NL-06-0312, Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries

Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries

ML060530028
Person / Time
Site:	Hatch
Issue date:	02/20/2006
From:	Sumner H Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-06-0312
Download: ML060530028 (16)

Text

H. 1.Sumner, Jr. Southern Nuclear Vice President Operating Company, Inc.

Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 February 20, 2006 COMPANY Energy to Serve Your Worldw Docket Nos.: 50-32 1 NL-06-03 12 50-366 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Revort of Facility Changes, Tests. and Exveriments Safety Evaluation Summaries Ladies and Gentlemen:

Enclosed is the 24 month report of facility changes, tests, and experiments safety evaluation summaries in accordance with the requirements of 10 CFR 50,59(d)(2).

This letter contains no NRC commitments. If you have any questions, please advise.

H. L. Surnner, Jr.

Enclosure:

Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries cc: Southern Nuclear Operating Comvanv Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, General Manager - Plant Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch

Enclosure Edwin I. Hatch Nuclear Plant NRC Docket Nos.: 50-321 and 50-366 Operating Licenses: DPR-57 and NPF-5 Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries

Enclosure Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS ABN as-built notice AC alternating current ADS automatic depressurization system AHU air handling unit ALARA as low as reasonably achievable APLHGR average power linear heat generation rate APRM average power range monitor ARI alternate rod insertion ARM area radiation monitor ARTS average power range monitor, rod block monitor, and Technical Specifications ASME American Society of Mechanical Engineers ATWS anticipated transient without scram ATWS-RPT anticipated transient without scram-recirculation pump trip BHD bottom head drain BOP balance of plant BWR boiling water reactor BWROG Boiling Water Reactor Owners Group CFR Code of Federal Regulations COLR Core Operating Limits Report CRD control rod drive CS core spray CST condensate storage tank DAS data acquisition system DBA design basis accident DBE design basis earthquake DC direct current DCB double cantilever beam DCR design change request DCS dry cask storage DHR decay heat removal dP differential pressure ECCS emergency core cooling system ECP electrochemical potential EDG emergency diesel generator EFCV excess flow check valve EFPD effective full power days EFPH effective full power hours

Enclosure Page 2 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Chan~es,Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS EHC electrohydraulic control ELI Equipment Location Index EM1 electromagnetic interference EOC-RPT end of cycle-recirculation pump trip EOF Emergency Operations Facility EPA Environmental Protection Agency ERFDS Emergency Response Facility Display System ETS Environmental Technical Specifications EQ Environmental Qualification FHA Fire Hazards Analysis FPC fuel pool cooling FSAR Final Safety Analysis Report GE General Electric GL Generic Letter GPC Georgia Power Company HCU hydraulic control unit HNP Hatch Nuclear Plant HPCI high pressure coolant injection HVAC heating, ventilation, and air-conditioning HWC hydrogen water chemistry I&C instnunentation and control IE inspection and enforcement IGSCC intergranular stress corrosion cracking ILRT integrated leak rate test IRM intermediate range monitor ISFSI independent spent fuel storage installation IS1 inservice inspection IST inservice testing LAN local area network LC0 limiting condition for operation LDS leak detection system LDCR license document change request LLRT local leak rate test LLS low-low set LOCA loss of coolant accident LOSP loss of offsite power

Enclosure Page 3 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS LPAP low power alarm point LPCI low pressure coolant injection LPM loose-parts monitor LPRM local power range monitor LPSP low power setpoint MCC motor control center MCPR minimum critical power ratio MCR main control room MCRECS main control room environmental control system MDC minor design change MG motor generator MPC Multi-Purpose Canister MOV motor-operated valve MPL master parts list MSIV main steam isolation valve MS SRV main steam safety relief valve MSL main steam line MSLRM main steam line radiation monitor MSR moisture separator reheater NMA noble metals addition NPSH net positive suction head NRC Nuclear Regulatory Commission NSSS nuclear steam supply system ODCM Offsite Dose Calculation Manual OPDRV operations with the potential to drain the reactor vessel OPRM oscillation power range monitor PAM post accident monitoring PASS post accident sampling system PCIS primary containment isolation system PCIV primary containment isolation valve P&ID piping and instrumentation diagram PLC programmable logic controller PPC plant process computer PRB Plant Review Board PRNM power range neutron monitor PSW plant service water PSW plant service water

Enclosure Page 4 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS QA quality assurance RBM rod block monitor RCIC reactor core isolation cooling RCPB reactor coolant pressure boundary RCS reactor coolant system REA Request for Engineering Assistance RES Request for Engineering Services RFI radio frequency interference RFP reactor feed pump RFPT reactor feed pump turbine RG Regulatory Guide RHR residual heat removal RHRSW residual heat removal service water RMCS reactor manual control system RPS reactor protection system RPT recirculation pump trip RPV reactor pressure vessel RRS reactor recirculation system RSCS rod sequence control system RWCU or reactor water cleanup RWC RWCS reactor water cleanup system RWE rod withdrawal error RWM rod worth minimizer SAER Safety Audit and Engineering Review SAT station auxiliary transformer SBGT or standby gas treatment SGTS or SGT SCM stress corrosion monitor SDC setpoint design change SED System Evaluation Document SJAE steam jet air ejector SLMCPR safety limit minimum critical power ratio SNC Southern Nuclear Operating Company SoRA Summary of Required Actions SPDS Safety Parameter Display System SRB Safety Review Board SR Surveillance Requirement SRM source range monitor SRV safety relief valve

Enclosure Page 5 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS SSAR safe shutdown analysis report SSC system, structure, or component TBWD thrust bearing wear detector TCV turbine control valve THV torus hardened vent TIL Technical Information Letter TIP traversing incore probe TLD thennoluminescent dosimeter TM Temporary Modification TRM Technical Requirements Manual TS Technical Specifications TSV turbine stop valve Version

Enclosure Page 6 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries 10 CFR 50.59 SUMMARIES TEMPORARY MODIFICATIONS TTM)

APC 1-04-033 This Temporary Modification via control of an Annunciator and Plant Component sheet (APC) is to disable the "ROD DRIFT" alarm function for rod 26-39 and rod 34-39 via installation of a jumper on the Probe Buffer Card. The FSAR mentions that a drifting rod is "indicated by an alarm and a red light in the MCR. The rod drift condition is also monitored by the process computer." As a result of this APC, the alarm will not come in for the three rods identified. This TM will allow the drift alarm to function for all other control rods for which the alarm is not disabled. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.

APC 1-04-089 This Temporary Modification via control of an Annunciator and Plant Component sheet (APC) is to disable the "ROD DRIFT" alarm function for rod 26-39, rod 34-27 and rod 34-39 via installation of a jumper on the Probe Buffer Card. The FSAR mentions that a drifting rod is "indicated by an alarm and a red light in the MCR. The rod drift condition is also monitored by the process computer." As a result of this APC, the alarm will not come in for the three rods identified. This TM will allow the drift alarm to function for all other control rods for which the alarm is not disabled. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.

TM 1-04-021, Rev. 0 This activity is for binding closed or "gagging" 1E1 1-F200B, which is the minimum flow valve for RHRSW pump 1E 11-COO 1B. The FSAR states that the design function of this valve is a low-flow bypass. Since the valve will be bound closed, no low-flow bypass will be available to the affected pump. This constitutes a change to the FSAR.

The valve addressed by this TM cannot form any part of a sequence of events which could cause an accident. Hence, no change to this valve could affect the probability of occurrence of an accident.

Enclosure Page 7 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries DESIGN CHANGE REOUESTS (DCR)

DCR 90-028. Rev. 0 This DCR provides the design for replacing nine Allis Chalmers safety related starters. Siemens starter pan assemblies will be used as replacements. The replacements are functionally equivalent and testing will be performed to assure the replacements are seismically and environmentally qualified for worst case conditions. The margin of safety of the Motor Control Centers will not be affected by the starter assembly change out.

DCR 98-047, Rev. 0 This DCP will replace RMS-9 trip devices on selected (based on load category) frames of 600V Busses 2A, 2B, 2AA and 2BB with MVT+ trip units. All the busses and their respective loads are nonsafety-related. The function of the 600V switchgear will not change. The breaker trip devices are being replaced to enhance 600V distribution system reliability by reducing spurious breaker trips.

The analysis of trip unit failure presented in the Discussion section of the safety evaluation addresses the potential impact of EMIIRFI conducted and radiated emissions from the new trip units on any safety-related equipment or system. The conclusion is that no safety-related equipment or system would be adversely affected by the installation of these trip units. Therefore, the probability of an accident previously evaluated in the FSAR will not increase.

DCR 99-050, Rev. 0 This DCR provides the design for the replacement of the Safety Parameter Display System (SPDS) and the Emergency Response Facility Display System (ERFDS). The design functions of the existing SPDS will be retained in the new system, although they will be accomplished using different software and microprocessor-based hardware. The SPDS has no control functions. It is a monitoring system only. The same information, including calculated values, will be displayed to the plant operators as the existing system. This DCR is considered to be a digital upgrade. EMYRFI testing has demonstrated that the RTP racks are not expected to adversely impact any safety related or important to safety system.

All interconnections between Class 1E equipment and non-class 1E components for the new SPDS will be properly isolated.

DCP 1H03-009, Rev. 3 This design change will provide adequate margin for the Unit 1 reactor building crane main hoist to lift and lower the rated 125 ton loads, improve crane

Enclosure Page 8 of 14 Edwin I. Hatch Nuclear Plant Rwort of Facility Changes, Tests. and Experiments Safety Evaluation Summaries reliability, and bring the crane into compliance with OSHA requirements. The addition of a pulley on the main hoist drum,which is coupled to an auxiliary shaft via a gear belt, introduces a new potential failure. This change may be considered adverse. However, there are no design basis accidents in the FSAR for which the crane is the initiator. The pulley driven overspeed switch provides a backup method for preventing a load drop. The Unit 1 Reactor Building Crane is a single failure proof crane for which a load drop is not considered a credible event. The new switch, belt and pulley assembly is considered to be as rugged as the prior system. Because the potential failures are similar, no new possibility of a malhction of an SSC important to safety with a different result than previously evaluated in the Updated FSAR is expected by this modification.

DCP 1H03-026T, Rev. 0 This DCR adds a blind plate in place of the 1P41-Dl66 orifice plate, located in the 1B diesel generator room, and remove a section of the 12" shield piping in the 2G switchgear room and relocate service water vent and drain valves. This design change will not affect the service water supply to any component except the 1B diesel generator. This change will remove the Division I back-up supply. This does not reduce the reliability of any component. It only reduces the redundancy in available back-up cooling water supplies. PRA evaluation shows that there is no increase in the average risk following implementation of this design change.

MDC 03-5009, Rev. 0 This MDC will permanently remove the automatic PSW transfer and isolation logic for the 1Z41-B008B MCR AC unit. This is a change to the plant as described in the Unit 1 FSAR section 10.7.6. This logic automatically transfers PSW supply fiom Div. I to Div. I1 in the event of a low flow condition in Div. I in conjunction with a loss of offsite power or a loss of coolant accident. Therefore, this logic is designed to function following a LOCA or LOSP and the subsequent occurrence of a problem causing a low flow in Div. I of the PSW system. This change has no impact upon the probability of a problem occurring. The overall system operation will remain as described in the FSAR. Therefore, the probability of a system failure remains unaffected as well.

DCP 1040113801, Ver. 1 This DCP is to implement setpoint and calibration changes to facilitate implementation of the 10 PSI reactor pressure increase to allow the achievement of 100 percent of the rated power approved under Appendix K uprate. Because of a slight increase in rated containment pressure the License Amendments for REA 00-650lRER 2003-254 (RPV 10 PSI Increase, LDCR-2003-077 (Tech Spec Changes) & LDCR 2004-040 (FSAR Revisions) that allows the increase must be approved prior to implementation of this DCP. LDCR 2004-041 which is a result

Enclosure Page 9 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Exveriments Safetv Evaluation Summaries of the change in steam density is included with this DCP. This DCP is to implement the setpoint and calibration changes associated with the 10 PSI increase. There are no impacts to the frequency or occurrence of accidents, likelihood of occurrence of a malfunction of a structure, system, or component (SSC), consequences of accidents previously evaluated, consequences of SSC malfunctions, possibility for the creation of an accident of a different type, possibilities for malfunctions of SSC's, impact on the fuel cladding, reactor coolant pressure boundary, or containment, or changes in the method of evaluations due to the changes associated with this package. Were the 10 PSI increase not to occur and these changes were implemented the result would be a reduction in operating margin rather than safety margin.

DCP 2040 113901, Ver. 1 This DCP is to implement setpoint and calibration changes to facilitate implementation of the 10 PSI reactor pressure increase to allow the achievement of 100 percent of the rated power approved under Appendix K uprate. Because of a slight increase in rated containment pressure the License Amendments for REA 00-650RER 2003-254 (RPV 10 PSI Increase, LDCR-2003-077 (Tech Spec Changes) & LDCR 2004-040 (FSAR Revisions) that allows the increase must be approved prior to implementation of this DCP. LDCR 2004-039 which is a result of the change in steam density is included with this DCP. This DCP is to implement the setpoint and calibration changes associated with the 10 PSI increase. There are no impacts to the frequency or occurrence of accidents, likelihood of occurrence of a malfunction of a structure, system, or component (SSC), consequences of accidents previously evaluated, consequences of SSC malfunctions, possibility for the creation of an accident of a different type, possibilities for malfunctions of SSC's, impact on the fuel cladding, reactor coolant pressure boundary, or containment, or changes in the method of evaluations due to the changes associated with this package. Were the 10 PSI increase not to occur and these changes were implemented the result would be a reduction in operating margin rather than safety margin.

LICENSING DOCUMENT CHANGE REOUESTS (LDCR)

LDCR 2003-076, Rev. 0 This proposed change is to revise U1 and U2 TS Bases to remove the discussion of the automatic swap of PSW cooling water to the "B" Control Room AC unit; add a description to the U 1 FSAR for the manual action, in place of the automatic action, for providing cooling water to the MCR AC unit 1241-BOO8Bfrom the PSW Division 11; and add component "condensing unit," its malfunction and

Enclosure Page 10 of 14 Edwin I. Hatch Nuclear Plant Report of Facilitv Changes, Tests, and Experiments Safety Evaluation Summaries comments to the MCR HVAC Systems Failure Analysis Table as the result of the manual action. This a change to the plant as described in the U1 FSAR section 10.7.6 and U1 TS Bases B 3.7.5. The automatic transfer function is merely a design feature of the B MCR AC train that is included for additional defense-in-depth. Retention of this feature is not necessary for operability of the MCR AC units.

Probability of failure of the PSW function is unaffected. The overall system operation will remain as described in the FSAR. Therefore, the probability of system failure remains unaffected as well.

LDCR 2004-006, Ver. 1 I. The LDCR addresses Unit 1 TRM TLCO 3.3.1 0 in that a one-time extension of the completion time for the LC0 is being proposed from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days.

2. On January 30,2004, the Unit 1 turbine "A" Master Trip Solenoid failed to function during the performance of TSR 3.3.10.1. This failure placed the unit in a required action to isolate the turbine from the steam supply within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The initial apparent cause of the failure is due to sticking of the solenoid.

The safety basis for the proposed one-time change is provided as follows:

With the master trip solenoid inoperable, electrical overspeed protection for the turbine is not available. However, the mechanical overspeed trip is unaffected by this component failure.

The surveillance on the mechanical overspeed trip is up-to-date.

The FSAR (HNP- 1-FSAR-7.11.3) identifies the mechanical overspeed as the protection feature credited with preventing catastrophic overspeed of the turbine.

The backup overspeed trip is credited only when the mechanical overspeed trip is locked out for testing.

Significant margin (-50%) exists between the mechanical overspeed trip setpoint and the speed at which the overstress could possibly lead to failure.

The incremental probability of occurrence of an overspeed event during the seven days allowed by this change is judged to be very small.

The following compensatory measures are recommended for the seven-day period:

In order to decrease even M e r the potential for grid-induced events that might result in a generator load rejection, switchyard work involving breakers should be curtailed. No work that would result in a change in EOOS color is permitted.

Enclosure Page 1 1 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries Operations personnel should brief at the beginning of each shift that the Master Trip Solenoid Valve (MTSV) is unavailable and prompt operator action may be required to cause turbine trip.

GENCOM should be apprised of the turbine status (without electrical overspeed protection) and asked to minimize any action that could lead to a load reject.

LDCR 2004-01 1, Rev. 1

1. Revision 1 addresses Unit 1 TRM TLCO 3.3.10 in that a one-time extension of the completion time for the LC0 is being proposed from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 8 days.

2. On February 06,2004, the Unit 1 turbine " B Master Trip Solenoid failed to h c t i o n during the performance of TSR 3.3.10.1. This failure placed the unit in a required action to isolate the turbine from the steam supply within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The initial apparent cause of the failure is due to sticking of the solenoid.

3. The safety basis for the proposed one-time change is provided as follows:

With the master trip solenoid inoperable, electrical overspeed protection for the turbine is not available. However, the mechanical overspeed trip is unaffected by this component failure.

The surveillance on the mechanical overspeed t i p is up-to-date.

The FSAR (HNP- 1-FS AR-7.1 1.3) identifies the mechanical overspeed as the protection feature credited with preventing catastrophic overspeed of the turbine.

The backup overspeed trip is credited only when the mechanical overspeed t i p is locked out for testing.

Significant margin (approximately 50%) exists between the mechanical overspeed trip setpoint and the speed at which the overstress could possibly lead to failure.

The incremental probability of occurrence of an overspeed event during the 8 days allowed by this change is judged to be very small.

The following compensatory measures are recommended for the 8-day period:

In order to decrease even further the potential for grid-induced events that might result in a generator load rejection, switchyard work involving breakers should be curtailed. That is, no work that would result in a change in EOOS color is permitted.

Operations personnel should brief at the beginning of each shift that the MTSV is unavailable and prompt operator action may be required to cause turbine trip.

Enclosure Page 12 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes. Tests, and Exveriments Safetv Evaluation Summaries GENCOM should be apprised of the turbine status (without electrical overspeed protection) and asked to minimize any action that could lead to a load reject.

LDCR 2004-014, Rev. 0 This proposed change is to revise the Unit 1 TRM TSR 3.9.3.1 to remove the requirement to perform a hoist limit loaded interlock test for the refueling platform fuel grapple and the auxiliary hoist every 7 days after its initial performance. This is a change to the Unit 1 TRM that relaxes the surveillance frequency requirements on the loaded interlock surveillances for the refuel platform fuel grapple and the auxiliary hoist.

The loaded interlocks setpoint surveillance in the TRM insures that the fuel grapple and the auxiliary hoist are capable of detecting when a fuel bundle has been lifted by the refuel platform. That fuel loaded signal is a part of the refueling interlocks. For example, when the fuel grapple is loaded, a rod withdrawal block will engage if the refuel platform is near the core and the mode switch is in the refuel position. This particular TRM surveillance verifies that the setpoint on the fuel grapple loaded signal is adequate. A separate Technical Specifications surveillance (not affected by this TRM revision) will insure that the integrated signals together provide the necessary signal (in the example above, the rod block).

Furthermore, the nature of the refueling interlocks is such that any problems with the interlocks will be evident to the refuel platform operator. As a result, increased frequencies for their surveillances are of questionable value. For example, when the fuel grapple is loaded with a fuel bundle, a "fuel grapple loaded" annunciator is provided to the operator in the platform cabin. It is therefore likely that any failure of this interlock will be obvious to the operator.

This point is noted in the Technical Specifications Bases for the refueling interlock surveillance SR 3.9.1.1 : "The 7 day frequency is based on engineering judgment and is considered adequate in view of other indications of refueling interlocks and their associated input status that are available to unit operations personnel."

No other plant systems are involved in the TRM change.

For the above reasons, the likelihood of occurrence of a previously evaluated system or component malfunction is not increased.

Enclosure Page 13 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes. Tests, and Experiments Safety Evaluation Summaries LDCR 2004-022, Ver. 2 This is a proposed change to the Unit 1 TRM and FSAR to increase the acceptance criterion on the RCIC AC inboard valve from 20 seconds to 25 seconds.

The safety function of this AC valve, in conjunction with the outboard DC valve, 1E5 1-F008, is to automatically isolate on a RCIC steam supply line break in the reactor building. During the 2004 Spring Refueling outage, a gearing change was made to the valve operator and, as a result, the as left close stroke time increased to 19.9 seconds. This meets the acceptance criterion of 20 seconds, but obviously leaves very little margin. It is therefore desired to explore the possibility of increasing the stroke time acceptance criterion.

Rupture of the RCIC steam line is one of the high energy line breaks (HELB) analyzed in the FSAR. Other examples include rupture of a main steam line, the High Pressure Coolant Injection (HPCI) steam supply line the RWCU supply line.

These breaks are non-limiting with respect to fuel limits and vessel inventory because, unlike the breaks in the primary containment, they isolate. The HELB safety analysis calculates the mass and energy escaping from the break into the reactor building from event initiation until the valve is fully closed. This mass and energy release is used to calculate the peak temperatures and pressure differentials within the reactor building and is also the basis of the environmental qualification (EQ) temperature profiles in the reactor building.

The calculation for the RCIC HELB has been reviewed and it was determined that an increase in the isolation stroke time acceptance criterion from 20 to 25 seconds can be done without affecting the reactor building temperatures, pressures, or the EQ temperature profile in the reactor building.

CAUTION TAGS 1-CA 1E 11-00 144. Rev. 0 1-CA-04-1E 11-00147, Rev. 0 This activity is for binding closed or "gagging" 1E l 1-F200B and 1El 1-F200C, which are the minimum flow valves for RHRSW pump 1El 1-COO 1B and 1E11-COO1C. The purpose of maintaining these valves in the closed position is to ensure that sufficient flow can be developed from the RHRSW at the worst-case conditions of river level. The FSAR states that the design h c t i o n of these valves is a low-flow bypass. Since the valves will be bound closed, no low-flow bypass will be available to the affected pump. This constitutes a change to the FSAR.

Enclosure Page 14 of 14 Edwin I. Hatch Nuclear Plant Rmort of Facility Changes. Tests, and Experiments Safety Evaluation Summaries The valves addressed by this activity cannot form any part of a sequence of events which could cause an accident. Hence, no change to these valves could affect the probability of occurrence of an accident.

2-CA-04-2E11-00059, Rev. 0 This activity is for binding closed or "gagging" 2E 11-F200B and 2E11-F200C, which are the minimum flow valves for RHRSW pump 2E 11-COO1B and 2E 11-COO1C. The purpose of maintaining these valves in the closed position is to ensure that sufficient flow can be developed from the RHRSW at the worst-case conditions of river level. The FSAR states that the design h c t i o n of these valves is a low-flow bypass. Since the valves will be bound closed, no low-flow bypass will be available to the affected pump. This constitutes a change to the FSAR.

The valves addressed by this activity cannot form any part of a sequence of events which could cause an accident. Hence, no change to these valves could affect the probability of occurrence of an accident.