NRC 2009-0067, 10 CFR 50.59 Summary Report for 2008
| ML092430137 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/28/2009 |
| From: | Jim Costedio Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2009-0067 | |
| Download: ML092430137 (11) | |
Text
August 28,2009 NRC 2009-0067 I 0 CFR 50.59(d)(2)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 10 CFR 50.59 Summarv Report for 2008 NextEra Energy Point Beach (NextEra), LLC, is submitting this 10 CFR 50.59 Summary Report for the Point Beach Nuclear Plant (PBNP), Units 1 and 2, for calendar year 2008. contains descriptions of facility changes, tests and experiments evaluated in accordance with 10 CFR 50.59 during 2008. Enclosure 2 contains commitment change evaluations completed in 2008.
This letter contains no new commitments and no revisions to existing commitments.
Very truly yours, NextEra Energy Point Beach, LLC Licensing Manager Enclosures (2) cc: Administrator, Region Ill, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Project Manager, Point Beach Nuclear Plant, USNRC PCSW NextEra Energy Point Beach, LLC,6610 Nuclear Road, Two Rivers, WI 54241
ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 10 CFR 50.59
SUMMARY
REPORT FOR 2008 MODIFICATIONS, FSAR CHANGES AND OTHER EVALUATIONS Modification, Removal of Leak Chase Channels, Unit 1 Activitv
Description:
Engineering Change (EC 11836) removed existing leak chase channels (LCCs) from Unit 1 penetrations 1CPP-28 and 1CPP-34. Removal of the LCCs resolved a process pipe thermal overstress issue affecting Unit 1 containment piping penetrations 1CPP-28 and 1CPP-34, and the hot process piping running through these penetrations. The LCC at the interface between the process pipe and the containment piping penetration closure head inside containment was removed. The LCCs were also removed from the spare and cold process piping running through these penetrations inside containment for consistency.
The LCCs were installed over welds to provide for local leakage testing of the welds. NRC Safety Evaluation (SE) dated September 18, 1989, approved the redefinition of containment pressure boundary to the LCCs and LCC welds. The LCCs do not need to be vented while performing leakage rate tests.
Summary of 10 CFR 50.59 Evaluation: NRC SE dated September 18, 1989, "Containment Liner Leak Chase Channel Venting," stated, "The original purpose of the leak chase channels was to have the ability to pressure-test the liner plate or penetration welds for leaks without pressurizing the full containment structure. Since these leak chase channels are not to be vented during subsequent Type A tests, they are considered as an integral part of the liner plate and therefore a part of the leak-tight containment pressure boundary." The SE concluded that,
"(a) the channel welds at the redefined pressure boundary are qualitatively equivalent to those for the primary containment liner welds and are acceptable," and "(b) the channels are capable to withstand the loading conditions of a postulated design basis accident, as well as during normal operation and maintain their structural integrity at all times." The SE further stated, "The staff therefore concurs with the licensee that it is not necessary for Point Beach plants to vent the containment liner weld leak chase channels during a containment integrated leakage rate test (CILRT), provided that the licensee commits to comply with the requirements of 10 CFR Part 50 Appendix J, including a visual inspection of readily accessible areas prior to each subsequent Type A test."
This SE is related to the use of the LCCs and LCC welds as a DBLFPB and that the LCCs did not need to be vented during a CILRT. The liner plate and its anchorage system for structural and leak-tight integrity are not affected by the removal of these LCCs. Removing the LCCs redefines the containment liner and weld under the LCCs back to the original pressure boundary, there is no change in the license basis for the containment pressure boundary.
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FSAR changes were required to reflect the removal of the LCCs from 1CPP-28 and 1CPP-34.
These changes assure that the design information regarding the ability to perform local leak rate tests (LLRTs) of the liner plate weld seams at these penetrations are accurate.
(EVAL 2008-006)
Modification, Removal of Leak Chase Channels, Unit 2 Activitv
Description:
Engineering Change (EC 11837) removed existing LCCs from Unit 2 penetrations 2CPP-28 and 2CPP-34. Removal of the LCCs resolved a process pipe thermal overstress issue, discussed in CAP 1115726 (OPR), regarding Unit 2 containment piping penetrations 2CPP-28 and 2CPP-34 and the hot process piping running through these penetrations by removing the LCCs at the interface between the process pipe and the containment piping penetration closure head inside containment. In addition, the LCCs were removed from the spare and cold process piping running through these penetrations inside containment for consistency.
The LCCs were installed over welds to provide for local leakage testing of the welds. NRC SE dated September 18, 1989, approved the redefinition of containment pressure boundary to the LCCs and LCC welds. The LCCs do not need to be vented while performing leakage rate tests.
Summarv of 10 CFR 50.59 Evaluation: NRC SE dated September 18, 1989, "Containment Liner Leak Chase Channel Venting," stated, "The original purpose of the leak chase channels was to have the ability to pressure-test the liner plate or penetration welds for leaks without pressurizing the full containment structure. Since these leak chase channels are not to be vented during subsequent Type A tests, they are considered as an integral part of the liner plate and therefore a part of the leak-tight containment pressure boundary." The SE concluded that,
"(a) the channel welds at the redefined pressure boundary are qualitatively equivalent to those for the primary containment liner welds and are acceptable," and "(b) the channels are capable to withstand the loading conditions of a postulated design basis accident, as well as during normal operation and maintain their structural integrity at all times." The SE further stated, "The staff therefore concurs with the licensee that it is not necessary for Point Beach plants to vent the containment liner weld leak chase channels during a containment integrated leakage rate test (CILRT), provided that the licensee commits to comply with the requirements of 10 CFR Part 50 Appendix J, including a visual inspection of readily accessible areas prior to each subsequent Type A test."
This SE is related to the use of the LCCs and LCC welds as a DBLFPB and that the LCCs did not need to be vented during a CILRT. The liner plate and its anchorage system for structural and leak-tight integrity are not affected by the removal of these LCCs. Removing the LCCs redefines the containment liner and weld under the LCCs back to the original pressure boundary, there is no change in the license basis for the containment pressure boundary.
FSAR changes were required to reflect the removal of the LCCs from 2CPP-28 and 2CPP-34.
These changes assure that the design information regarding the ability to perform LLRTs of the liner plate weld seams at these penetrations are accurate. (EVAL 2008-002)
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, Revision of FSAR Sections 6.2 and 9.2, Remove Requirement to Backseat Normally Open Valves Activitv
Description:
The activity changed FSAR Sections 6.2 and 9.2 to clarify that the backseats of emergency core cooling system (ECCS) valves, do not have control functions and are not normally relied upon as primary leakage barriers.
Summarv of 10 CFR 50.59 Evaluation: ECCS valves do not perform a control function on their backseats. When an ECCS valve is open but not backseated, the valve is relying on its packing to prevent leakage. This evaluation concluded that the only credible failure mode of an open valve that is not backseated is packing leakage. (EVAL 2008-008)
Modification, Replacement of Unit 1 Feedwater Heaters 1HX-020 N B and 1HX-021 A/B Activitv
Description:
This modification (EC12033) replaced the existing Unit 1 Nos. 4 and 5 feedwater heaters (FWHs), 1HX-020NB and 1HX-021N B , with new FWHs designed to accommodate the future extended power uprate (EPU), along with updating associated plant documents. Inherent in the design of the replacement FWH is an increase in thermal performance and feedwater flow characteristics. This activity modified condensate, feedwater and feedwater heater drains system piping, as well as modifying the turbine building structural members to support installation of the new FWHs. The activity replaced the existing pneumatic level controls and mercury level instrumentation with digital technology. Revision 1 of EC 12033 incorporated the ability to operate with the fifth feedwater point heater bypass to maintain the final temperature between 425°F and 430°F.
Summarv of 10 CFR 50.59 Evaluation: This modification resulted in a change in the full power operating final feedwater temperature to the steam generators. This resulted in the potential adverse effects on the results of the Loss of Normal Feedwater (LONF), and main steam line break (MSLB) dose analyses. The change also involved an increase in the uncertainty in the leading edge flow meter (LEFM) feedwater flow measurement. The potential adverse conditions on the results of the Chapter 14 analyses and the reactor thermal output (RTO) power measurement uncertainty were evaluated. The evaluations determined that the applicable acceptance criteria for affected analyses continue to be met, the doses are sufficiently below the regulatory limits and would not result in a more than minimal increase in the consequences of an accident and the licensed power measurement uncertainty of 0.6% is not affected.
The design parameters of the installed components are consistent with current design standards for the operating conditions and will not create new or different failure modes, and will not result in an increase in probability of failures or accidents other than those previously evaluated in the FSAR.
The increase in the dose for the MSLB outside containment has been evaluated and the increase in dose does not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR. The modification and the associated evaluations do not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses. (EVAL 2008-016)
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Calculation Revision, Evaluation of the Impact of Revised Containment Heat Sink Paint Thicknesses on Containment lntegrity Analyses (MSLB and LOCA) for Units 1 and 2 Activitv
Description:
The activity revised the thicknesses of containment heat sink coatings in the loss of coolant accident (LOCA) and MSLB containment integrity analyses to bound as-found plant conditions.
Surnmarv of 10 CFR 50.59 Evaluation: The increased coating thickness reduces the rate of heat transfer to the passive structural heat sinks in containment, and results in higher peak pressures and temperatures for both LOCA and MSLB design basis accidents. The calculated peak pressures and surface temperatures remain below the design pressure and temperature of the containment buildings. The revised calculations for LOCA and MSLB were preformed using codes and methods previously approved and utilized at Point Beach Nuclear Plant (PBNP) for containment integrity analysis. The higher peak pressures and temperatures are bounded by the pressure and temperature profiles used in the environmental qualification of equipment inside containment. The DBLFPB for the containment pressure of 60 psig was not exceeded or altered by this activity.
The increased containment atmospheric temperature has a minimal effect on sump temperature, chemical effects, ECCS heat exchanger duty and ECCS piping and support stresses. There is no effect on the available net positive suction head (NPSH) for the residual heat removal (RHR) pumps.
The activity did not involve accident initiators, so the frequency of existing design basis accidents is not impacted and there are no new accidents created. (EVAL 2008-018)
Calculation Revision, Evaluation of Revised Containment Heat Sinks for MSLB Containment Integrity Analysis for Units 1 and 2 Activitv
Description:
An analysis was performed to determine the effect of using revised surface areas, volumes, and thermal conductivities of passive heat sinks inside containment for the MSLB containment integrity analysis. The changes more closely reflect current physical installations.
The net result of the combined changes was an increase in the effectiveness of the credited heat sinks inside containment. This resulted in an increased energy removal and therefore lower predicted peak pressure and temperature inside containment for the MSLB containment integrity analysis.
Analysis of the MSLB accident with the updated heat sink input values, results in a reduction in peak containment pressure and temperature. A MSLB accident with the updated heat sink input values results in peak containment pressure of 56.81 psig at 266 seconds. This is 3.03 psi lower than the 59.84 psig at 276 seconds predicted in the FSAR as based upon the analysis, and 3.16 psi lower than the 59.97 psig at 275 seconds predicted in the analysis, when the thicker coatings are considered.
Summarv of 10 CFR 50.59 Evaluation: The calculated peak pressure and temperature remain below the design pressure and temperature of the containment buildings. The revised calculation for MSLB was preformed using codes and methods previously approved and utilized at PBNP for containment integrity analysis. The resulting peak pressure and temperature are bounded by the pressure and temperature profiles used in the environmental qualification of Page 4 of 7
equipment inside containment. The DBLFPB for the containment pressure of 60 psig was not exceeded or altered by this activity. Therefore, no SSCs important to safety (including the containment buildings and components inside containment) are impacted by this change and the probability, consequences and results of SSC malfunctions are not impacted. The analysis did not involve accident initiators, so the frequency of existing design basis accidents is not impacted and no new accidents are created. (EVAL 2008-0019)
Calculation Revision, Impact of Feedwater Temperature and Volume Increases with Replacement of FWHs on MSLB Containment Response for Units 1 and 2 Activity
Description:
This activity analyzes the impact of the replacement of the Nos. 4 and 5 FWHs, HX-20NB and HX-21NB on both units with new FWHs designed to accommodate the future EPU.
The effect of the new FWHs at the current licensed core power level is an increase in the final feedwater temperature supplied to the stream generators at 100% power (from a current operating value of -430°F to a calculated value of -436"F), and an increase in the feedwater fluid volume upstream of the feedwater regulating valves.
Summarv of 10 CFR 50.59 Evaluation: This evaluation is limited to the impact of the proposed activity on the containment integrity analysis for an MSLB event as described in FSAR Chapter 14. Other considerations were evaluated and reviewed in EVAL 2008-016 and 10 CFR 50.59 screening SCR 2008-190.
The FWHs do not perform a safety-related function. The replacement FWHs perform the same function as the previous FWHs.
The containment integrity analysis was re-performed using revised inputs for FWH volume and final feedwater temperature. The analysis was performed consistent with the existing license basis methodology. The results demonstrated that while the changes result in an increase in peak containment pressure and temperature, both remain within the design limits of the containment building and components important to safety contained therein.
The replacement FWHs have been designed and fabricated in accordance with applicable codes and standards to ensure reliable operation, and do not pose an increased risk of a malfunction that may initiate an accident or transient. (EVAL 2008-020)
, Increase of the Maximum Time Delay for the 41 60 V Bus Undervoltage Reactor Trip Activitv
Description:
This change permitted increasing the maximum time delay for the 41 60 V bus A01/A02 undervoltage (UV) reactor trip from 1.5 seconds to a maximum of 2.5 seconds.
Summary of 10 CFR 50.59 Evaluation: The increase in the delay time is acceptable because the resulting minimum departure from nucleate boiling (DNB) ratio remains above the design basis limit for the fuel cladding, a fission product barrier. The reactor coolant system (RCS) pressure transient remains less than the design basis limit for the RCS fission product barrier.
The supporting analyses were performed consistent with the established current licensing basis methods and assumptions, and are not a departure in the approved method of analysis.
(EVAL 2008-0022)
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, Temporary Water Supply to the Water Treatment Plant Activitv
Description:
This change temporarily supplied water to the water treatment plant from the fire water system. The water treatment plant is normally supplied from the service water (SW) north header. During a portion of the U2R29 refueling outage, the north SW header was removed from service for several days to support maintenance activities. Continued operation of the water treatment plant was desirable to support other planned outage activities and the operation of Unit 1. The water was routed via temporary hoses, fittings and piping, through a duplex basket strainer and a pressure reducing valve, to the normal SW supply header for the water treatment plant.
Summarv of 10 CFR 50.59 Evaluation: The proposed activity affected the total capacity of the installed fire protection system by diverting up to -200 gpm of flow while the temporary modification was installed. The flow diverted was well within the demonstrated capacity of either fire pump. Fire pump capacity tests confirmed that both fire pumps are capable of producing more than 800 gpm in excess of the required flow rate for fire protection requirements and approximately 1500 gpm in excess of the required flow rate for safe shutdown requirements. There were no potential radiological consequences associated with this change, and this change did not involve a change in the method of analysis. (EVAL 2008-004)
Modification, Installation of Flood Barriers and Fire Door Upgrades Activitv
Description:
This modification (EC11416) revised the design basis for fagade flooding, installed 36" barriers, removed grout, and upgraded doors between the El. 8' of the primary auxiliary building and the facades.
Summary of 10 CFR 50.59 Evaluation: This modification provided a safe flood volume of 158,815 gallons in the Unit 1 facade, and 142,168 gallons in the Unit 2 facade. This change included the evaluation of several non-seismic flood sources in the facades which were not previously recognized as postulated flood sources. This resolved a long-standing conflict between the plant configuration, and the stated configuration in the FSAR with respect to the fagade flooding evaluation. This evaluation concluded that the new design features allow sufficient time to secure the source of flooding thus preventing damage to equipment required to achieve safe shutdown. (EVAL 2007-005)
Procedure Change, Circulating Water Pump House Post-Accident Operator Actions Activitv
Description:
This activity revised several emergency and abnormal operating procedures to open the circulating water pump house rollup doors following a loss of ventilation coincident with a design basis accident (DBA) in order to maintain acceptable environmental conditions for the service water pump motors.
Summarv of 10 CFR 50.59 Evaluation: This evaluation concluded that the activity was not a time-critical operator action because of the short time it takes to perform the activity as well as the difference between the time it is expected to be completed and the time to completion. It can be performed with minimal plant staff within the time available. (EVAL 2009-001)
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, Revisions to FSAR A.7, Plant Internal Flooding
- These revisions to FSAR Section A.7 credited the failure of the turbine building rollup door instead of the seismic walls, doors, large floor area and floor drain to maintain the allowable flood level in the auxiliary feedwater (AFW) pump rooms. Other changes were based on crediting the slots with rubber flaps in one of the bottom ventilation louvers for mitigating the consequences of a EDG rooms circulating water (CW) loss of integrity flood, instead of counterweighted ventilation louvers.
Summarv of 10 CFR 50.59 Evaluation: It was concluded that the CW loss of integrity flood will remain below the allowable flood level in the AFW pump rooms. In the EDG rooms the flood water will leave the room at a faster rate than the potential influx from a CW loss of integrity flood. Therefore, the equipment in both of these rooms is not affected. (EVAL 2009-003)
Calculation Revision, Evaluation of the Use of Existing Craneway Bolts Activitv
Description:
This activity evaluated the use of the primary auxiliary building (PAB) crane, with identified structural deficiencies (such as, missing lower connection bolts at certain craneway girders, several missing bearing stiffeners, and several smaller diameter bolts than the original design that are installed in the craneway upper lateral restraint).
Summarv of 10 CFR 50.59 Evaluation: It was concluded that restricted load use of the PAB crane was structurally acceptable. Use of the PAB crane, is acceptable for load lifts up to 15,000 Ibs. between column lines G through R (includes the transfer canal and to the east of the transfer canal) and up to 250,000 Ibs between column lines R through U (includes the spent fuel pool and to the west of the spent fuel pool). (EVAL 2009-005-01)
Procedure Change, Implementation of Procedure AOP-30, Temporary Ventilation for Vital Areas Activitv
Description:
This activity implemented a new procedure to be used in the event of a sustained loss of HVAC cooling to areas containing vital equipment that may be vulnerable to elevated temperatures. The direction included blocking open doorways and setting up temporary portable blowers and spiral ducting between the vital areas and adjacent spaces that have adequate coolinglventilation. This will force circulation between the cooledlventilated spaces and the vital areas to ensure adequate cooling is available until such time as normal ventilation can be restored.
Summarv of 10 CFR 50.59 Evaluation: This evaluation concluded that the equipment is available and staged for use in responding to fires. The procedural direction was derived from existing procedures that direct the same actions for fire events. (EVAL 2008-021)
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ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 10 CFR 50.59
SUMMARY
REPORT FOR 2008 COMMITMENT CHANGE EVALUATIONS B.5.b Command and Control: The original commitment required the Security Shift Supervisor to attempt to contact the control room to validate the loss of command and control. If normal command and control is not available, the Security Shift Supervisor would report to the muster location. Commitment change evaluation (CCE) 2008-004 revised the commitment to allow a Security individual to report to the muster location.
Justification for Chancre: The original muster location was within the protected area and the location had been changed to outside the protected area. This change of muster location does not allow the Security Shift Supervisor to report to the muster location. The commitment was revised to allow a Security individual to report to the muster location. (CCE 2008-004)
B.5.b Command and Control: The original commitment required training of appropriate plant personnel to report to a designated location, which included a backup location, in the event that a catastrophic event occurred that eliminated normal on shift command and control structure.
CCE 2008-005 removed the backup location. The original muster location was changed from two areas inside the Protected Area to a single location outside the Protected Area.
Justification for Chanqe: The Phase 2 and 3 Mitigation Strategy requirements directed a muster location greater than 100 yards from the target area. Using a single muster location simplifies the response and enhances command and control by eliminating the potential of personnel reporting to two locations. (CCE 2008-005)
B.5.b Mitigation Strategies: The original commitment was to ensure that a memorandum of understanding (MOU) included assistance for cooling the spent fuel pool (SFP) as well as reducing radioactive releases by spraying. This commitment was dispositioned as a list item.
CCE 2008-007 removed the commitment to carry this strategy as a listed strategy. Listed strategies are those strategies that were not going to be implemented.
Justification for Chanqe: This listed strategy is being met through the SFP spray - external strategy (Table A.2.2-3) commitments and capability. The mutual aid agreements between the local fire department and other area fire departments are sufficient to bring the required apparatus to PBNP to implement the SFP spray strategy. Equipment compatibility and a demonstration of the ability of one local fire department to work with PBNP to deploy a spray strategy were demonstrated at PBNP on August 18, 2008. (CCE 2008-007)
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- The original commitment stated that a flow rate estimated at more than 200 gprn will be provided from Lake Michigan (unlimited capacity) via the fire water system or directly from the forebay. CCE 2008-008 revised the original commitment of more than 200 gpm, to an initial flow rate of 200 gprn decreasing to a lower flow rated in accordance with the guidance of WCAP-16800-NP, "Insights for Operating Steam Generators to Minimize RCS Inventory Loss Following a Loss of All AC and DC Power (PA-OSC-0356), Rev 0, November 2007."
Justification for Chanqe: With the development of WCAP-16800-NP (Revision 0, dated November 2007), additional guidance was developed to aid pressurized water reactor (PWR) licensees in a loss of all AC and DC scenario as it relates to use of the steam generators (SGs) as heat sinks. Providing a straight 200 gprn response can result in an overcooling situation that could lead to nitrogen injection into the reactor coolant system (RCS), inhibiting or stopping natural circulation flow in the RCS and worsening RCS control. WCAP-16800-NP provides a methodology for control of heat removal to avoid an overcooling situation. (CCE 2008-008)
- The original commitment states that in the event an internal and external water makeup is insufficient or unavailable, PBNP would deploy an external SFP spray capability. The strategy uses a portable, diesel-powered water pump and hoses to provide flow to a spray nozzle capable of delivering at least 400 gprn of spray over the spent fuel. The external pump will be deployed in the vicinity of an available suction source at a location to be determined based on plant conditions. Hose connections and sufficient lengths of fire hose will be available to spray from approachable sides of the buildings. The discharge of the pump will be connected to an offsite fire apparatus or mobile crane. CCE 2008-009 revised the commitment of one spray nozzle to two spray nozzles.
Justification of Chanqe: To meet the strategy that required delivery of greater than 400 gprn of water to the SFP in such a manner that all the fuel assemblies are covered by the water stream, it was identified that two nozzles would be needed to cover the rectangular area of the SFP.
(CCE 2008-009)
B.5.b Mitigation Strategies: The original commitment states that the unlimited water supply is Lake Michigan via the fire water system or directly from the forebay. CCE 2008-010 revised the commitment by adding additional capabilities for accessing Lake Michigan as a water source.
Justification of Chanqe: As the B.5.b project progressed, additional capabilities to access Lake Michigan as a source of water were developed. (CCE 2008-010)
B.5.b Mitiqation Strategies: The original commitment for makeup to condensate storage tank (CST) equipment locations states that the storage location of the hose to CST connectors would be stored locally at the CST. The storage location for the portable pump hydrant connector would be El. 26' of the PAB. CCE 2008-01 1 revised the location of the hose to the CST connectors and the hydrant connector to greater than 100 yards from the target area.
Justification of Chanqe: Per NEI 06-12, the two strategies (makeup to CST and containment flooding with portable pump) require the adapters or connection devices to be stored greater than 100 yards from the target area. The decision was made to store the CST connectors in a job box that is located greater than 100 yards from the target area. The decision was also made to store the hydrant connector (containment spray adapters) in the same job box. This job box Page 2 of 3
is kept on the B.5.b truck that is stored greater than 100 yards from the target area.
(CCE 200-011)
- The original commitment states that PBNP will use a portable diesel-driven water pump and hoses to supply at least 300 gpm flow to containment. The suction source will be Lake Michigan, or refueling water storage tank (RWST), if available, similar to the SFP external strategy. The pump discharge will be connected to the containment spray headers via existing vents and drains or by the use of a special connecting flange which would be installed in place of the containment spray header check valve bonnets.
CCE 2008-012 removed the RWST as a suction source.
Justification of Chanqe: The original commitment could imply that the RWST would be a possible suction source for the portable diesel driven water pump and hoses to supply at least 300 gpm flow to containment. However, the RWST will not be used as suction. If available, the RWST will be used as the normal source of water to spray into containment with the containment spray pumps. (CCE 2008-012)
B.5.b Mitiuation Strateuies: The original commitment provides makeup to the SFP using reactor makeup water to the SFP vial hose connections. This may provide 100 gpm to the SFP and was dispositioned as implement. CCE 2008-013 removed the commitment to carry this strategy as an implement strategy.
Justification of Chanqe: The NRC safety evaluation states that an implement strategy provides additional tangible makeup capability beyond that already available (e.g., a substantial fraction of 500 gpm). Thus, while potentially feasible, a strategy that provides a low makeup rate would not be considered viable. This particular implementation strategy provides makeup of only an estimated 100 gpm. Therefore, it is considered not viable. (CCE 2008-013)
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