Text

1994 Annual Operating Period
	ML20079A342
Person / Time
Site:	Hatch
Issue date:	12/31/1994
From:	Beckham J; GEORGIA POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	HL-4787, NUDOCS 9502280404
	Download: ML20079A342 (80)
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4-Georgia Power Comp.ny L;

. 40 lavimess Centar Parkway !

Post Ofhcs Box 1295 '

Birmingham. Alabama 35201 ;

Telephone 205 877 7279

> J. T. Beckham. Jr. Georgia Power '

Vice President - Nuclear Hatch Project !!* 80Uttern f]*ctnc system ,

February 24, 1995 i Docket Nos. 50-321 HL-4787 I 50-366 .

I U.S. Nuclear Regulatory Comnussion q

' ATTN: Document Control Desk :

Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant l Annual Operating Report for 1994 t

Gentlemen-Enclosed is the 1994 Annual Operating Report for Edwin I. Hatch Nuclear Plant Unit 1, l Docket No. 50-321, and Unit 2, Docket No. 50-366. This report is submitted in !

accordance with the requirements of 10 CFR 50.59(b)(2) and Technical Specifications ;

6.9.1.4 and 6.9.1.5. -l Sincerely, j

(

[ J. T. Beckham, Jr.

OCV/ld

Enclosure:

1994 Annual Operating Report for Plant Hatch Units 1 and 2 I

cc: Georgia Power Comjsuty Mr. H. L. Sumner, Nuclear Plant - General Manager !

NORMS ;

U.S. Nuclear Regulatory Commission. Washington. D.C. E Mr. K. Jabbour, Licensing Project Manager - Hatch !

i U.S. Nuclear Regulatory Commission. Washington. D.C. i Mr. S. D. Ebneter, Regional Administrator i Mr. 8. L. Holbrook, Senior Resident Inspector - Hatch 9502290404 941231 :

PDR ADOCK 05000321 :

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ENCLOSURE EDWIN I. IIATCH NUCLEAR PLANT- UNITS 1 AND 2 NRC Docket Nos. 50-321 and 50-366 Operating Licenses DPR-57 and NPF-5 t

ANNUAL OPERATING REPORT !

1994 ;

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TABLE OF CONTENTS fait n

Glossasy...................................................................................................................... ii Introduction........................................................................................................ I g Safety Relief Valve Challenges for 1994..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Design Changes and Tests or Experiments .. ....... ... . ...... . ... . . ... ............ ..... .. ..... . 3 Unit 1/ Common Design Changes (Safety Related) .... ........ . . ....... ............ .............. ... 5 c Unit 2 Design Changes (Safety Related) .. ..... ........ .. ..... ... . .. ..... . ..... . . . . . . 27 Unit 1/ Common Design Changes (Nonsafety Related) ... ... . .. .......... . .... .. ...... . .. . 45 Unit 2 Design Changes (Nonsafety Related) .. .... ....... .. ............ .. .. .. ......................... 50 Unit 1/ Common 10 CFR 50.59 Safety Evaluations Written as a Result of Activities (LARs, Procedures, etc.) AfTecting FSAR/ FHA in 1994.. ...................... .. ... 53 Unit 210 CFR 50.59 Safety Evaluations Written as a Result of Activities (LARs, Procedures, etc.) AfTecting FSAR/ FHA in 1994........ ... . ... . .... ... . ... . ....... 63 Unit 1 Test or Experiment Requests .......... ... . . . ............. ..... .......... ............ . . . . . 68 Unit 2 Test or Experiment Requests .. ...... .... ...... ... ..... . ... .. .... .... .. .... ......... .... ... 69 Data Tabulations and Unique Reporting Requirements........ .. . .............. . ................ ...... 70 Occupational Personnel Radiation Exposure for 1994....... . .... .. .. ...... . .. . .. ... . ...... 71-Regulatory Guide 1,16 Information End of Year Report - 1994...... . . . ........ ..... ..... ..- 72 React or Coolant Chemi st ry. . .. . . . . .. . . .. . . . . . . .. . . .. . . . . . . .. . . . . . .. . . . . ... . . . . . .. . . . . . . . . . . . . .. . .. . . . . . . . . . . '73

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i GLOSSARY ;

ACRONYMS AND ABBREVIATIONS ,

l ABN as-built notice '

A/C air conditioning AC alternating current ADS automatic depressurization system '

AHU air handling unit -l A/E- architect / engineer ALARA as low as reasonably achievable ANSI American National Standards Institute AOV air-operated valve APRM average power range monitor ARTS APRM Rod Block Technical Specifications ASCO Automatic Switch Company .

ASME American Society ofMechanical Engineers j ATTS analog transmitter trip system ATWS anticipated transient without scram BOP balance of plant BWR . boiling water reactor BWROG Boiling Water Reactor Owners Group CAD containment atmosphere dilution !

CAV crack arrest verification CFR Code of Federal Regulations COLR Core Operating Limits Report CPIS containment purge and inerting system -

CRD control rod drive CRDA control rod drive accident CS core spray CST condensate storage tank l

~DBA design basis accident I

DBE design basis earthquake DAAS data acquisition and analysis system DC direct current DCR design change request dP differential pressure

~ECCS emergency core cooling system ECP electrochemical potential EHC electrohydraulic control EQ environmental qualification 1

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GLOSSARY ACRONYMS AND ABBREVIATIONS i i

FHA fire hazards analysis l FPC fuel pool cooling and cleanup :

FSAR final safety analysis report GDC general design criterion GE General Electric ,

GL Generic Letter ;

GPC Georgia Power Company

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HELB high energy line break HNP Hatch Nuclear Plant HPCI high pressure coolant injection HVAC heating, ventilation, and air-conditioning HWC hydrogen water chemistry .

I&C instrumentation and control IE inspection and enforcement !

IGSCC intergranular stress corrosion cracking l ILRT integrated leak rate test ;

IRM intermediate range monitor ;

ISI inservice inspection l IST inservice testing i LAR- licensing action request LCO ~ limiting condition for operation LDS leak detection system LED light emitting diode LER Licensee Event Report l LLRT localleak rate test LLS low-low set LOCA loss of coolant accident !

LOSP' loss of offsite power .l LPCI low pressure coolant injection ;

LPRM local power range monitor ;

i MCC motor control center MCPR minimum critical power ratio MCR main control room MCRECS main control room environmental control system MG motor generator .

MOV motor-operated valve i Page iii

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-l GLOSSARY !

i 1 ACRONYMS AND ABBREVIATIONS MPL master parts list - . _

MSIV main steam isolation valve- 'l'

- MSL. main steam line i

MSR moisture separator reheater NFPA National Fire Protection Association NPSH net positive suction head .

NRC Nuclear Regulatory Commis:, ion j

-NSSS nuclear steam supply system j PASS post accident sampling system !

PCIS primary containment isolation syrdem ;

PCIV primary containment is-lation valve PCRS piocess computer replacement system P&ID piping and instrumentation diagram ;

PRB Plant Review Board PSW plant service water !

QA quality assurance l RBCCW reactor building closed cooling water RBM rod block monitor RCIC ' reactor core isolation cooling RCPB reactor coolant pressure boundary .

RFPT reactor feed pump turbine -

RHR residual heat removal :i RHRSW residual heat removal service watu l RPIS rod position indicating system j RPS- reactor protection system . ,

RPV reactor pressure vessel 1 RRS reactor recirculation system !

RTD resistance temperature detectors l RWCU reactor water cleanup l RWCS reactor water cleanup system RWE rod withdrawal error !

4 M- rod worth minimizer ,

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SAT station auxiliary transformer l SBGT standby gas treatment l SBLC standby liquid control l SDV scram discharge volume ;

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GLOSSARY I i

ACRONYMS AND ABBREVIATIONS i

SFP spent fuel pool SFPCCS' spent fuel pool cooling and cleanup system SJAE steamjet air ejector >

SLCS standby liquid control system i SNC Southern Nuclear Operating Company SPDS safety parameter display system SRM source range monitor SRV safety relief valve TBCCWS turbine building closed cooling water system TCV turbine control valve ;

THV torus hardened vent

TIP traversing incore probe TOL thermal overload TSV turbine stop valve l

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INTRODUCTION The Edwin I Hatch Nuclear Plant is a two-unit facility located approximately 11 miles north of Baxley, Georgia on U.S. Highway 1. The plant consists of two light water reactors each licensed to operate at a power level of 2436 MWt. The maximum dependable capacities for 1994 were 741 net MWe for Unit I and 765 net MWe for Unit 2. General Electric Company furnished the boiling water reactor, the nuclear steam supply system, the turbine, and the generator for both units. The plant was designed by Southern Company Services, Inc., with assistance provided by Bechtel Power Corporation. The condenser cooling method employs induced-draft cooling towers and recirculating water systems with normal makeup supplies drawn from the Altamaha River.

The plant is a co-owned facility with ownership delegated as fnllows:

Georgia Power Company 50.1 %

Oglethorpe Electric Membership Cooperation 30.0%

Municipal Electrical Authority of Georgia 17.7 %

City of Dalton, Georgia 2.2%

Licensing information for the units is as follows:

Unit 1 ILnitj Docket Number 50-321 50-366 License Issued 08/06/74 (DPR-57) 06/13/78 (NPF-5)

Initial Criticality 09/12/74 07/04/78 Initial Synchronization 11/11/74 09/22/78 Commercial Operation 12/31/75 09/05/79 Georgia Power Company has sole responsibility for overall planning, design, constmetion, operation, maintenance, and disposal of the Hatch Nuclear Plant.

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i SAFETY RELIEF VALVE CIIALLENGES FOR 1994 -

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6 No SRV challenges occurred this year. ,

Unit 2 t

D_ alt Valves 08/30/94 2B21-F013B, C, D, F, G, H, K, and L ,

A Reactor Protection System (RPS) actuation caused by a momentary loss of power due to overtravel of the RPS power supply transfer switch during a ,

transfer from the alternate to the normal power source resulted in an :

automatic shutdown of the reactor. All systems functioned as designed. ,

Eight Safety Relief Valves (SRVs) opened to reduce pressure. The SRVs -

subsequently closed at the appropriate pressures as the pressure decreased.

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DESIGN CIIANGES AND TESTS OR EXPERIMENTS :

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t SAFETY EVALUATIONS FOR ALL SAFETY-RELATED DESIGN CIIANGES, NONSAFETY-RELATED DESIGN CIIANGES, AS-BUILT NOTICES, AND OTHER ACTIVITIES RESULTING IN FSAR/ FHA UPDATES IN 1994 t

Pursuant to 10 CFR 50.59, the following is a brief description and summary of the safety l evaluation for each change made to safety-related systems and components, and each test or experiment performed during 1994. The safety evaluation summaries address the three criteria used to determine whether a proposed change, test, or experiment involves an u reviewed safety question, i.e.: ,

1. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR may be increased.

2. If the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR may be created. l l

3. If the margin of safety as defined in the bases of the Technical Specifications is reduced.

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1994 ANNUAL OPERATING REPORT EDWIN L IIATCII NUCLEAR PLANT UNIT I/ COMMON DESIGN C1IANGES (SAFETY RELATED) 13A-013 Update the Units 1 and 2 FSARs to document the essential MOV requirements for TOL relay alarms, trips, and calibrations.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. MOV TOL alarms provide information only and do not have an impact on valve operation, response, or failure modes. These alarms do not significantly contribute to the detection of marginal or failed essential MOVs. The present surveillance, setup, and testing practices are sufficient to detect marginal or failed conditions. Bypassing the TOL relay trip protection increases the probability of a MOV completing its designated safety function. TOL relay setpoint calibration is not a factor in valve operation because the TOL trip function is bypassed during power operation.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The MOV failure modes and accident mechanisms remain the same as those evaluated in the FSAR, with or without MOV TOL alarms or calibration checks. Removal of MOV TOL trips reduces the failure modes and consequent accident possibilities. Bypassing the TOL relay trip function increases the probability of a MOV completing its designated safety function.

3. The margin of safety as dermed in the bases of the Technical Specifications is not reduced.

The change does not increase any acceptance limits or decrease any failure points. The ;

Technical Specifications do not specifically address the requirements for MOV TOL alarms, l trips, or calibration. i l

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Revise the tables in the Units I and 2 FS ARs which identify the load distributions on the i emergency buses to add < symbols in front of each value listed for emergency 600-V loads, as well as the total kilowatts. Add a footnote stating the values shown are acceptable analyzed !

values supported by calcu'.ations and the present actual loads are equal to or less than these values.

This reduces the need for updating the tables if the new loads are within the values shown in the !

tables.

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1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased The operation of the diesel generators is not a precursor to any evaluated accident. The diesel generators are not i used for normal operation. The response, operation, and reliability 1of the diesel generators are not affected by this change because both the actual loads and the loads shown in the tables do i not exceed acceptable analyzed limits. No failure modes are impacted. j l

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1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED)

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2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. Since the loads on the diesel generators are still maintained within acceptable analyzed limits, there is no effect on operation or reliability such that any new accident mechanisms or failure modes are created.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The loads on the diesel generators are still maintained within acceptable analyzed limits as defined by the Technical Specifications. No failure point is affected because operation and reliability are not impacted.

89-0145. Rev. O Add sets of flanges to provide spool pieces that can be removed, allowing easy access to the RHRSW pumps for maintenance. Replace flex conduit.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification does not change the operation or function of the RHRSW system. The msterials and design of the flanges meet the original design requirements. Replacement of the flex conduit does not affect the operation of the pump motors.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The function and operation of the RHRSW system are unaffected.

Flanges currently exist in the RHRSW system; therefore, failure of flanges has been previously evaluated. No new modes of failure are created. ,

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The addition of flanges and replacement of flex conduit do not affect any parameters discussed in the Technical Specifications, nor do they add any that should be included. No setpoints are changed, and no limits are exceeded.

91-0039. Rev. O I

Modify HPCI steam exhaust drain pot level switch stilling well to allow the addition of two new level switches. Modify the logic of the exhaust drain bypass valve. Inhibit annunciation for both the supply and exhaust drain pot high water level alarms during HPCI operation. Replace steam traps with orifice plates. This change ensures the HPCI turbine steam exhaust drain pot is drained prior to HPCI turbine operation and prevents spurious level alarm annunciation.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This change meets or exceeds applicable design, material, and construction standards. The function and operation of the HPCI system are not affected. No new system interfaces are created. No new operational Page 6 l

7 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED) parameters that would increase the probability of water hammer, vibration, fatigue, corrosion, or thermal cycling are introduced. This modification does not change, degrade, or prevent the accident response or safety function of any system or component. The drain pot piping alterations do not change any fission product barriers or any assumptions made in evaluating the radiological consequences of an accident.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new failure modes are introduced. Bypasses or alternate flow paths are still available. Inhibiting the drain pot high water level alarms upon HPCI initiation ,

has no effe et on HPCI system availability.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ;

This design change does not affect the LCOs or surveillance requirements for the HPCI system or any other system important to safety. No acceptance limits are increased, and no failure points are decreased.91-066. Rev. O Provide adequate support for the conduit above the RCIC pump and turbine skid in the southwest diagonal of the reactor building, el. 87 feet, per seismic criteria, by adding new suppons and modifying existing ones to meet maximum span requirements.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The support modifications satisfy current design criteria for conduit span lengths. The addition and modification of conduit supports will not increase the consequences of any accident analysis.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new types of accidents or failure modes are introduced by the addition of the supports and the modification of existing supports. !

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3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No system design failure points or system limits are decreased, nor are any acceptance limits !

increased by this change to conduit supports. j i

92-0053. Rev. O j 1

Remove the trip functions of the MSL radiation monitors for MSIVs and MSL drain valves l closure, reactor scram, and MCR pressurization to minimize inadvertent reactor scrams and safety feature actuations. Other functions of the MSL radiation monitors, including isolation of the sampling valves, remain operational. This Technical Specifications change was reviewed and l approved by the NRC.

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i UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED)

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No equipment associated with the control rods or their drives are affected. These logic changes do not affect i the operation of any equipment having the potential to cause a CRDA. With the trips removed, adequate controls are in place so that high radiation in the MSLs can be monitored and controlled with annunciator response procedures.

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2. The possibility of an accident or malfunction of a different type than any evaluated previously I in the FSAR is not created. All existing design, construction, and inspection requirements are j met by this change. No new modes of failure or accident mechanisms are introduced. This logic change does not affect the operation of other equipment or systems necessary for the prevention or mitigation of an accident. ,

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3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

For a CRDA with the elimination ofMSIV isolation, Plant Hatch remains well within 10 CFR 100 limits. Annunciator response procedures and abnormal operating procedures provide -

guidance to ensure increases in MSL radiation levels are promptly addressed. Removing the MCRECS pressurization mode is acceptable, because sufficient redundancy exists with ESFs j and procedural controls.

92-0056. Rev. O Redesign the existing thermocouples to measure temperatures within the pipeline thereby improving the monitoring of recirculation water temperatures. This modification provides a faster i response time to temperature changes, allowing the operator more time to establish necessary !

heatup rates Use the two thermowells presently containing the RTD assemblies for the new design. Remove the RTD assemblies, and replace with combination RTD and thermocouple !

i assemblies.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The installation of the new assemblies in the existing thermowells does not impact the system pressure boundary.

System function is unchanged. The new equipment performs no safety-related function nor does it interface with any safety-related equipment. The new instrumentation is seismically acceptable.

l l 2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. This change enhances operator interface with :he reactor recirculation system by providing faster response time to temperature change, more accurate temperature monitoring, and a wider range of temperature indication. No new modes of failure are introduced.

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X 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED)

3. The margin of safety as dermed in the bases of the Technical Specifications is not reduced.

This modification does not change any operating parameters or system response described in the Technical Specifications. No acceptance limits are increased and no failure points are decreased.

92-0089. Rev. O Replace the RCIC flow controller with a Yokogawa microprocessor-based programmable controller and replace five GE flow indicators with Dixson indicators. Install a power line filter and a surge absorber on the power circuit for the controller.

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1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Since the new controller is more reliable and the total number of components is reduced, RCIC availability is enhanced. Replacement of the indicators has no impact on accident or transient scenarios.

The controller is qualified for electromagnetic interference and seismic response in this environment. The amount of controller-operator interaction is not increased. Response times ;

and other system performance characteristics are not adversely affected.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new types of system level failure modes are created by this modification. Because RCIC is a single loop system, common-cause failure is not a concern. ,

RCIC system function is unaffected, and reliability is enhanced. The operating system software has been validated and verified consistent with industry standards to assure no ;

malfunction of a different type is introduced. ,

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No safety limits, setpoints, plant parameters, or failure points are affected. This modification causes no adverse impact on instrument channel response time, trip accuracy, indicated ar. curacy, or any plant transient response.

92-0120. Rev. O Alleviate potential noncondensable gas buildup in the RPV level instrument cold reference legs by providing a continuous injection of CRD fluid drive water at a rate greater than the condensing rate and less than the rate determined that will induce an unacceptable amount of error in the instrumentation utiliz.ing the reference leg.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Components are not added which would degrade the integrity of the pressure boundary or controls which are in place to protect the pressure boundary. The only postulated failure is a line break or high flow rate, which might impact water level indication. However, the system is designed to protect Page 9

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r UNIT I/ COMMON DESIGN CHANGES (SAFETY RELATED) against these failure modes. The induced error to the reference leg instrumentation is insignificant when compared with the instrument loop uncertainties. This design meets !

existing seismic design criteria. ,

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2. ' The possibility of an accident or malfunction of a different type than any evaluated previously f in the FSAR is not created. Due to the extremely low flow rate and isolation capability, no L new accident scenarios or failure modes are created. The design is valid for protection against nozzle failure due to thermal stresses, line break, high flow rates, inadvertent excess flow !

check valve closure, and check valve leakage.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The effects on channel uncenainties potentially affected by this change are insignificant ,

compared to instrument loop uncertainties. This change does not affect any acceptance limits i or failure points. ,

92-0125. Rev. O Replace the operators on RHR MOVs lEl 1-F016A-B to reduce the final thrust and remain within :

the Technical Specifications defined stroke time of 11 seconds.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This design does not i change the operating parameters of the MOVs, but rather replaces the existing operators with ;

ones that perform better within established guidelines. The RHR system contribution to previous accident analyses is not altered. All aspects of this design change meet or exceed existing system design criteria.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The replacement operators perform the same function as the existing operators. No unanalyzed accidents or new failure modes are introduced. The requirements of all necessary codes and standards are met.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. I The Technical Specifications criteria for the valves and associated system remain unchanged, and the valves perform within these guidelines.

92-0138 Rev. O Remove existing RHRSW air release valves and cap off piping at those locations. Install four new l air release valves with isolation and surge check valves. Locate one at the discharge of each of the !

RHRSW pumps upstream of the discharge check valve. This will prevent the air release valves from being flooded by the upstream system static head, allowing them to release the air from the pump columns as the pump columns fill with water.

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1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. The new valves and ,

piping are seismically qualified and designed in accordance with ASME Code,Section III, ['

Class 3. The modification is designed to eliminate damage to the system by water hammer.

Overall system operation and performance are not adversely affected. Failure results of the .

replacement valves are identical to those of the existing valves. {

2. The possibility of an accident or malfunction of a different type than any evaluated previously ,

in the FSAR is not created. No accidents of a different type or new failure modes are j introduced. This design change meets the requirements of the necessary codes to preclude the i possibility of adversely affecting any other safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ,

The operating characteristics and accident response of the RHRSW system are unaffected by this design change. No acceptance limits are increased, and no failure points are decreased. (

l 92-0149 Rev. O Lower the reactor trip setpoint value for the low reactor water level 3 scram and isolation i functions per approved Technical Specifications Amendment 173. I 1

1. The probability of occurrence or the consequences of an accident or malfunction of equipment )

important to safety previously evaluated in the FSAR is not increased. Circumstances j initiating the loss of feedwater transient and LOCA events remain unaffected. The new ]

setpoint provides several additional seconds for operator actions during a loss of feedwater i transient, and may avert an unnecessary reactor scram. No change to system or component maintenance or testing is involved. All equipment and instruments remain the same. j l

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The safety-related systems and components whose operation may be initiated by a low water level 3 signal still operate in the same manner as before, and no new j failure modes are created.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Technical Specifications Amendment 173 addresses the impact of reduction from the previous level 3 setpoint and confirms an adequate margin to the safety limits.

92-0158. Rev. O Installjumpers across the trip output contacts of reactor building radiation monitors IDll-K609A-D, and refueling floor radiation monitors IDIl-K611 A-D to provide electrical isolation of the trip circuits during testing.

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i 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CIIANGES (SAFETY RELATED)

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification does not affect the original design or operation modes of the subject monitors. The change provides test points for the monitors. No trip logic is altered, and no accident analysis is affected by this change.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The modification does not affect the function of the monitors as previously designed.

3. The margin of safety as dermed in the bases of the Technical Specifications is not reduced.

The wiring of the monitors is changed to meet a commitment in LER 92-17 and to comply with the existing Technical Specifications during testing.92-038, Rev. 2 Add mass to diesel fuel oil transfer pumps columns and motors to shift the natural fiequencies out of resonance with the operating frequencies to reduce vibrations to within acceptable limits. Add pressure gauges to assist in systua operation surveillance. Replace discharge swing check valves ,

with lift check valves to reduce leakage.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment l important to safety previously evaluated in the FSAR is not increased. These modifications !

wae designed in accordance with Seismic Category I criteria and thus, will not adversely 1 affect the operation of any equipment. The reliability of the pumps and motors is improved. l The consequences of safety-related equipment failures would be the same as before the equipment alterations. ;

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2. The possibility of an accident or malfunction of a different type than any evaluated previously ;

in the FSAR is not created. The changes meet all existing design, construction, and inspection l requirements, and no system operation or function will be altered. No new modes of failure or i accident mechanisms are introduced by this change. 1

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No acceptance limits or failure points are affected. Overall system design and operation will not be changed.

93-0008 Rev. O Add 3.5-second time delays in the operation of undervoltage, underfrequency, and overvoltage trip circuits in safety-related panels and overvoltage trip circuits in nonsafety-related panels of the RPS.

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1994 ANNUAL OPERATING REPORT EDWIN 1. HATCH NUCLEAR PLANT j L UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED) p

1. The probability of occurrence or the consequences of an accident or malfunction of equipment j important to safety previous); evaluated in the FSAR is not increased. The time delay reduces the propensity for spurious tripping introduced by momentary changes in voltage or frequency !

that are not of sufficient duration to cause component damage. This modification does not change frequency or voltage protection values. ]

2. The possibility of an accident or malfunction of a different type than any evaluated previously :

in the FSAR is not created. This modification does not change the function of any RPS component. The introduction of time delays in the trip circuits does not cause any component damage. No new modes of operation for any system are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This modification does not change frequency or voltage protection values. This delay does :

not cause damage to any component associated with the RPS. This modification is considered l an improvement in safety since the introduction of a time delay reduces the challenge to safety systems by eliminating unnecessary trip actions. ;

93-0n14. Rev. 0 l l

Add a sensing line from the torus to the commercial grade analyzer panel so both the torus and the drywell atmospheres may be sampled for oxygen content by the commercial grade analyzer, as required by Technical Specifications.

1 The probability of occurrence or the consequences of an accident or malfunction of equipment !

important to safety previously evaluated in the FSAR is not increased. Oxygen monitoring in '

the torus and drywell, or failure to do so, is not a precursor to any previously analyzed accident. The new installation meets the design requirements of the existing H202 sample line j such that probability of physical failure is not increased. The function of the post accident H20 2system remains unchanged, and use of the commercial grade analyzer for normal sampling increases the availability and reliability of the H 202 system. The H202 panels are not affected when the commercial grade system is performing a toms sample.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. Minimizing the use of the H202 sampling system during normal i i

plant operation increases its reliability and availability, while the function remains the same.

The H202 system is adequately isolated from the nonsafety-related O2 analyzer by a normally l closed manual valve. No new failure modes or accident scenarios are created. j

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. l' The accuracy of the nonsafety-related analyzer is allowed for in measuring the oxygen level so .

that the Technical Specifications requirement is still met. The Technical Specifications do not :

require any change.

I Page 13 ;

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l 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CIIANGES (SAFETY RELATED) 93-0017. Rev. O Replace the existing nonsafety-related displacer-type circulating water fiume level switch with a nonsafety-related level transmitter and associated level indicating switch. The new level transmitter is less susceptible to fiume turbulence. Locate the level indicating switch on safety-related MCR panel 11111-P650 and provide alarm contacts to operate the existing low level alarm and a new high level alarm to allow operators to monitor fiume level from the MCR during chlorination.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This change has no !

impact on the reliability or on any operating or response parameters of the circulating water system. Modifications to the MCR panel have no adverse impact on other components within the panel. The new level indication instruments do not interface with or perform any safety- '

related function.

2. The possibility of an accident or malfunction of a different type than any evaluated previously -

in the FSAR is not created. These modifications enhance operator interface with the circulating water system by providing real-time indication of flume level and annunciation of high flume level to allow better level control. No new modes of failure are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ;

The circulating water system is not addressed in the Technical Specifications. These ,

modifications do not change operating parameters or system response.

93-0028. Rev.1 Reroute cable trays TFQS-01 and TFR3-01, which supply power to safety-related equipment, outside the cable spreading room to remove the requirement for fire barriers in the cable spreading room.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. All cables and t

raceways are rerouted and installed per plant requirements for separation and seismic qualifications for Class lE equipment. This modification decreases the probability of cable !

failure due to fire exposure in the cable spreading room and prevents the malfunction of !

equipment resulting from cable damage. This change has no effect on any failure mode which !

could affect the release ofradioactive material.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. Rerouting the power cables outside the cable spreading room does not affect any parameter used to determine any equipment failure mode or accident possibility.

Page 14

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l1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT. -l f

s UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED)

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This modification does not affect the safety limits of any equipment and ensures the operation i of equipment required to achieve safe shutdown.

93-0029. Rev. O j Replace LPCI discharge check valves 1El1-F050A-B to improve LLRT performance, reduce ;

repair costs and man-rem exposure, and eliminate impact on the outage critical path. ;

1. The probability of occurrence or the consequences of an accident or malfunction ofequipment ;

important to safety previously evaluated in the FSAR is not increased The new installation .

meets all applicable codes. The capability of the RHR system to provide emergency cooling to the core in a LOCA is not affected. The increase in pressure drop as a result of replacing the :

existing check valves is insignificant. The function, operation, and reliability of the j replacement valves are equivalent to the original valves. The consequences of any additional unidentified drywell leakage are not increased over those described in the FSAR.

' 2. The possibility of an accident or malfunction of a different type than any evaluated previously l in the FSAR is not created. No changes in the original design criteria for the RHR system ;

result from this modification. The replacement valves meet all applicable constmetion codes. i No other equipment important to safety is impacted.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ;

The original design bases for the RHR system and the LPCI mode of operation, and limits on ,

unidentified leakage are not impacted. Operation and availability are not affected. LCOs and !

i surveillance requirements remain the same. No safety limits or failure points are changed.

i 93-0030. Rev. O ;

E Replace the local starters for HPCI MOVs IE41-F006, IE41-F007, and 1E41-F008 due to obsolescence. Since the new starters are approximately 11 inches larger, some conduit and circuitry work is required. .

1. The probability of occurrence or the coasequences of an accident or malfunction of equipment j t

important to safety presiously evaluated in the FSAR is not increased. The replacement of the local starters has no effect on the operation of any equipment or system assumed to function in j response to an accident and which could affe.ct the release of radioactive efIluents. The new starters are qualified to 10 CFR 50.49 requirements and will improve the reliability of the I

MOVs. The operation and potential failure mechanisms of the HPCI valves are not adversely affected.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The replacement of the local starters with starters qualified for Page 15

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o 1994 ANNUAL OPERATING REPORT- EDWIN I. IIATCII NUCLEAR PLANT .

t UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED) i safety-related use does not introduce any new malfunction initiators or failure modes. The ;

operation and reliability of the HPCI valves is no dversely impacted.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The replacement starters are qualified for use in safety-related applications. No failure points or safety limits are adversely affected.

93-0032, Rev. O Add new electrical containment penetration assemblies in spare nozzles. Each penetration contains two SRM and four IRM cables. Each inboard and outboard junction box has a vertical metal barrier to enhance channel separation. Install new nitrogen monitors.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This change to the SRM and IRM circuits neither p events the reactor from being scrammed on appropriate RPS inputs nor prevents control rod ehJrawal block. The new penetration assemblies meet or exceed all applicable codes and standards. Loss of a penetration is highly unlikely due to the .

construction of the penetrations and the design measures used to maintain separation. The l assemblies are tested ar.d approved for the same pressure and temperature as the drywell, and !

are capable of withstanding the thermal and mechanical stresses encountered during all modes !

of operation and DBAs.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The penetrations are emironmentally qualified for all modes of ,

operation and DBAs and they are designed to withstand a total integrated radiation dose of normal plus accident accumulations. This change does not introduce any new failure modes to the RPS or the p-imary containment electrical penetrations. The RPS will continue to be fail-safe.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The assemblies are approved for the same pressure and temperature as the drywell, and are l 1

capable of withstanding the thermal and mechanical stresses encountered during all modes of operation and DBAs. Relocating the detector circuits from the existing penetrations to the new penetration assemblies will not alter their automatic scram response. j 93-0034. Rev. O i l

l Modify the HPCI logic to prevent the speed control circuitry of the ramp generator signal !

converter from energizing until both the HPCI turbine stop valve and steam supply valve have left ;

the fully closed position to avoid severe turbine-pump acceleration transients. Change the logic of i the water to tube oil cooler MOV to prevent it from opening on a high drywc!! pressure or low reactor water level until the HPCI turbine stop valve and steam supply valve have started open.

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i 1994 ANNUAL OPERATING REPORT EDWIN I. HATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED)

This prevents water from the HPCI pump suction from flowing back into the barometric condenser and to the HPCI turbine when HPCI is not running should the barometric condenser !

condensate pump fail. l

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. These changes do not impact any precursors or parameters for any accidents previously analyzed in the FSAR. They >

do not change, degrade or inhibit the accident response of any system or component important l to safety. The modifications meet the requirements of applicable Plant Hatch design criteria. l

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The modifications do not adversely change the HPCI system ;

i operation, but ensure the HPCI system functions to meet its design objectives by preventing :

undesirable operational occurrences. No other systems or components important to safety are affected, and no new modes of failure are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

These design changes do not affect the LCOs or surveillance requirements for the HPCI system or any other system important to safety. No acceptance limits are increased, and no failure points are decreased by the logic changes. >

i i

93-0035. Rev. O Replace the voltage dropping resistors in the HPCI and RCIC turbine control systems with larger !

wattage resistors per the manufacturer's recommendation to improve reliability. Add cutouts in i

the turbine control panels for ventilation purposes.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment j important to safety previously evaluated in the FSAR is not increased. Functional operation of 1 the HPCI and RCIC systems is not affected. All components added are of similar operating characteristics and higher ratings than original components. The temperature rise within the ,

panel is not increased by the resistor replacements. The louvered openings increase air flow through the panels, ensuring a reduction in the internal panel temperature without impacting .

the seismic integrity of the panels. The control panels are located in a mild environment, and EQ concerns are adequately addressed. Applicable design criteria are met.

2. The possibif.ty of an accident or malfunction of a different type than any evaluated previously i in the FSAR is not created. These changes do not alter system operation. The replacement !

components possess similar operating characteristics and meet the same EQ criteria as the !

original components. The seismic qualification of the panels is not adversely affected. j I

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1954 ANNUAL OPERATING REPORT EDWIN 1. IIATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CIIANGES (SAFETY RELATED)

3. The margin of safety as Mmed in the bases of the Technical Specifications is not reduced.

System operation is not altered. The replacement components possess similar operating characteristics to those of the original components.

93-0041. Rev, O Bypass the torque switch in the open and close circuits for five specific butterfly valves in the PSW system to provide maximum assurance the valves will close.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the FSAR is not increased. The modification ensures the design function or the valves is maintained by utilizing the limit switches in the valve circuits. No equipment function is affected by the modification.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created. The systems continue to perform their original design functions under normal and accident conditions while providing an increased reliability of operation. In case of a valve malfunction, the redundant division will be available to perform the valve's safety function.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The modification ensures reliable valve operation. No acceptance limits or failure points are affected.

93-0051. Rev, O Eliminate the humidity controllers in the SGT system to prevent the possibility of failed humidity sensors afTecting heater operation. Interlock the heaters in the SGT filter units with the fan flow switches so the heaters are activated whenever the SGT system fans are operated.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The heaters will continue to perform their intended function, and the removal of the humidity controls for the heaters will not degrade the ability of the carbon filters to perform their safety-related functions. The SGT system is not a precursor to any accident evaluated in the FSAR. All other interlocks and indications will be unaffected by this modification.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The SGT system will continue to operate as required, and the efliciency of the charcoal absorber will continue to meet design requirements. The heaters will be interlocked with the fans, and all existing temperature interlocks will remain to ensure proper heater operation.

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y 1994 ANNUAL OPERATING REPORT EDWIN L IIATCII NUCLEAR PLANT ;

UNIT I/ COMMON DESIGN CIIANGES (SAFETY RELATED) i

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ,

The SGT system trains will still perform their intended function afler modification of the heater controls. This modification will not affect any allowable limit or failure point of any safety-related system or equipment altered by this design change.

93-0058. Rev. O Modify the existing HWC system to allow higher flow rates ofoxygen and hydrogen at levels recommended by GE, thereby protecting reactor internal component protection from IGSCC. t Install an in-core stress corrosion monitoring system with ECP sensors and double cantilever beam

. crack growth sensors by modifying two LPRMs to determine the amount of hydrogen needed for protection inside the reactor core regions.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The deletion ofone l LPRM neutron detector does not challenge the effectiveness of the APRM on accident mitigation, or affect the detection of transients or accidents. The modified LPRM :over tubes are stronger than and perform equally as well as the standard tubes. The injection of hydrogen reduces the potential for corrosion and the probability of eguipment or piping failure within the RCPB. Safety-related equipment outside of the HWC environment is not affected. If a break of any of the sensor cables to the modified flanges is postulated, this event is bounded by the analyzed design basis LOCA. Should the new probes and sensors break, they are sufficiently :

small they would not cause unacceptable coolant flow blockage or malfunction of equipment important to safety, or be of concern during or after any previously evaluated accident. The :

increased levels of N-16 in the main steam lines are negligible when compared with the -

calctiwd doses from the evaluated accidents. ;

2. The possibility of an accident or malfunction of a different type than any evaluated previously ;

in the FSAR is not created. The deleted LPRM neutron detector does not reduce APRM effectiveness. The two modified LPRM cover tubes are stronger than the standard design.

The flanges housing the probes and sensors are designed to the requirements of an RCPB ,

component. The probes and sensors can be postulated to break with the broken parts not mting an accident or malfunction of a different type. Hydrogen injection, while not changing reactor operating conditions, reduces the potential for corrosion. j

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. !

The single deleted LPRM has no impact on the effectiveness of the APRM system. The two modified LPRM assemblies are designed to assure readings during actual reactor operating conditions are not affected. l

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1994 ANNUAL OPERATING REPORT EDWIN 1. IIATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CIIANGES (SAFETY RELATED) ,

T 93-5017. Rev. 0 ,

Replace the drywell air cooler discharge check valves in the PSW system with stainless steel ,

piping. 1

1. The prchability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The drywell air coolers ;

are not required for post-accident operation and have no safety function. Valve position is not i critical for post-accident system operation, and the valves do not have a safety position. The ;

check valves prevent backflow through isolated coolers. Isolating the coolers for maintenance purposes is provided by manual globe valves downstream and air-operated isolation valves j upstream of each cooler. Each outage an LLRT is performed to ensure system integrity. i Replacing the check valves with stainless steel pipe has a negligNe effect on the seismic analysis.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The check valves prevent backflow to the isolated drywell air coolers. Neither the coolers nor the valves have perform a safety function. The portion of ,

PSW in the dryweF receives an LLRT each outage to ensure containment integrity.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. }

This change, which removes check valves having no safety function, has no impact on any safety limits or settings.

93-5030. Rev. O Change the reset value of the diesel generator fuel oil low pressure alarm from 15 to 12 psi (low pressure alarm setpoint is 10 psi) to be within the pressure switch maximum differential of 2.8 psi. t

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Except for this change, diesel fuel oil system operation will not be affected by this change. There is no change to any accident analysis contair,ed in the FS AR. s

2. The possibility of an accident or malfunction of a different type than any evaluated previously ,

in the FSAR is not created. This change does not create any new or unevaluated accident ;

possibilities. Since design and operation are unaffected, no new failure modes or accident mechanisms are introduced. !

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ,

The change of the fuel oil pressure switch reset value will not impact any margins of safety as defined in the bases for the Technical Specifications.

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r 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT ;

UNIT I/ COMMON DESIGN CHANGES (SAFETY RELATED) 93-5040, Rev. O i Replace SLCS Lonergan SRVs with more reliable Anderson Greenwood valves, which are flanged l for maintenance and testing purposes.

1, The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The new Anderson i Greenwood SRVs have the same critical operating parameters as the Lonergan valves. These ,

replacements improve the reliability of the valves. 1

2. The possibility of an accident or malfunction of a different type than any evaluated previously t in the FSAR is not created. This is a like-kind valve replacement, and no different accident or l malfunction types are introduced. ,

i

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ;

This modification does not change any safety settings or limits in the Technical Specifications. t r

93 5054, Rev. 0 Replace the PSW discharge Mission Duo-check valves in diesel generators I A and IC with i stainless steel Enertech KRV nozzle check valves. !

1. The probability of occurrence or the consequences of an accident or malfunction of equipment l important to safety previously evaluated in the FSAR is not increased. No seismic difference exists since the weight reduction is insignificant and has a negligible effect on the stress :

analysis. The difference in valve Cvs marginally affects flow rates, but does not have a [

negative effect on the system. The replacements improve reliability, because the nozzle check ;

valves have fewer moving parts and are of sturdier design. The new valves meet all original l design criteria and are fabricated to ASME Code,Section III, Class 3, requirements. Stainless steel valves are flanged with insulating bolt sleeves and appropriate gasket material to prevent interaction between the stainless and carbon steel interfaces. Stainless steel valves are less likely to erode or corrode in the PSW system. t

2. The possibility of an accident or malfunction of a different type than any evaluated previously [

in the FSAR is not created. The more reliable replacement valves do not create any new !

accident mechanisms or failure modes different from the original valves.

r

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ;

This modification does not change any safety settings or limits.

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1994 ANNUAL OPERATLNG REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY IiELATED) 93-5065. Rev. O i

Replace the Powell lift check valve in the RWCU return to feedwater header line with an Enertech nozzle check valve.

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the FSAR is not increased. The weight increase has a negligible effect on the stress analysis. The nozzle check valve spring is sized for the minimum and normal flow requirements. The nozzle check valve provides the same function as before and does not negatively affect the system. The replacement valve meets all original design criteria, is fabricated to ASME Code,Section III, Class I requirements, and complies with 10 CFR 50, Appendix J, leak rate testing requirements. The valve is moved to a vertical <

position to enhance LLRT performance. The new valve is more reliable, has less moving parts, and is of a sturdier design.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The more reliable replacement valve does not create any new accident mechanisms or failure modes different from the original valve.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This modification does not change any safety settings or limits.

94-0005. Rev. O Due to an increase in valve factors and torque requiremems for the PSW turbine building isolation MOVs, HPCI turbine steam tupply MOV, and HPCI pump discharge MOV, modify these valves j to meet the increased loads. Replace the motors on the PSW MOVs and the HPCI pump i discharge MOV, and modify the actuators on all the valves. Bore a hole in the downstream side of the replacement disc of the HPCI pump discharge MOV to provide pressure relief to prevent potential pressure locking of the valve. With these modifications, the valves meet GL 89-10 requirements.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The modifications to these valves increase their capability and improve system reliability. The components meet all the standards applicable to the affected systems. The functions of these valves are unchanged.

The modifications include a change in the valve stroke times. The increased response time for the HPCI system is still well within the response time stated in the GE SAFER /GESTR repon and does not impact any accident or transient analysis. The HPCI response time for an ATWS condition is longer, but delayed HPCI initiation is I eneficial for an ATWS event. The hole in the replacement disc of the HPCI pump discharge MOV does not reduce the strength of the valve disc or have any other negative impact on valve operation. The valve continues to provide an adequate seal for isolation. These modifications meet the PSW flow requirements Page 22 j

P 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED) for safety-related systems and equipment required for a design basis event. No fission product barrier is adversely affected. Malfunction consequences remain the same as before the modifications. :

t

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. These modifications meet all design and construction requirements for safety-related components. No new accident mechanisms are introduced.

No new failure modes are created by replacement of the motors with higher torque motors, since all potential weak links connected with valve operation are also replaced as required.

3. The margin of safety as defined in the bases of the Technical Spec;fications is not reduced.

This modification meets all design criteria and applicable code requirements. No acceptance limits or failure points are affected.

94-0019. Rev. 0 l In response to NRC GL 94-03, install a pre-emptive repair of the reactor core shroud to permanently replace the 304 stainless steel circumferential welds which are susceptible to IGSCC. .

The repair design consists of a set of four mechanical stabilizer assemblies to structurally replace all horizontal welds H1 through H8, without credit being taken for actual weld conditions. Each ,

stabilizer attaches to the top of the shroud and to a shroud support gusset at the bottom.,

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The stabilizer assemblies do not significantly affect the flow within the downcomer and do not adversely affect the performance of any reactor internal. The water inventory in the downcomer with the stabilizers installed exceeds the volume used in the existing analyses. The leakage flow from the shroud repair holes and the replacement access hole covers is sufficiently small that the steam separation system performance, jet pump performance, fuel thermal margin, and fuel ,

cycle length remain within design limits. The impact on ECCS performance is insignificant.

The stabilizers and affected shroud and RPV components satisfy structural and seismic requirements. The stabilizer components are essentially not subject to fatigue. The stabilizers assure the shroud, even if cracked, performs its safety functions. The stabilizers are designed as safety-related components. ,

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The stabilizers are designed to meet all applicable FSAR criteria.

The stabilizers are fabricated from stress corrosion resistant material and have low applied stresses during normal operation. Welding is not used in their construction or installation. A single failure of a stabilizer is highly unlikely and will not lead to the malfunction of other ,

safety-related equipment.

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m 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 1/ COMMON _ DESIGN CIIANGES (SAFETY RELATED) p-.

3. The margin of safety as dermed in the bases of the Technical Specifications is not reduced.

The stabilizers are designed as safety-related components and assure the shroud, even if cracked, will perform its safety functions. .

94-0039. Rev. 0 -

Modify Udts I and 2 SBGT system logic such that an accident signalin any given unit starts both ;

Units I and 2 SBGT systems. Change Unit 2 to Unit I divisional starts to include Unit 2 LOCA and Unit 2 reactor building high radiation initiation signals. Change the Unit 2 SBGT system damper logic to open both the reactor building and refueling floor dampers on SBGT system i initiation. Alter the containment isolation logic such that SBGT system initiation on either unit isolates both the Unit I and Unit 2 normal IIVAC systems in the reactor building and on the refueling floor. L

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. These modifications meet all requirements for divisional separation and provide greater assurance that the SBGT :

system is available to perform its safety-related function. These changes do not adversely affect any failure mode which could impact the release of radioactive material, or any parameter used to determine the consequences of an accident. ;

2. The possibility of an accident or malfunction of a different type than any evaluated previously f in the FSAR is not created. These modifications do not adversely affect any parameter used to determine the possibility of an accident, since the SBGT system does not initiate any accidents. ;

It only mitigates the consequences of accidents. These changes have no adverse effect on any l equipment failure mode.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No acceptance lilAits or failure points are affected by these modifications.

94-0042. Rev. O Change the settings on IAC overcurrent relays and microversa trip devices in 600-V buses 1C and ID to achieve coordination with the downstream 225 amp molded-case circuit breakers in essential cabinets l A and IB.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment )

important to safety previously evaluated in the FSAR is not increased. This modification I enhances coordination between the 600-V buses and the 120/208-V essential cabinets, thus improving their reliability, operation and response.

2. The possibility of an accident or malfanction of a different type than any evaluated previously in the FSAR is not created. This modification improves the reliability of these systems and Page 24

s 1994 ANNUAL OPERATING REPORT- EDWIN I. IIATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CIIANGES (SAFETY RELATED) ensures all involved systems function as designed. No new failure modes are created by [

enhancing coordination.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Changing the settings of the trip devices in the 600-V buses enhances coordination, thus

- improving their reliability. No acceptance limits or failure points are affected by this ;

modification.

94-5001. Rev. 0 i

Add overlay domes to the reactor building tornado roof vents. Delete the differential pressure electrical switch opening of the reactor building, turbine building, and control building tornado roofvents.

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Existing shear bolts, manual releases, and fusible links will operate the tornado roof vents as originally designed. ;

Continued operation during and after a seismic event is unchanged. No chemical intrusion or .

I detrimental effects will result should any acrylic material be introduced into the fuel pool or reactor vessel. Maintenance and modifications to the tornado roof vents will not adversely affect any accident analyses or impact the operation or reliability of any equipment important !

to safety. No response to any evaluated accident is impaired. The radiological consequences of safety-related equipment failures are not increased or affected since secondary containment will be maintained, except in a tornado. !

2. The possibility of an accuent or malfunction of a different type than any evaluated previously in the FSAR is not created. No new type of accidents will be created since the installation and operation of the overlay domes meet the original design criteria. No chemical intrusion or detrimental effects will result should any acrylic material be introduced into the fuel pool or the reactor vessel. No new modes of failure are introduced. Operation of the tornado roof vents when the pressure difTerential reaches the maximum design allowable will be dependent upon the shear bolts breaking rather than activation of the pressure switch. l

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The secondary containment will be maintained as presently required by the Technical Specifications. Structural integrity will be maintained as originally intended.

94-5031. Rev. 0 l

l I

Remove RHR heat exchanger shell side thermal relief valves lEl1-F3078A-B and install blind flanges.

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1994 ANP"UAL OPERATING REPORT EDWIN L IIATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (SAFETY RELATED)

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. Valves lEl1-F055 A-B serve the same function as lEl1-F3078A-B, respectively, and receive IST. All valves have the same relieving set pressure. Deleting the valves and installing blind flanges do not have a '

negative effect on RHR system performance or any seismic impact. Adequate relief protection is provided by the larger relief valves lEl 1-F055A-B.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The remaining relief valves will continue to provide adequate relief protection. Valve failure modes are the same in their effect as before this modification.

3. The margin of str tye as defined in the bases of the Technical Specifications is not reduced.

This change does not impact any safety settings or limits.

F10A-004 Increase the maximum permissible loading specified in the FHA for the intake structure to prevent the need for administrative approval in increasing the combustible loading.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The increase is insignificant since the fire loading remains well below the standard FHA guidelines for a structure of this type. This change does not alter any current fire watch requirements. The status of fire loading in the intake structure does not affect the operation of any equipment important to safety. The maximum permissible fire loading does not alter any installed fire protection devices.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created. This change does not increase the maximum permissible loading ,

beyond that specified in the FHA for a structure of this type. It does not constitute any activity resulting in an accident or malfunction of a different type.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The Technical Specifications do not address the fire protection issues applicable to this change.

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1994 ANNUAL OPERATING REPORT - EDWIN L HATCH NUCLEAR PLANT !

UNIT 2 DESIGN CIIANGES (SAFETY RELATED) 12D-011 Remove the statement regarding PSW piping heat tracing from Unit 2 FSAR paragraph 9.2.1.4.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment ,

important to safety previously evaluated in the FSAR is not increased. For the pipe sizes of i interest, frozen equipment is not a concern for any credible freeze event. No physical changes are made to the plant. Accordingly, no accident analysis is changed. 1

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created. Freezing of pipe and equipment is not a credible scenario for the pipe sizes ofinterest.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No l changes are made to the physical plant. The Technical Specifications do not address PSW heat tracing.

12D-026 i

Revise Unit 2 FSAR paragraph 7.5.1.4.2 to clarify which controls on the remote shutdcwn panel i are required to comply with GDC 19.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. No instruments are added to, or deleted from, the remote shutdown panel. Neither the operation nor ma'mtenance 1 of any remote shutdown component is changed. The probability and consequences of a control room evacuation event are not increased.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The proposed change does not affect the operation of the plant.

As a result, no unanalyzed modes of operation are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This change is being made in support of the Improved Technical Specifications, which will include those instruments on the remote shutdown panel required to comply with GDC 19. i The FSAR will identify what instruments are required to comply with the GDC.

87-0178. Rev. O Add decontamination ports to the FPC system piping so crud can be removed by spraying high pressure water ram."v as a hose is run in and out of a pipe segment (hydrolasing).

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- 1994 ANNUAL OPERATING ~ REPORT EDWIN L HATCH NUCLEAR PLANT

/> UNIT 2 DESIGN CHANGES (SAFETY RELATED)

1. ' The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The additions will be made to comply with the appropriate codes and standards. Although the FPC system ties into the RHR system, which is safety-related, the change will not affect any system equipment important to safety.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new modes of failure or accident mechanisms are introduced.

The modification will not alter the FPC system operation or function. The change will only reduce personnel exposure.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No acceptance limits or failure points are affected.

88-0106. Rev. O Delete the MSIV leakage control system and utilize existing alternate leakage paths to the condenser to handle post-LOCA radioactive effluents.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment i important to safety previously evaluated in the FSAR is not increased. The leakage control I system is operated after an accident, and use of the alternate existing leakage path is limited to a post-LOCA. ASME Section III, Class I and 2 requirements are followed, and all modifications have been seismically reviewed. An analysis of the alternate paths from the MSL to the main condenser indicated that no adverse effects, such as pipe failure, should be experienced. Radiological consequences will not be significantly higher than previously analyzed and are bounded by the present LOCA analysis.

2. The possibility of an accident or equipment malfunction of a different type than any evaluated previously in the FSAR is not created. Radiological dose contributions due to MSIV leakage are bounded by a LOCA. A LOCA analysis in which the main steam piping and the main j condenser were used as the treatment method indicated that the alternate method meets i regulatory requirements. Failure modes of the caps on the MSLs would be the same as those !

for the MSLs and have been previously evaluated in the FSAR. Failure of the existing l alternate drain path will not adversely affect any equipment important to safety.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. l Deletion of Technical Specification 3/4.6.14 does not involve a significant reduction in the i margin of safety. The function of the leakage control system will be performed by using the alternate paths via the MSL drains and the main condenser. This method is effective for treatment of MSIV leakage over an expanded leakage range. The increased public dose is less L

than 1 percent of the approved offsite dose rate.

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f i i 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLIAR PLANT UNIT 2 DESIGN CHANGES (SAFETY RELATED)88-035. Rev. O Provide a permanent power feed through reactor building penetration number 137 for use during plant outages. ;

1. The probability of occurrence or the consequences of an accident or malfunction of equipment ;

important to safety previously evaluated in the FSAR is not increased. The penetration is sealed to the same standards as it was initially, and the components are seismically mounted.

The integrity of the penetration remains unaffected by the design change.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The penetration is sealed to the same standards as it was initially to maintain integrity of the secondary containment. The components are mounted per the acceptable seismic requirements. ;

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The penetration is adequately sealed and the components are seismically installed. Therefore, no offsite radiation exposure is increased, and no margin of safety as defined in the Technical Specifications is reduced.

89-0146. Rev, o Install an additional set of flanges in the minimum flow line of the RHRSW pump discharge piping to provide a spool piece that can be removed to allow easy access to the pump.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification does not change the operation or function of the RHRSW system, and the materials and design of these additional slanges meet the original design requirements of the RHRSW system.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The function and operation of the system are unaffected, and the design and materials meet the original design requirements of the system. Failure of the new flanges is bounded by the loss of the RHRSW system as previously evaluated.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. I This modification does not affect any parameters discussed in the Technical Specifications, nor does it add any that should be included. No setpoints are changed, and no limits are exceeded.

90-0109. Rev. O Replace existing Class 1E local staners for MOVs 2El1-F008,2E41-F006,2E41-F007, and 2E41-F008 with Class lE qualified starters.

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w 1994 ANNUAL OPERATING REPORT. EDWIN I. IIATCH NUCLEAR PLANT 4 UNIT 2 DESIGN CHANGES (SAFETY RELATED)

(

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The replacement improves reliability. MOV safety-related equipment or equipment important to safety is affected by the modification. The change does not alter the FSAR analysis associated with the e consequences of malfunction of equipment important to safety. l

2. The possibility of an accident or malfunction of a different type than any evaluated previously ,

in the FSAR is not created. The new starters are qualified to the same standards as the replaced starters; therefore, no new malfunction initiators or failure modes are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No acceptance limits are increased, and no failure modes are decreased due to implementation of the modification. .

90-0130. Rev. 0 l Replace the existing RHRSW heat exchanger control valves 2El1-F068A & B with new valves designed to resist cavitation.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Valve response characteristics will remain the same. The new valves will not affect the contribution the RHRSW system has to any accident analysis, and as a result, no accident analysis action' or -

assumption previously evaluated is affected. All aspects of the design change meet or exceed existing system design criteria. Failure of the replacement valves would result in the same effects as failure of the existing valves.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The overall response characteristics of the valves remain unchanged. The failure modes of the replacement valves are the same as those for the existing valves. The design change meets the requirements of required codes and standards to preclude the possibilities ofintroducing any new accidents or adversely affecting any other safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This design change will improve the reliability of the RHRSW system. The Technical Specifications Bases, Section 3/4.7.1, will not be affected by this design change.

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1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 2 DESIGN CIIANGES (SAFETY RELATED) 91-0047. Rev. O Install a drain to the off-gas sample chamber,to assure proper radiation monitoring. Provide a condensate pot with a level switch and two solenoid valves to avoid condensate accumulation within the chamber

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification will provide the capability of draining condensate from the chamber, thus increasing its reliability.

If some of the newly installed components fail to function, the chamber will function the same as before the modification. The chamber is not associated with any accident or equipment >

imponant to safety previously evaluated in the FSAR.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. If the added drain fails to open, the chamber will function the same as before the modification. The worst case scenario would be if the level switch were frozen closed; in which case, only a minute amount of air would be admitted into the main condenser. The chamber is not associated with, nor will its function impact, any equipment important to safety.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The modification increases the availability and accuracy of the monitoring of the gases exiting the main condenser. Although the chamber does no' perform any safety-related function, off-gas pretreatment radiation monitoring is provided to ensure the limits ofTechnical Specifications Section 3.ll.2.l(a) are not exceeded. If an added component fails, the chamber will function the same as before the modification.

91-0144, Rev. 0 Upgrade the existing anchorage for MCC 2R24-S027 located in the diesel generator building by using plates to weld the MCC base to the existing embedded channels and bolting the tops of adjacent MCC cabinets together.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The modification ensures compliance with the requirements of the Seismic Qualification Utility Group Generic Implementation Procedure and the Seismic Margin Assessment methodology. Neither the function of the MCC nor the function of any safety-related equipment is affected by the !

modification. ,

2. The possibility of an accident or malfunction of a different type than any evaluate ( previously in the FSAR is not created. The modification upgrades equipment anchorage. The design change does not affect the function of any equipment.

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I 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT t UNIT 2 DESIGN CIIANGES (SAFETY RELATED)

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced, l since the requirements for seismic design are not defined in the Technical Specifications.

91-0145. Rev. 0 ,

Upgrade the existing anchorage for MCC 2R24-S025 located in the diesel generator building by using plates to weld the MCC base to the existing embedded channels and bolting the tops of adjacent MCC cabinets together.

t

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The modification :

ensures compliance with the requirements of the Seismic Qualification Utility Group Generic i Implementation Procedure and the Seismic Margin Assessment methodology. Neither the !

function of the MCC nor the function of any safety-related equipment is affected by the modification.

2. The possibility of an accident or malfunction of a different type than any evaluated previously I in the FSAR is not created. The modification upgrades equipment anchorage. The design change does not affect the function of any equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced, !

since the requirements for seismic design are not defined in the Technical Specifications.

91-0175. Rev. 0 l Replace existing Target Rock solenoid valves 2B21-F013 A-H and air operators with a new upgraded model which draws three times more current than the older model. Replace 24 Agastat relays which have contacts in the control circuitry of LLS and ADS SRVs with a model containing a magnetic blow-out device to allow proper operation of the solenoids without causing damage or malfunction of the relay contacts. i

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification will ,

not change the existing function or operation of the SRVs, the ADS, or the LLS system. The new solenoid valves and air operators are fully qualified and more reliable than the existing !

valves and operators. The replacement of the relays will ensure relay contacts will not be j overloaded by the increased load of the solenoid. The new relays will provide the same logic ,

as the existing relays with added assurance the SRVs will operate as designed. !

2. The possibility of an accident or malfunction of a different type than any evaluated previously ;

in the FSAR is not created. This modification does not affect the safety-related function or method of operation of the SRVs, ADS or LLS system. Reliability is improved by 1

i Page 32 l l

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L - 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 2 DESIGN CHANGES (SAFETY RELATED) replacement with upgraded models. The new relays will perform the same function as the existing relays. No new modes of failure are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. .

The replacements do not affect SRV setpoints, function, or operation. No acceptance limit is :

increased, and no failure point is decreased. .

92-0077. Rev. 2 i Remove and plug the packing leakofflines and valve bonnet vent lines of the recirculation suction and discharge valves 2B31-F023 A & B and 2B31-F031 A & B. Also, remove and plug the pre. king leakoffline of RHR shutdown cooling isolation valve 2El 1-F009 and RHR LPCI check valve 2El1-F050B. ,

t r

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The operation and failure of these vent and leakofflines are not factors in or precursors to any previously analyzed accidents. The design change meets or exceeds existing system design criteria. Any .

contribution that a failure of the plugged vent or leakofflines would have to any accident is bounded by the existing LOCA and shutdown cooling malfunction analyses. A failure of the plugs would result in the same effects as failure of the existing valves.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The design change does not affect the operation of the subject systems. The failure modes of the systems remain unchanged. i i

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The operating limitations of the Technical Specifications are not affected by the modification.

No acceptance limits are increased, and no failure points are decreased.

92-0086 Rev.1 Add a circuit to provide a redundant power source to increase the reliability of the condensate and feedwater systems. Power the flow controllers and switches, which provide an input to the ,

reactor recirculation system, by the new circuit. Replace the existing flow controllers for the condensate, condensate booster pump, and reactor feed pump minimum flow control valves with new programmable controllers. Replace feed pump flow indicators with new Dixson type .

i BB10lP indicators. Replace the SJAE condensate flow controller with a new controller to provide improved automatic operation.

1. The probabihty of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification will decrease the probability of a transient being initiated due to the condensate and feedwater l

l Page 33 l l

d 1994 ANNUAL OPERATING REPORT EDWIN L HATCII NUCLEAR PLANT !

~ UNIT 2 DESIGN CHANGES (SAFETY RELATED) l minimum flow valve failing open due to a loss of power. The seismic integrity of the MCR panels is not adversely affected. The condensate and feedwater system is not required to mitigate the consequences of any accident evaluated in the FSAR.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The plant responses which will be affected by this change are ;

confined to the condensate and feedwater system. The loss of feedwater flow has been i evaluated in the FSAR. This modification will not alter or potentially affect any safety-related ,

equipment or any equipment required to support the operation of safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No safety limits or setpoints of any safety equipment are affected. No acceptance limits will be ,

increased, and no failure points will be decreased.

92 0137. Rev. 0 Install four new air release valves in each of the four RHRSW pump discharge lines to replace existing RHRSW air release valves 2El1-F0209 A & B which have experienced damage to their l floats and seats due to water hammer during pump start.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The modification installs valves designed to eliminate the potential of damage to the system due to water hammer. Overall system performance is not affected. The new valves are procured as safety- ;

related and are designed to Seismic Category I requirements. A failure of the replacement valves would result in the same effects as failure of the existing valves.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No accidents of a different type are introduced. The modification meets necessary code requirements to preclude the possibility of adversely affecting any other safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The operating requirements and accident response of the RHRSW system are unaffected by .

the design change. No acceptance limits are increased, and no failure points are decreased.

92-0159. Rev. O Installjumpers across the trip output contacts of reactor building radiation monitors 2D11-K609A-D and refueling floor radiation monitors 2D11-K611 A-D to provide electricalisolation of the trip circuits during testing. !

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l 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT l 1

UN.IT 2 DESIGN CIIANGES (SAFETY RELATED) l

- l. The probability of occurrer.ce or the consequences of an accident or malfunction of equipment l important to safety previous:y evaluated in the FSAR is not increased. The modification does !

not affect the original design or operational modes of the subject monitors. The change provides test points for the mon' tors.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. Thr, modification does not affect the function of the monitors as previously designed.

3. The margin of safety as defin:d in the bases of the Technical Specifications is not reduced.

The change modifies the subject monitors' wiring logic to meet Licensee Event Report 92-17 commitment and comply with the existing Technical Specifications during testing.92-01M. Rev. O Relocate and reorient the IIPCI it.jection to feedwater system isolation valve 2E41-F006 to a location further upstream from the feedwater system connection.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previousi) evaluated in the FSAR is not increased. The modification does not change the function or operation of the valve or system. The pipe class and pipe supports are adequate for the modification. The failure mode for the valve at the new location remains unchanged. The change does r.ot atreet any fission product barrier.

2. The possibility of an accident or n'alfunction of a difTerent type than any evaluated previously in the FSAR is not created. The design change does not prevent any safety-related component from performing its designated function. The failure mode for the valve is not affected.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The operation, function, acceptance limits, failure points, and avaihbility of the affected systi:ms remain unchanged following implementation of the modification.92-039 Rev. 2 Add mass to diesel fuel oil transfer pumps columns and motors to shift the natural frequencies out of resonance with the operating frequencies to reduce vibrations to within acceptable limits. Add pressure gauges to assist in system operation surveillance. Replace discharge swing check valves with lift check valves to reduce leakage.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. These modifications were designed in accordance with Seismic Category I criteria and thus, will not adversely affect the operation of any equipment. The reliability of the pumps and motors is improved.

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7 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT ,

UNIT 2 DESIGN CHANGES (SAFETY RELATED)

The consequences of safety-related equipment failures would be the same as before the equipment alterations. ;

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The changes meet all existing design, construction, and inspection requirements, and no system operation or function will be altered. No new modes of failure or accident mechanisms are introduced by this change.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No acceptance limits or failure points are affected. Overall system design and operation will not be changed.

93-0015. Rev. 0 Add a sensing line from the torus to the commercial grade analyzer panel so that both the torus ,

and the drywell atmospheres may be sampled, as required by the Technical Specifications, for oxygen concentration by the commercial grade analyzer, Revise Unit 2 FSAR paragraph 7.6.4.3.2 to state that the commercial grade oxygen analyzer samples both the drywell and torus during normal operation. Also, revise Unit 2 FSAR figures 6.2-44 and 11.4-6 to reflect the commercial grade oxygen analyzer tap into the primary containment atmosphere hydrogen and oxygen monitoring system.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Allowing the

commercial grade analyzer to perform the required containment sampling from both the torus and drywell, or failure to do so, is not a precursor to any previously analyzed accident. The new installation will meet the design requirements of the existing hydrogen and oxygen sample line such that the probability of a physical failure is not increased. The hydrogen and oxygen sampling system is isolated by a normally closed manual valve and its function and operation remain the same. The hydrogen and oxygen panels will not be affected when the valve is open ;

and the commercial grade system is performing a torus sample. I

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The function of the hydrogen and oxygen sampling system remains the same. No new accident scenarios or failure modes are created by providing a cross-tie from the torus hydrogen and oxygen sample line to the nonsafety-related panel. The hydrogen and oxygen sampling system is not adversely affected. Minimizing the use of the post accident hydrogen and oxygen sampling system will increase its reliability and availability. I

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The accuracy of the nonsafety-related analyzer has been .*llowed for in the measurement of the oxygen level and meets the Technical Specifications requn : ment.

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1994 ANNUAL OPERATING REPORT EDWIN I. HATCll NUCLEAR PLANT UNIT 2 DESIGN CHANGES (SAFETY RELATED) 93-0020. Rev. O Replace RCIC system turbine steam supply valve 2E51-F045 Limitorque operator SMB-0-40 with an SB-0-40 operator to reduce the final thmst and ensure the FSAR RCIC system design flowrate response time of 45 seconds.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The replacement .

operator reduces thmst and torque on the valve; therefore, the change enhances system reliability. The overall system response time remains within the FSAR specified limit of 45 seconds. System operation and function remain unchanged. )

l

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new modes of operation for any system are introduced by the ;

modification. The function and operation of the system are not affected.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The modification does not change system operation and design flowrate response time. ,

93-0021. Rev. 0

, Replace the internal gearing parts and motors for RHR outboard injection valves 2El1-F017A &

B to comply with NRC Generic Letter 89-10, Supplement 3, guidelines.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. The modification enables the valve to function within its design requirements. An increase in valve stroke time does not affect the RHR response because the valve remains in its open position the same as before the design change. Valve replacement parts are of the same material as before, and the valve motors are qualified to the requirements oflEEE 323-1974. No other equipment important to safety is affected by the design change. Any contribution to the consequences of failure of safety-related equipment due to malfunction of the subject valves is the same as ,

before implementing the design change.

2. The possibility of an accident or malfunction of a different type than any evaluated previously l in the FS AR is not created. Valve function remains the same, and the availability of both RHR l loops is not affected. Material for the valves and motors remains the same as per the original i design. No new equipment malfunction possibilities are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. l Valve stroke time is not addressed in the Technical Specifications or the FSAR. System response is not affected. No failure points are decreased, and no acceptance limits are increased by the modification. 4 l

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1994 ANNUAL OPERATING REPORT EDWIN L IIATCH NUCLEAR PLANT UNIT 2 DESIGN CHANGES (SAFETY RELATED) 93-0021, Rev. O Replace the internal gearing parts and motors for RHR outboard injection valves 2El1-F017A &

B to comply with NRC Generic Letter 89-10, Supplement 3, guidelines.

L

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The modification enables the valve to function within its design requirements. An increase in valve stroke time does not affect the RHR response because the valve remains in its open position the same as before the design change. Valve replacement parts are of the same material as before, and the valve motors are qualified to the requirements oflEEE 323-1974. No other equipment important to safety is affected by the design change. Any contribution to the consequences of failure of safety-related equipment due to malfunction of the subject valves is the same as before implementing the design change.

2. The possibility of an accident or malfunction of a differem type than any evaluated previously in the FSAR is not created. Valve function remains the same, and the availability of both RHR loops is not affected. Material for'the valves and motors remains the same as per the original design. No new equipment malfunction possibilities are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Valve stroke time is not addressed in the Technical Specifications or the FSAR. System response is not affected. No failure points are decreased, and na acceptance limits are increased by the modification.

93-0022, Rev. O Add a new neutron monitoring system electrical penetration assembly to the nozzle currently designated as MPL No. 2T52-X100C. The penetration will contain two source range monitor cables and four intermediate range monitor cables for trip channels 1 A and IB.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The new penetration assembly meets or exceeds all applicable codes and standards. This change does not prevent the RPS from scramming the reactor or blocking control rod withdrawal on appropriate inputs. Even with a penetration failure, the RPS maintains its ability to insert all control rods.

This change has been reviewed against the applicable separation and single-failure criteria and found to be acceptable. The new penetration assembly will be seismically and environmentally qualified.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. While thil change increases the number of components that may Page 38

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l 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCH NUCLEAR PLANT ;

UNIT 2 DESIGN CHANGES (SAFETY RELATED)

L be affected by a single failure in the penetration, the affected components are part of the fail- ,

L safe RPS. This change does not introduce any failure modes that would prevent the reactor

from scramming on appropriate RPS inputs or prevent control rod withdrawal block.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Source range monitor and intermediate range monitor circuitry, which is part of the RPS, is designed as fail-safe. Relocating the detector circuits from penetrations 2T52-X100A and ,

2T52-X100B to the new penetration assembly will not alter the auto scram response.

93-0036. Rev.l L

Modify the HPCI logic to prevent the speed control circuitry of the ramp generator signal converter from energizing until both the turbine stop valve and the steam supply valve have left the fully closed position to avoid severe turbine and pump acceleration transients. Change the logic of the water to tube oil cooler MOV to prevent it from opening on high drywell pressure or low reactor water level until the HPCI turbine stop valve and steam supply valve have started to open. This will prevent water from the HPCI pump suction from flowing back into the barometric condenser and to the HPCI turbine when HPCI is not running should the barometric condenser condensate pump fail. Modify the water level controls and add new drain logic for the HPCI turbine steam exhaust drain pot to reduce the chance of water hammer in the turbine exhaust piping.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. These changes do not impact any precursors or parameters for any accidents previously analyzed in the FSAR. They only ensure proper operation of the HPCI system as required by the FSAR. These changes do not alter, degrade, or prevent the accident response of any system or component. The '

modifications will meet the requirements of applicable Hatch design criteria.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. These modifications will not adversely change HPCI system operation, but will ensure the HPCI system will function to meet its design objectives by preventing undesirable operational occurrences. New components meet all standards required by Hatch design criteria. No new modes of failure are introduced. No other system or component important to safety is affected.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This design change does not affect the limiting conditions for operation or surveillance !

requirements for the HPCI system or any other system important to safety. No acceptance limits are increased and no failure points are decreased.

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l UNIT 2 DESIGN CHANGES (SAFETY RELATED) ;

93-0037. Rev. O j Replace existing voltage dropping resistors in the HPCI and RCIC turbine control panels with l larger wattage resistors per manufacturer recommendations. Make cutouts in the panels for ventilation purposes. ;

1. The probability of occurrence or the consequences of an accident or malfunction of equipment ,

important to safety previously evaluated in the FSAR is not increased. Functional operation of ;

the HPCI and RCIC systems is not affected, and the ability of the turbine control systems to perform their intended function will not be degraded. All components added by the change have similar operating characteristics and higher ratings than original components. These changes meet the requirements of Plant Hatch design criteria. The louver plates will not impact the seismic integri:y of the panels.

2. The possibility of an accident or malfunction of a different type than any evaluated previously 3 in the FSAR is not created. These changes will not alter system operation. Replacement components possess similar operating characteristics to those of existing components and will ^

meet the same equipment qualification criteria as the original components. The seismic qualification of the panels is not adversely affected. ,

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

These activities do not alter system operation. Replacement components possess similar ;

operating characteristics to those of existing components.

93-0062. Rev. O Replace class IE station service batteries 2A and 2B due to cracking celljars and a concern about j remaining battery capacity. Replace cable, trip devices, breakers and fuses to ensure adequate coordination of protective devices. Revise Unit 2 FSAR paragraph 8.3.2.1.1 which references the station service battery's continuous discharge ratings.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment i imponant to safety previously evaluated in the FSAR is not increased. These batteries are j purchased to specifications that allow no effect on any other system or component beyond what has been previously evaluated. The replacement equipment is functionally equivalent to the existing equipment with equal or enhanced operational capabilities, and the engineered safety feature responses are unchanged.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The replacement equipment is functionally equivalent to the existing equipment, and the function and operation of the systems involved are unchanged.

No new modes of failure will be introduced.

Pag:40 d - - - - - - - _ _ -'.-_- _ .- - ___ F

1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 2 DESIGN CHANGES (SAFETY RELATED) F

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The replacement equipment has equal or enhanced operational capabilities which meet !

Technical Specifications requirements without changing failure points or acceptance limits.

93-0069. Rev. O Modify the reactor water level instrumentation floodup range condensing chamber and piping to be similar to Unit 1. Lower the condensing chamber below the head vent nozzle and slope the condensing chamber inlet piping to about one halfinch per foot up from the condensing chamber t to the inlet nozzle. Condensing will occur prior to the condensing chamber, thus allowing condensation to fill up to the inlet nozr.ie.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Loss of the floodup reference leg would not increase the probability of an accident because this indication is used in a post accident condition to provide the operator with RPV level indication up to the MSLs.

This modification will increase the reliability of the floodup range level indication. The replacement condensing chamber inlet piping will be fabricated according to ASME Section III, Class 1, criteria. The design will be installed to meet the existing design seismic criteria.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. This modification will not affect any engineered safety feature.

The replacement condensing chamber inlet piping will be fabricated according to ASME Section III, Class 1, criteria. The design will be installed to meet existing design seismic criteria. No new failure modes are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Changing the elevation of the floodup reference leg does not impact the margin of safety. This modification will not affect any other instmment or information defined in the Technical Specifications.93-042. Rev. O Bypass the torque switch in the open and close circuits for five specific butterfly valves in the plant service water system. Bypass the torque switch in the open circuits for 16 specific gate valves in the piant service water and drywell hydrogen recombiner systems.

1. The probability of occurrence or the consecuences of an accident or malfunction of equipment important to safety previously evaluated in t he FSAR is not increased. The modification ensures that the design function of the valve;s is maintained by utilizing the limit switches in the valve circuits. No equipment function is affected by the modification.

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1994 ANNUAL OPERATING REPORT EDWIN I, IIATCII NUCLEAR PLANT UNIT 2 DESIGN CIIANGES (SAFETY RELATED)

2. The possibility of an accident or malfunction of a different type than any evaluated previously

. in the FSAR is not created. The systems will continue to perform their original design functions under normal and accident conditions while providing an increased reliability of operation. In case of a malfunction of the valve, the redundant division would be available to perform the safety function of the valve. .

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The modification ensures a reliable operation of the valve. No acceptance limits or failure points are affected. l 93-5015. Rev. 0 Remove the HPCI room cooler discharge check valves from the PSW system. ;

1. The probability of occurrence or the consequences of an accident or malfunction of equipment ,

important to safety previously evaluated in the FSAR is not increased. The valves are located where the water supply line pressure should always be greater than the discharge pressure, and ,

backflow is not possible. Since their safety significance is to open and they are not needed due to system design pressures, these check valves are not required. Isolation of the coolers for ;

maintenance is provided by other existing valves.

2. The possibility of an accident or malfunction of a different type than any evaluated previously '

in the FSAR is not created. Since their safety function is to open and backflow is not possible, these check valves are not required. .

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This change has no impact on any safety limit or setting.

i 93-5023 Rev. O Replace PSW supply header swing check valves 2P41-F0ti4 and 2P41-F065 with flanged nozzle check valves due to wear.

l. The probability of occurrence or the consequences of an accident or malfuncti6h of equipment important to safety previously evaluated in the FSAR is not increased. The weight increase is acceptable in respect to stress considerations. Flow rates are marginally enhanced due to improved flow characteristics of the replacement valves which are fabricated to the same constmetion code and with the same materials as the original valves. The replacement of these )

valves improves reliability. j

)

2. The possibility of an accident or malfunction of a different type than any evaluated previously l in the FSAR is not created. This is a valve replacement, and no different accident or malfunction types are introduced.

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1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT -

UNIT 2 DESIGN CIIANGES (SAFETY RELATED)

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This change does not alter any safety setting or limit.

93-5039 Rev. 0 Remove the lateral brace on main steam piping hanger 2N11-MS-R57. The pipe support is attached to two separate seismic structures. The main frame is attached to the control building and one brace is attached to a turbine building column. The two structures have different seismic displacements and analysis shows the base plates cannot meet the required safety margin. Current ;

analysis indicates the lateral brace is not needed.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This modification will assure the hanger will function as designed. As a result, seismic acceptance criteria will be i met. This change will eliminate any overstress and ensure an unanalyzed condition does not exist.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The remaining pipe support is adequate for the piping seismic load. An unanalyzed condition does not exist. ,

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Seismic criteria are met and overstress is precluded by this change. No unanalyzed condition exists due to this modification.

93 5041 Rev. 0 Replace SBLC system pumps discharge piping and line accumulators Lonergan safety relief valves with functionally equivalent Anderson Greenwood valves. The Lonergan valves are unsuitable due to poor reliability, maintenance history, and availability of replacement parts and valves.

1. The probability of occurrence or Ge consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The new safety relief valves have the same critical operating parameters (relieving pressure and capacity) as the existing valves. This valve replacement improves reliability.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. This is a like-kind valve replacement; thus, no different accident or malfunction typs are introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. l This change does not alter any safety setting or limit. I 1

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1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT ,

II q v

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UNIT 2 DESIGN CIIANGES (SAFETY RELATED)

I 93-5050. Rev. O ;

Remove spare GE thermal overload relays from safety-related motor control centers 2R24-S021 !

and 2R24. " 022 in response to GE Letter PRC-17 concerning a possible problem with these types i ofrelays.

r

1. The probability of occurrence or the consequences of an accident or malfunction of equipment i important to safety previously evaluated in the FSAR is not increased. These spare relays are ,

not connected to load and do not perform any function; therefore, their removal does not affect any system operation. The de power system continues to operate as described in the Unit 2 Technical Specifications. ;

i The possibility of an accident or malfunction s *'a different type than any evaluated previously l in the FSAR is not created. The subject thern. d overload relays are spares. !

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No !

instrument setpoint or setting is changed.

i l

i I

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y 1994 ANNUAL OPERATING REPORT EDWIN I. HATCH NUCLEAR f'LANT UNIT 1/ COMMON DESIGN CHANGES (NONSAFETY-RELATED)81-107. Rev. I Remove spring supports from air handler no. 2 of the temporary cooling system installed in the northeast quadrant Unit I reactor building, el.130 feet. Hard mount the air handler to the existing foundation. Iastall a structural steel frame around air the handler to reinforce the sheet metal housing and support the new duct.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. For the components installed inside the reactor building, Seismic Category I design criteria are used for all piping and duct supports and in mounting the air handler to the floor. Due to the location of the unit with regard to safety-related equipment, during a seismic event, the unit would not adversely affect any safety-related equipment.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The new duct is 10 gage and seismically supported, and for extra conservatism during a seismic event, the framing is added around the air handler. These actions preclude the possibility the equipment could fall and damage any safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. The redundancy and separation of the safety-related equipment ensure the capability of safely shutting down the plant is maintained.89-281 Rev. O Replace the temporary Unit I reactor building chillers and associated equipment with permanent chillers and equipment. Install four 200 ton air cooled chillers outside near the Unit I condensate storage tank. All power will be provided from the supplemental power system. System control will be accomplished using a microprocessor controlled HVAC management system. The system is not safety-related; however, this change will require revisions to Unit 1 FSAR figures 10.9-1 and 10.9-2, and FHA sections 7.0,8.0, and 11.7.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The Unit I reactor building chiller equipment is not required to function after an accident, and does not perform a safety-related function. This design will incorporate applicable codes and standards commensurate with reactor building HVAC system design criteria. All components installed in Category I structures will be analyzed for Seismic II over I. The addition of this load does not negatively impact the r3pplemental power system or any load it presently serves. If the permanent system fails, the impact to safety-related equipment would be as if the existing temporary system failei Page 45

y 1994 ANNUAL OPERATING REPORT EDWIN L HATCII NUCLEAR PLANT i~

UNIT 1/ COMMON DESIGN CHANGES (NONSAFETY-RELATED) ,

E 2 The possibility of an accident or malfunction of a different type than any evaluated previously .

in the FSAR is not created The design of the new equipment meets the existing design l criteria, and all components installed in Category I structures will be analyzed for Seismic II j over I. Should the permanent system fail, the effects to safety-related equipment will not be ;

different than before the modification. This change does not introduce any new type of failure !

modes to any safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This change only improves system reliability, and will not decrease any failure point or increase any acceptance limit of any Technical Specifications required equipment.

90-0053. Rev. 0 Eliminate six snubbers and replace eight snubbers with rigid struts on the main steam piping [

between the main turbine control valves and the high pressure turbine to minimize premature wear !

of support pins and end attachments. Stress analysis eliminates the requirement for the deleted snubbers. :

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. The design change '

does not reduce the reliability of the main steam system or impact the safety-related function of any system. Although the removal of the snubbers does not significantly increase the loads ;

on the lines and supports, the additional loading is determined to be acceptable by calculation. !

2. The possibility of an accident or malfunction of a different type than any evaluated previously i in the FSAR is not created. The design parameters and operation of the main steam system are not altered by this change. The pressure boundary of the main steam piping is not altered, and the integrity of the system is not affected.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Plant operation is not affected by this modification. No limits or setpoints are exceeded. The pressure boundary of the MSLs is not altered by the snubber removal. l 92-0028. Rev. O Modify the startup level control system to a one-valve configuration identical to the Unit 2 system which successfully controls feedwater flow. Remove the low flow control valve and replace the :

high flow control valve internals and actuator to enable operation over the full range of flow rates ,

and differential pressures. -

1. The probability of occurrence or the consequences of an accident or malfunction of equipment t important to safety previously evaluated in the FSAR is not increased. This change impro"es .

Page 46 f

1994 ANNUAL OPERATING REPORT EDWIN L IIATCII NUCLEAR PLANT UNIT 1/ COMMON DESIGN CIIANGES (NONSAFETY-RELATED) the reliability and performance of the startup level control system. System function remains the same. The valves involved are not safety-related.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. This design change is necessary to improve reliability and enhance maintenance on the startup level control system. It does not affect plant stanup or compromise system integrity.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The modification improves reliability and performance by utilizing a single-valve configuration for flow control.93-031. Rev. O Provide the remainder of the equipment, piping, electrical, and civil design necessary to suppon

, the decay heat removal (DHR) system, which may be operated to provide cooling capacity for the SFP of either Unit I or 2 during refueling outages, thus allowing the RHR system and/or the normal fuel pool cooling system to be taken out of service. (A portion of the piping, electrical conduit, wiring, and terminal boxes for the DIIR were previously installed by DCR 93-001.)

1. The probability of occurrence or the consequences of an accident or malfunction of equipment impon =nt to safety previously evaluated in the FSAR is not increased. There is no accident analysis in the Unit I or 2 FSAR concerning the storage of spent fuel or a full core ofiload in the SFP. The analyses of SFP cooling in the Units 1 and 2 FSARs fully encompass any effect that the proposed activity could have. Complete loss of the DHR system is no more likely than the concurrent failure of both Units 1 and 2 SFP cooling systems.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The proposed activity can affect only the storage of spent fuel or the temporary storage of a full core ofiload in the SFP. The analyses of SFP cooling in the Units I and 2 FSARs fully encompass any effect the proposed activiti could have.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The Units 1 and 2 Technical Specifications prohibit loads in excess of 1600 pounds from travel over the SFP, and only the suction and discharge pipe spools, which weigh less than that, will be transponed over the spent fuel assemblies. The Units 1 and 2 Technical Specifications set

< the minimum limit for SFP water level in relation to the spent fuel rods, and the design of the suction and discharge spools prevents siphoning below that limit.

94-5014. Rev. 0 l

Add a tool case frame ulull in the dead-end hallway in the diesel generator building to provide a secure area to store toof and equipment.

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6 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON DESIGN CHANGES (NONSAFETY-RELATED) -

1. The probability of occurrence or the consequences of an accident or malfunction of equipment j i

important to safety previously evaluated in the FSAR is not increased. Due to its location in a dead-end hallway far removed from any safety-related equipment, this wall has no impact on any previously evaluated accident. A seismic II/I evaluation ensures the stmetural integrity of the building is unaffected for a seismic event. ;

2. The possibility of an accident or malfunction of a different type than any evaluated previously :

in the FSAR is not created. The installation of this wall in no way affects the operation of any safety-related equipment.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The installation of this wall has no affect on the margin of safety. It provides a secure area for gang boxes already located in this area.

94-5043. Rev. O Cut and separate MSIV LLRT lines that stem from valves IB21-F026A-D so multiple tests can be performed concurrently.

1. The probability ofoccurrence or the consequences of an accident or malfunction of equipment '

important to My previously evaluated in the FSAR is not increased. The lines are modified in a fashion to preserve seismic integrity. The lines, which are not safety-related, have no ;

active safety function. This modification has no effect on the operation or response of any system.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The lines are modified in a fashion to preserve seismic integrity.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The Technical Specifications do not address LLRT test connections or their use. ,

F10A-003 l i

Revise the fire door tables in FHA Appendix B to properly align the columns to identify the fire . j area in which the detection system associated with each door is located.

l. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FS AR is not increased. The change does not :

alter the status of any fire door as it relates to detection systems and fire areas. The status of a i fire door regarding surveillance criteria does not affect on the operation of any equipment :

important to safety.

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1 1994 ANNUAL OPERATING REPORT- . EDWIN I. IIATCH NUCLEAR PLANT l 1

1 UNIT 1/ COMMON DESIGN CHANGES (NONSAFETY-RELATED) l

. 2. The possibility of an accident or malfunction of a different type than any evaluated previously l in the FSAR is not created. This change serves to clarify the fire door association with the fire areas and suppression systems it separates. This does not constitute any activity resulting in an ,

accident or malfunction of a different type.

-]

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ;

The Technical Specifications do not address fire doors or their impact on any margin of safety.

F10A-007 ;

Revise FHA Appendix B to clarify the compensatory action statement relative to fire . hose stations and ensure the corrective action improves overall plant safety. i

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Fire hose stations are passive equipment requiring manual action to operate. This editorial change clarifies the ;

compensatory action for inoperable fire hose stations, and does not affect operating equipment.

2. The possibility of an accident or malfunction of a difTerent type than any evaluated previously in the FSAR is not created. This editorial change clarifies the compensatory action for passive and inoperable manual fire hose stations.

3. The margin of safety as defmed in the bases of the Technical Specifications is not reduced. ;

The Technical Specifications do not address fire hose stations. ,

I i

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1994 ANNUAL OPERATING REPORT EDWIN L IIATCII NUCLEAR PLANT UNIT 2 DESIGN CIIANGES (NONSAFETY-RELATED) 84-0275. Rev.1 Reroute the sourc of purge water for the MSR hot well level transmitter reference leg to the condensate booster pump header in order to maintain purge flow during all phases ofMSR operation.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This change improves the capability of the level controls to maintain normal MSR hotwell level. The improved level controls reduce the likelihood of an abnormal MSR hotwell level, and the consequences of a turbine trip transient resulting from a high MSR hotwell level are the same as before this modification.

2. The oossibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. Malfunction of the MSR hotwell level reference leg purge can result in loss of control of MSR hotwell level, but no different accident could result from this.

Since the reference leg purge can influence only the MSR hotwell level controls, it can have no impact on the possibility of a different malfunction of any equipment or system important to safety.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No operating parameters or system safety limits are changed. !90-093. Rev, 0 l Replace drywell equipment drain leak rate timers with timers having a wider span than the existing timers. Modify the setpoints of both timers to reduce the occurrence of nuisance annunciations caused by improper timer setpoints. These nonsafety-related alarms monitor leakage with respect to Technical Specifications limits.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment i important to safety previously evaluated in the FSAR is not increased. This modification has l no effect on the accident analysis of the liquid radwaste tanks. Radwaste system responses are unaffected. This change will not affect the consequences of the failure of a radwaste storage tank.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The radwaste system will continue to perform its original design function after the modification. No new types of 31ure modes are introduced.

3. The margin of safety as defined in the bases of the fechnical Specifications is not reduced.

The operability of essential equipment as defined by the Technical Specifications is not affected. No acceptance limits are increased, and no failure points are decreased.

i Page 50 I

D, 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 2 DESIGN CHANGES (NONSAFETY-RELATED) 91-0112. Rev, 0 - >

Modify the RP!S probe buffer card logic to allow the closure of the "00" switch of a control rod to produce a more reliable full-in indication which is used by the refueling interlocks and the .

operator to confirm all rods are full-in, panicularly following a scram. Card replacement may be done when the system is not required and whether RPIS cabinet power is maintained or removed.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment :

important to safety previously evaluated in the FSAR is not increased. The RPIS is not a safety-related system, and its production of full-in signals was not considered in the accident evaluations. The full-in indications are Type "B" indications used to provide indication of safety system operation. This change makes the operation of full-in signals more reliable and reduces the probability of malfunction. 1

2. The possibility of an accident or malfimetion of a different type than any evaluated previously in the FSAR is not created. Full-in indication occurs under the same rod positions as before the modification. Only the RPIS operation is affected. The effects of RPIS failures were evaluated, and no new effects are created by this .-hange.

1

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. !

Full-in indication is still received for the same control rod positions as before the change.

92-0130. Rev. O !

l Add manual isolation valves and test connections upstream of RBCCW containment isolation valve 2P42-F051 and downstream of RBCCW containment isolation valve 2P42-F052. The manual valves will isolate the RBCCW ponion supplying the primary containmeni allowing for normal operation and avoiding RHR alignment to the FPC during an LLRT.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment ;

important to safety previously evaluated in the FSAR is not increased. This modification does j not change the results of any accident previously evaluated in the FSAR. Since the modificatior k located outside the primary containment isolation boundary, it will not affect l l

any safety-reimed function. These additions will not change any function of the RBCCW system.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created. The additions will not affect the operability or integrity of the l- RBCCW containment isolation valves or piping. In case of an accident, the nonsafety-related ponion of the RBCCW system may be closed manually to isolate the nonsafety-related pan of the RBCCW system. The addition of the manual isolation valves and test connections, all passive components, will not introduce any new failure mechanism.

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1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 2 DESIGN CHANGES (NONSAFETY-RELATED)

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The RBCCW system does not perform any safety-related function other than primary containment isolation. Its operation is not addressed in the Technical Specifications. These additions will not affect the requirements for containment isolation as addressed in the Technical Specifications.

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'1994 ANNUAL OPERATING REPORT EDWIN 1. IIATCII NUCLEAR PLANT UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS l r .

l 12B-024 1

Reload the reactor core for Cycles 14 and 15 with the following changes

a. reconstitution of two Siemens Nuclear Power (SNP) Lead Fuel Assemblies (LFAs)(cycle 14 l only). ,

b. Zircaloy-2 Thermal Size Annealed fuel channels.

c. water rod replacement in one H1-R11 fuel bundle.

d. core loading pattern for Cycle 14 including 172 fresh bundles.

e. core loading pattern for Cycle 15 including 164 fresh bundles and 4 Lead Use Assemblies.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment l important to safety previously evaluated in the FSAR is not increased based on the following: l l

a. The reconstituted bundle performance is bounded by the GE bundle for steady-state and anticipated occurrences. The loss ofcoolant and fuel handling accidents are more favorable than the original SNP LFA, and sufficient margin to the 280 cal /gm limit is maintained in the rod drop accident. The SNP LFAs remain seismically qualified, and there are no adverse effects on the performance or reliability of equipment important to safety.

b. There is no consequential difference in the bundle structural integrity during fuel handling or in reactor performance during normal operation or transient and accident conditions.

The channels are seismically qualified and do not create any direct or indirect adverse effects on the performance or reliability of equipment important to safety.

c. New water rod disengagement probability is negligibly small and significantly less than the previous GE fuel product line. The modified bundle is functionally identical to the original one, and its performance is unchanged under normal operation and transient and accident conditions. The bundle remains seismically qualified.

d. Conclusions reached in the FS AR for the initial core remain valid for the final Cycle 14 core loading provided that power distribution limits in the COLR are not violated. The radiological consequences of an accident as evaluated in the FSAR remain bounding relative to plant operation with the Cycle 14 fuel mix and loading pattern. The design basis performance of the safety systems is not degraded.

e. The conclusions reached in the FSAR for the initial core remain valid for the Cycle 15 core loading provided that the power distribution limits in the COLR are not violated. The radiological consequences of an accident or malfunction of equipment important to safety Page 53

+

~1994' ANNUAL OPERATING REPORT EDWIN I. HATCH NUCLEAR PLANT-UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS l y as evaluated in the FSAR remain bounding relative to plant operation with the Cycle 15

. l

~

core loading. There are no direct or indirect effects on safety system performance or increased challenges to safety systems. l

2. The possibility of an accident or malfunction of a different type than any evaluated previously I in the FSAR is not created based on the following: ;

a. Since the reconstitution of two SNP LFAs is performed using replacement rods with .

j identical components as the original rods, the reconstituted bundle is functionally identical i to the original one. The upper end cap of the replacement rods is lengthened to ensure : !

engagement of the rods in the upper tie plate throughout their design life. ;

b. The physical and mechanical behaviors ofZircaloy-2 and Zircaloy-4 are alnmst identical. l There is no consequential difference in the bundle structural integrity during fuel handling !

or in reactor performance during normal operation or transient or accident conditions. _

l

c. The replacement water rod is considered a "like-kind" replacement. The modified bundle is functionally identical to the original design. (

l

d. No changes in plant design or equipment are made except for core loading which was explicitly modeled using NRC-approved methods and procedures.

e. No changes in the plant design or safety systems were made. No new accident initiators j or egtipment failure modes are introduced with the Cycle 15 core loading pattern and fuel mix. ,

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced i based on the followmg: ;

i

a. For anticipated operational occurrences, the modified bundle is bounded by the GE bundle f which it emulates. For postulated accidents, the reconstituted bundles show an ;

improvement in the LOCA and fuel handling accident events, while sufficient margin is -l calculated to the deposited enthalpy limit for the rod drop accident. !

i

b. There are no changes to the channel dimensional design or structural propert:es which l adversely affect any margin of safety. No additional margin of safety is required for the )

MCPR safety limit. I

c. The replacement water od is consideied a "like-kind" replacement. The modified bundle is functionally identical to the original design.

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1994 ANNUAL OPERATING REPORT EDWIN 1. IIATCII NUCLEAR PLANT UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS L

d. MCPR and overpressurization safety limits are protected for all transient conditions, L provided the unit is operated in conformance with Unit 1 Technical Specifications and the k power distribution limits in the Cycle 14 COLR.

c. MCPR and overpressurization safety limits are conservatively protected under all limiting transient conditions, provided the plant is operated in conformance with Unit 1 Technical Specifications and the power distribution limits in the Cycle 15 COLR. Safety design margins for cold reactivity limits are met for cold shutdown margin and standby liquid control. Fuel thermal and mechanical overpower limits are met for all transients analyzed.

12C-006 Revise Unit I rdAR paragraph S.3.3.3 to reflect the design basis for the SGT system which should have the necessary capacity of two combined SGT system filter trains to reduce and hold -

the refueling floor and the reactor building at a sub-atmospheric pressure under neutral wind conditions.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This change revises the FSAR to be consistent with the design documents and to identify the intended design basis for the SGT system. -

2. The possP.;ility of an accident or malfimetion of a different type than any evaluated previen:y ;

in the FSAR is not created. The proposed change has no impact on the intended desif,n i function of the SGT system. No new accident scenarios or failure mechanisms are created. j

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ,

The Technical Specifications are not impacted by this change. j 12C-007 I

Revise Units 1 and 2 FSARs to eliminate references to the reactor recirculation flow control's i capability of functioning in an automatic control and load following mode.

1. The probability of occunence or the consequences of an accident or malfunction of equipment ,

important to safety previously evaluated in the FSAR is not increased. No plant equipment ;

with the responsibility of maintaining safe operation is dependent upon the recirculation flow ;

control system's automatic load following capabilities. No contributing impact upon plant i operations regarding the absence of automatic control capability occurs as it relates to a feedwater controller failure or a recirculation flow controller failure. Plant operators are still ;

able to manually compensate for changes in the cere power due to loss of a feedwater heater :

by means of power monitoring alarm annunciation, If the neutron flux level exceeds its alarm i

Page55 ;

1994 ANNUAL OPERATING REPORT EDWIN L IIATCII NUCLEAR PLANT UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS setting, scram safety actions are set in motion by the operation of the neutron monitoring, reactor protection, and CRD systems.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. Only the accidents identified in the Unit 1 FSAR regarding the recirculation flow control system manual mode ofoperatire are applicable. The new controllers are more reliable than the old ones. The Unk. I FC AR only addresses the manual mode of operation.

3. The margin of safety as defined in the bases of the Technir.al Specifications is not reduced. No

. safety limits or setpoints of any safety equipment are affected. There are no adverse effects on any plant transient response. No acceptance limits will be increased, and no failure points will be decreased.

12C-004 Revise the Unit 1 FSAR to reflect the plant's capability to cope with a loss of offsite ac power to the essential and non-essential electrical buses concurrent with a turbine trip and the unavailability of the redundant onsite emergency ac power systems for a duration of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Emergency diesel generator IB, in conjunction with the Class IE 125/250-V station service batteries, is capable of powering the required loads for the electrical and instrumentation components needed for core cooling and decay heat removal following an SBO. Areas of the plant housing components required for SBO coping have environmental conditions, which are either below the component environmental qualification design, or are only slightly above design and are well below the minimum generic limit established in NUMARC 87-00. Plant procedures were established to ensure that necessary doors are opened and remain open during an SBO to prec;ude MCR heatup. To limit a temperature rise in the MCR during an SBO, " egg crate" ceili: g tiles were installed in the MCR. Equipment operability during accidents other than SBO will not be impacted

2. The possibility ofV. accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The detailed SBO coping analysis does not identify any new events, other than SBO that are not identified in the FSAR. No malfunction of equipment important to safety of a different type has been identified in the analysis that has not been previously evaluated in the FSAR. NRC final review and acceptance of the SBO event evaluation has been prosided.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. 1 The plant response to SBO for the required duration is within the analyzed conditions for system capabilities and equipment design. The basis for an existing Technical Specification i

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Page 56

1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS will not change as the result of this revision. The present safety design basis for any system or component is unchanged.

12C-008b Revise the Unit 1 FSAR to reflect the plant's capability to cope with a loss of offsite ac power to the essential and non-essential electrical buses concurrent with a turbine trip and the unavailability of the redundant onsite emergency ac power systems for a duration of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Emergency diesel generator IB, in conjunction with the Class IE 125/250-V station service batteries, is capable of powering the required loads for the electrical and instrumentation components needed for core cooling and decay heat removal following an SBO. Areas of the plant housing components required for SBO coping have environmental conditions, which are either below the component environmental qualification design, or are only slightly above design and are well below the minimum generic limit established in NUMARC 87-00. Plant procedures were established to ensure that necessary > ors are opened and remain open during an SBO to preclude MCR heatup. To limit a temperature rise in the MCR during an SBO, " egg crate" ceiling tiles were installed in the MCR. Equipment operability during accidents other than SBO will not be impacted

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The detailed SBO coping analysis does not identify any new events, other than SBO that are not identified in the FSAR. No malfunction of equipment important to safety of a different type has been identified in the analysis that has not been previously evaluated in the FSAR. NRC final review and acceptance of the SBO event evaluation has been provided.

3. The margin of safety as dermed in the bases of the Technical Specifications is not reduced.

The plant response to SBO for the required duration is within the analyzed conditions for system capabilities and equipment design. The basis for an existing Technical Specification will not change as the result of this revision. The present safety design basis for any system ci component is unchanged.

12C-011 Modify Units I and 2 FS ARs to clarify the existing plant configuration regarding MOV limit and torque switches.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated la the FSAR is not increased. Conformance to design requirements ensures the required equipment response for events analyzed by the FS AR.

Page 57

1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS Proper operatien for safety-related MOVs is assured. No change in equipment or operations is introduced.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new configurations or scenarios that could result in accidents not previously evaluated are introduced. The design configuration and setting requirements '

were fully evaluated to assure proper operation of safety-related MOVs.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No acceptance limits are increased, and no failure points are decreased. Improved documentation and the potential for improved configuration control provide greater assurance that the margin of safety as identified in the Technical Specifications is maintained.

12C-012 7 Revise Unit 1 FSAR subsection 10.4.4 to state how the SFP switch and skimmer surge tank ,

switches are actually tested.

l. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The function of the levelinstruments remains the same. These instruments are not safety-related and are still periodically tested to verify system operability. This change only clarifies the type of testing performed.

2. The possibility of an accident or malfunction of a different type than any evaluated previously ,

in the FSAR is not created. All previous accident scenarios evaluated in the FSAR remain the same, and no other accident scenarios are possible. The instruments are not safety-related, and their function remains the same.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The SFP margin of safety as defined in the Unit 1 Technical Specifications is not impacted by this change.

12C-022 Revise the Unit 1 FSAR to clarify the method by which visitors in the visitors center are notified ,

of plant emergencies. ,

l. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No system or .

I equipment is being changed. The assumptions or operator actions of any previously evaluated accident or transient are not affected by this change.

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1994 ANNUAL OPERATING REPOllT EDWIN I. HATCH NUCLEAR PLANT '

UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The revision does not involve changing the operation of the plant.

As a result, no unanalyzed modes of operation are being introduced.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This revision does not involve the Technical Specifications. The actions being added to the FS AR are from the Hatch Emergency Plan, which has been previously reviewed and approved .

by the NRC.

12 D-012 Revise Unit 2 FS AR sections 13.1 and 17.2 to change "SNC President and Chief Operating Oflicer" to " President and Chief Executive Officer," and " Chief Executive Officer" to " Board of Directors" where applicable. ,

L

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This administrative change does not involve physically alter the plant or revise any setpoint or operating parameter. The operation, maintenance, and testing of the plant are not changed, nor are any i

changes made to any current procedures addressing plant operations. The accident and malfunction analyses in the FSAR remain unchanged. The potential for a dose above 10 CFR 100 limits is not increased.

2. The possibility of an accident or malfunction of a different type than any evaluated previcWy in the FSAR is not created. No limiting condition for operation, limiting safety system setting, or safety limit is affected by this administrative change, since neither the physical design nor the operation of the plant is affected.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Adequate processes are still in place to assure line management remains focused on the i operational support of HNP.

12 D-029 !

j Add historical information regarding river water temperature to Unit 2 FSAR table 2.4-2.

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This administrative change does not physically alter the plant or revise any setpoint or operating parameter. The operation, rnaintenance, and testing of the plant are not changed. The highest maximum river water temperature in the time frame added is less than the highest maximum river water temperature previously recorded in the FSAR. The accident and malfunction analyses in the Page 59 ,

1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS 1

FSAR remain unchanged, and the potential for a dose above 10 CFR 100 limits is not increased. 1 I

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. This administrative change does not affect any limiting condition i for operation, limiting safety system setting, or safety limit.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This administrative change does not affect the safety aspects of plant operation.

12D-031 Revise Unit 2 FSAR Figure 13.1-7 to reflect the new position of Vice President and Corporate Counsel for SNC.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This administrative change involves no physical alteration to the plant or changes to setpoints or operating parameters. The physical operation, maintenance, and testing of the plant are not changed.

Accident and malfunction analyses remain unchanged and the potential for a dose above the 10 CFR 100 limits is not increased.

2 The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The limiting conditions for operation and safety limits are not affected by this administrative change. No potential exists for a dose increase above the 10 CFR 100 limits.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No physical design or operation of the plant is changed. In accordance with Section 6 of the Technical Specifications, clear lines of authority continue to exist. This position is not in '

direct line management for day-to-day decisions involving plant operations. Control functions are thus unaffected. The Technical Specifications are not affected.

13A-029 Update the resume for Manager Health Physics and Chemistry in Unit 2 FSAR table 13.1-1.

1. The probability of occurrence, or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This administrative change does not affect the design or operation of any plant system, component, or stmeture.

The inclusion of a current resume in chapter 13 does not afTect the design or operation of any facet of the plant. This change does not introduce or create a new failure mode.

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1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT

. UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS.

2. The possibility of an accident or malfunction of a different type than any evaluated previously ;

in the FSAR is not created. This administrative change does not involve any plant system, i component, or structure. j i

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.. !

This administrative change is not addressed in the Technical Specifications. ;

F9C-004 l i

Revise FHA section 4.2 to show that the use of Knowool, FP-60, or solid metal covers for cable i trays is acceptable for achieving the 30 minute fire rating required for maintaining divisional i separation and reducing combustible loading and, therefore, can be used interchangeably for these ,

. applications. Revise FHA Appendix F to reflect the general design criteria of sections 1 and 2 as ;

they relate to NFPA Codes 72E and 72D. Delete section 3 since other sections of the FHA and i the fire area discussions provide sufficient information to identify the selection and location of fire j detection devices for a given fire area. l

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The overall reliability i of the Plant Hatch Fire Protection Program will not be adversely affected by this change. The j

. design intent of maintaining divisional separation is not compromised. Thew changes are +

within acceptable design limits. l t

j

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new accident scenarios are created. Overall reliability of the !

Plant Hatch Fire Protection Program will not be adversely affected by this change. This i revision protects safety-related raceways to support conformance with the single-failure !

criteria design basis of the plant. !

i

3. The margin of safety as defined in the bases of the Tedmical Specifications is not reduced. No j margin of safety is affected because no failure points are decreased and no acceptance limits l are increased.

l REA HT-93630 l l

Determine the minimum requirements for the use of kaowool as a fire resistive barrier at Plant ;

Hatch, and document these locations on design drawings. Due to the impact this REA has on i safety-related raceways, perform calculations to address the effects of cable ampacity derating and !

seismic loading. 'l l

1. The probability of occurrence or the consequences of an accident or malfunction of equipment i important to' safety previously evaluated in the FSAR is not increased. The appropriate l analyses were performed to support any plant modifications resulting from this REA. Analysis ;

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Pagc 61 t

L 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT

- UNIT 1/ COMMON 10 CFR 50.59 SAFETY EVALUATIONS determined that these changes to the FSAR are within acceptable design limits. Kaowool helps to mitigate the propagation ofinternally generated raceway fires from affecting both divisions of safe shutdown equipment in support of the single failure criteria design basis of the plant. Thus, it aids in preventing a malfunction of safety-related equipment. ;

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. This REA identifies and provides for protection of safety-related raceways to support conformance with the single failure criteria design basis of the plant. No new accident scenarios are created as a result of this REA.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Determining the minimum requirements for the use of Kaowool and documenting the design drawings does not affect the margin of safety as defined in the bases for any Technical Specification.

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s 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT f UNIT 210 CFR 50.59 SAFETY EVALUATIONS !

12B-002 Revise several plant position titles to reflect current site organization. Change the responsibility

_ for maintaining plant training and indoctrination records from the manager training and emergency ;

preparedness to the plant administration manager. Also, make editorial and grammatical corrections to enhance clarity and readability. ;

1. The probability of occurrence or the consequences of an accident or malfunction of equipment j important to safety previously evaluated in the FSAR is not increased. This change is strictly administrative and editorial in nature and has no effect on plant design or operation. ;

2. The possibility of an accident or malfunction of a difTerent type than any evaluated previously in the FSAR is not created. The proposed change does not affect the design or operation of ;

any plant system, structure, er component.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. .

The proposed change is administrative and editorial in nature and does not affect the design or !

operation of any plant system or component.

12B-013 l Revise Unit 2 FSAR reference figure 9.2-1 to indicate that the recirculation MG set coolers are '

serviced by the RBCCW system, not the PSW system.

i

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No design drawings or accident analyses are affected by this change. The applicable PSW system P& ids indicate the

- PSW system does not service the recirculation MG set coolers. This is an FSAR documentation change which does not impact the operation or function of the PSW system.

2. The possibility of an accident or malfunction of a different type than any evaluated previously ,

in the FSAR is not created. This is an FSAR reference figure documentation change. No new l accident scenarios or modes of failure are introduced by this change.

3. The margin of safety as defined in the bases of the Technical Specificatioru is not reduced.

This is an FSAR documentation change to a reference figure. No Technhl Specifications operating limits or surveillance requirements are impacted by this change.

128-016 Revise Unit 2 FSAR paragraph 8.3.2.2.1 to state that battery performance discharge testing is performed in accordance with section 5.2 ofIEEE 450-1980, instead of section 4.2.

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3 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT UNIT 210 CFR 50.59 SAFETY EVALUATIONS 1

1. The probability of occurrence or the coraequences of an accident or malfunction of equipment i

important to safety previously evaluated in the FSAR is not increased. No equipment or system function is changed, and operating conditions as previously evaluated are also unchanged. This action does not degrade safety system performance during an accident.

2. The possibility of an accident or malfunction of a different type than any evaluated previously -

in the FSAR is not created. This change does not create any new unevaluated accident

_ possibilities. Since no equipment changes are made, no new malfunction is created.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

This change does not have an impact on physical parameters or conditions, nor is it reflective of a plant condition that is in conflict wjth the Technical Specifications.

12B-025 Revise Unit 2 FSAR sections 1.2,4.4, 5.5,7.1,7.7,10.2, and 15.1 to clarify that the RFCS '

reactor recirculation master flow controller has been replaced with a digital controller, and is programmed to function only in the manual mode.

1. The probability ofoccurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the FSAR is not increased. For generator load rejection, although the master controller will function only in the manual mode, compensation l for an increase in main generator speed will continue to be accomplished. With loss of a l feedwater heater, the thermal power monitor remains intact to monitor the neutron flux and l the heat flux for any adverse effect. Since the only mode of operation is manual, operator )

action remains as previously evaluated. The operator reduces core flow when the neutron flux j or heat flux alarm is activated. No equipment used to maintain safe operation is dependent on l RFCS automatic load following capabilities. System response is not adversely affected by the modification.

2. The possibility of an accident or malfunction of a different type than any evaluated previously {

in the FSAR is not created. The RFCS continues to perform its original design function while l providing for increased reliability of operation. Experiences at other power plants with l controllers of this type have shown them to be no more sensitive to interference than the )

analog controllers they are replacing. The new controllers will operate only in the manual mode, which is addressed by the Unit 2 FS AR.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. No safety limits or setpoints of any safety equipment are afTected. There are no adverse effects on any plant transient response. No acceptance limits will be increased, and no failure points will be decreased.

Page 64

1994 ANNUAL OPERATING REPORT EDWIN 1. HATCH NUCLEAR PLANT ;

UNIT 210 CFR 50.59 SAFETY EVALUATIONS 128-028 ,

Revise Unit 1 FSAR section G.5 and Unit 2 FSAR supplement 15C to make the allowable repair time statements consistent with the Technical Specifications.

1. The probability of occurrence or the consequences of an accident or malfunction' of equipment !

important to safety previously evaluated in the FSAR is not increased. All out-of-service times :

are within the limits specified in the Technical Specifications.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. All out-of-service times are within the limits specified in the '

Technical Specifications.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. All out-of service times are within the limits specified in the Technical Specifications.

12C-001 ,

Update Unit 2 FSAR figures 2A-17 and 2A-18 to reflect changes in building settlements which '

have been evaluated as acceptable compared ta previously established allowables. Marker location discrepancies are corrected, and the format of the curves is changed to facilitate updating and enhance readability.

1. The probability ofoccurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Differential and total settlement values are within established acceptance limits. Plant design and operation are unaffected.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created. The new settlement values are within established acceptance limits. No new accident mechanisms or failure modes are created for any system or structure.

System operation is unaffected. ;

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

Building settlement is not addressed in the Technical Specifications. No acceptance limits or t failure points of any system or structure are affected.

F9B-001 Revise FHA Appendix D to reflect that the fire brigade is trained once per calendar quarter in t accordance with NFPA guidelines and per Units I and 2 Technical Specification 6.4.2.

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n 1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 210 CFR 50.59 SAFETY EVALUATIONS

1. The probability of occurrence or the consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the FSAR is not increased. This activity is being done in order to clear up a discrepancy between the Technical Specifications and the FHA.

The plant currently meets Technical Specifications criteria (based on NFPA guidelines) for fire brigade training. By nature, this activity has no impact or influence on the operation of any plant equipment. Therefore, there is no basis for which any type of accident or equipment malfunction could occur as the result of this activity.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created. By nature, this activity has no impact or influence on the operation of any plant equipment. Therefore, there is no basis for which any type of accident or equipment malfunction could occur as the result of this activity.

3. The margin of safety as dermed in the bases of the Technical Specifications is not reduced.

This activity does not change any Technical Specifications criterion regarding fire brigade training.

F9B-006 Revise FHA sections 10.1,11.1,11.2, and 11.3 to remove required manual actions for two locations which have no emergency lighting. A switch required manipulation to assure power available at MCC R24-S018 A, but an alternate manual action in an illuminated location removed the requirement. RHR valve 2El1-F065B was to be locally verified opened or manually opened if it had closed, but reevaluation determined it would fail to its required position. Also, a requirement to protect a redundant circuit to the torus water temperature recorder was removed.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the FSAR is not increased. The safe shutdown operation of the equipment will not adversely change, and the existing function of the equipment will remain the same. This change will not alter, degrade, or prevent any required accident response, nor will it degrade any fission product barriers or change any assumptions made in evaluating the radiological consequences of an accident. Also, it does not impose any additionalloads on the affected systems, downgrade any support system performance, affect the frequency of operation of any equipment, or impose increased or more severe testing requirements.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. No new modes of failure are introduced. The required operational control of the components will be unchanged, and analysis has shown their safe shutdown operation to be acceptable. I

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced. ,

This change does not affect the limiting conditions for operation or surveillance requirements !

Page 66

V rs 1994 ANNUAL OPERATING REPORT EDWIN 1. IIATCII NUCLEAR PLANT i

UNIT 2 to CFR 50.59 SAFETY EVALUATIONS for any_ system important to safety. No acceptance limits are increased, and no failure points ;

are decreased. .

13A-004 Revise the calculated accident doses in the Unit 2 FSAR section 15.1 accident analyses tables.

l

1. The probability of occurrence or the consequences of an accident or malfunction of equipment '

important to safety previously evaluated in the FSAR is not increased. The change only affects the method of calculating the doses resulting from postulated accidents. The change in dose calculation methodology is approved by the NRC. The revised calculation doses are well within regulatory limits. This revision does not affect the availability or performance of ;

equipment important to safety.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FS AR is not created.. The change involves only calculations of radiological doses resulting from postulated accidents. Equipment important to safety is not affected.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The change involves only the calculation of radiological doses resulting from postulated accidents. The revised calculated doses are within regulatory limits.

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i Page 67

1994 ANNUAL OPERATING REPORT EDWIN I. IIATCII NUCLEAR PLANT UNIT 1 TEST OR EXPERIMENT REQUESTS 94 001. Rev. O Determine the ability of the Unit I and Unit 2 Standby Gas Treatment (SGT) Systems to draw and maintain less than 1/4" vacuum (water gauge) in less than 120 seconds.

The test was designed to start both Unit I trains and one Unit 2 train. In order to test the ability of two fans to pull a vacuum on both containments, one fan from each Unit was started for a portion of the test. The running SGT fans in both cases were lined up to the refueling floor and their respective reactor buildings. No systems or components were made inoperable.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Secondary containment isolation signals, which took the ventilation systems to their fail safe configuration, were ,

simulated. SGT and/or normal ventilation are not precursors to any type of accident.

Secondary containment was isolated, and no unanalyzed ventilation lineups were introduced.

None of the ventilation logic isolation signals were blocked; thus, no equipment was made inoperable.

l

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. SGT and/or normal ventilation systems are not precursors to any accident. No unanalyzed lineups were perfoimed. Critical temperatures were monitored, and procedural instruction was provided to terminate the test should any critical temperature approach trip setpoints.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The operability of safety-related systems or components was unaffected. Secondary containment was isolated, and ventilation systems were taken to their fail safe configuration.

Page 68

1994 ANNUAL OPERATING REPORT EDWIN !. IIATCII NUCLEAR PLANT UNIT 2 TEST OR EXPERIMENT REQUESTS94-001. Rev. O Determine the ability of the Unit I and Unit 2 Standby Gas Treatment (SGT) Systems to draw and maintain less than 1/4" vacuum (water gauge) in less than 120 seconds.

The test was designed to start both Unit I trains and one Unit 2 train. In order to test the ability of two fans to pull a vacuum on both containments, one fan from each Unit was started for a portion of the test. The running SGT fans in both cases were lined up to the refueling floor and their respective reactor buildings. No systems or components were made inoperable.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Secondary containment isolation signals, which took the ventilation systems to their fail safe configuration, were simulated. SGT and/or normal ventilation are not precursors to any type of accident Secondary containment was isolated, r.nd no unanalyzed ventilation lineups were introdued.

None of the ventilation logic isolation signals were blocked; thus, no equipment was made inoperable.

2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. SGT and/or normal ventilation systems are not precursors to any accident. No unanalyzed lineups were performed. Critical temperatures were monitored, and procedural instruction was provided to terminate the test should any critical temperature approach trip setpoints.

3. The margin of safety as defined in the bases of the Technical Specifications is not reduced.

The operability of safety-related systems or components was unaffected. Secondary containment was isolated, and ventilation systems were taken to their fail safe configuration.

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'1994 ANNUAL OPERATING REPORT EDWIN 1. IIATCII NUCLEAR PLANT DATA TABULATIONS AND UNIQUE REPORTING REQUIREMENTS Page 70

1994 ANNUAL OPERATING REPORT EDWIN L IIATCH NUCLEAR PLANT OCCUPATIONAL PERSONNEL RADIATION EXPOSURE FOR 1994 This section satisfies the requirements of Edwin I. Hatch Nuclear Plant Units 1 and 2 Technical Specifications Section 6.9.1.5 and assures compliance with the Code of Federal Regulations as set forth in pertinent sections of Title 10. Special attention was afforded to the methods prescribed by the Commission in Regulatory Guide 1.16 in order that the intent, as well as the letter of these laws, might be fulfilled by providing meaningful information as to the degree ano circumstances of all exposure of personnel at this facility. An indication of the effectiveness of the plant radiation ;

program may be inferred from the large number ofind viduals with no measurable exposure or with minimal dose.

The time period covered by this tabulation extended from January 1,1994 through December 31, 1994. Individual exposures as indicated by Electronic Direct Reading Dosimeters (EDRDs) were recorded daily with use of an ALARA Computer System. These exposures and the differences between these readings and the most restrictive exposure limit were tabulated, printed, and posted on a daily basis and upon individual request. The corresponding EDRD results, as recorded on the dosimetry files, were supplanted by thermoluminescent dosimeter measurements made over a period of approximately one calendar quaner as the data became avaliable.

Each person listed in ti e dosimetry files was assigned a usual job category based on daily activities. The sixjob categories of this nature are identified in the following table. Running totals of doses acquirec' in each of these categories were placed in each person's dosimetry file.

Each dosimeter reading, in addition to being retained for exposure records, was added for indisidual exposure records and to the total representing the cumulative dose in the appropriate '

job category.

The implicit assumption involved in this method of accounting for exposure in different tasks is that all exposure acquired in job categories other than the usual will be documented by a Radiation Work Permit. This circumstance should prevail in all cases. .

Further delineation to the number of persons and amount of exposure to individuals in difTerent job categories by various personnel categories is indicated by the standard reporting format of Regulatory Guide 1.16. Each personnel dosimetry file contains the personnel category information required to accomplish this completion. The individual running dose totals for each job were used by the ALARA Computer to compute the number of man-rem indicated in each group. Backup disc files and hard copy records, as printed by the ALARA Computer, were maintained.

Using the ALARA Computer, dosimetry information was compiled, retained, and tabulated in such a manner as to satisfy the peninent Federal Regulations and Plant Technical Specifications.

The system is organized to provide the information in the format specified by these requirements and the suggestions of the Regulatory Guides.

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o e 1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT ,

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Regulatory Guide 1.16 Report Information for 1994 ~

Georgia Power Company - Nuclear Generation ,

L Plant E.I. Hatch - Licenses: DPR-57 , NPF-5 ,

No. of Personel > 100 mrem Total Man-Rem 1 Task and Job Function Station Utility Contract Station Utility Contract l REACTOR OPERATIONS AND SURVEILLANCE !

Maintenance & Construction 119 2 56 33.52 .507 29.45 l Operations 81 0 0 47.074 0 .124

- Health Physics & Laboratory 53 3 45 19.593 .432 16.912 :

Supervisory & Office Staff 33 2 7 11.444 .426 2.278 :

Engineering Staff 12 0 5 4.05 .129 2.546 !

ROUTINE PLANT MAINTENANCE Maintenance & Construction 180 12 671 114.932 3.981 322.337 ,

Operations 8 0 1 4.491 0 .196 Health Physics & Laboratory 6 0 6 4.444 .111 2.474 Supervisory & 7Cfice Staf f 10 1 17 3.312 .603 6.436 ;

Engineering Stat? 12 1 36 6.242 .346 20.046 INSERVICE INSPECTION Maintenance & Construction 4 0 61 2.267 .06 32.57 operations 0 0 0 .004 0 0 '

Health Physics & Laboratory 9 0 23 9.389 .033 21.605 Supervisory & Office Staff 3 2 0 1.217 .244 .546 Engineering Staff 2 1 22 .8 .262 8.101 SPECIAL PLANT MAINTENANCE Maintenance & Construction 38 1 98 18.646 .205 49.473 Operations 2 0 0 3.484 0 0 Health Physics & Laboratory 1 0 7 .406 0 1.827 Supervisory & Office Staff 5 0 2 3.444 .042 .799 Engineering Staff 1 1 13 .603 .282 4.907 ;

WASTE PROCESSING J Maintenance & Construction 2 0 27 1.091 .061 8.413 Operations 1 0 0 .727 0 .08 '

Health Physics & Laboratory 11 0 18 3.252 0 6.187 Eupervisory & Office Staff 1 0 0 .253 .038 .079 Engineering Staff 0 0 0 .036 0 .073 I

REFUELING OPERATIONS Maintenance & Construction 0 0 55 .571 .083 18.436 Operations 3 0 0 2.096 0 .016 Hetlth Physics & Laboratory 0 0 0 .011 0 .174 Supervisory & Office Staff 0 0 1 .423 .018 .21 Engineering Staff 0 0 20 .239 .008 5.253 SITE TOTALS Maintenance & Construction 343 15 968 171.023 4.898 460.702 Operations 95 0 1 57.875 0 .416 Health Physica & Laboratory 80 3 99 37.095 .576 49.19E Supervisory & Office Staff 52 5 27 20.116 1.384 10.362 Engineering Staff 27 3 96 11.969 1.027 40.921 Grand Totals: 597 26 1191 298.078 7.885 561.600 :

Site Man-Rem Total: 867.563 l I

Page 72 i

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1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT REACTOR COOLANT CIIEMISTRY ,

Tabulations on a monthly basis of SJAE isotopic values and reactor coolant parameters, as required by section 4.6.F.1 for the Unit 1 Technical Specifications, are found in the following

. tables. Unit 2 values are also shown, although it is not required they be reported. Isotopic values l lis:ed as "0" are less than the lower limit of detection for the counting system. l f

i f

I l

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i l

i, v

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l Page 73

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1994 ANNUAL OPERATING REPORT EDWIN L HATCH NUCLEAR PLANT .

l UNIT I l 1994 SJAE ISOTOPICS uCi/SEC P

DATE 1994 atWT Xe-133 Xe-135 Xe-138 Kr-85m Kr-87 Kr-88 I6 ,

Jan 6 2436 8.77 E-1 1.22 El 2.55 E2 2.19 E0 1.42 El 8.11 EO 3.15 E2 Feb 3 2436 7.44 E-1 1.23 El 2.95 E2 2.36 E0 1.60 El 8.41 EO 3.35 E2 l h1ar 3 2436 8.50 E-1 1.23 El 3.18 E2 2.40 E0 1.80 El 8.20 EO 3.59 E2 l i Apr 7 2436 6.15 E-1 9.14 E0 2.44 E2 1.76 E0 1.17 El 7.42 E0 2.75 E2 51ay 5 2436 5.74 E-1 8.33 E0 2.09 E2 1.63 E0 1.05 El 5.71 E0 2.36 E2 June 2 2436 6.39 E-1 9.04 E0 2.08 E2 1.74 E0 1.11 El 6.14 E0 2.37 E2 July 7 2436 6.00 E-1 9.10 E0 2.11 E2 1.70 E0 1.20 El 6.20 E0 2.41 E2 Aug4 2436 6.09 E-1 9.16 E0 2.39 E2 1.64 E0 1.32 El 6.07 E0 2.70 E2 Sep1 2436 6.07 E-1 7.93 E0 1.85 E2 1.50 EO 8.59 EO 5.40 E0 2.09 E2 Oct 10 0 0 0 0 0 0 0 0 i Nov11 2436 3.12 E-1 5.42 E0 1.47 E2 1.41 EO 9.51 EO 5.12 E0 1.68 E2 Dec 1 2436 4.80 E-1 6.20 E0 1.71 E2 1.40 EO 9.86 EO 5.25 E0 1.95 E2 REACTOR CIIEMISTRY IODINES uCi/ml DATE 1994 51WT I-131 1-132 I-133 I-134 I-135 del-131 Jan 6 2436 1.80 E-6 1.31 E-4 5.51 E-5 5.56 E-4 1.52 E-4 4.36 E-5 Feb 3 2436 0 8.08 E-5 4.42 E-5 3.52 E-4 8.19 E-5 2.76 E-5 51ar3 2436 5.44 E-6 1.37 E-4 4.88 E-5 5.60 E-4 1.37 E-4 4.45 E-5 Apr 7 2436 0 1.47 E-4 5.78 E-5 6.22 E-4 1.61 E-4 4.49 E-5 Stay 5 2436 2.98 E-6 133 E-4 5.26 E-5 5.06 E-4 1.46 E-4 4.28 E-5 i June 2 2436 3.14 E-6 1.48 E-4 5.21 E-5 6.31 E-4 1.42 E-4 4.51 E-5 July 7 2436 3.86 E-6 1.29 E-4 5.04 E-5 5.06 E-4 1.41 E-4 4.25 E-5 Aug 4 2436 u 1.01 E-4 4.68 E-5 3.81 E-4 1.19 E-4 3.26 E-5 SepI 2436 0 1.15 E-4 4.54 E-5 4.84 E-4 1.11 E-4 3.39 E-5 Oct 10 0 0 0 0 0 0 0 Nov11 2436 2.43 E-6 6.32 E-5 4.91 E-5 4.01 E-4 !.32 E-4 3.58 E-5 Dec 1 2436 0 9.31 E-5 3.51 E-5 3.89 E-4 9.74 E-5 2.76 E-5 I

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1994 ANNUAL OPERATING REPORT EDWIN L IIATCII NUCLEAR PLANT ,

UNIT 2 1994 :

SJAE ISOTOPICS uCi/SEC l DATE 1994 MWT Xe-133 Xe-135 Xe-138 Kr-85m Kr-87 Kr-88 I6 !

Jan 7 2071 7.23 El 1.04 E3 2.47 E4 1.96 E2 1.01 E3 7.29 E2 2.79 E4 '

Feb 4 1705 5.20 El 6.35 E2 1.44 E4 1.10 E2 7.04 E2 4.13 E2 1.63 E4 Mar 4 1705 5.10 El 6.21 E2 1.49 E4 1.09 E2 7.18 E2 3.61 E2 1.68 E4 Apr 0 0 0 0 0 0 0 0 May 11 2436 1.38 El 2.70 E2 6.66 E3 5.27 El 3.22 E2 1.73 E2 7.50 E3 June 2 2436 2.52 El 2.96 E2 7.77 E3 5.45 El 3.47 E2 1.97 E2 8.69 E3 July 1 2436 2.01 El 2.97 E2 7.32 E3 5.13 El 3.42 E2 1.86 E2 8.22 E3 Aug 5 2436 1.99 El 2.84 E2 6.52 E3 5.00 El 3.23 E2 1.88 E2 7.39 E3 Sep 16 2436 1.66 El 2.60 E2 6.35 E3 4.63 El 2.87 E2 1.72 E2 7.13 E3 j Oct 7 2436 2.08 El 2.88 E2 7.18 E3 5.18 El 3.20 E2 1.70 E2 8.03 E3 Nov 4 2436 2.31 El 2.86 E2 6.64 E3 4.86 El 3.07 E2 1.86 E2 7.49 E3 .

Dec 2 2436 1.54 El 2.86 E2 7.42 E3 4.86 El 3.24 E2 1.58 E2 8.25 E3 REACTOR CIIEMISTRY IODINES uCi/mi DATE 1994 MWT I-131 1-132 1-133 I-134 I-135 DEI-131 Jan 7 2071 1.94 E-4 8.65 E-3 3.70 E-3 4.81 E-2 1.02 E-2 3.17 E-3 Feb4 1705 1.60 E-4 6.78 E-3 3.09 E-3 3.50 E-2 8.02 E-3 2.50 E-3 Mar 4 1705 1.80 E-4 7.88 E-3 3.41 E-3 3.98 E-2 9.15 E-3 2.82 E-3 Apr 0 0 0 0 0 0 0 ;

May II 2436 2.06 E-4 2.92 E-3 2.08 E-3 2.25 E-2 4.45 E-3 1.66 E-3 l June 2 2436 9.20 E-5 4.19 E-3 1.60 E-3 2.46 E 2 4.34 E-3 1.45 E-3 l July 1 2436 6.84 E-5 3.80 E-3 1.45 E-3 2.00 E-2 3.94 E-3 1.26 E-3 l Aug 5 2436 1.18 E-4 4.28 E-3 1.51 E-3 2.58 E-2 3.97 E-3 1.45 E-3 l l

Sep 16 2436 6.79 E-5 3.98 E-3 1.45 E-3 1.94 E-2 3.72 E-3 1.24 E-3 Oct 7 2436 8.17 E-4 3.59 E-3 1.46 E-3 1.80 E-2 3.88 E-3 1.24 E-3 Nov 4 2436 1.79 E-4 3.90 E-3 1.86 E-3 1.93 E4 4.44 E-3 1.52 E-3 Dec 2 2436 5.91 E-5 3.13 E-3 1.14 E-3 1.71 E-2 2.97 E-3 1.02 E-3 l l

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HL-4787, 1994 Annual Operating Period

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