCI booster and main pumps to ensure that the system would be able to provide the expected flow rate at the required pressure; (3) a verification that suction sources for the HPCI pump would be available during accident/special event conditions and the affect on lube oil cooling while taking suction from the torus; (4) a review of the technical adequacy of net positive suction head curves utilized in the emergency operating procedures for both the RHR and HPCI booster pumps; (5) a review of the technical adequacy of the maximum design basis differential pressures assumed for selected risk significant valves; and (6) a review of the B & C RHR pump runout protection and low pressure coolant injection loss of offsite power selection modification.
Additional mechanical design aspects reviewed included design documentation, drawings, HPCI operability determinations, calculations of RHR system capacity, RHR pump minimum flow and runout protection, and adequacy of the high pressure service water cross-tie capability for containment flooding. The impact on RHR net positive suction head (NPSH) due to the installation of the suction strainers in the torus was
reviewed. The team reviewed the availability and reliability of the RHR room heating, ventilation, and air conditioning (HVAC) equipment to provide adequate equipment space environmental conditions during normal and accident conditions. This included a review of room heat load calculations for accident conditions, and performance history of the required equipment. The team performed field walkdowns of the accessible RHR piping and HPCI equipment for Unit 2 and 3 to assess the material condition and verify that the installed configuration was consistent with design drawings and design inputs to calculations. Additionally, the team reviewed the potential for common cause failure of the RHR pumps due to potential flooding in the equipment spaces.
The team reviewed the design and performance capabilities of the electrical and instrumentation and control systems to support the operation of the RHR and HPCI systems under accident and transient conditions. These reviews included verification that selected design requirements and commitments contained in the Updated Final Safety Analysis Report (UFSAR), design documents, and industry standards were being fulfilled. Documents reviewed included drawings, calculations (including instrument setpoint and loop uncertainty calculations), engineering analyses, accident analyses, work orders and hardware modifications. The team reviewed electrical testing and operating procedures to verify selected design parameters were being tested and that recommendations and restrictions contained in the vendor technical manuals for selected components had been incorporated. For example, the team reviewed the battery testing procedures against the technical specification requirements and IEEE Standards. Additionally, vendor information in the form of service information letters (SILs) were reviewed to ensure the licensee properly evaluated and incorporated applicable recommendations.
The team evaluated the adequacy of the circuit protection features and performed independent calculations and analyses to verify that the values utilized in the licensees computer generated calculations were correct. The independent calculations included evaluations of circuit data based on conductor size and length and the type and resistance of the contractors and protective devices. The acceptability of the circuit breaker, fuse and thermal overload coordination related to selected system components was evaluated. The team also reviewed the direct current (DC) system voltage regulation to ensure adequate voltage levels were available at required loads under normal test and accident conditions. The adequacy of voltage supplied to the HPCI turbine controls was reviewed in detail.
Components selected for detailed review in the electrical area included the 125/250 volt batteries, the HPCI auxiliary oil pump (AOP), the HPCI steam admission valve (MO-14),
and the RHR pump motors. To ensure the capability of the RHR pump motors the team reviewed emergency diesel generator (EDG) loading and evaluated the circuitry utilized to accomplish load stripping along with applicable procedures to ensure the circuitry was being adequately tested. The HPCI turbine AOP and MO-14 circuit coordination calculations were evaluated and selected inputs were verified by independent calculation.
The team reviewed the procedures used to operate and test the RHR and HPCI systems during normal and accident conditions. The types of procedures reviewed included: system operating procedures, abnormal and emergency operating procedures,
alarm response cards, and surveillance tests. In particular, the impact of power re-rate on system design margins was reviewed for both the RHR and HPCI systems to verify the adequacy of acceptance criteria in system testing. The team reviewed the training lesson plans for the systems to ensure they appropriately described the design features of the systems.
The team reviewed the station blackout (SBO) procedures with respect to the assumption of placing an RHR pump in service within one hour of the blackout condition. The review included the original design, modifications related to adding the 3EA transformer as an option during maintenance, and the associated procedures. The inspector also walked down the SBO lineup in the plant, at the SBO switchyard, and at the Conowingo Hydro-Electric station to verify the capability of placing an RHR pump back in torus cooling within the one hour time-frame assumed in SBO analyses.
The team selected a sample of condition reports and action requests associated with the selected systems to verify the licensee was identifying and correcting design issues at an appropriate threshold, entering them in the corrective action program, and taking appropriate corrective actions. Documents reviewed and personnel interviewed during the inspection are listed in Attachment A.
b.
Findings
.1 High Pressure Coolant Injection Function-ATWS Analyses Introduction The inspection team identified a finding concerning an inadequate emergency operating procedure (EOP) for returning the suction of the high pressure coolant injection pump to the condensate storage tank (CST) to ensure the self cooled HPCI lube oil temperatures would remain within the analyzed limit. This issue was associated with the HPCI safety function during a postulated anticipated transient without scram. The issue was considered to be of very low safety significance (Green) since there was no actual loss of the HPCI system, and was determined to be a non-cited violation (NCV) of the Peach Bottom Technical Specifications, Section 5.4.1.b., Procedures.
Description During a review of the power re-rate analysis along with discussions with licensee personnel, the team noted that an assumption in the anticipated transient without a scram scenario was that the suction for the HPCI pump was always from the CST. The reason to maintain the suction from the preferred CST source for as long as possible, in the event of a main steam isolation valve (MSIV) closure ATWS, is because of the increasing temperature of the torus water. During this ATWS, the safety relief valves (SRVs) lift due to high pressure in the reactor vessel; the SRVs discharge to the torus (also referred to as the suppression chamber). The torus water was calculated in the licensees plant specific re-rate analysis to be as high as 188 degrees Fahrenheit. The HPCI pump is self-cooled, and the lube oil for the control system, pump and turbine bearings had been analyzed for suction water temperatures of up to 180F.
The team reviewed the TRIP procedures (Transient Response Implementation Plan -
Peach Bottoms term for the EOPs) to determine how this assumption was translated into the TRIPs. The team noted that the TRIP procedure for an ATWS (T-117, Level/Power Control) directed the use of HPCI, with the statement -CST SUCTION IS PREFERRED, DEFEAT HIGH TORUS LEVEL SWAP OVER USING T-226 IF NECESSARY. Secondary TRIP procedure T-226, Defeating HPCI High Torus Level Suction Transfer, assumed that the swap over from the CST to the torus had not already occurred. During interviews with the HPCI system engineer, the Operations Support Manager, and the EOP Program Manager, the team determined that the licensee personnel thought that there was sufficient time to defeat the swap over.
The inspectors questioned whether there was enough time for the operators to implement T-226 in the event of an ATWS. Subsequently, the licensee ran several ATWS scenarios in the simulator and found that an automatic swap over due to increasing torus level occurred between 3 and 9 minutes; i.e., before the operators would be directed to implement T-226. Although T-117 contained guidance for defeating the swap over, there was no guidance to switch the HPCI pump suction back to the CST if it had already automatically transferred to the torus on high torus water level. The licensee initiated a condition report and planned to revise T-226 to incorporate the steps for returning the HPCI suction to the CST. The inspectors considered the planned action to be reasonable.
Analysis The lack of appropriate procedural direction to maintain HPCI on its preferred suction source was considered to be more than minor because the increased suction temperature could affect the availability and reliability of the HPCI system, a mitigating system of the reactor safety cornerstone; specifically, the finding was associated with the procedure quality attribute associated with a mitigating system and affected the objective of ensuring the capability of the HPCI system. The issue was screened green in phase 1 of MC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This issue was determined to be of very low safety significance (Green) because HPCI would have remained available following an ATWS event. The conditions where HPCI may have become unavailable following an ATWS were based on a worst case analysis that assumed adverse conditions that were not currently applicable (i.e 5% of the RHR heat exchanger tubes plugged and fouling factors at design values). Additionally, the frequency of ATWS events is very low and even if HPCI was unavailable, operators could still provide vessel makeup by depressurizing the reactor and injecting with low pressure pumps.
Enforcement Technical specification (TS) 5.4.1.b requires that written procedures shall be established, implemented, and maintained covering emergency operating procedures.
Contrary to the above, the TRIP procedures were inadequate in that procedural direction did not ensure the continued operation of the HPCI system consistent with
plant specific analysis assumptions (i.e., ensuring that the HPCI pump would always take a suction from the CST during an MSIV closure ATWS condition by either defeating the high torus water level swap or by giving adequate guidance to return the suction to the CST after it had swapped). This was determined to be a violation of TS 5.4.1.b.
This issue was associated with an inspection finding that was characterized by the Significance Determination Process as having very low risk significance (i.e., Green)
and is being treated as a Non-Cited Violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue is in the licensees corrective action program as Condition Report CR-112172. (NCV 50-277;278/02-011-01)
.2 High Pressure Coolant Injection Surveillance Testing Introduction The team identified that HPCI and Reactor Core Isolation Cooling (RCIC) surveillance procedures incorporated steps which cycled 12 HPCI system valves and 8 RCIC valves, some several times, before the ASME in-service timing test. The team determined that this practice had the potential to mask the as found condition of the valves. The team noted that the valve test sequence was not in accordance with the definition of acceptable preconditioning identified in Inspection Manual Part 9900 technical guidance (Maintenance-Preconditioning of Structures, Systems, And Components Before Determining Operability), which was referenced in Inspection Procedure (IP) 71111.22, Surveillance Testing. The issue was determined to be a finding of very low safety significance (Green) because there was no actual loss of a valve safety function.
Description During a review of surveillance testing procedures ST-0-023-301-2/3, HPCI Pump, Valve Flow & Unit Cooler Functional & IST, for proper acceptance criteria, the team noted that the HPCI procedure cycled 12 valves before the ASME In service timing test.
The observation was compared to information notice (IN) 97-16, Preconditioning Of Plant Structures, Systems, And Components Before ASME Code In service Testing Or Technical Specification Surveillance Testing, and NRC Inspection Manual (IM) Part 9900 guidance on preconditioning. The surveillance activity matched the definition of preconditioning of components which was; the alteration, variation, manipulation, or adjustment of the physical condition of a component before technical specification surveillance or ASME Code testing. The basis for the valve exercising sequence in the surveillance procedures had not been evaluated for the effects of preconditioning and did not meet the conditions for acceptable preconditioning defined in technical guidance document IM Part 9900. The team found that both air operated and motor operated valves in the HPCI and RCIC systems were cycled prior to the ASME time testing. The team noted that cycling of air operated valves prior to stroke time testing can bypass or mask the as-found condition of these components. The team was concerned that motor operated stroke times may also change after initial cycles of the valves for standby systems. The relevant technical issues include pressure locking, mechanical drag when clearing the backseat, operator gearbox lubrication, and valve stem lubrication. The team noted that direct current motor speed can be affected by the above issues and
therefore the as-found condition of the valves may be masked by the current test sequence for the HPCI and RCIC systems.
The licensee indicated that procedure changes which created the testing sequence were made in response to a related significant operational experience report 81-13 and GE SIL #336, Surveillance testing recommendations for HPCI & RCIC systems, which discussed preventing preconditioning of the HPCI and RCIC turbines. However, the team determined that the majority of valves identified do not have the potential to influence turbine startup testing as discussed in the referenced documents, if they were stroke timed prior to the turbine runs. The valves included the HPCI gland seal condenser condensate pump discharge line to radwaste, the CST suction valves, torus suction valves, minimum flow line valves, and the test return valves. The licensees test sequence was not identified as a violation of a specific ASME code testing requirement.
However, IM Part 9900 has a documented position on preconditioning as it relates to the wording of 10 CFR50.55a regarding operational readiness.
Analysis Appendix E of Manual Chapter 0612 was not applicable with regard to this finding. The unrecognized valve preconditioning was determined to be more than minor because it was associated with the procedure quality attribute for the HPCI and RCIC mitigating systems and affected the mitigating system objective to ensure the availability and reliability of the applicable valves. The finding was associated with testing performed to determine the operability and reliability of the HPCI and RCIC systems and therefore was processed by Manual Chapter 0609, the significant determination process (SDP).
The team reviewed available valve data, including design margin, preventive maintenance history and actual historical valve test results and concluded that there was no indication that the associated valves could not support their safety functions. The issue was determined to be a finding of very low safety significance (Green) through a phase 1 SDP review because there was no actual loss of a safety function of a system.
Enforcement No violation of regulatory requirements occurred. Exelon entered this issue into the corrective action system as Condition Report (CR) # 00111936. (FIN 50-277;278/02-011-02)
4. OTHER ACTIVITIES (OA)
4OA6 Exit Meeting Summary The team presented the inspection results to Mr. Gordon Johnston and other members of the licensees staff at an exit meeting on June 21, 2002. Proprietary information
examined during the inspection was identified and returned to the licensee at the conclusion of the inspection.
ATTACHMENT 1 SUPPLEMENTARY INFORMATION Key Points Of Contact Exelon Generation Company M. Alfonso, Director Training J. Armstrong, Nuclear Oversight Manager C. Behrend, Branch Manager-NSSS P. Davison, Site Engineering Director M. Delowery, Senior Manager Plant Engineering D. Falcone, Operations Support Manager B. Hanson, Operations Director J. Heyne, Maintenance Support Manager G. Johnston, Plant Manager J. Jordan, Manager Mechanical Design T. LaMontange, Reactor Operator J. Lyter, EOP Program Manager J. Pomeroy-Senior Reactor Operator D. Warfel, Senior Manager Design J. Zardus, HPCI System Engineer United States Nuclear Regulatory Commission M. Buckley Resident Inspector L. Doerflein Chief, Systems Branch, R1 DRS A. McMurtray Senior Resident Inspector List of Items Opened, Closed, and Discussed Opened/Closed 50-277;278/02-011-01 NCV Trip Procedures Inconsistent With Plant Specific Analysis 50-277;278/02-011-02 Finding Preconditioning of HPCI, RCIC Valves prior to IST
LIST OF DOCUMENTS REVIEWED PROCEDURES:
A-C-1, App 3, Exh 11 Preparation of Rapid Response Cards, Revision 0 A-C-226 TRIP & SAMP Procedures Program, Revision 1 A-C-226-01 TRIP & SAMP Procedures Writers Guide, Revision 2 AO-32.2-2 HPSW Injection into the Reactor Vessel, Revision 1 GP-2 Normal Plant Start-Up, Revision 99 GP-3 Normal Plant Shutdown, Revision 90 HU-AA-104-101 Procedure Use & Adherence, Revision 0 LS-AA-105 Operability Determinations, Revision 0 LS-AA-125 Corrective Action Program (CAP) Procedure, Revision 2 NOM-C-10.2 Operations Section Performance Standards (OSPS) Introduction and Overview, Revision 0 NOM-C-7.1 Procedure Use, Revision 2 OP-PB-108-101-1001 Simple Quick Acts / Transient Acts, Revision 0 PLOR-00-05C Training Material: Station Blackout Modification P00907, Revision 0 PLOR-087P Training Material: Defeating HPCI High Level Torus Level Suction Transfer, Revision 11 PLORT-02-01B Training Material: Summer Readiness, Revision 0 PLOT-1555 Training Material: Special Events (SE), Revision 5 PLOT-2111 Training Material: T-111, Level Restoration, Revision 0 PLOT-5051 Training Material: Substations, Revision 2 PNLO-3115 Training Material: T-200 & T-300 Trip Procedures, Revision 1 PNLOC-00-03C Training Material: Mod P00907 Susquehanna 351/191 Distribution Line Enhancements, Revision 1 PNLOC-00-06B Training Material: LOOP (Back Feeding and Attachments W & Z),
Revision 0 PSEG-0215R Training Material: ATWS [T-117], Revision 14 PSEG-0417R Training Material: LOOP with no DGs Available, Revision 2 PSEG-0514L Training Material: SE-11.1, Revision 0 PSTG-A-Cautions Operator Precautions - Appendix A, Revision 3 PSTG-A-Intro Introduction, Revision 1 PSTG-A-T-101 RPV Control Guideline, Revision 7 PSTG-A-T-117 Contingency #5 Level/Power Control, Revision 9 PSTG-B-Cautions Operator Precautions - Appendix B, Revision 5 PSTG-B-Intro Introduction, Revision 1 PSTG-B-T-101 RC/Q RPV Control Guideline, Revision 7 PSTG-B-T-101 RC/RL RPV Control Guideline, Revision 8 PSTG-B-T-101 RC/P RC Guideline Part RC-P - Appendix B, Revision 5 PSTG-B-T-117 Contingency #5 Level/Power Control, Revision 10 PSTG-Cautions Operator Precautions, Revision 2 PSTG-T-101 RPV Control Guideline, Revision 6 PSTG-T-117 Contingency #5 Level/Power Control, Revision 8 RRC-10.1-2 RHR System Torus Cooling During a Plant Event, Revision 0 RRC-10.2-2 RHR System LPCI Manual Start During a Plant Event, Revision 0 RRC-11.1-2 Standby Liquid System Initiation During a Plant Event, Revision 0 RRC-13.1-2 RCIC System Operation During a Plant Event, Revision 0
RRC-14,1-2 Core Spray Manual Initiation During a Plant Event, Revision 0 RRC-16.1-2 Bypass & Restore Instrument N2 Supply to Drywell, Revision 0 RRC-1G.1-2 Automatic Depressurization System, Revision 0 RRC-1G.2-2 Relief Valve Manual Operation During a Plant Event, Revision 1 RRC-23.1-2 HPCI System Operation During a Plant Event, Revision 2 RRC-3B.1-2 Alternate Rod Injection During a Plant Event, Revision 0 RRC-44A.1-2 Maximize Drywell Cooling, Revision 2 RRC-53.1-2 Unit 2 House Loads Transfer During a Plant Event, Revision 0 RRC-55.1-2 Cross-Tie of 480V Load Centers During a Plant Event, Revision 0 RRC-7J.1-2 Drywell & Torus H2/O2 Sampling Startup During a Plant Event - CAD Mode, Revision 0 RRC-94.1-2:1 URO Scram Reports, Revision 0 RRC-94.1-2 Reactor Operator Scram Actions, Revision 0 RRC-94.2-2 Plant Reactor Operator Scram Actions, Revision 0 RRC-94.2-2:1 PRO Scram Reports, Revision 0 SAMP-1 Bases RPV & Primary Containment Flooding Control, Revision 1 SE-1 Plant Shutdown from the Remote Shutdown Panel, Revision 16 SE-1 Bases Plant Shutdown from the Remote Shutdown Panel, Revision 16 SE-10 Alternate Shut Down, Revision 11 SE-10 Att 9 HPCI Operations from the Alternative Shutdown Panel, Revision 1 SE-10 Bases Plant Shutdown from the Alternative Shutdown Panels, Revision 12 SE-11 Bases Loss of Off-Site Power, Revision 11 SE-11 Loss of Off-Site Power, Revision 12 SE-11.1 Operating Station Blackout Line During a LOOP Event, Revision 3 SO-10.1.A-2 RHR System Set-Up for Automatic Operation, Revision 3 SO-10.1.A-2A COL RHR System Set-Up for Automatic Shutdown, Revision 18 SO-10.1.A-2B COL RHR System Set-Up for Automatic Shutdown, Revision 13 SO-10.1.B-2 RHR System Shutdown Cooling Mode Manual Start, Revision 28 SO-10.1.C-2 RHR System Precise Reactor Temperature Control, Revision 4 SO-10.1.D-2 RHR System Torus Cooling, Revision 15 SO-10.2.A-2 RHR System LPCI Shutdown & Return to Standby, Revision 2 SO-10.2.B-2 RHR Shutdown Cooling Mode Shutdown, Revision 17 SO-10.3.A-2 RHR System A Loop Filling & Venting, Revision 11 SO-10.3.C-2 Manually Venting of the RHR LPCI & Containment Spray Line Vent Accumulator Lines, Revision 2 SO-10.3-2 RHR System Fuel Pool Cooling Mode, Revision 3 SO-10.5.A-2 RHR System Piping Flush, Revision 2 SO-10.7.B-2 RHR System Automatic Response During LOCA and Manual System Initiation upon Automatic Injection Failure, Revision 6 SO-10.7.D-2 RHR Shutdown Cooling Operation Through MO-2-10-020 RHR Loop X-Tie, Revision 0 SO-10.8.A-2 RHR System Routine Inspection, Revision 2 SO-14A.1.A-2 Torus Water Cleanup and Level Control, Revision 8 SO-23.1.A-2 HPCI System Setup for Automatic or Manual Operation, Revision 10 SO-23.1.B-2 HPCI System Manual Operation, Revision 15 SO-23.2.A-2 HPCI System Shutdown, Revision 14 SO-23.7.A-2 HPCI System Automatic Initiation Response, Revision 7 SO-23.7.B-2 Transfer of HPCI Pump Suction from CST to Torus, Revision, 4
SO-23.7.C-2 HPCI System Recovery from System Isolation or Turbine Trip, Revision 7 SO-32.1.A-2 HPSW System Startup and Normal Operations, Revision 11 SO-32.2.A-2 HPSW System Shutdown, Revision 6 SO-32.3.A-2 HPSW Filling and Venting the HPSW Side of the RHR Heat Exchangers, Revision 3 SO-32.7.A-2 Placing Unit 2 HPSW Loops In-Service Using Unit 3 HPSW Pumps, Revision 10 SO-32.8.A-2 HPSW System Routine Inspection, Revision 5 SO-51H.7.A SBO Breaker Rack-Out/Rack-In, Revision 2 SO-51H.7.B SBO Disconnect Switch Operations, Revision 5 SO-53.7.C Response to a Loss of #2 Off-Site Startup Source, Revision 27 SO-53.7.D Response to a Loss of #343 Off-Site Startup Source, Revision 25 SO-53.7.G Off-Site AC Power Restoration Following Loss of Grid, Revision 11 SO-53.7.P Response to a Loss of #3 Off-Site Startup Source, Revision 13 T-100 Scram, Revision 9 T-101 Bases RPV Control, Revision 21 T-101 Reactor Control, Revision 17 T-102 Primary Containment Control, Revision 13 T-102 Bases Primary Containment Control, Revision 15 T-103 Secondary Containment Control, Revision 14 T-104 Radioactivity Release, Revision 7 T-111 Level Restoration, Revision 11 T-111 Bases Level Restoration, Revision 10 T-112 Bases Emergency Blowdown, Revision 14 T-112 Emergency Blowdown, Revision 14 T-116 Bases RPV Flooding, Revision 11 T-116 RPV Flooding, Revision 11 T-117 Level/Power Control, Revision 13 T-117 Bases Level/Power Control, Revision 13 T-203-2 Initiation of Torus Sprays using RHR, Revision 1 T-204-2 Initiation of Drywell Sprays using RHR, Revision 1 T-205-2 Initiation of Containment Sprays using HPSW, Revision 1 T-226-2 Defeating HPCI High Torus Level Suction Transfer, Revision 3 T-231-2 HPSW Injection into the Torus, Revision 2 T-233-2 CST Makeup to the Torus via HPCI Minimum Flow Line, Revision 1 T-240-2 Termination and Prevention of Injection into the RPV, Revision 6 T-245-2 HPSW Injection into the RPV, Revision 2 T-250-2 RPV Pressure Control Using HPCI, Revision 3 T-251-2 RPV Pressure Control Using RCIC, Revision 3 T-BAS (Intro)
Introduction to TRIPs & SAMPs - Bases, Revision 5 T-BAS (Trip)
TRIP/SAMP Curves, Tables, and Limits - Bases, Revision 5 TSG-3.1 TRIP/SAMP Action Timing, Revision 0 IC-11-00388, rev. 6 Calibration of HPCI Turbine Generator Control System for PB SI2P-2-404-A1C2,2 Calibration Check of Reactor Pressure Loop Instruments PT/PR SI2P-10-100-B1C2 Calibration Check of Drywell Pressure Instruments
SI2P-10-120-C1CQ, Calibration Check of RHR Pump C discharge pressure switch FI3M-23-GOV-XXC2 Calibration Check of HPCI Turbine Governor RT-X-023-210-3 HPCI Flow Control Stability Test NUCLEAR OPERATIONS MANUAL NOM-C-5.2 Resetting Protective Devices/Restoring Power, Revision 0 NOM-P-5.5 Fuses and Quality Parts, Revision 1 ALARM RESPONSE CARDS 00C135-D-4 Battery Room Vent Trouble / Hi / Lo Temperature, Revision 1 207-20C236L-A-1 System I Torus Water High Temperature/Failure, Revision 3 207-20C236L-A-2 System I Torus Water High High Temperature, Revision 1 207-20C236L-A-3 System I Torus Water High High High Temperature, Revision 1 212-20C205RR-A-5 Unit 2 HPSW Bay Level High-Low, Revision 5 212-20C205RR-B-5 Emergency Cooling Tower Reservoir Level High-Low, Revision 6 221-20C204B-B5 HPCI Inverter Power Failure, Revision 2 226-20C203D-A-4 Torus Water Level out of Normal Range, Revision 1 333-30C004BX-D-1 HPCI Pump Suction Pressure Low, Revision 0 333-30C004BX-E-1 HPCI Turbine Exhaust Pressure High, Revision 0 ROUTINE & SURVEILLANCE TESTS:
RT-I-033-631-2 RHR Room Cooler ESW Heat Transfer Test, Revision 6 RT-O-010-304-2 RHR/HPSW System Valves Alternate Control Testing, Revision 5 RT-O-010-415-2 HPSW to RHR Emergency Cross-Tie Valve Functional Test, Revision 1 RT-O-023-750-2 HPCI Functional Test from Alternative Control Panels, Revision 12 RT-O-023-760-2 HPCI Valve and Component Test from Alternative Control Panel, Revision 7 RT-O-032-300-2 HPSW Pump, Valve and Flow Functional Test, Revision 10 RT-O-51H-900-2 Station Blackout Line Loading Verification, Revision 3 ST-O-023-301-3 HPCI Pump, Valve, Flow and Unit Cooler Functional and In-Service Test, Revision 30 ST-O-023-350-2 HPCI Valve Alignment and Filled and Vented Verification, Revision 1 ST-O-032-350-2 HPSW Valve Alignment Verification, Revision 0 ST-O-51H-200-2 Station Blackout Line Operability Verification, Revision 5 ST-0-054-751-2 E12 Bus LOCA/LOOP functional test, Revision 14 ST-0-054-752-2 4kV Bus LOCA/LOOP Functional Test ST-M-57B-731-3 3A Battery Performance Test Conducted 10/11/99 ST-M-57B-732-3 3B Battery Performance test conducted 9/10/99 ST-M-57B-733-3 3C Battery Performance Test conducted 9/30/99 ST-M-57B-734-3 3D Battery Performance Test conducted 9/16/99 ST-M-57B-741-3 3A Battery Service test conducted 10/18/99 ST-M-57B-742-3 3B Battery Service test conducted 10/13/99 ST-M-57B-743-3 3C Battery Service test conducted 10/16/99 ST-M-57B-744-3 3D Battery Service test conducted 10/7/99 ST-1-101-105-2 RHR Loop B LSFT ST-0-013-302-2 RCIC, Pump, Valve and Flow test
DRAWINGS:
95418C Woodward Governor Wiring 9970-613 HPCI Diagram Wiring Schematic Walworth Valve Dwg A-4088-M-23A, Cast Steel Globe Valve Bechtel Specification 6280-M-77, Special fan Cabinets DWG, E-2007085-908-008, PB units 2 and 3 RHR strainer module general arrangement E-1, Single Line Diagram Station (Sheet 1 of 4), Revision 40 E-5-7 sh.1 Schematic Diagram EDG Exciter-Regulator, Rev. 49 E-5-32 Schematic Diagram EDG Exciter-Regulator E-8 Single line meter and relay diagrams Unit 2 4160 volt systems E-10 Startup and Emergency Power, Revision 25 E-12 Unit 3, 4160 volt system schematic E-26 125/250 Volt DC System E-27 Unit 3 125/250 Volt DC System, Rev. 72 E-71 Schematic Diagram Aux. Swgr. Source Circuit Breaker, Rev. 38 E-71sh. 2 Schematic Diagram Aux. Swgr. Source Circuit Breaker, Rev. 31 E-188 sh. 1-8 Schematic Diagrams 4KV emergency bus relaying E-189 sh. 1-2 Schematic Diagrams Emergency Transformer Relaying E-188 sh. 1-8 Schematic Diagrams EDG Circuit Breakers E-405 sh. 1 HPCI Connection diagram, Rev. 34 E-405 sh. 2 HPCI connection diagram, Rev. 23 E-1615 Single Line Meter & Relay Diagram-Unit 2 MCCs, Rev. 66 E-1617 Single Line Meter & Relay Diagram-Unit 2 E-1619 Single Line Meter & Relay Diagram-Unit 2 EDGs E-1715 Single Line Meter & Relay Diagram-Unit 3 MCCs, Rev. 61 E-1717 Single Line Meter & Relay Diagram-Unit 3, Rev. 53 E-2892 sh 1 HPCI Alternative Shutdown Connection Diagram, Rev. 8 E-2894 sh. 1 Terminal Box J-1267 Connections, Rev. 2 E-2903 sh.2 Schematic Diagram U2 Alternative Control Instrumentation, Rev. 1 E-5343 Station Blackout Substation Single Line Diagram, Rev. 11 M-1-EE-222 1 Sh. 1 Connection Diagram Panel 9-48, Rev. 38 M-1-EE-222 Sh. 2 Connection Diagram Panel 9-48, Rev. 34 M-1-S-36 Sheets 1-12 HPCIS Schematic Diagrams M-1-S-65 Sheets 1-33 RHRS Schematic Diagrams M-1-S-40 Sheet 1A Automatic Actuation Relays, rev. 38 M-1-S-40 Sheet 2 Automatic Actuation Relays, Rev. 51 E-5343 Station Blackout Substation Single Line, Revision 11 M-315 Emergency Service Water & High Pressure Service Water Systems, Revision 64 M-361 Residual Heat Removal System, Revision 78 M-365 High Pressure Coolant Injection System, Revision 62 CALCULATIONS:
Loop Uncertainty for LS-3-23-074, Rev. 1, CST low level Loop Uncertainty for LS-3-23-091A, Rev. 1, HPCI Pump Source Suction Transfer Loop Uncertainty for PT-3-10-100A, Rev.00, Drywell Pressure instrumentation
Loop Uncertainty for LS-3-02-3-404A, Rev. 01, RHR pressure PE-0017 125/250 Volt DC Battery Capacity & System Voltage Analysis, Rev. 11 PE-0093 Class 1E 480 V Transformer Tap Settings, Rev. 21 PE-0121 Voltage Regulation Study, Rev. 6 PE-0155 Seismic Evaluation of Battery Racks, Rev. 1 PE-0166 EDG Loading Analysis, Rev. 5 PE-0182 125 Volt DC Voltage Analysis, Rev. 12 PE-0196 125/250 Volt DC System Coordination, Rev. 2 PE-0155 Seismic Evaluation of Battery Racks ME-693, WS13,Determination of Vortex Limits for LPCI, HPCI, CS, RCIC PM-1010 RHR Pump NPSH, Rev. 5 PM-1013 Minimum Containment Pressure Available, Rev. 3 ME-507 acceptance criteria for the ST of RHR pumps to meet the TS section 4.5.A.d PM-846 Elimination of ESE Flow requirements to RHR pumps PM-958 Calculate RHR/CS room temperature post LOCA MISLCalc 64 Verification of ECCS Strainer Pressure Drops for PB Units 2 and 3 ME-293 Pressure Drop for HPCI Injection at Flow rate of 5000 gpm, Rev. 0 ME-537 NPSH for HPCI & RCIC, Rev. 1 ME-534 Determination of vortex limits, unit 2 ME-693 Determination of vortex limits, unit 3 ME-502
- tubes allowed plugged in unit 3 HPCI lube oil cooler PM-138 Determination of dedicated CST volume for HPCI, RCIC suction, Rev. 1 PM-1013 Minimum Containment pressure available, Rev. 3 18247-M-001 Maximum torus temperature allowed for the ECCS systems 18247-M-035 CST minimum water level to prevent vortex PB-99-1212-000 Station Blackout Request Ability to Use 3EA for Loads PE-0154 Station Blackout Voltage Regulation Conowingo Source, Revision 5 VENDOR TECHNICAL MANUALS M-1-JJ-30 Terry Turbine Manual E-5-166 Fairbanks-Morse Vendor Manual E-5-167 Operation and maintenance manual for EDGs E-13-123 Exide Instructions, installing and operating station batteries
MODIFICATIONS ECR 93-03983, Replace RCIC Flow Controller FC-2-13-091 ECR98-02298, HPCI Resolution of Various Mechanical Issues; 02/02/99 ECR 98-02299,HPCI Resolution of Various Mechanical Issues; 12/07/99 ECR 99-00284, As found gaps on HPCI Cooling Water Pipe Supports PB 01-01268, Gag RHR pump suction relief valves RV-2-10-072 PB 95-05469, Removal of HPCI and RCIC unit coolers from support of TSs PB 94-05346, Replace Motor oil with Mobil SHC 624 Oil OTHER DOCUMENTS:
Teleconference Memorandum R. McNabb/T. Cabrey dated 4/28/89, DC Voltage Adequacy Eval GE SIL No. 351, Rev. 2 HPCI and RCIC turbine control system calibration GE SIL No. 480, HPCI System Startup Transient Improvement Design Basis Document (DBD) P-S-03, Rev. 19, HPCI system NUMARC 87-00 Appendix G, Rev. 1, Topical Report on Assessments of Equipment Operability in Dominant Areas Under Station Blackout conditions GE Letter to PECO dated May 25, 1972, HPCI/RCIC suction water temperature limits System No. 10, RHR 4th quarter, 2001, Units 2 &3 System No. 23, HPCI, 4th quarter 2001, Units 2 &3 System No. 10, PB Maintenance Rule Bases information System No. 23, PB Maintenance Rule Bases information Performance Indicators for HPCI and RCIC 01/98 thru 04/02 Design Reviews: Evaluation for Power Rerate, Appendix V, RHR, Appendix VII, HPCI, ATWS P-S-09, RHR system design baseline document UFSAR Section 4.8, RHR system, Rev. 18 P-S-08C, Reactor Building HVAC DBD, Rev. 11 Test Specification 6280-34, Demonstration test report, HPSW 10CFR50.59 for: T-226-2, Defeat of HPCI High Torus Level Suction Transfer, Revision 0 BWROG EPGs/SAGs, Revision 1 BWROG EPGs/SAGs Bases, Revision 1 Degraded Equipment Log, dated 06/07/02 GENEDC-32230P, Reactor Safety Performance Evaluation [Power Uprate], Section 9.3.1 NUMARC 87-00, Guidelines & Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, November 1987 PBAPS Technical Specifications PBAPS DBD No. P-S-03, High Pressure Coolant Injection System, Revision 19 PBAPS DBD No. P-S-09, Residual Heat Removal System, Revision 16 PECo Letter to NRC, dated August 6, 1992, Station Blackout Response to NRC Questions Concerning the Use of the Conowingo Hydroelectric Power Station as the Alternate AC Power Source UFSAR, Section 6.0, Core Standby Cooling Systems UFSAR, Section 4.0, Reactor Coolant System UFSAR, Section 13.0, Conduct of Operations USNRC Regulatory Guide 1.155, Station Blackout, August 1988 CONDITION REPORTS:
70324 87532 87936 98027 103009 103753 112172 10012379 00112521 198915 193919 199124 ACTION REQUESTS:
A1243766 A1246921 A1253678 A1263593 A1354409 A1367719 A1269128 A1274064 A1310094 A1317983 A1328987 A1332748 A1332921 A1333654 A1336837 A1344638 A1350942 A1365067 A1369351 A1370913 A1211046 A1373248 A1238474 A1348993 A1176584 WORK ORDERS R0547635 R0547879 R0550462 R0645522 R0744366 R0477376 NCR PB 97-02609 001 NCR-PB-94-00005 R0020268 R0029609 List Of Acronyms AOP Auxiliary Oil Pump ATWS Anticipated Transient Without Scram CST Condensate Storage Tank DBA Design Basis Accident DBD Design Bases Document DC Direct Current EDG Emergency Diesel Generator EOP Emergency Operating Procedure HPCI High Pressure Coolant Injection HPSW High Pressure Service Water HVAC Heating, Ventilation & Air Conditioning IPE Individual Plant Examination LOCA Loss of Coolant Accident LOOP Loss of Offsite Power LPCI Low Pressure Coolant Injection MSIV Main Steam Isolation Valve NPSH Net Positive Suction Head PBAPS Peach Bottom Atomic Power Station PECO Peco Energy RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RPV Reactor Pressure Vessel SAMP Severe Accident Management Plan SBO Station Black-Out
SDP Significance Determination Process SIL Service Information Letter SRV Safety Relief Valve TRIP Transient Response Implementation Plan UFSAR Updated Final Safety Analysis Report USNRC United States Nuclear Regulatory Commission V&V Verification & Validation