Text

1991 Annual Rept Part 2
	ML20095K764
Person / Time
Site:	Vogtle
Issue date:	12/31/1991
From:	Mccoy C; GEORGIA POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	ELV-03678, ELV-3678, NUDOCS 9205060042
	Download: ML20095K764 (173)
	v • d • e

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April 29, 1992 ELV-03678 001609 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen: V0GTLE ELECTRIC GENERATING PLANT 1991 ANNUAL REPORT - PART 2 In accordance with the applicable regulatory requirements, Georgia Power Company hereby submits Part 2 of the 1991 Annual Report of operating information. It includes the re n itider of the 1991 reports not previously submitted. Sincerely,

                                                        . f C. K. McCoy CKM/JLL/gmb

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Enclosure:

Annual Report - Part 2 xc: Georaia Power Company Mr. W. B. Shipman Mr. M. Sheibani NORMS U. S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle

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l GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET NOS. 50-424 AND 50 425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 1991 ANNUAL REPORT - PART 2 TABLE OF CONTENTS

1. INTRODUCTION II. PLANT MODIFICATIONS AND TESTS OR EXPERIMENTS o PLANT MODIFICATIONS o TESTS OR EXPERIMENTS 111. EMERGENCY CORE COOLING SYSTEMS OUTAGE DATA REPORT IV. ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT V. ANNUAL ENVIRONMENTAL OPERATING REPORT

I GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 INTRODUCTION The_ Vogtle Electric Generating Plant Units 1 and 2 are powered by pressurized i water reactors, each rated at 3411 megawatts thermal. It is located on the Savannah River in Burke County Georgia, 34 miles southeast of Augusta. The Unit 1 operating _ license was received on January 16, 1987 and commercial operatinn started on May 31, 1987. Unit 1 is operating in its fourth fuel cycle. Unit 2 received its operating license on February 9,-1989, and began commercial' operation on May 20, 1989. Unit 2 is in a refueling outage, preparing to enter

  .its_ third fuel cycle.

l l I

4 - a 4 11 GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY- OPEPATING LICENSE NOS. NPF-68 AND NPF-81 PLANT MODIFICATIONS AND TEST OR EXPERIMENTS

4 , s. s - 2 II 1991 ANNUAL REPORT - PART 2 l 10 CFR 50.59 (b) PLANT MODIFICATIONS

P 1 II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT PLANT DESIGN CHANGES

Subject:

DCP 87-V1E0098, Revision 0, Sequence 1 Description : This change replaces the ultrasonic transmitter _ system for crud tank level indication with a pressure transmitter system. This system will measure tank head pressure and provide indication and annunciation at the remote backfluchable i filter panel for a more reliable level indicating system. Safety Evaluation: This change provides local indication of Crud Tank level prior to starting the Crud Tank Pump. This change will have no effect on accidents described in the FSAR nor will any new accident scenarios be introduced. The margin of safety is increased by providing a more reliable back flushable crud tank level indicating system.

Subject:

DCP 87-V1E0099, Revision 0, Sequence 1

Description:

This change' deletes containment Isolation Phase "A" function from radiation monitors 1RE- 0005 &OOO6 and the associated annunciation with reset functions. Annunciation windows-will be blanked and spared along with the reset switches. Safety Evaluation: Deletion of this input from the Containment isolation phase A does not-affect the accident evaluations of FSAR chapter 15. This change does not increase the chance for any accident, there is sufficient input from several redundant features that will isolate containment prior to a high radiation signal. This change does not disable the function of the radiation monitors and they will provide the required level of accident L monitoring. l 1

Subject:

DCP 87-V1NO362 Revision 0, Sequence 1

Description:

This change adds a temperature switch to the Auxiliary / Turbine Building Train A Electrical Tunncl and removes the automatic operation of the ventilation system for this area. This is required due to the tunnel flow path being blocked by fire doors at either end. Safety Evaluation: The change does not affect the ability of the ventilation system to maintain the tunnel temperature below the environmental qualification temperature. The temperature switches will provide an alarm in the control room to alert plant operators to take action.

Subject:

DCP 87-V1N0462 Revision 0, Sequence 1

Description:

This change replaced two 3/4" - 1500 p.s.i. flanges located on the end of a double valved test vent located on the Safety Injection line "B" train. This change was required to enable the vent piping to withstand OBE (Operating' Basis Earthquake) stresses within the code allowances. L Safety Evaluation: This change was made between outside containment isolation valve HV-8802B and Train "B" S.I. pump. This change does not affect the Containment test valves or the testing acceptance criteria, l l l ! 4 l'

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 87-V1N0471 Revision 0, Sequence 1

Description:

This change adds a protective relay to the "B" train RHR pump room cooling fan control circuit to prevent loss of control power in case of a fire. This change was required due to both trains of RHR pump room temperature controllers being located in the same fire area. Safety _ Evaluation: This change provides' control power protection and does not affect.the design operation of the .an. All modifications are located inside the existing MCC cubicle. The change involves only fire event safe shutdown and assures that the RHR pump room cooler system is available.

Subject:

DCP 88-VCN0025 Revision 0, Sequence 1

Description:

This change. replaces 11 portable dry chemical fire extinguisher, intended for suppression of control, remote shutdown, and computer rooms, with portable Halon 1211 fire extinguisher. Safety Evaluation: -This change adds a new type of fire extinguisher in the plant'in order to mitigate the effects of-extinguisher discharge on sensitive equipment. This change affected no equipment assumed to function in an-accident.

                     .FSAR figures 9A-19, and 9A-23 required revision as a result of this change.

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5 II 1991 ANNUAL-REPORT - PART 2-

                                                    ; 10 CFR50.59(b) REPORT

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Subject:

DCP 08-V1N0046 Revision 0,- Sequence 1

             --Description:                  This change extended-fire suppression system
                                             /079 by extending coverage to-Control building room R-178. This is required because of.the:significant amount of safety related equipmenttin the room and its inaccessibility.

for: manual -fire suppression. Safety: Evaluation: This change lis necessary to meet the. requirements of CMEB 9.5-1. and provides the necessary protection for cable tray in the room. The-following FSAR pages required revision:

Page 9A.1.86-3, Table 9.5.1-10b(2-of 5),

Figure 9A-21, Page 9.5.1-11, Figure 9.5.1-1 (8_&-10)Tas did the fire protection

                                            ; pre-plans.

Subject:

DCP 88-V1N0068 Revision 0, Sequence 1

Description:

.               'This; change. eliminates the' actuation of sprinkler system 1-2301-S4-077 upon an alarm from fire protection. zone "133A". This restores the protection / scheme to the design

~ configuration of; system -077 serving zones "133B"-and "176" and-system -108 serving zone "133A". L Safety Evaluation:- This chan.gejto.the fire protection local zone L indicacing panel restores the system to.the

                                            ! design configuration as described.in the FSAR. This t:hange will not reduce-the ability ofJsafety related components to function as required to mitigate the consequences of an 3

accident. l. I i l' 6

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l t II-1991" ANNUAL REPORT - PART 2 > 10 CFR50. 59 (b) ~ REPORT

Subject:

. .DCP 88-V1N0077 Revision 0, Sequence 1

Description:

This change modifies the Data Processing _ Module counting function _and high and low alarm-default parameters for the plant radiation monitors. This change is to reduce the high. number of false and spurious radiation alarms'being generated. The count changed from 16 counts'to 1024 counts to provide a more representative sample, the counting period remained at 245 ms.-The Data Processing Module-(DPM)

                                . default value for high and low radiation alarms was changed to 7.77 E 29 to reduce the-           -

probability of high radiation alarms when-the DPM-is' turned-on or reset after a power failure. Safety Evaluation:' This change-does not affect the ability of

                                 ,the radiation monitors to function as.

designed. Unnecessary operator action and erroneous input into the SSPS-are eliminated. The new default settings are displayed on a

                                -loss of-power or DPM' reset to indicate to-the1 operator that:a DPM reset has-occurred. This change will notiaffect:the response times:of
                                 .any equipment assumed,to function in an-accident.

Subject:

DCP 88-VCE0110 Revision 0,' Sequence 1

Description:

- This change modifies the fire detection systemito reflect the completion of Unit 2 as follows: changes fire protection computer addresses;1 allows actuation of Unit-2

sprinkler, system'013 by Unit 1 zone 12;-

reconnects Unit 2 detection zones connected to Unit 1-during construction. Safety. Evaluation: ~ This change does not!does notLaffect any safety related equipment nor does it impact any. safety / accident-analysis described in-the FSAR. The change is a-completion of the original design intent. i 7 ! l

l II 1991 ANNUAL REPORT - PART 2 10 CFR50. 59 (b) REPORT

Subject:

DOP 89-V1N0003 Revision 0, Sequence 1 l

Description:

This change replaces the' existing solenoid operated Nuclear Sampling System globe valve 1HV-8220 with a solenoid operated gate valve. This change will minimize in-line leakage and provide-more reliable position indication. Safety Evaluation: This new solenoid operated gate valve meets all design requirements of the original valve, provides greatly reduced in line leakage and provides more reliable valve indication. The intended function of the J I valve as described in the FdAR is not affected. This change does not affect the valve closure time specified in FSAR Table 6.2.~4-1. FSAR-Table 6.2.4-1(3 of 10), Figure 6.2.4(11 of 12), and Figure 9.3.2-1(2 of 2) required revision as a t#sult of this change.

Subject:

DCP 89-V1N0015 Revision 0, Sequence 1

Description:

This change installed pressure gauges and isolation valves on the MSIV's actuator's hydraulic and nitrogen systems. This was to enhance the-operator's ability to determine the hydraulic and nitrogen system pressure in a more expediticus manner and determine the cause of control room alarms. Safety Evaluation: This change does not effect the operation of the MSIV's or their ability to respond as required by the FSAR. The change enhances the trouble shooting of alarms to aid the operator in determining plant status and corrective actions. This modification maintains the design criteria and the qualification'of the MSIV's. 8

y II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 89-V2E0018 Revision 0, Sequence 1

Description:

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 89-V2N0075 Revision 0, Sequence 1

Description:

This change adds provisions for an electrically powered air compressor to be used as_the supply of breathing air in lieu of bottled air. This also deletes the control room annunciation f' aw air pressure. Safety Evaluation:- This change does not affect any system assumed to function during an accident. The breathing air system is only in service during plant outages.The low breathing air pressure alarm will be replaced with local alarms on portable bottled backup supplies. FSAR Section 9.3.1, Tablec 3.2.2-1,and 6.2.4-1 required revision to reflect this change.

Subject:

DCP 89-VCE0077 Revision 0, Sequence 1

Description:

This change modified the access and egress into the Radiologically Controlled Area (RCA). This was accomplished by the addition of removable partitions resulting in separate ! pathways into and out of the RCA. Safety Evaluation: This change is a is a physical modification to the facility but does not affect the plant description of FSAR Section 1.2. The change is located in a seismic category 2 area and will have no effect on any safety related equipment or plant response to an accident. 12

This change removes the protective ring from the Reactor Internals Lifting Device. The protective ring is used during refueling to protect the Reactor Vessel flange 0-Ring surface. Safety Evaluation: The removal of the ring does not affect the accident analysis of fuel damage. The ring is to protect the flange not to absorb the impact of dropping internals. The flange O-Ring surface is inspected for defects prior to vessel assembly per plant procedure. 13

l l i II 1991 ANNUAL REPORT - PART 2 10 CFR50. 59 (b) REPORT

Subject:

DCP 89-V2N0116 Revision 0, Sequence 2 l

Description:

This change adds three 480 volt, 3 phase electrical power feeders to provide permanent ; supply points for outage needs. One supply is l located at the Auxiliary Maintenance Building and the other two side U2 Containment Building level 1. Sr.fety 6 valuation: The additional temporary electrical feeder to the Auxiliary Maintenance Building is powered from 2NB02. This_ supply is non-1E from a seismic category 2 structure and will have no effect on the plant. The power supplies to the containment required _a change out of overcurrent protection to a Microversa Trip instead of the ECS type. This required revision of FSAR Figure 8.3.1-7(sheet 9 of 19)-to demonstrate proper penetration overcurrent protection coordination. Table 16.3.5 was revised to add circuit breakers 2NB0902 and 2NB0914 as the penetration overcurrent protection devices. The feeder is not required during plant operation and was installed using qualified

DCP 89-V2N0291 Revision 0, Sequence 1

Description:

This change added packing leakoff drains and welded a disc nut retainer to the disc stem of the Extrrction Steam check valves. This change was made to comply with the vendor's recommendations. Safety Evaluation: Extraction Steam is a non-safety related system which is enveloped by the accident analysis of FSAR sections 3.5, 15.3, and 15.5. This change makes the check valves more reliable and will not adversely affect their operation. 16

II 1991 ANNUAL REPORT - PART 2 10 CFRSO.59(b) REPORT

Subject:

DCP 89-V1N0294 Revision 0, Sequence 1

Description:

This change replaced two existing swing check valves with sliding plate check valves on the discharge of the Unit 1 and Common reciprocating air compressors. These compressors are used to supply Instrument and Service Air systems. This change improved reliability and reduce maintenance on the check valves.(ref. SCER 86-03). Safety Evaluation: Although there are safety-related air operated valves that are supplied air by the affected systems, none of these devices require a source of air to perform their safety related function. The system function is not affected and these changes should increase the reliability of the check valves and decrease the maintenance demands on them

Subject:

DCP 89-V1NO313 Revision 0, Sequence 1

Description:

This change Provides for the replacement of Steam Generator Feed Pump Turbir: (SGFPT) 15V DC and 30V DC power supplies with General Electric supplied equivalent power supplies. This is required because the original power supplies were discontinued. Safety Evaluation: The replacement of the ISV DC and 3G V DC power supplies does not alter the function or operation of any existing safety related system. This change is to non-safety related, non 1E equipment, whose function or operation does not change 17

8 ; >, . 11 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT l Subject; DCP 89-V2NO314 Revision 0, Sequence 1

DCP 90-V1N0019 Revision 0, Sequence 1

Description:

This change removed the internals from the check. valves for NSCW pump motor bearing cooling lines. This was to prevent loss of cooling to the pump due to a check valve failure. Safety Evaluation: Removal of the check valve internals does not alter the system operation but does remove the possibility of malfunction as a result of a-valve sticking shut. . FSAR Figures 9.2.1-1, sneets 1 & 2 of 5; 9.2.1-1 andLTable 3.9.B.3-9, sheet 1 of 6 were revised to reflect the check valve internals removal. 20

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II 1941 ANNUAL REPORT - PART 2 M CFR50.59(b) REPORT

Subject:

                 ' W- 40-V1N0040 Revision 0, Snquence 1 99scription:         This change allows replacement of the existing Crosby Watts safety valves with consolidated safety valves for the diesel generator air start compressor safety valves.

These valves provide overpressure protection for the compressor, aftercooler and the interconnecting piping. Safety Evaluation: The Consolidated valves meet or exceed the design operating parameters of the original valves. Due to a much higher temperature rating and the use of stronger materials, those replacement valves will provide more reliable service. This. change will have no affect on the capability of the safety-related portions of the diesel generator starting air syctem to perform its function.

Subject:

DCP 90-V2H0041 Revision 0, Sequence 1

Description:

This change allows replacement of the existing Crosby Watts. safety valves with Consolidated safety valves for the diesel generator air start. compressor safety valves. These valves provide overpressure protection for the compressor, aftercooler and the interconnecting piping. I safety Evaluation: The consolidated valves meet or exceed the design operating parameters of the original valves. Due to a much higher temperature rating and the use of stronger materials, these replacement valves will provide more reliable service. This change will have no affect on the capability of the safety-related portions of the diesel generator starting air system to perform.its function. l 21 L L

i II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

90-V1H0061 Revision 0, Sequence 1

Description:

This change replaced Auxiliary Bldg. doors l 14C and 207 with fire and pressure rated doors capable of withstanding the pressure and temperature required for the High Energy Line Break Analysis (M*LBA) d conditions in the rooms they access. Safety Evaluationt These replacement doors are not addressed in chapter 15 of the FSAR and are not a factor in the evaluation of any accidents described in the section. The design function of the doors as pressure and fire barriers will-mitigate any incident at the door openings in : regard to the communication-of steam, heat, or radiation from a postulated pipe break or fire as described in Sections 3.6, 3.11, and 9.5.1 of the FSAR. Furthermore, the doors will ensure that the door openings conform to the design basis described in the FSAR.

Subject:

DCP 90-VIN 0064 Revision 0, Sequence 1

Description:

This change replaced the thormal relief valves on high pressure .cedwater heaters 6A and 6B with more reliable _ pilot operated relief valves. This is to resolve reoccL; ring leakage problems with the existing valves. Safety Evaluation: The replacement valves weet the requirements of thu original specification and the valve setpoints do not change. The replacement valves will enhance the system reliability and performance because they are better suited to handle the rapid pressure pulsations and rescat following the return to normal system pressure. 22 . ~ __ . _ _ _ . _ _ . . _ _ _ _ . _ . _ . _ _ _ _ . _ _ _ _-

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Subjects DCP 90-V1N0067 Revision 0, Sequence 1

Description:

This change removed the internal components of check valves associated with the unit 1 ESF chilled water system (trains A & B). This was done because these check valves were not required for the chilled water system. Safety Evaluations These valves were originally provided in the chilled water system as standard engineering practice and not to prevent any particular backflow concern. The prevention of backflow is not a concern with the ESF chilled water system. There are no normally open parallel flow paths around-the single chilled water pump and the system is a closed loop. FSAR Table 3.9.B.3-9, 9.2.9-3 and Figure 9.2.9-1 required revision to reflect this change. 23 a__ a.

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Subjects DCP 90-V2N0072 Revision 0, Sequence 1 l Descriptions The lower taps for steam generator narrow range level instruments were lowered tu locations below the transition cones of the steam generators. This stabilized the steam generator narrow range level system at low power levels and lowered the potential for level-related reactor trips. Safety Evaluationt No new performance requirements are being imposed on the level system or steam generator such that any design criteria will be exceeded. System and component integrity are maintained commensurate with criteria discussed in the FSAR. The predicted doses presented in the FSAR for such transients as rod ejection, steam

                                                      -generator tube rupture and LOCA remain valid since the mass release will not exceed that which is currently assumed in the FSAR.

Fission product barrier integrity is not affected by this modification nor is any equipment which is assumed to mitigate the radiological consequences of an accident. No new failure modes were defined for any system or component modification nor has any new single limiting single failure been identified. Technical Specification Table 2.2-1, low-low l steam generator water level trip setpoint and l Table 3.3-3, ECF actuation system setpoints were revised to reflect this modification. t 24

II 1991 ANNUAL REPORT - PART 2 10 CFR50. 59 (b) REPORT ]

The following FSAR sections required revision as a result of this chnnge Table 15.0.3-2 (sheet 1 of 3); Table 25.0.6-1; Section 15.1.2.1, 15.1.2.2.1, 10,1.2.2.2, 15.1.2.3; Section 15.1 reference; Figure 15.1.2-1, 15.1.2-2; Table 15.1.2-1 (sheet 1 of 2); Section 15.2.6.'1, 15.2.6.2.1, 15.2.6.2.2,

15.2.6.4; Figure 15.2.6-1, 15.2.6-2; Table 15.2.3-1 (sheet 3 of 5), 15.2.3-1 (sheet 2 of -

5); Section 15.2.7.1,-15.2.7.2.1, 15.2.7.2.2; Figure 15.2.7-1, 15.2.7-2; Section 15.2.8.1, 15.2.8.2.2, 15.2.8.3; Section 15.2 reference; Table 15.2.3-1 (sheet 3 of 5), 15.2.3-1 ; (sheet 4 of 5), 15.2.3-1 (sheet 5 of 5); Figure 15.2.8-1, 15.2.8-2, 15.2.8-3, 15.2.8-4,15.2.8-5, 15.2.8-6, 15.2.8-7.

Subject:

DCP 90-VIN 0073 Revision 0, Sequence 1

Description:

This change replaced the existing blind flange on the water drain connection on the Diesel Fuel Oil Storage (DFOS) Tanks with a pipe flange, pipe spool picco, and threaded cap. This change was made to facilitate performance of the Tech Spec Surveillance requirements for these tanks. Safety Evaluation: The new fittings were installed such the connections were water tight and will not allow for water intrusion into the tanks. The modification will perform the same function as the original installation and will not change the operat!.on of the system. FSAR Figure 9.5.4-1: required revision to reflect this change. l l I l l l 25

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Subjectt DCP 90-V2N0074 Revision 0, Sequence 1 Descriptions- This change replaced the existing blind flange on the water drain connection on the , Diesel Fuel Oil Storage (DFOS) Tanks with a l pipe flange, pipe spool piece, and threaded cap. This change was made to facilitate ; performance-of the Tech Spec Surveillance requirements for these tanks. i Safety Evaluation The new fittings were installed such the connections were water tight and will not allow for water intrusion into the tanks. The modification will perform the same function as the original installation and will not change the operation of the system. FSAR Figure 9.5.4-1 required revision to reflect this change. t 26 -, i -

l I l II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 90-VIN 0075 Revision 0, Sequence 1

== Description:== This change deleted the main and auxiliary feodwater temperature monitoring system by deleting all syster. temperature indication, differential temperature indication, and l differential temperature alarms from the main control board. The remaining system thermovells, temperature elements and cabling up to and including the boards in the QBCP associated with the temperature indications, will remain in place. Safety Evaluation: The Feedwater Temperdture monitoring system is non-safety related and is utilized only during low power operations. Its deletion will therefore, have no effect on.the safety function of the Auxiliary Feedwater system. Although the potertial for feedwater system water hammer has not been entirely eliminated, the possible existence of conditions in the feedwater piping necesscry for water hammer have been significantly reduced by a combination of feedwater piping configuration, steam generator design, and procedural changes. Therefore the monitoring of the feedwater temperature ad j acent to the steam generator-is no longer required. The worst case design basis accident (loss of feedwater due to a line break) remains unchanged. Therefore the deletion of the main and auxiliary feedwater temperature monitoring system is enveloped by existing system accident analyses. 27

                       - -     . .        .- ~ - - -        _. _~ -- . - . . - . . . - . --.. .-

This change added two manual vents to the Spent Fuel Pool Cooling system return lines from the train A and B spent fuel pit heat exchangers. This was done to allow maintenance on the heat exchangers and .

                                   . downstream components without lowering the level in the spent fuel pool below Tech Spec limits.

Safety Evaluation: The vents are safety-related, project class l 313, once installed they do not perform any l active safety function and serve to maintain l pressure integrity. The installation of these L vents complies with the design criteria l- applicable to the Spent Fuel Pool Cooling I system.

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28 y1-Mw --- -

1 1 II 1991 ANNUAL REPORT - PAkT 2 , 10 CFR50.59(b) REPORT j

Subject:

DCP 90-V1N0096 Revision 0, Sequence 1 l

Description:

This DCP involved several changes: A) Relocated the reactor protection system (RPS) / turbino emergency trip supply t (ETS) fluid pressure transmitters IPT-6161, 6162, 6163 senoir.g points from the sensing port on the fast-acting solenoid valves to the supply side tubing of the control valve. B) Included an isolation and calibration valve at each pressure transmitter installation. C) Relocated the pressure transmitters from the control valve to tubing supports near the control valves to reduce vibration influence. D) Replaced a portion of the rigid electro-hydraulic control (EHC) tubing near the control and stop valves with 24 flexible Teflon hose assemblies with stainless steel braided outer jacket. E) Added isolation valves in the ETS and fluid actuation supply (FAS) tubing to provide isolation of EHC fluid for control and stop valve maintenance. ' F) Changed the project class designations . for the EHC hydraulic fluid power units and control coolers in FSAR Table 3.2.2-1 from Class 424 to Class 626. Safety Evaluation: A) The new connections meet the same classification requirements as the original tap points. This change removes the direct reactor trip above 50% power when the main control valves are closed without a turbine trip. The reactor trip on turbine ' trip will still occur above 50% power when a valid turbine trip signal is generated and the ETS. pressure is relieved. This change required revision of FSAR section 10.1.2. B)_ The added valves are environmentally qualified and mounted in accordance with plant design criteria. The instrumentation installation, including all tubing connections meet the project class requirements for their installation. Consequently, this change ments all current l design criteria for the link between the I transmitters and the sensing points. L 29 u

_ . - ~ - _ . _ - - - . _ . .- . - - . . . . - II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT C) Location of the pressure transnitters do not require seismic mounting since they are located in the turbine building. The tubing and connections required to support this l relocation meet the original design and i project class requirements. i D) The flexible hose installations meet the requirements of the original rigid tubing installations. This change will affect neither the operation nor response of the EHC system, including the related turbine overspeed protection system. E) The isolation valves are suitable for hydraulic service and can operate with a process pressure of 1600 psi and temperature of 180 degrees Fahrenheit. The valves provide positive indication of position and may be locked in position. Isolation of an ETS or ' FAS line before the affected main stop or control valve is closed results in a condition outside the analyzed configuration of the turbine overspeed analysis. ' Administratively controlling the position of the isolation valves will assure the isolation valves are not closed when the main stop or control valves ove open. Any time an ETS or FAS fluid line is required to be isolated, the affected main stop or control valve must be clcaed first. F) The EHC system hydraulic fluid power units, control coolers and some piping were engineered, procured and installed to project class 424 requirements, this is an. incorrect project class designation for this system. Project class 424 components also moet the requirements for class 626, which is the correct classification of the system. Ctanging the project classification will not h ve any effect on the syste:n as the existing 424 components meet the requirement for 626 classification. 30-5)

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT ibject: DCP 90-V2N0097 Revision 0, So.mence 1

Description:

This DCP involved : A) Relocated the reactor protection system (RPS) / turbilee emergency trip supply (ETS) fluid pressure transmitters 2PT-6161, 6162, 6163 sensing points from the sensing port on the fast-acting solenoid valves to the supply side tubing of the control valvo. B) Included an isolation and calibration valve at each pressure transmitter installation. Safety Evaluation: A) The new connections meet the same classification requirements-as the original tap points. This change removes the direct

                                                                                                    -reactor trip above 50% power '. hen the main control valves are closed without a turbine trip. The reactor trip on turbine trip will still occur above 50% power _when a valid turbine trip signal is generated and the ETS pressure is relieved.

This change required revision of FSAR section 10.1.2. B)-The added valves are environmentally qualified and mounted in accordance with plant design criteria. The instrumoistation instal ~tation, includlag all tubing connections meet the project class requirements for their installation. Consequently, this change meets all current denign ~riteria for the link between the transmitters and the sensing points. l 31

II 1991 ANNUAL REPORT - PART ? 10 CFR50.59(b) REPORT Subject DCP 90-V1N0103 Revision 0, Sequence 1

Description:

This change added a retaining bolt through the crank-arm spring pin fastener of damper 1-TV-12097A (Diesel Generator ESF HVAC System) actuator linkage to prevent dislocation of the pin and subsequent dampar actuation failure. This damper is located in the Unit 1 Train A Diesel Generator Building. Safety Evtluation: The addition of the spring pin retainer will prevent damper failure caused by dislocation of the spring-pin, and maintain the original design _ function of the Diesel Generator Building ESF HVAC System.

Subject:

DCP 90-V2E0113 Revision 0, Sequence 1 .

Description:

This chb. ige installed flange joints down stream of the drain valves to the steam generator blowdown heat exchargers. This will al,'.ow removal'of the Steam Generator blowdown heat exchanger heads without cutting the drain lines. Safety Evaluatict.: Tha portion of the system that was modified is outsido containment and is classified as non-safety. It performs no function related to the safe shutdown of the plant and is not involved in any accident described in the FSAR. The modification meets all criteria for its project class. FSAR Figure 10.4.8-1 was revised to reflect 1 this chat.go. L 32 L L~

II- I 1991 ANNUAL REPORT - PART 2 l 10 CFR50.59(b) REPORT Subject; DCP 90-VIE 0115 Revision 0, Sequence 1

Description:

This change provided the option to use e split or cartridge type mechanical seal or the existing packing on the Turbine Plant Cooling Water (TPCW) Pumps. This change also routed utility water to the TPcw pumps to une as cooling and flush water for the mechanical seals. Safety Evaluation: The TPCW pumps serve no safety function nor do they impact the operation of any safety related components. Therefore, this modification will have no impact on the plant accident response. This changa revised FSAR Figures 9.2.11-1

(sheet l'of 3), and 10.4.5-1 (sheet 2 of 2).

33

 .c.             -

l II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT i

Subject:

DCP 90-V2N0119 Revision 0, Sequence 1

Description:

This change added a time delay to the Steam Generator Sample line isolation valves to allow them to be opened 30 seconds after an Auxiliary Feedwater actuation. This is to allow the plant operator a positive method of determining whether a steam generator tube leak has occurred. Safety Evaluation: Installation of a time delay relay in the control circuitry of the steam generator sample line to allow opening of the valve following an AFW auto-start signal will have L no effect on the ability of the AFW system to perform its intended safety function. Sufficient margin exists following the opening of one steam generator sample valve to ensure that adequate AFW is delivered to the intact steam generators. Plant Emergency' Operating procedures were revised to reflect this change and to limit the sample line opening to one line at a time. FSAR Sections 7.3.7.1.and'9.3.2.2.3 required l revision to reflect this change. This modification will not effect the ability of the steam generator sample line to close upon receipt of the AFW auto-start signal and due to the time delay relay can only be opened 30 seconds after valve closure. This change did not cffect the control room indication or' ability of the plant operator to manually close the steam generator sarple valve at any time. L l i-1 L 34

l II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b). REPORT

Subject:

DCP 90-V1N0123 Revision 0, Sequence 1

Description:

This DCP replaced ASCO solenoid valves on Instrument Air System Dryers with ASCO long life valves. This will decrease down time and maintenance. Safety Evaluation: This change enhances the performance of the instrument air system. This' system is non-safety related and will not affect the required response of any air operated safety related component as these components are designed to perform their safety function without air pressure available.

Subject:

DCP 90-V2N0124 Revision 0, Sequence 1 l rescription: This DCP replaced ASCO solenoid valves on

                              ,    Instrument Air System Dryers with ASco long life valves. This will decrease down time and maintenance.

Safety Evaluation: This change on!.ances the performance of the instrument air system. This system is non-safety related and will not affect the required response of any air operated safety related component as these components are designed to perform-their asfety function without air pressure available. i e 35

II 1991 ANNUAL REPORT - PART 2 10 CFR50. 59 (b) REPORT

Subject:

DCP 90-V1N0125 Revision 0, Sequence 1

Description:

This change modified the three Unit 1 condensers as follows: A) A Aew sparger and supports was added to the main steam drain pot manifold (connection 61 on condenser "B") B) The connections on condensers "A" & "C" corresponding to connection 61 on "B" condenser were renumbered to differentiate them. These will continue to have an internal baffle only. C) A note was added to the condenser drawings to allow installation of the hot well suction

                                                                                                                                       . screens in sections (condensers "A", "B",   &
                                                                                                                                        "C").

Safety evaluation: These changes were made to correct internal damage discovered during 1R1 refueling outage. These changes were designed and installed to the original design and project class criteria. They will not affect the operation of the condensers, design capacities, or radiation monitoring of the air removal.

Subject:

OCP 90-V1N0130 Rcvision 0, Sequence 1

Description:

This change added strut support members to RHR pump motor tag number 1-1205-P6-002-H01 to reduce operating vibration level. Safety Evaluation: The additional strut support members will not violate the seismic qualification of the RHR pump and motor assembly or support. This modification will not adversely affect the operation of the RHR or any other safety-related system. 36

l II 1991 ANNUAL REPORT - PART 2 l 10 CFR50.59(b) REPORT l

Subject:

DCP 90-V1N0130 Revision 0, Sequence 1

Description:

This chtnge allowed the use of 1 7/8" diameter SA-193 Gr. B7 hold down bolts on RHR l pump tag number 11205P6002 in lieu of the 2" I diamator SA-193 Gr. B7 bolts originally l supplied. The new bolts were torqued to 240-250 ft-lbs. Safety Evaluation The equipment manufacturrro (Westinghouse and Ingersoll-Rand) were contacted concerning the

                        -use of smaller hold down bolts. Westitighouse stated that the original seismic, analysis assumed a 1 7/8" diameter SA-193 Gr. B7 bolt.

Therefore, the 1 7/8" diameter bolts meet the original analysis and will not adversely affect the RHR system response.

Subject:

DCP 90-V1N0134 Revision 0, Sequence 1

Description:

This DCP documents the "as-found" cold setting of snubbers and modifies pipe support (V1-1201-117-H601) to facilitate the snubber design cold setting, located on the branch lines (RTD return and Reactor Coolant Loop Drain) of the RCS Loop 2 crossover leg. Safety Evaluation: 7he cold setting position of the snubbers on the RCS branch line piping has been analyzed and conforms to the original design criteria. Any redistribution of piping loads within the system as a result of the "as-found" cold position of snubbers were evaluated and compensated for, where necessary, with verification to the existing pipe support system. All supports within the piping system were reviewed to assure adequacy of pipe support design as a result of changes in loade/ displacements-resulting from the piping evaluations. = 37

1 i 1 II 1991 ANNUAL REPORT - PART 2 i 10 CFR50.59(b) REPORT Subject DCP 90-V1N0141 Revision 0, Sequence 1 d

== Description:== As.a result of the Station Blackout Analysis the 30 amp breakers 1AY1A-05 and 20 amp breakers 1CY1A-03 and 1DY1B-05, in the Vital 120V AC distribution panels 1-1807-Q3-VII, VI2,VI3, and VI4 were replaced with 35 and 30 amp breakers respectively. Similarly 15 amp breakers.1AD12-08 and 1BD12-03 in the 125V DC distribution panels 1-1806-03-DA2 and DB2 were replaced with 20 amp breakers. Safety Evaluation: The Breaker ratings breaker I coordination (including short circuit analysis) and the associated cable sizes were analyzed for adequacy and safety in the and found acceptablo. This modification will prevent inadvertent tripping of the breakers during Station Blackout condition due to loss of ventilation to the AC and DC switchgear rooms. 38

II i 1991 ANNUAL REPORT - PART 2 l 10 CFR50.b9(b) REPORT

Subject:

DCP 90-VIN 0143 Revision 0, Sequence 1

Description:

This change added uninterruptibio power supply (UPS) units for the Emergency Notification Network (ENN), Emergency Notification System (ENS), Administrative Decision Line (ADL), and PABX system Merlin Equipment. Additionally, duplex Wall receptacles with twist lock receptacles were added to prevent inadvertent removal of power to the UPS units. The plug for the control room ENN will be replaced with a NEMA LS-ISP plug to prevent inadvertent power removal. Safety Evaluation: A fault in a UPS unit cannot cause a malfunction in any safety-related power source. The panels feeding the UPS units are not safety-related. Separation criteria between the non-lE lighting panels and the safety-related load centers which feed them is maintained. The batteries in the UPS units are the sealed maintenance-free type and produce no gasses when charging. The UPS unit and battery module in the control room will be positioned to prevent contact with the QEAB panel during-a seismic event by placing the units side by side on a skid resistant mat. Due to the dimensions and weight of the equipment, non-seismic supports utilized. i l V 39 l

s . _ _ _ _ _ - _ __.. . _ _ _ _ . - - . . _ - - - _ _ - _ . . . _ _ _ _ ._ II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 90-V2N0149 Revision 0, Sequence 1

== Description:== This change inverted the reactor vessel level instrumentation system (RVLIS) reactor vessel head pressure transducer. This was to reduce calibration inaccuracies caused by air in-leakage during refueling operations. { l Safety Evaluation: The RVI.IS provides control room operators ! with information to monitor and assess the ! reactor coolant inventory following an I accident to ensure that adequate core cooling is available. By-re-orienting the pressure , transducer and minimizing the effects of air ' in-leakage, which affect instrument calibration, the safety-related function of the system will be enhanced. Remounting of the pressure transducer was performed in accordance with the original design details so as not to affect the seismic qualification of the equipment.

Subject:

90-VIN 0153 Revision 0, Sequence 1

== Description:== This change corrected the control and coordination logic.of the Regenerative Heat Exchanger inlet and outlet isolation valves and reduced the possibility of damage to the Regenerative Heat Exchanger. Sa.ety Evaluation: No new components are being added by this change, only the timing of valves 1HV-8149A, B, & C are being changed. Lotdown isolation, containment isolation, and HELB isolation on this line are not affected.

Subject:

90-V1N0162 Revision 0, Sequence 1

== Description:== This change remodeled and refurbished the control room kitchen. Safety Evaluation: The kitchen equipment used was reviewed for affects to other systems, equipment, and components with no impacts found. The new installations meet all original and current design standards and applicable codes. 40

l i II 1991 ANNUAL REPORT - PART 2 . 10 CFR50.59(b) REPORT

Subject:

90-VCN0177 Revision 1, Sequence 1

Description:

Subject:

DCP 90-V2N0197 Revision 0, Sequence 1

Description:

This change stiffened motor supports for Auxiliary building normal HVAC supply units 2-1551-A7-001 & 002 to reduce fan motor vibration. Safety Eva?.uation: The normal HVAC system has no safety design basis and is not required to function after a safe shutdown 1 earthquake (SSE) . This change will allow the air handling units to operate within the original limitations for maximum allowable vibration and will not affect the normal system operation or response. 1

s 43

f. )

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT l Subject DCP 90-V2N0198 Revision 0, Sequence 1 Jescription: This change relocated the pressure l transmitters for the Reactor Protection ) System from the control valve to tubing j supports near the control valves to reduce vibration influence. Replaced a portion of the rigid electro-hydraulic control (EHC) tuoing near the control and stop valves with 24 flexible Teflon hose assemblies with stainless steel braided outer jacket. Safety Evaluation: Location of the pressure transmitters do not require seismic mounting since they are located in the turbine building. The tubing and connections required to support this relocation meet the original design and i I project class requirements. ' The flexible hose installations meet the requirements of the original rigid tubing installations. This change will affect neither the operation nor response of the EHC system, including the related turbine overspeed protection system. 1 44

11 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 91-VIN 0015 Revision 0, Sequence 1

Description:

This change provided for one of the following options to be performed on the steam line drain pot high level drain valves:

1. a) change out existing stainless steel

                             -trunion and bonnet bushings with aluminum-bronze bushings. b) Change out existing ribbon-wound graphite packing to teflon packing. c) Upgrado existing actuator spring assembly with a "tiffer spring assembly.

2.) a) Change out existing stainless steel trunn:.on $nd bonnet bushings with aluminum-bronze bushings. b) replace existing Fisher actuator with Orbit actuator equipped with gas-over-oil reservoir. 3.) Change out existing stainless steel trunnion and bonnet bushings with aluminum-bronze bushings. Safety Evaluation: The drain valves modified by this package perform no functions which are required to mitigate the consequences of an accident evaluated by the FSAR. These valves are non-safety related and do not interface with any equipment important to safety.

Subject:

DCP 91-V1N0031 Revision 0, Sequence 1

Description:

Safety Injection Accumulator level system bellows will be-rotated 180 degrees (calibration fitting pointing up). This is to allow complete venting of tha bellows and less drift of the transmitters. Safety Evaluation: The output of Accumulator Tank level transmitters is.non-safety related. However the diaphragms of the bellows sensing sJatem must maintain the pressure boundary of the Safety Injection Accumulator Tanks. Rotation ! of the bellows by 180 degrees does not does not affect the pressure boundary integrity. l l l u 45

                                                                       )
                                   -II 1991 ANNUAL REPORT - PART 2 10 CFR50.59 (b) REPORT

Subject:

DCP 91-V1N0054 Revision 0, Sequence 1

Description:

This change replaced five chemical and Volume control System (CVCS) diaphragm valves with gate s41ves to reduce leakage past the valvaa. Safety Evaluation: The new gate valves neet the design, material, quality and construction ctandards applicable to the CVCS. The repl. aent valves are functionally equivalent to the diaphragm valves. The additional weight of each gate valve has been evaluated fer impact on pipe stress analysis and support loading a.4d been found acceptable. FSAR Figure 9.3.4-1 (sheet 3 of 6) required revisien to reflect this change.

Subject:

DCP 91-V1N0056 Revision 0, Sequence 1

Description:

This change. modified the Main Turbine Lube Oil Reservoir and Conditioner to provide continuous oil flow to the conditioner from the reservoir and to provide a permanent hookup point.for a portable oil polisher for moisture removal. Safety Evaluation: The components affected by this change are not safety related. Malfunction of the lube oil system may result in loss of be.aring oil to the turbine uhich would-result in a turbine trip, however these changes will not affect the turbine lube oil reservoir level. FSAR Figure 10.2.2-1 (sheets 5 and 6) required revision as a. result of this change. i l r i ! 46

l II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Subjects- DCP 91-V1N0057 Revision 0, Sequence 1

== Description:== This change modified the Steam Generator Feed Pump Turbine (SCFPT) Lube 011 Reservoir and conditioner to provide continuous oil flow to the conditioner from the reservoir and to provide a permanent hookup point for a portab;e oil polisher for moisture removal. Safety Evaluation: The Steam Generator Feed Pumps / Turbines and the associated lube oil systems are not - safety related and are not relied on for safe shutdown of the plant during an accident. Low lube oil will result in a SGFPT trip. This change had no adverse affect on reservoir lube oil level since the flow tubes are set

above the reservoir low level alarm. Thic will prevent any adverse affect on the SGFPs due-to oil level. FSAR Figure 10.2.2-1 (sheets 5 and 6) required revision as a result of this change.

Subject:

DCP 90-V1N0070 Revision 0, sequence 1

== Description:== This' change eliminated the redundant Main turbine Trip provided via the 386M lockout relay located in protective relay panel 1-1816-U3-008. The protective relay panel is located in the main control room. Safety Evaluation:- Neither the overspeed protection system nor the turbine trip logic is affected by this change. This change.will not affect the ability of the--386M relay to perform its intended function. The 386M contacts which input to the mechanical and electrical turbine trip circuits perform-no useful function since the turbine has already received a trip signal and is locked out prior to 386M actuation. This change has no affect on the circuits which trip the turbine on a reactor trip or-trip the reactor on a turbine trip.- FSAR Sections 7 & 10 were revised as a result of drawing corrections made during research of the DCP. 47

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 91-V1N0075 Revision 0, Sequence 1

Description:

This change modified the return line from the Spent Fuel Pool Cooling and Purification System to the refueling cavity. This modification consists of the addition of removable piping segments interconnected by quick disconnect fittings. These removable sections and their supports will be installed during refueling outages with the reactor in modes 5 or 6 and will be removed following refueling for decontamination and storage outside containment. SE ation: The Failure modes of the additional piping were evaluated for postulated accident scenarios which could occur during the period this piping is installed. This evaluation revealed no accident scenarios that were not bounded by existing evaluations. The piping is equipped with a siphon-vent to prevent backflew from reducing refueling canal inventory. Subjects DCP 91-V1N0081 Revision 0, Sequence 1

Description:

This change modified the closure permissive on the Steam Generator Feed Pump Turbine (SGFPT) exhaust valve from a main turbit.e trip to the associated feedwater turbine trip. Safety Evaluation: The permissive to isolate SGFPT from the main condenser can occur while the main turbine continues to run. This design change does not affect any transients or accidents that have been postulated which could result in a reduction of the capacity of the secondary system to remove heat generated in the reactor coolant system. 48

This change removed the internals from check-valves-associated with the Feedwater Heater

Moisture-Separator drain tank's high level j dump line.

Safety evaluation: This change does not impact any system assumed to function in order to mitigate the consequences of an accident. 49

y II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

. DCP 91-V1N0113 Revision 0, Sequence 1

Description:

This change modified three areas of the Emergency Diesel Generators (EDG): A) Converted the high temperature jacket water engine trip on each train of the EDGs to trip on normal start only (group II). B) Removed the low pressure jacket water. alarm from the first-out circuitry. Low pressure jacket water trip was not removed. C) Added test aonnections and manual isolation valvt to the engine trip sensor's air supply pneumatic lines. Safety Evaluation: A) This change enhances the performance of the EDG by decreasing the probability of an inadvertent diesel generator trip during an

                    . emergency start. The manual trip function is still available to the operator should a high jacket water temperature alarm be generated and will provide adequate engine protection.

B) The disconnection of the low pressure jacket water alarm from the first-out circuitry will not affect the performance or response of the diesel generator. The first-out annunciator circuitry for the low pressure jacket water trip will-remain. C)The addition of the valved tees and manual isolation valves in the air supply lines only affects the leak testing of the associated lines and sensors. During normal EDG operation, these components will not be active. The following FSAR sections required revision: A) FSAR Sections 8.3.1.1., 8.3.1.1.3 (Table 8.3.1-1) 9.2.1.3 (Table 9.2.1-2), 9.5.5.2.2, 9.5.5.3 (Table 9.5.5-2), 9.5.5.5, ' 9.5.5.7.2.2, 14.2.8.1.64, Figure 8.3.1-3 I l I l 50 l l

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 91-V1N0115 Re\'aion 0, sequence 1

Description:

This change replaces the shuttle valves on the pneumatic control board number 1A-6952 on each Emergency Diesel Generator (EDG) with an "OR" element. Th9 second part of this change added a two second time delay to the jacket water temperature circuits.- Safety Evaluation: Replacement of the shuttle valves with a more reliable "OR" element will enhance-the

                       . performance of the EDG by increasing the start-up reliability. The tubing and logic changes required to make this change will meet all the criteria of the original design.

Description:

- This change' repaired the rotor shaft of ths Steam Generator _ Feed Pump Turbine (SGFPT) 2A. The original shaft was machined at the low pressure end and a new. stub shaft installed in accordance with vendor-drawings. The new

                          -shaft is similar to the original with minor modifications to prevent a similar-, failure.

Safety Evaluation:- This_ change only affected the 2A-SGFPT and

the--ability.to drive the 2A'feedpump. The

                          =SGFP-turbines and_ pumps are not assumed to
                          -function in-accidents analyzed in the FSAR.

The engineering for this-change was provided

                           .by the original equipment. manufacturer and.

the modification meet the original design-

                          -criteria.

l

                                                                                                 -l J

i

                                                                                                 -l 1

j 1 52

r , F II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 91-VIN 0161 Revision 0, Sequence 1

Description:

This change revised the overpower delta-T and overtemperature-delta-T setpoints to support VANTAGE 5 fuel upgrade. These setpoints are calculated and used in the reactor trip system to ultimately prevent the onset of departure from nucleate boiling (DNB). Safety Evaluation: The overpower delta-T and overtemperature delta-T setpoints are being revised to support the VANTAGE 5 fuel upgrade. The revision of these setpoints does not create the possibility of a new or different type of accident from those previously evaluated. It adjusts fuel related parameters related to the VANTAGE 5 fuel upgrade.

Subject:

- DCP 91-V1N068 Revision 0, Sequence 1

Description:

This change disconnected cables 1NRV57EXA & 1XRV574EXB, which run between the Steam Generator Blowdown Panel and the inoperable Solidification Processing Panel. This removed instruments LI-1165B & PI-1166B and made loops.1165 and-1166 operable with the Radwaste Building abandoned. Safety Evaluation: This change removed the instruments from the inoperable Solidification Processing panel and did not affect any safety system

                           -postulated to function in any FSAR accident analysis.

FSAR Figure 10.4.8-1(sheet 2/2) required revision. I i I

53 !

L i

      +                                                                   ,

l

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

DCP 91-V1N0169 Revision 0, Sequence 1

== Description:== This change disconnected cables 2NRV57EXA & 2XRV574EXB, which run between the Steam Generator Blowdown Panel and the inoperable Solidification Processing Panel. This removed instruments LI-1165B & PI-1166B and made loopa 1165 and 1166 operable with the Radwaste Solidification Building abandoned. Safety Evaluation: This change removed the instruments from the inoperable Solidification Processing panel . and did not affect any safety system ; postulated to function in any FSAR accident analysis. l

Subject:

DCP 91-V1N0176 Revision 0, Sequence 1 i

== Description:== This change added a variable spring support cn1 the 10 inch Steam Packing Exhaust Blower Discharge Piping in the Turbine Building in order to prevent damage to this equipment. Safety Evaluation: The variable spring support does not affect the radiation monitor in the condenser vacuum exhaust filter system. The new support will reduce the stress in the piping and decrease the compressive load on the blowers.

Subject:

DCP 91-V1N0179 Revision 0, Sequence 1

== Description:== This change raised the high voltage tap

                  -setting on Reserve Auxiliary Transformer (RAT) 1NXRA and 1NXRB from 98.75% to 100% as evaluated in the Unit 1 load study (calculation X3CA18).

Safety Evaluation: Changing the tap setting from 98.75% to 100% on RATS'1NXRA and 1NXRB will be a beneficial change in that the negative effects of overvoltage on class 1E electrical equipment will be reduced. According to calculation X3CA18 the RAT tap change from 98.75 to 100% will not affect the ability of class 1E motors to operate when required including those class 1E motors required for Loss of Coolant Accident mitigation. 54

II 1991 ANNUAL REPORT - PART 2 10-CFR50.59(b) REPORT

Subject:

DCP 91-V1N0180 Revision 0, Sequence 1

Description:

This change replaced four foodwater flow elements with elements equipped with inspection / cleanout ports. The inspection / cleanout ports will be plugged with a carbon steel plug secured with a 900 lb blind flange connection.

 -Safety Evaluation:   The operation of the flow element remains unchanged, the consequence of a failure remains unchanged. The new flow element is designed to the same parameters as the original flow element.-This will not affect the operation of the Condensate and Feedwater system. The cleanout port flange / plug assembly is designed not to disrupt the flow in the finw element. The port will only-be used for cleaning / inspection activities, and at other times the operation of the flow will be the same as the original installation.

Subject:

-           DCP 91-V2N0182 Revision 0, Sequence 1

Description:

This change replaced the Backflushable Filter Crud Tank level indicator with a digital indicator with a display range of 0-200 inches of water. Safety: Evaluation; The crud tank level indicator does not

                      ~ provide any setpoint alarms or control functions which may affect any plant operation or any equipment assumed to function in an accident analyzed in the FSAR.

This design did not degrade-the original design intent of functional capabilities. 55

        , , - , - .            .,   -                - . - .-          .    .     -        - - . -        _       -_ - .-- . . . . .~ .-.~,. -. -
        /r 4f/                                                                             II 1991 ANNUAL REPORT --PART 2 10 CFR50.59(b) RIPORT                                                                                 ,

Subject:

DCP 91-V1N0183 Revision 0, Sequence 1 Descriptions- This chan'ge replaced the underfrequency relays for the Reactor Coolant Pump Motors with functionally identical but improved relays. Safety Evaluation: The new relays did not. affect the-system performance or operation. The RCP trip for an; underfrequency condition greater than approximately 2.4 HZ-will not be changed. The . new relays are certified to the original requirements and will not adversely affect the seismic ~or environmental qualification of the 13.8 KV switchgear.

Subject:

- 91-V1N0195 Revision 0, Sequence 1

Description:

This; change replaced two 3/4" check valves with a' functionally equivalent 3/4" check

                                                                     . valve, suitable for use in the~ Reactor Coolant. System due to the unavailability of
                                                                                                   ~

identical' replacement valves . Safety' Evaluation: The replacement check valves satisfy all

_ original design and functional' requirements.

The installation of the new valves-did not adversely impact the seismic qualification oor allowable stresses of the~ existing piping installation. J l. L L 56 l_ . [&g ir- +.---.---r--+- - - m ,i- m -. 7 - --y-. y - - - --+-w + -y ,' m.mg e

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

                    - cooling water valve to travel in open direction thereby causing the controlled process temperature to decrease. NRC Letter ELV-00503 requires that operatir    1 convention of these OIM's be such that op .acion of raise. push button.causes.an increase in the controlled parameter and vise-versa for the down push           1 button.

Safety Evaluation: The proposed change involves physical change in the facility, however, the portion of the facility involved is not described in FSAR sections 10.2.2-and 10.4.7.- Subject OIMs are not mentioned any where in Technical l Specifications. The procedures covering these OIMs'are not described in FSAR.

Subject:

MDD # 89-V2M094

Description:

This MDD reverses the operation of five manual push button controllers. They are TIC-5498, 5499, 7097, 7116 and 7355. Presently operation

of raise push-button causes the associated cooling water valve to travel in open direction thereby. causing the controlled process temperature to decrease. NRC Letter ELV-00503 requires that operational convention of these OIM's be such that operation of raise push button causes an increase in the controlled parameter and vise-versa for the down push button.

59

a II 1991 ANNUAL REPORT - PART 2 10 CFR50.59 (b) REPORT Safety Evaluation: The proposed change involves physical change

                                   -in the_ facility, however, the portion of the

facility involved is not described in FSAR '

sections 10.2.2 and 10.4.7. Subject OIMs are not mentioned any where in Technical Specifications. The' procedures covering taese OIMs are not describer in FSAR.

Subject:

MDD # 89-VAM098

Description:

Reduce false alarms on UPS system (Safeguards) Safety _ Evaluation: TheLalarms generated or-the power supply componentsfin this change were not referenced by any FSAR or Technical Specification sections.- No change to these documents is. required. It does not affect any procedures.

Subject:

MDD # 89-V2M102

Description:

- During normal-operation rapid pressure fluctuations cause controller to improperly position valves. This change adds pressure snubbers to common sensing line of heater drain pump seal injection controller 2PC-4380

                                   .and 2PC-4386 for pump B.

Safety-Evaluation: This change does-not affect the: facility as discussed in FSAR section 10.3.6. _The snubbers appear on installation drawings only. P& ids in FSAR da not change. . There are no procedures affected by this change.- The heater drain pumps are not. safety _related and as such are not addressed in Technical' Specifications. s .

Subject:

MDD:/ 89-V1M116

Description:

Thisl change' installs additional bolted-hardware on room cooler 1555-A7-010-000 to reduceLaxial vibrations at the outboard end of fan housing.

             - Safety Evaluation:   Technical Specifications 3/4.7.11 provides        ,

r operability requirementsLfor the room coolers. This MDD does not affect performance of room coolers. The description of room coolers contained in FSAR is of General Nature (section 9.4.3.2.2.2.C). No change to FSAR or procedures is required. 60

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

MDD # 90-V1M100

Description:

- The existing hole (2" dia.) of the boraflex coupon tree hanger / hook is too small for lifting the coupon tree out of the spent fuel pool for removal the boraflex test coupons. Design an optional boraflex (neutron absorbing material) coupon tree hook for 1-2202-A6-001 and A6002 that has a larger diameter " lifting eye" (coupon tree hook / crane hook interface) and to provide for a larger hook for hanging the coupon tree in the spent fuel pool. Safety Evaluation: The fuel storage and handling system is described in FSAR section 9.1. The FSAR does not provide the level of detail that would encompass this change. The change represented by this modification does not affect any procedures or evaluations contained in the FSAR. The portion of the fuel storage and handling system addressed in this change is not the subject of any Technical Specification.

Subject:

MDD 90-V1M101 \ l

Description:

Existing flow switches 1FSH-11734 and l 1FSH-11776 for the NSCW cross-tie high flow alarm are being replaced with an ITT Barton Model 321 Blind D/P switch. This was necessary due to the calibration problems (switch setpoint was in close proximity to the low range of the existing switch) that had been encountered on the previously installed switch. The new switch has a range that will allow the setpoint to actuate at approximately mid-span. Safety Evaluation: The NSCW system is discussed in FSAR section 9.2. The replacement'of the NSCW cross-tie high flow alarm D/P switch does not change the function of the switch or system operation as described in the FSAR. The alarm setpoint was not altered by this change. This change does impact-any safety related ft ition of the NSCW system. The flow switches and alarm function are not the subject of any Technical Specification. l l 61

r II. 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

MDD / 90-V2M102

Description:

Existing flow switches 2FSH-11734 and 2FSH-11776 for the NSCW cross-tie high flow , alarm have been replaced with an ITT Barton Model 321 Blind D/P switch. This was necessary due to the calibration problems (switch setpoint was in close proximity to the low range of the existing switch) that had been encountered on the previously installed switch. The ne$- switch has a rar.ge that will allow the setpoint to actuate at approximately mid-span. Safet'f Evaluation: The NSCW system is discussed in FSAR section 9.2. The replacement of the NSCW cross-tie high flow alarm D/P switch does not change the function of the switch or system operation as described in the FSAR. The alarm setpoint was not altered by this change. This change does impact any_ safety related function of the NSCW system. The flow switches and alarm function are not the subject of any Technical Specification.

Subject:

MDD # 90-V1H105

Description:

The change revises relay setpoints, adds delay relays and defeats ground fault trips for selected breakers to obtain proper protective relay coordination for the common 4160V swgr. The instantaneous ground fault overcurrent trip for common swgr feeder breakers 1NA01-11, 1NA04-05 and 1NA04-12. In addition, the phase instantaneous overcurrent' trip setpoint was decreased and a time delay installed to add 16 cycles to the instantaneous trip for the overcurrent relays of the common swgr feeder breakers 1NA01-11, 1NA04-05, 1NA04-12, 2NA01-11,'2NA04-05 and 2NA04-12. This results in an increase in the instantaneous overcurrent trip time delay for swgr feeder breakers 1NA01-01 & 03, INA04-01

                       & 03, 2NA01-01 & 03 and 2NA04-01 ^ 03.

62

T. I 1991 AFNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Safety Evaluation: On-site power systems are addressed in FSAR section 8.3. The change initiated by this modification involves the non-class 1E on-site power system only. The changes will enhance the protective relay reliability and coordination for_the common 4160 swgr. The operation and function of the common 4160V

                    -swgr is not affected by this change. The non-class 1E switchgear are not the subject of any Technical Specification.

Subject:

MDD / 90-VCM111

Description:

The mo'dification replaced the antenna system 50 ohm attenuators in control building R-115 and auxiliary building R-D104 and R D105 with quarter wave, unity gain, magnetic mount antennas. The installation of the antennas will improve the radio transmission and reception at the H.P. Control Point Area (Control _ Building., R115) and 480V swgr room (Aux. Bldg., R-D105) and meet the design requirements of the emergency plan and the fire event safe shutdown. Safety Evaluation: The addition of the antennas in the specified areas will improve operation of the radio communications system. This addition will not result in a change to the plant or to plant procedures as described in FSAR sections 9.5, 13.5, 18.1 or the Emergency Plan section F and appendix 8.0. The radiax antenna system'does not impact any safety-related equipment. In plant radio communications is not the subject of any Technical specification.

Subject:

MDD # 90-V1M116

Description:

The setpoints of the CCW pressure safety valves (PSVs) at the RHR heat exchanger and the spent fuel pool cooling heater exchanger have been raised from 135 i 4 psig to 145 5,-0 psig. This will prevent the inadvertent opening of the CCW PSV's upon CCW pumps start. l 63

     --                =                         .-

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Safety Evaluation: A review of FSAR sections 5.4, "RHR"; 9.1

                            " Spent Fuel Pool Cooling and Purification" and 9.2, "CCW" indicates that there is no affect on these systems as described in these sections. Section 9.2 specifically addresses the thermal relief valves affected by this change however actual setpoint specifics are not presented. The-system will continue to operate as before. _The change does not impact any procedures as described in the FSAR.

Technical Specifications do not address the PSV's affected by this change.

Subject:

MDD # 90-V2M117 L

Description:

The setpoints of the CCW pressure safety valves (PSVs)'at the RHR heat exchanger and the spent fuel pool cooling heater exchanger have been raised from 135 i 4 psig to 145 i 5,-0 psig. l This will prevent the inadvertent opening of the CCW PSV's upon CCW pumps start. Safety Evaluation: A review of FSAR sections 5.4, "RHR"; 9.1

                            " Spent Fuel Pool Cooling and Purification" and 9.2, "CCW" indicates that there is no affect on these' systems as described in these sections. Section 9.2 specifically addresses the thermal relief valves affected by this change however actual setpoint specifics are not presented. The system will continue to operate as before. The change does not impact any procedures as described in the FSAR.

Technical Specifications do not address the PSV's affected by this change.

Subject:

MDD / 90-VCM121 i

Description:

Upgrade the alarm station operator keyboards. l (Safeguards) l Safety Evaluation: There is no impact to FSAR, Technical L Specifications or change to procedures in FSAR. l l l 64

1 II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

MDD # 90-V1M122

Description:

Coupling guards have been fabricated and installed on both the NSCW pumps and the NSCW tr3nsfer pumps. Coupling guards were installed by drilling and tapping four holes on each side of the pump stand around u.e coupling openings and attach using 1/4" bolts. Safety Evaluation: The NSCW system is described in FSAR section 9.2. The modification involved drilling and tapping holes in a section of the pumps which does not serve as part of the same pressure boundary. The effect of this change on the structural integrity and operability of the pumps during a seismic event was evaluated in calculation X4 CPS.0075.311. The calculation indicated that the change would have no adverse impact on the availability of the NSCW system during an accident. System operation is not affected by this change. Technical Specifications place requirements on NSCW system availability during various modes of plant operation. As the availability has not been affected by the change, there is no impact on Technical Specifications.

Subject:

MDD / 90-V1W123

Description:

Coupling guards have baen fabricated and installed on both the hsCW pumps and the NSCW transfer pumps. Coupling guards were installed by' drilling and tapping fou.- holes on each side of the pump stand around the- coupling openings and attach using 1/4" bolts. Safety Evaluation: The NSCW system is described in FSAR section 9.2. The modification involved drilling and tapping holes in a section of the pumps which does not serve as part of the same pressure boundary. The effect of this change on the structural integrity and operability of the ! pumps during a seismic event was evaluated in calculation X4 CPS.0075.311. The calculation L indicated that the change would have no l adverse impact on the availability of the NSCW system during an accident. System operation is not affected by this change. l. I 65 l

4 II 1991 ANNUAL REPORT - PART 2 10 CFR50.59 (b) REPORT Technical Specifications place requirements on NSCW system availability during various modes of plant operation. As the availability has not been affected by the change, there is no impact on Technical Specifications.

Subject:

MDD # 90-V1M126

Description:

Wiring configuration for control room annunciator 1ALB01A05, "Servair Swing Compressor Misaligned" was corrected to permit annunciation when the propor conditions are met (Air compressor number 4 selected to Unit 2 and valve 2-2401-U4-510 not 100% open) . This change was accomplished by changing the AX-3 relay in the annunciator circuit from a normally open contact to a normally closed contact. Safety Evaluation: Section 9.3 discusses basic system operation.

                     -Receipt of this alarm initiates an operator response. In this case the response ic not affected as the change has validated the alarm.

Figure 9.3.1 (Sheet 1 of 9) identifies the existence of the misalignment alarm only and does not describe its actuation conditions or operation. The change corrects a condition which prevented proper operation of the alarm. The control room annunciator is not addressed

                     -in any Technical Specification.

Subject:

MDD / FO-V1M128 Description- Cables 1NCAAB04 BUR and 1NCAAB04 BUS coming from the high voltage switchyard to the under frequency panel were only grounded at one end. T.his change provided for the grounding of these cables in the under frequency panel therefore providing these cables with greater shielding protection against spuriously induced signals. This improves the stability and reliability of the under frequency relay. f Safety Evaluation: FSAR sections-8.0, "Elcctrical Power Systems" and 10.2, " Turbine Generator" describe the l on-site and off-site electrical power distribution systems. Grounding requirements for the shielded cables addressed by this change are not included as part of the FS' I description. The change will not affect system 66 L

II 1991 ANNUAL REPORT - PART 2 10 CFRSO.S9(b) REPORT operation or the manner in which the system is operated.- The particulars of this change are not the subject of any Technical Specification.

Subject:

MDD / 90-V1M129

Description:

The modification consist of changing the flow direction through the heater drain tank dump valves (LV-4333 and LV-4334) from the normal

                     " forward" direction to the normal " reverse" direction. This was accomplished by rotating the valve body 180 degrees. The change was implemented to reduce bushing wear and subsequently reduce valve leakage which has been a' continuing problem.

l Safety-Evaluation: The-heater drain tank high level dump valves l are generally discussed in FSAR section 10.4. l The internals of the valve are not included as part of this discussion. Rotation of the valves does not affect system operation or any operating or maintenance procedure associated with the heater drain system operation. The heater drain system is not the subject of any Technical Specification.

Subject:

MDD 90-V2M130

Description:

The modification consist of changing the flow direction: through the heater drain tank dump valves (LV-4333 and LV-4334) from the normal

                     " forward" direction to the normal " reverse" l

direction. This was accomplished by rotating the valve body 180 degrees. The change was implemented to reduce bushing wear and subsequently reduce valve leakage which has been a continuing problem. Safety Evaluation: The heater drain tank high level dump. valves i are generally discussed in FSAR section 10.4. The internals of the valve are not included as part of this discussion. Rotation of the valves does not af fect system operation or any operating or maintenance procedure associated with the heater drain system operation. The heater drain system is not the subject of any Technical Specification. l 67

m < o , f a s ki' II-1991/ ANNUAL REPORT - PART 2. w .10 CFR50.59(b). REPORT

-/,

Subject:

MDD-90-V2M134

Description:

The steam generator top head insul'ation support ring bolted connection has been modified to-

allow for the use of. hex head bolts with nuts ; tack welded to the support ring in lieu of hex l head-bolts tapped into-the support ring and to allow for the.use of carbon steel or stainless steel hex head bolts / nuts with coarse threads in-lieu of stainless steel hex head bolts with fine threads.

              !S afety Evaluation:       FSAR section 5.4.2 discusses =the design and operation _of.the steam generators.--The steam                  J
                                      -generator top 1 head insulation ~is not-specifically. addressed in this section. The change does not-impact any plant procedures,                     i The-steam generator-top head-insulation.is'                      '

not addressed in the Technical Specifications.

Subject:

MDD # 90-VCM135

Description:

The change involved _the: addition of globe

                                      . valves =and flange. connection points to the
                                      = recirculation returnilines of the clean-and dirty: lube: oil storage tanks.(A-1307-T4-001-
                                         & 002)Lto allow for: return-flow,to the
                                      . respective tankffromLthejtemporary vacuum polisher.

Safety Evaluation: The: clean and dirty'lubeDoll storagefsystems

                                      !areJdiscussed'in FSAR'section 10.2.2.
                                                        ^
                                                                                               - The
                                      - addition of the return flow connections does

not= change systemLoperation-as described-in

                                      -this.section however-figure 10.2.2-1' required a revision :to . depict; the. added- valves. --

operating. procedures for these systems were " ' revised to reflect the(current' configuration. l of the systems. The lube oil' systems affected: by this change are not addressed in~the Technical Specifications. Failure of lube oil

                                      . storage can not affect any' safety related component.

68

                                 -g w  -        .              .-.  - +           ,&-i- w A._w       - -

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT

Subject:

MDD # 90-V1M136

Description:

The internals of check valves 1-1305-U4-501, 502, 508 and 509 in the feedwater heater drain system have been removed. The valves are located in the high level drain lines off of each of the moisture separators. Problems had been encountered with the check valves as a result of their cyclic action during operation. This resulted in stem failure. The check valves are not necessary due to the low pressure and low energy conditions in the main condenser which is not conducive to creating a backflow in these lines. Safety Evaluation: Sections 10.2 and 10.4.7 of the FSAR discuss the turbine generator and condensate and feedwater systems. The modified check valves are not specifically addressed in either of the discussions. These check valves do not affect the operation or function of any safety related equipment and are not required to operate in the event of an accident. Removal of the internals does not impact any operating procedure. System operation is not affected by the change. The affected valves are not the subject of any Technical Specification.

Subject:

.            MDD # 90-V2M137

Description:

The elementary diagram and annunciator window ALB20F03 engraving involving the turbine bearing oil header pressure were revised to reflect the proper switch contact configuration and window engraving as specified by the General Electric (Vendor) turbine control diagram (2X4AA11-281). Safety Evaluation: The turbine generator and its support systems are addressed in FSAR section 10.2. The annunciators on the main control boards are discussed in section 18.1 of the FSAR. The implementation of this change does not affect the description presented in these sections. The function of the low bearing oil header 69

n v

           -                                      II
                                  -1991 ANNUAL REPORT      PART 2
                                       -10 CFR50.59(b) REPORT pressure switch remains the same. Operator actions upon receipt of the annunciator are not affected by this change. The pressure switch does not impact any safety related components. The change does not affect the turbine cycle or turbine operation and
                                   -therefore-     does    not    affect    Technical                j Specifications.                                                 1

Subject:

MDD # 90-V1M138

Description:

The pressure switches that were originally used for;1PDS-5200A/B and 1PDS-5201A/B, " Feed Pump _

                                    ' Discharge pressure" are no longer available through the manufacturer, SOR, Inc. SOR
                                                        ~

provides a direct replacement for the discontinued switch. The change permits the installation of the.new model switch as-necessary to replace the existing switches.

             ' Safety Evaluation:    The condensate and feedwater systems are discussed in FSAR-section 10.4. The specifics of the discharge pressure switches are not included in this discussion. The new replacement: switch is-equivalent in design except for the switch housing and functions                   -

in the-same manner.as the previously installed switch. Portions of the condensate and feedwater systems applicable;to this change are not the_ subject of any Technical Specifications.

Subject:

MDD / 90-V2M193

Description:

The pre-modification shutdown logic design on the dieselLgenerators employed a 0.028 flow: restricting orifice to provide enough pressure ~

The high vibration trips on reciprocating air compressors #3 (2-2401-C4-503 & 504) are defeated to eliminate spurious actuations due high vibrations associated with vibration switches. Safety Evaluation: The FSAR states that compressed air systems have automatic protection trips. The high vibration alarms will remain functional to warn operators can then take compensatory actions. 73

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT The annunciator response proceduro will be reviewed but procedurch in FSAR are not affected. This change does not involve any requirements of Technical Specifications.

Subject:

MDD # 91-V2M011

The air regulator pressure for valves 1LV-4331

and 1LV-4332 is being increased from 35 psig to 40 psig. The orifice size in ILY-4331 and ILY-4332 is being increased from 3/32" to 1/8".

Description:

In normal operating conditions, when switching from one ACCW pump to the standby pump, both pumps are running simultaneous for a short period of time. At this time the flow rate reaches 247 gpm, causing . valve HV-2041 to close._ Raising the setpoint for the loop associated with FT-2043, which is the transmitter' controlling HV-2041, to 253 gpm-will eliminate valve isolation while switching

                                   - ACCW pumps.
     - Safety Evaluation            The valves and their function are described in sections 7.6.6.4, 9.2.8, and the active valve table 3.9.B,3-9.                                            Va ves 1/2HV-2041 are designed to close on high -flow or- high pressure, however.the specific transmitter setpoint is not referenced.- Isolation of the ACCW in the event of a thermal barrier leak is not specifically referenced in the discussion of RCS pipe breaks in section 3.6.

g 82

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s.emr--yw - 7*y' er'-M #Tr'-F 7 d* ?-?

                                                                                                                                ,+y,- W- mr-rre v y=9 y r - g g+9 *rw -ve-y

                                        ,- =_..,..._- - - - - - - - -...~ - -...-

I l II 1991 ANNUAL REPORT - PART 2 i 10 CFR50.59(b) REPORT Sections 15.6.2 and 15.6.5 discuss loss of coolant inventory and small treak LOCA's, however no reference to the thermal barrier isolation flow is made. Therefore, this change 4 does not represent a change to the plant as described in the FSAR. Plant operating procedures are not affected by sotpoint change. Technical Specification section 3/4.7.12 refers to the isolation of the ACCW in the event of a thermal barrier leak. The isolation function is designed to prevent.a spill of the reactor coolant from a - p o s t u l a t e d - b r e a c h er" therttal i barrier should a break occur in the i safety related ACCW piping downstream of th' solation , valve. As stated above, the valve r : point is not addressed. The isolation valve is designed to close in the event of-an RCS leak which would result in a flow rate much higher than

                               -the normal ACCW flow. Raising the setpoint to allow for normal flow transients will not require           a       change                          to                 the Technical Specifications..                 Review of the FSAR sections

MDD # 91-V1M094

      -Description:              This change will address several modifications due- to damage discovered. during visual inspection of the expansion bellows assoc.iated with the 24 inch extraction steam piping inside each main. condenser.
                                                                       =

Safety Evaluation: -The modifications being performed under this design change will-not represent a change to the FSAR. This conclusion was-reached based upon review of FSAR sections 1.2.4 (Steam and power conversion), 10.1 (Summary description of steam and power) ,10.2 (Turbine-generator) , E 10.4 (Other features of_ steam and power conversion). D This change will not require a change to L procedures described in the FSAR since the L extraction. steam system is not described to this level.of detail. 4 87 w ;-

               ,. i,-,-;- ,.-,  -.     ,

l l o II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT This change will not require a change to the Technical Specifications and/or environmental protection plan. The replacement expansion bellows has been reviewed by SCS and approved as meeting or exceeding the original design criteria for this joint. Based on review of the Technical Specifications and the environmental protection plan the extraction steam system is not described in sufficient detail to warrant a change. Reference Technical Specification section 3/4.7 (Turbine Cycle).

Subject:

MDD / 91-V1M108

Description:

The addition of sleeves to the MSR Pocket Line Drains (575-1", 576-1") in the areas of heavy pipe wall erosion. This will provide additional wall thickness in the event that erosion rates increase greatly, and to prevent possible pressure houndary failure during this fuel cycle. Safety Evaluation: The facility is not effected as described or implied in the FSAR. Reference a review of FSAR sections 10.2-Turbine / Generator,10.3-Main i Steam Supply System, 10.4.1-Main Condensers. Change does not effect any procedure described or implied in the FSAR. Reference FSAR Chapter 10 and 15 review. Change does not effect any safety related system or component nor does the system failure effect any safety related system or component nor does the system feilure effect any safety related system or component based upon a review of Technical Specification 3/4.7-Plant Systems.

Subject:

MDD / 91-V1M111

Description:

The installation of a stem packing injection pathway into the stuffing box of valve 1LV-6187. This change will provide a connection point for a stem packing injection to the valve on-line: or when the valve is repacked conventionally, a threaded plug to seal the valve stuffing box. 88

       -         _      ~ . . - ..       _ _ . _ - __       ._-_.

II 1991 ANNUAL REPORT - PART 10 CFR50.59(b) REPORT Safety Eva)uation: The facility is not changed as described or implied in the FSAR. Reference a review of section 10.3-Main Steam Supply System and 10.4.1-Main Condensers. Valve will continue to serve its design function. Changa does not effect any procedursi described , or implied in the FSAR. Change does not effect ! the operation of the valve or its purpose to provide water induction protection of the Main Steam Dump Spargers and condensers / turbine. Change does not effect any safety related component or system based upon a review of Technical Specifications 3/4.7-Plant Systems.

Subject:

MDD / 91-V1M112

Description:

The installation of a stem packing injection pathway into the stuffing box of valve 1-HV-6203. This change approves the drilling of an inje tion pathway into the stuffing box and installation of an injection valve or plug as appropriate. Safety Evaluation: This activity does not impact the function of 1HV-6203 nor is tho' valve specifically described in FSAR. For these reasons, this activity does not require a revision-to FSAR. Sections 10.2, " Turbine Generator", and 10.3, "MMin Steam Supply System", reviewed. This activity does not impact procedures described or referenced in FSAR. This activity does not involve or impact any safety related components. This design ! change does not require a revision to l Technical Specifications or the Environmental Plan. i { i 89 l

       .                                                               - = _ . . -

II , 1 l l 1991 AliliUAL REPORT - PART 2 10 CFR 50.59(b) TESTS OR EXPERIl4EliTS

This procedure describes the method for flushing the liquid radwaste microfiltration system installed per DCP 90-VAN 0108. Safety Evaluation: The Liquid Waste Gas System (1901) is non safety related. Its failure does not impact any safety related systems or equipment. There is no need to change the Technical Specification since this is an enhancement over the existing system. Affected portions of FSAR have been reviewed and changes submitted. 1 .

Subject:

T-ENG-91-09

Description:

The proposed procedure tests the Proteus Hardware and Sof tware modification made to Unit 1 and Unit 2. systems under MDD 90-V1M131 and 90-V2H132. The procedure performs system diagnostics for new hardware and performs sof tware tests to prove operability as required by FSAR and Technical Specification. Safety Evaluation: The modification does not require change to FSAR or Technical Specifications. The system function complies with requirements of both the FSAR and Technical Specification. L

Subject:

T-ENG-91-011 Descriptiont- Functional' test of DCP 89-V2NO305. The test consists of verification that the simulated signal for CVI is blocked by the switches. Safety Evaluation: No change to the Technical-Specification is required; however, changes to FSAR to show the blocking switches are~ needed. No procedures in FSAR require a change but plant annunciator response procedure will require a change. The design change does not affect the function of equipment or component assumed to be functional in accident analysis in FSAR.

Subject:

T-ENG-91-012

       )escription:        This test is conducted to collect electrical data on certain equipment during the ESFAS test. Non-intrusive test equipment is used to record data for Train A.

Safety Evaluation: Use of non-intrusive test equipment during the performance of ESFAS procedure will have no impact on either the ESFAS procedure or the plant. This procedure does not alter the intent of-ESFAS procedure. This test is not described in FSAR therefore considered a special test. Review of - Technical Specification reveals no impact from the non-intrusive installation of aquipment. i l l 3

II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Performance of this test does not create malfunction or failure not alcoady evaluated in FSAR.

Subject:

T-ENG-91-013

Description:

This test is conducted to collect electrical data on certain equipment during the ESFAS test. Non-intrusive test equipment is used to record data for Train B. Safety Evaluation Use of non-intrusive test equipment during the performance of ESFAS procedure will have no impact on either the ESFAS procedure or the plant. This procedure does not alter the intent of ESFAS procedure. This test is not described in FSAR therefore considered a special test. Review of Technical Specification reveals no impact from this non-intrusive installation of test equipment. Performance of this test does not create malfunction or failure already not evaluated in FSAR. l 4

7 II 1991 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT T-OPER-91-004 This procedure was intended to identify and stop RCS pressure boundary leakage from pressurizing the Safety injection header through SI cold leg second isolation check valves. This involves running both SI pumps individually on miniflow and running one SI pump to initiate flow through each SI cold leg isolation check ' valve to the SI test header back to the RWST. The leak will be stopped by attempting to rescat the check valves.

1. The test does not affect FSAR analysis and does not create an accident not previously evaluated.

If an SI actuation occurred during the test, the SI test header valves would automatically isolate containment and the additional flow path back to the RWST. In addition, the SI system remains capable of delivering borated water to the RCS throughout the test. Therefore, this test does not decrease the margin of safety defined by Technical Specifications. T-OPER-91-006 This procedure performs the required ECCS flow balance test as specified by Technical Specifications 4.5.2g and 4.5.2h. In addition, Centrifugal Charging Purp operating

                 -characteristics were observed between 470 gpm and 575 gpm, and the suction boost provided to the Centrifugal Charging Pumps during recirculation mode will be determined.      The later two portions of the test are separate from the ECCS flow balance sections of this test and were recommended by Westinghouse.

1. This test does not deviate or create an accident or malfunction different-from FSAR analysis.

During the test, shutdown cooling was maintained by the RHR train not being tested. In addition, adequate procedural steps were included to insure action is taken if CCP abnormal operating , characteristics are observed. Also, steps were included to prevent overpressurization of the ECCS ring header during the suction boost section. 5 j

I l II 1991 ANNUAL REPORT - PART 2 10 CFR50. 59 (b) REPORT

2. Technical Specification 4.5.2g and 4.5.2h were I met as required and therefore do not deviate from j

                    -the margin of safety. The CCP and suction boost portions of the test are performed when in mode 6, with the Reactor head off, when ECCS systems are not required for ECCS injection mode. With these criteria met and one boration flowpath available, the limitations of Technical Specifications are satisfied, i

I i L 6 l-

A .;_-ah._.%2_ m44 - a. 4 .M =b.-,E A.-,e-- - - --un. s ._,-a ,A.a-M.--ea,J Cm 44%,. .eaA &. ---.. d--*r-* 5 l t 111 GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET N05.- 50 424 AND 50-425 FACILITY OPERATING LICENSE N05. NPF-68 AND NPF 81 EMERGENCY CORE COOLING SYSTEMS OUTAGE DATA REPORT I o i

                     --w                -             ,r -- -.- , , -
                                                                .          .p

III 1991 ANNUAL REPORT - PART 2 ECCS OUTAGE DATA REPORT UNIT 1 (CONTINUED)

6. a) 3-5-91 8 hours and 40 minutes b) & c) Train B RHR system removed from service for preventative maintenance on bypass valve 1-FV-619B.

d) Maintenance completed, surveillances performed ) and system restored to service.

7. a) 3-12-91 35 hours and 41 minutes b) & c) Train A CCP system removed from service for 1 preventative maintenance and MOVATs. I d) Maintenance completed, surveillances performed l and system restored to service. !

8. a) 3-21-91 28 minutes b) & c) Valve 1-HV-8485A closed for surveillance '

14608-1 rendering train A CCP inoperable. i d) Surveillance performed and system restored to ' service.

9. a) 3-21-91 47 minutes b) & c) Valve 1-HV-8485A closed for surveillance 14608-1 rendering train A CCP inoperable.

d) Surveillance performed and system restored to service.

10. a) 3-22-91 7 hours 51 minutes b) & c) Train A RHR system removed from service for Channel calibration and AOV checkout for RHR Heat Exchanger outlet valve HV-606.

d) Surveillances performed and system restored to service.

11. a) 4-2-91 .35 hours and 11 minutes b) &'c) Train B SI ' pump removed from service for scheduled maintenance.

d) Maintenance completed, surveillances performed and system restored to service.

12. a) 4-7-91 8 minutes b) & c) Train A RHR pump was placed in pull-to-lock for valve stroking of heat exchanger inlet valve.

d) Surveillance performed and system restored to service. 1 a 2 L

III 1991 ANNUAL REPORT - PART 2 ECCS OUTAGE DATA REPORT UNIT 1 (CONTINUED)

13. a) 4-14-91 9 minutes b) & c) Train A RHR pump was placed in pull-to-lock for performance of surveillance 14825-1.

d) Surveillance performed and system restored to service.

14. a) 5-1-91 5 hours and 46 minutes b) & c) Train B RHR pump was removed from service for preventative maintenance, loop calibration, on heat exchanger outlet HV-607 procedure 24369-1.

d) Maintenance completed surveillance performed and system restored to service.

15. a) 5-24-91 8 minutes b) & c) Train B RHR system was removed from service for performance of surveillance 14825-1.

d) Surveillance performed and system restored to service.

16. a) 6-4-91 1 hour and 33 minutes b) & c) Train A CCP for performance of surveillance 14808-1.

d) Surveillance performed and system restored to service.

17. a) 6-5-91 5 hourc and E minutes

     - b) & c)  Train A SI, CCP and RHR systems removed from service for maintenance on the A Train NSCW system.

d) Maintenance completed and system restored to service.

18. a) 6-13-91 2 minutes b) & c) Train A CCP pump discharge closed for functional test of alarm circuit.

d) Functional test performed and system restored to service.

19. a) 6-13-91 1 hour and 41 minutds b) & c) Train A CCP for performance of surveillance 14608-1.

d) Surveillance performed and system restored to service. ! 3 L t i

III 1991 ANNUAL REPORT - PART 2 ECCS OUTAGE DATA REPORT UNIT 1 (CONTINUED)

23. a) 6-28-91 18 minutes b) & c) Train A SI pump was placed in pull-to-lock for performance of surveillance T-OPER-91-003.

d) Surveillance performed and system restored t; service.

21. a) 7-7-91 18 minutes b) & c) Train A RHR pump was placed in pull-to-lock for performance of surveillance 14825-1.

d) Surveillance performed and system restored to i service. 1

22. a) 8-2-91 7 minutes )

b) & c) Train A SI pump was placed in pull-to-lock for l performance of surveillance T-OPER-91-003. l d) Surveillance performed and system restored to service.

23. a) 8-3-91 2 minutes b) & c) Train B CCP system was removed from service due to closing 1-HV-8438 for performance of survaillance 14825-1.

d) S;rveillance performed and system restored to service.

24. a) 8-8-91 9 minutes b) & c) Train A SI pump was placed in pull-to-lock for performance of surveillance T-OPER-91-003.

d) Surveillance performed and system restored to service.

25. a) 8-15-91 9 minutes b) & c) Train A SI pump was placed in pull-to-lock for performance of surveillance T-OPER-91-001.

d) Surveillance performed and system restored to service.

26. a) 8-16-91 37 minutes b) & c) Train B CCP system was removed from service due to closing pump discharge valve 1-HV-8485 for performance of surveillance 14609-1.

d) Surveillance performed and system restored to servi.e. 4

III 1991 ANNUAL REPORT - PART 2 ECCS OUTAGE DATA REPORT UNIT 1 (CONTINUED)

27. a) 8-22-91 3 minutes b) & c) Train A SI pump was placed in pull-to-lock for performance of. surveillance T-OPER-91-003.

d) Surveillance performed and system restored to service.

28. a) 8-27-91 23 minutes b) & c) Train A CCP pump was removed from service due to closing discharge valve 1-HV-8485A for o -performance of surveillance 14808-1.

d) Surveillance performed and system restored to service.

29. a) 9-6-91 19 minutes b) & c) Train A CCP pump was removed from service due to closing discharge valve 1-HV-8485A for performance of surveif. lance 14608-1.

d) Surveillance performed and system restored to service.

30. a) 11-16-91 38 minutes b) & c) Train 1A SI rendered inoperable due to clocing 1-HV-8821A while filling accumulators.

d) System re-tored to service.

31. a) 11-29-93 42 minutes b) & c) Train B CCD pump was removed from service due to closing oischa; ge valve 1-HV-8485B for

_ performance of snrveillance 14609-1. d) Surveillance perfcemed and system restored to service.

32. a) 11-30-91 45 minutes bl & : c) Train B SI pump war placed in pull-to-lock for performance of surveillance 14619-1.

d) Surveillance performed and system restored to service. 33, a) 12-l-91 2 hours and 19 minutes b) & c)- Train A CCP pump was removed from service due to closing discharge valve 1-HV-8485A for performance of curveillance 14808-1. d) Surveillance performed and system restored to service. 5

A

III-1991' ANNUAL ~ REPORT'- PART 2 ECCS-OUTAGE DATA REPORT UNIT 1 (CONTINUED)

34. a) 1-91_ 2 hourc and 19 minutes b) & c) Train A CCP pump was removed from service due to closing discharge valve .1-HV-8485A for.

performance of: surveillance 14808-1. d) -Surveillance performed and system restored to service.

35. = a)' .

12-22-91 10'w.inuto-b ,' '& c) Train A ks.." system was removed from cervice _ s for surveillance 14825-1.- d) Surveillanca performed and system restored to service, a s 6

III 1991 ANNUAL REPORT - PART2 ECCS' OUTAGE DATA REPORT

                                - UNIT 2 -

Unit 2 Emergency Core Cooling System components were out of service a total of 147 hours and 57 minutes in 1991.

1. a) 1-1-91 5 hours and 1 minute b) & c) Train B SI, CCP and RHR system.3 removed from service for me '

tenance on the B Train NSCW system .

d) Maintenance completed and system restored to service.

2. a) 1-4-91. 1 hours and 16 minutes b) & c) Train .B RHR system removed from service for l preventative maintenance on train B RHR I

miniflow valve, d) Maintenance completed, surveillances performed and system restored to service. 1

3. a) 1-20-91 13 minutes b) & c) Valve stroking for surveillance 14825-2 rendered train A RHR inoperable, d) Surveillance completed and system restored to service.

1 L

4. a) 2-1-91 41 minutes b) & c) Train A RHR system was removed from service for surveillance- 14805-2.

d) Surveillance performed and system restored to service. l 5. a) 2-1-91 29 minutes l b) & c) Train B RHR system was removed from service for surveillance 14805-2. , d) Surveillance performed and system restored to service.

6. a) 3-1-91 6 minutes b) & c) Train B RER pump was placed in pull-to-lock for performance of surveillance 14825-2.

d) Surveillance performed and system restored to service. l l 7

III 1991 ANNUAL REPORT - PART2 ECCS OUTAGE DATA REPORT UNIT 2 (CONTINUED)

7. a) 3-13-91 18 minutes b) & c) Train A CCP system was removed from service due to closing pump discharge valve 2-HV-8485A for performance of surveillance 14808-2.

d) Surveillance performed and system restored to service.

8. a) 3-19-91 65 hours and 7 minutes b) & c) Train A CCP system removed from service for scheduled maintenance.

d) Maintenance completed, surveillances performed and system restcred to service.

9. a) 3-22-91 1 hour and 28 cainutes b) & c) Train A SI rendered inoperable due to closing 2-HV-8821A while filling accumulators.

d) System restored to service.

10. a) 3-23-91 22 minutes b) & c) Train B RHP system was rendered inoperable due to valve realignment for surveillance 14450-2.

d) Surveillance performed and system restored to service.

11. a) 3-23-91 22 minutes b) & c) Train A RHR system was rendered inoperable due to valve realignment for surveillance 14450-2.

d) Surveillance performed and system restored to service.

12. c) 4-14-91 9 minutes b) & c) Train A RHR pump was placed in pull-to-lock for performance of surveillance 14825-2.

d) Surveillance performed and system restored to service.

13. a) 4-23-91 3 hours and 54 minutes b) & c) Train A RHR pump was declared inoperable due to low flow per 2-FI-618 with miniflow valve 2-FV-610 open. Miniflow was found to be isolated.

d) Proper valve alignment was restored, pump surveillances performed and the system restored to service. 8

III 1991 ANNUAL REPORT - PART2 I ECCS OUTAGE DATA REPORT 1 UNIT 2 (CONTINUED)

14. a) 5-12-91 51 minutes b) & c) Train A SI and CCP systems removed from service for surveillance on 14825.

d) Surveillances performed and system restored to service.

15. a) 5-24-91 36 minutes b) & c) Train 3 RHR system was removed from tservice for performance of surveillance 14825-2.

d) Surveillance performed and system restored to service. l L 16. a) 5-30-91 1 hour and 36 minutes b) & c) Train B CCP system was removed from service due to closing pump discharge valve 2-HV-8485B for performance of surveillance 14808-2. d) Surveillance performed and system restored to service.

17. a) 6-6-91 1 hour and 24 minutes b) & c) Train A CCP pump was-removed from service due to closing 2-HV-8485A for performance of surveillance 14808-2.

d) Surveillance performed and system restored to service.

18. a) 6-25-91 19 hours and 3 minutes b) & c) Train B SI system removed from service for preventative aaintenance on SI B pump motor, d) Maintenance completed, surveillances performed and system restored to service.

19. a) 7-16-91 2 minutes b) & c) Train B CCP pump removed from service due to discharge valve being closed to troubleshoot Group 1 monitoring light alarm circuit, d) Maintenance completed, surveillances performed and system restored to service.

2 0.- a) 7-17-91 5 minutes b) & c) Train B CCP ystem removed from service due to discharge valve being clo~ed to troubleshoot Group 1 monitoring light alarm circuit. d) Maintenance completed, surveillances performed and system restored to serv'ce. 9

l III 1991-ANNUAL REPORT - PART2 ECCS OUTAGE DATA REPORT UNIT 2 (CONTINUED)

21. a) 7-23-91 41 hours and 48 minutes b) & c) Train B CCP system tripped on overload during 14604-2 also subsequently tagged out to perform maintenance, d) Power restored maintenance completed, surveillances perforned and system restored to service.

22. a) 8-2-91 32 minutes b) & c) Train A SI pump placed in pull-to-lock for performance of surveillance T-OPER-91-003.

d) Surveillance performed and system restored to service.

23. a) 8-8-91 15 minutes b) & c) Train A SI pump placed in pull-to-lock for performance of surveillance T-OPER-91-003.

d) Surveillance performed and system restored to service.

24. a) 8-10-91 5 hours and 40 minutes b) & c) Train A SI, CCP and RHR systems removed from service for maintenance on the A Train NSCW system.

d) Maintenance completed and system restored to service.

25. a) 8-15-91 7 minutes b) & c) Train A SI pump placed in pull-to-lock for performance of surveillance T-OPER-91-003.

d) Surveillance performed and system restored to service.

26. a) 8-16-91 39 minutes b) & c) Train B RHR pump was removed from service r perforaance of surveillance 14825-1.

d) Surveillance pet-formed and system restored to service.

27. a) 8-22-91 1 hour and 11 minutes b) & c) Train A SI pump was placed in pull-to-lock for performance of curveillance T-OPER-91-003.

d) Surveillance performed and system restored to service. 10 1

III 1991 ANNUAL REPORT - PART2 ECCS OUTAGE DATA REPORT UNIT 2 (CONTINUED)

28. a) 9-4-91 23 minutes b) & c) Train A CCP system removed from service due to discharge valve HV-8485B being closed for OSP 14608-2.

d) Surveillances performed and cystem restored to service.

29. a) 9-17-91 1 hour and 4 minutes b) & c) Train B CCP system removed from service due to discharge valve HV-8485B being closed for OSP 14808-2.

d) Surveillances performed and system restored to service.

30. a) 9-29-91 5 minutes b) & c) TrMn A RHR pump was placed in pull-to-lock for performance of surveillance 14825-2.

d) Surveillance performed and system restored to service.

31. a) 11-8-91 19 minutes b) & c) Train B RHR system was rendered inoperable due I

to valve stroking for surveillance 14825-2. d) Surveillance performed and system restored to service. 32, a) 12-6-91 45 minutes b) & c) Train A RHR system was rendered inoperable due to valve realignment for surveillance 14805-2. d) Surveillance performed and system restored to service.

33. a) 12-6-93 40 minutes b) & c) Train 9 RHR system was rendered inoperable due to valve realignment for surveillance 14805-2.

d) Surveillacce. performed and system restored to service.

34. a) 12-22-91 1 hour and 2 minutes b) & c) Train A RHR pump was placed in pull-to-lock I for portormanco of surveillance 14825-2.

d) Surveillance performed.and system restored to service. 11 l l l

III 1991 ANNUAL REPORT - PART2 ECCS OUTAGE DATA REPORT UNIT 2 (CONTINUED)

35. a) 12-22-91 1 hour and 2 minutes b) & c) Train A RHR pump was removed from service due to heat exchanger outlet isolation valve preventativo maintenance.

d) Maintenance performed and system restored to service. 36, a) 13-31-91 1 hour and 22 minutes b) & c) Train B RHR pump was removed from service due to realigning valve 2-HV-8716A for surveillance 14805-2. d) Surveillance performed and system restored to service. 12 1

1 IV GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY OPEPATING LICENSE NOS. NPF-68 AND NPF-81 l ANNUAL. RADIOLOGICAL ENVIRONMENTAL SVRVEILLANCE REPORT CALENDAR YEAR 1991 i-i l I l I 1 l

V0GTLE ELECTRIC GENERATING PLANT

           -RADIOLOGICAL ENVIRONMENTAL SURVE!LLANCE REPORT TABLE OF CONTENTS S10110E TITLE                                                            PAGE

1.0 INTRODUCTION

, . . . . . . . . . . . . . . . . . . . . .         1-1 2.0

SUMMARY

DESCRIPTION . . . . . . . . . . . . . . . . . . 2-1 3.0 RESULTS

SUMMARY

       . . . . . . . . . . . . . . . . . . . . 3-1 4.0   DISCUSSION OF RESULTS . . . . . . . . . . . . . . . . .          4-1 4.1       Airborne . . . . . . . . . . . . . . . . . . . . . .       4-4 4.2       Direct Radiation . . . . . . . . . . . . . . . . . .       4-6 4.3       Milk _....,...................4-8 4.4       Vegetation    . . . . . . . . . . . . . . . . . . . . . 4-9 4.5       River Water . . . . . . . . . . . . . . . . . . . . .      4-11 4.6       Drinking Water       ...,...............                   4-13 4.7       Fish . . . . . .   . . . . . . . . . . . . . . . . . . 4-17 4.8       Sediment . . . . . . . . . . . . . . . . . . . . . .       4-19 5.0   INTERLABORATORY COMPARIS0N PROGRAM . . . . . . . . . . .         5-1

6.0 CONCLUSION

S .................... . 6-1 i l o l l'

LIST OF TABLES TABLF IIILE PEE 2-1

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 2-2 2-2 RADIOLOGlCAL ENVIRONMENTAL SAMPLING LOCATIONS 2-7 3-1 RADIOLOGICAL ENVIRONMENTAL MCNITORING PROGRAM ANNUAL

SUMMARY

3 4-1 LAND USE CENSUS RESULTS 4-3 _ S-1 CROSSCHECK PROGRAM RESULTS S-3 / 4 ii

LIST OF FIGURES FIGVRE. JJILE Pf_G1 2-1 TERRESTRIAL STATIONS NEAR SITE BOUNDARY 2-10 2-2 TERRESTRIAL' STATIONS BEYOND SITE BOUNDARY OUT-TO APPR0XIMATELY SIX MILES AND RIVER STATIONS 2-11 2-3 TERRESTRIAL STATIONS BEYOND SIX HILES 2-12

2-4 DRINKING WATER STATIONS 2-13 iii t . '.

b ACRONYMS CL Confidence Level El Environmental Laboratory (Georgia Power Company) EPA Eny'.conmental Protection Agency GPC Georgia Power Company b HNP Edwin I. Hatch Nuclear Plant _ LLD Lower Limit of Detection MDA Minimum Detectable Activity MDD Minimum Detectable Difference NA Not Applicable NDM No Detectable Measurement (s) NRC- Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual REMP Radiological Environmental Monitoring Program RL Reporting Level ~ TLD Thermoluminescent Dosimeter i TS Technical Specifications 4' , a L iv

V0GTLE ELECTRIC GENERATING PLANT RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT

1.0 INTRODUCTION

This is the fifth Annual Radiological Environmental Surveillance Report for the Alvin W. Vogtle Electric Generating Plant (VEGP). It covers activities of the Radiological Environmental Monitoring Program (REMP) during calendar year (CY) 1991. Hence all dates in this report are for 1991 unless otherwise indicated. The specifications for the REMP are provided by Section 3/4.12 of the Technical Specifications (TS). The objectives of the REMP are to ascertain the levels of radiation and the uncentrations of radioactivity in the VEGP environs and to assess any radiological impact upon the environment due to plant operations. A comparison between the results obtained during the preoperational and operational phases provides some basis for such an assessment. A comparisc*. between the results obtained at control stations (locations where radiological levels are not expected to be significantly affected by plant operations) and at indicator stations (locations where it is anticipated that radiological levels are more likely to be affected by plant operations) provides a further basis for this assessment. The preoperational stage of the REMP started in August of 1981 when the initial collections of the radiological environmental samples were made; there was a phase-in period of a few years before the preoperational program was fully implemented. The transition from the preoperational stage to the operational stage hinged about initial criticality for Unit ! I which occurred on March 9, 1987. A summary description of the REMP is provided in Section 2. This includes maps showing the sampling locations; the maps are keyed to a table indicating the oistance and direction of each sampling location from a point midway between the two reactors. An annual summary of the laboratory analysis results obtained from the main samples utilized for environmental monitoring is presented in Section 3. A discussion of the results including assessments of any radiological impacts upon the environment is provided in Section 4. The results of the Interlaboratory Comparison Program are presented in Section 5. The chief conclusions are stated in Section 6. 1-1 l l

2.0

SUMMARY

DESCRIPTION A summary description of the REMP is provided in Table 2-1. This table portrays the program in the manner by which it is being regularly carried out; it is essentially-a copy of Table 3.12-1 of the TS which delineates the program's requirements. Sampling locations specified by Table 2-1 are described in Table 2-2 and are shown on maps in Figures 2-1 tFrough 2-4. This description of the sample locations closely follows that found in the table and figures of Section 3.0 of the Offsite Dose Calculation Manual (00CM). It is stated in Footnote (1) of Table 3.12-1 of the TS that deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances, such as, hazardous conditions, seasonal unavailability, and malfunction of sampling equipment. Any deviations are accounted for in the discussions for each par Licular sample type in Section 4. For CY-91, all the laboratory analyses except for the reading of the thermoluminescent dosimeters (TLDs) were performed by Georgia Power Company's (GPC's) Environmental Laboratory (EL) in Smyrna, Georgia. The reading of the TLC: was provided by Teledyne Isotopes Midwest Laboratory in Northbrook, Illinois. 2-1 I

A-TABLE 2-1l(SHEETI0F5)- SUMARY DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRM SAMPLING AND TYPE AND FREQUENCY EXPOSURE PATHWAY lht.. .R OF REPRESENTATIVE. SAMPLES OF~ ANALYSIS

                                          'AND SAMPLE LOCATIONS          COLLECTION FREQUENCY AND/OR SAMPLE Quarterly-           Gamma dose quarterlyL

1. Direct Radiation Thirty-nine routine monitoring

                                -stations with two'or more dosimeters placed as follows:

An inner ring of stations, one in - each meteorological. sector _ in the

                                -general area of the site boundary; An outer ring of stat.uns, one in.

each meteorological sector in-the 6 mile range from the site; n3 g, and Special interest areas such as population centers, nearby residences, schools and control stations. 1 e e s l f s _ _ _ _ _ _ _ . _ _ _

{

                                                                                                                                                                ~
                                                                                                                                -~

TABLE 2-1 (SHEET 2 0F'5)L SUNIARY DESCRIPTION OF: RADIOLOGICAL. ENVIROMIENTAL MONITORING PROGRAM I 1 NUNBER OF REPRESENTATIVE. SAMPLES SAMPLING AND'- : TYPE-Afm FREQUENCY' EXPOSURE PATHWAY AND SAMPLE' LOCATIONS COLLECTION FREQUENCY- - '0F ANALYSIS L AND/OR SAMPLE

2. Airborne Samples from seven locations.- . Continuous sampler oper- Radiolodine Cannister:.

Radiolodine and- ation with sample collec . ' l-131 analysis weekly-

                        - Particulates-                                                                  tion weekly. or more frequently if required-by.

Five locations close to the - Particulate' Sampler: : site.: boundary in different dust loading sectors;; Gross beta analysis (1) s following iilter change and '

'                                                                                                                                            ' gasma' isotopic analysis (2)-
  -m                                                                                                                                           of composite :(by location) 4,                                                                                                                                          quarterly

~ A community having the' highest l calculated annual average ground-level D/Q; and i A control'1ocation in the vicinity of a populztton center; at a distance of about 15 miles. 9 9 4 If e

                                               ,                             g       - - - --

w --r .. g gg , -, . . 4

TABLE 2-1 (SHEET 3 0F 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIR0t' MENTAL HONITORING TYPE AND FREQUENC. SAMPLING AND NUMBER OF REPRESENTATIVE SAMPLES COLLECTION FREQUENCY OF ANALYSIS EXPOSURE PATilWAY AND SAMPLE LOCATIONS l AND/OR SAMPi.E

3. Waterborne Gama 1setopic analysis (2)

One samp12 upriver Composite sample)over 1-month period (4 monthly. Composite for '

a. Surface (3) tritium analysis quarterly Two samples downriver I-131 analysis on each Compcsite sample of sample when the dose

b. Drinking Two san,ples at each of the two river water near the nearest water treatment plants calculated for the that could be affected by plant intake at each water consumption of the water is treatment plant over discharges 2-week period (4) when greater than 1 mrem per i-131 analysis is year (5). Composite for Two samples at a control required to be performed gross beta and gama m location on each sample, monthly isotopic analyses (2) on raw a water' monthly. Gross beta, composite otherwise; and gamma isotopic and 1-131 grab sample of finished analyses on grab sample of water at each water finished water monthly.

treatment plant every 2 veeks or monthly, as Composite for tritium analysis on raw and finished appropriate water quarterly Gawa isotopic analysis (2) Semiannually One sample from downriver semiannually

c. Sediment from area with existing or potential Shoreline recreational value l

One sample from upriver area with existing or potential recreational value v l[ ! j hf

j- ' TABLE 2-1 (SilEET 1 Of 5) SUPMARY DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING Fibn4 SAMPLING AND TYPE AND FREQUENCY EXPOSURE PATliWAY NUMBER OF REPRESENTATIVE SAMPLES OF ANALYSIS i COLLECTION FREQUENCY ANJ SAMPLE LOCATIONS l AND/OR SAMPLE

4. Ingestion Biweekly Gamma

a. Milk Two samples from milking animals (6) analysisisotop)ic (2 7 biweekly at control locations at a distance of about 10 miles or more Semiannually Gama isotopic analysis (2)

At least one sample of any comer- on edible portions

b. Fish cially or recreat:enally important semiannually species in vicini:y of plant discharge area

'? At least one sample of any comercially or recreationally important species in an area not influenced by plant discharge During spring spawning Gama isotopic analysis (2) At least one sample of any on edible portions annually anadromous species in vicinity of season plant discharge Gama isotopic

c. Grass or Leafy One sample from two onsite locations Monthly during growing analysis (2,7) monthly near the site bouviary in different season Vegetation sectors

                 '                One sample from a control location at about 18 miles distance n

TABLE 2 1 (SHEET 5 0F 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTtt MONITORING PROGRAM TABLE NOTATIONS (1) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and thoron-daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. (2) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the. facility. (3) The upriver sample is taken at a distance beyond significant influence-of the discharge. The dcwnriver samples are be taken in areas beyond and near-the mixing zone. (4) Composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to_ assure obtaining a representative sample. (5) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM. (6) A milking animal is a cow or goat producing milk for human consumption. (7) If gamma isotopic analysis is not sensitive enough to meet the Lower Limit of Detection (LLD) for I-131, a separate analysis for I-131 will be performed. 2-6

TABLE 2-2 (SHEET 1 0F 3) RAJI 0 LOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station De, riptive Direction (2) Distance (2) Sample Number Tvoe (1) location (miles) Tvoe (3) 1 I Hancock Landing Road N 1.1 D 2 I River Bank NNE 0.8 0 3 1 Discharge Area NE 0.6 A 3 I River Bank NE 0.7 0 4 I River Bank ENE 0.8 0 5 I Rivc* Bank E 1.0 D 6 I Plant Wilson ESE 1.1 D 7 I Simulator Building SE 1.7 0,V,A , 8 I River Road SSE 1.2 0 9 i River Road S 1.1 D 10 I Met Tower SSW 0.8 A 10 1 River Road SSW l.1 0 11 1 River Road SW l.2 0 12 I River Road WSW l.2 D,A 13 I River Road W l.3 D 14 I River Road WNW l.8 D 15 I Hancock Landing Road NW 1.5 0,Y 16 I Hancock Landing Road NNW l.4 D,A 17 0 Savannah River Site (SRS) River Road N 5.5 0 18 0 SRS D Area NNE 5.1 D 19 0 SRS Road A.13 NE 4.7 0 20 0 SRS Road A.13.1 ENE 4.8 D 21 0 SRS Road A.17 E 5.8 D 22 0 River Bank Downstream of Buxton Landing ESE 5.2 0 23 0 Rive Road SE 4.6 D _ 24 0 Chance Road SSE 4.9 D 25 0 Chance Road near Highway 23 S 5.2 0 26 0 Highway ??- Mile 15.5 SSW 4.6 D 27 0 Highway 23, Mile 17 SW 4.8 0 28 0 Claybon Road WSW 5.0 0 29 0 Claxton-Lively Road W 5.0 D 30 0 Nathaniel doward Road WNW 5.0 0 31 0 River Road at Allen's Church Fork NW 5.0 0 32 0 River Bank KNW 4.8 D 33 0 Hunting Cabin SE 3.3 0 2-7

TABLE 2-2 (SHEET 2 0F 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station Descriptive Direction (2) Distance (2) Sample Number Tvoe (1) __Lqation (miles) Tvoe (3) 35 0 Girard SSE 6.6 D,A 36 C Waynesboro WSW 14.9 D,A 37 C Substation (Waynesboro) WSW 17.5 D,V 43 0 Employees Recreation Area SW 2.2 0 47 C Oak Grove Church SE 10.4 0 48 C McBean Cent'ery NW 10.3 0 80 C Augusta Water Treatment Plant NNW 27.5 W(4) 81 C Savannah River N 2.5 F(5),S(6) 82 C Savannah River (RM 151.2) NNE 0.8 R 83 I Savannah River (RM 150.4) ENE 0.8 R.S(6) 84 0 Savannah River (RM 149.5) ESE 1.6 R 85 I Savannah River ESE 4.3 F(5) 87 I Beaufort-Jasper County , Water Treatment Plant; Beaufort, SC SE 76 W(7) 88 I Cherokee Hill Water Treatment Plant; Port Wentworth, GA SSE 72 W(8) 98 C W. C. Dixon Dairy SE 9.8 M i 99 C Boyceland Dairy WNW 24.5 M TABLE NOTATION: (1) Station Types C - Control 1 - Indicator 0 - Other (2) Direction and distance are reckoned from a point midway between the two reactors (3) Sample Types A - Airborne Radioactivity D - Direct Radiation F - Fish M - Milk R - River Water S - River Shoreline Sediment W - Drinking Water V - Vegetation 2-8

1 TABLE 2-2 (SHEET 3 0F 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS TABLE NOTATIONS (Continued) (4) The intake for the Augusta Water Treatment Plant is located on the Augusta Canal, The entrance to this canal is at River Mi: (RM) 207 on the Savannah River. The canal effectively parallels the river. The intake to the pumping station is 3.6 miles down the canal and only a tenth of a mile across a narrow neck of land to the river. (5) About a five mile stretch of the river is generally needed to obtain adequate fish samples. Samples are normally gathered between RM 153 and 158 for upriver collections and between RM 144 and 149.4 for downriver collections. (6) Sediment is collected at locations with existing or potential recreational value. Because high water, shifting of the river bottom, or other reasons could cause a suitable location for sediment collection to becomc unavailable or unsuitable, a stretch of the river between RM 148.5 and 150.5 is designated for downriver collections while a stretch between RM 153 and 154 is designated for upriver collections, In practice, collections are normally made at RM 150.2 for downriver collections and at RM 153.3 for upriver collections. (7) The intake for the Beaufort-Jasper County Water Treatment Plae is located at the end of a canal which begins at RM 39.2 on the Savannah River. This intake is about 16 miles by line of sight down the canal from its beginning on the Savannah River. (8) The intake for the Cherokee Hill Water Treatment Plant is located on Abercorn Creek which is about one and a quarter creek miles from its mouth on the Savannah River at RM 29. 2-9 1

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3.0 RESULTS

SUMMARY

In accordance with Section 6.8.1.3 of the TS, summarized and tabulated results of all of the regular radiological environmental samples and radiation measurements taken during the year at the designatcd indi ator and control stations are presented in Table 3-1 in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. Results for samples collected at locations other than indicator or co6 trol stations or in addition to those stipulated by ' Table 2-1 are included in Section 4, the discussion of results section, for the type sample. Naturally occurring radionuclides which are not included in the plant's effluent releases are not required to be reported. Naturally occurring Be-7 is produced in the reactors; miniscule quantities are found in the _ liquid releases. No other naturally occurring radionuclides are known to be included in the plant's effluent releases. Hence, the radionuclides of interest for the radiological environmental samples monitoring liquid releases (river water, drinking water, fish, and sediment) are manmade radionuclides plus Be-7, while only manmade radionuclides are of interest for the other radiological environmental samples. n 3-1 2!

TABLE 3-1 (SHEET I 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia, c'lendar Year 1991 Location with Highest Control Locations Number of Medium or Type and Lower Limit All Indicator Reportable Pathway Sampled Total Nurt'r of Locations Annual Mean Mean (b) l Name Mean (S) Range Gccurrences (Unit of of Analy;e, Detection (a) Mean (b) Performed (LLD) Range Distance & Range (Fraction) Measurement) (Fraction) Direction (Fraction) l No. 7 19.8 19.2 0 Airborne Gross Beta 10 19.3 5-47 Simeslator 10-47 8-46 i Particulates 310 1.7 miles (52/52) l 3 (258/258) (51/51) (fCi/m ) SE T' Gamma Isotopic 28 NDM 0 50 NDM (c) NDM Cs-134 NDM NDM 0 Cs-137 60 NDM NDM NDM 0 Airborne I-131 70 NDM Radioipdine 310

         -(fCi/m )

No. 3 22.0 11.1 0 f Direct Gamma Dose NA (d) 16.9 13-26 River Bank 18-26 11-23 Radiation 80 (mR/91 days) (64/64) 0.7 miles (4/4) (16/16) NE

TABLE 3-1 (SHEET 2 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL St# MARY - Vogtle Electric Generating Plant, . Docket Nos. 50-424 and 50-425 Burke County, Georgia, Calendar Year 1991 j Medium or ' Type and Lower Limit All Indicator Location with Highest Control Locations Number of Pathway Sampled Total Number of . Locations. Annual Mean Mean (b) Nonroutine ; (Unit of of Analyses Detection (a) Mean (b) Name Mean (b) Range Reported i Measurement) Performed (LLD) Range Distance & Range (Fraction) Measurements (Fraction) Direction (Fraction) ,

Milk Gama Isotopic (pCi/1) 54 Cs-134 15 NA NDM NDM 0 Cs-137 18 NA No. 98 14.1 14. 0 t Dixons 14-14 14-14

y. 9.P <siles (1/27) (1/54) w SF Ba-140 60 NA NDM NDM 0

                                                        .La-140                       15             NA                               NDM         NDM                 0 i-131                     1              NA                               NDM         NDM                 0 54 Grass              Gama ' Isotopic (pCi/kg wet)      36 I-131                     60             NDM                              NDM         NDM                 0 Cs-134                     60             NDM                              NDM         NOM                 0

TABLE 3-1 (SNEET 3 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING' PROGRAM ANNUAL SUPMARY i Vogtle' Electric Generating Plant, Docket Nos. 50-424 and 50-425

Burke County, Georgia, Calendar Year 1991 ,

Mediuta or Type and. Lower Limit All Indicator Location with Highest Control Locations Number of , Pathway Sampled Total Number of Locations Annual Mcan Mean (b) Nonroutine (Unit of'. of Analyses Detection (a)- Mean (b) Name Mean (b). Range Reported Measurement) Performed (LLD) Range Distance & Range (Fraction) Measurements < (Fraction) Direction .(Fraction) Cs-137 80 35.3. No. 37 62.4 62.4 0 18-58 Substation '62-62 62-E2 (5/24) 17.5 miles (1/12) (1/12) WSW River Water Gamma Isotopic (pCi/1) 24 y Be-7 80 (e) ND*1 NDM NDM 0 i Mn-54 15 NDM NDM NDM 0 Fe-59 30 NDM NDM NDM 0 Co-58 15 NDM NDH NDM 0 i Co-60 15 NDM NDM NDM 0 l Zn-65 30 NDM NDM NDM 0 Zr-95 30 NDM NDM NDM 0 ! Hb-95 15 NDM NDM NDM 0 I-131 15 NDM NDM NDM 0

                    .                                                                                                                     ~

4 TABLE 3-1 (SHEET 4 0F 10) RADIOLOGICAL ENVIRO *01 ENTAL MONITORING PROGRAM ANNUAL SIM1ARY

'Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia, Calendar Year 1991 Medium or Type and Lower Limit All Indicator Location with Highest Control locations Number of l Pathway Sampled Total Number of Locations Annual Mean Mean (b) Nonroutine (Unit of- of Analyses Detection (a) Mean (b)-. Name Mean (b) Range Reported Measurement,' Performed (LLD) Range Distance & Range (Fraction) Measurements (Fraction) Direction (Fraction) i Cs-134 15 NDM NDM NDM 0 Cs-137 18 NUM NDM NDM 0 Ba-140- 60 NDH NDM NDH 0 La-140 15 NDM NDM NDM 0 Y'

w Tritium 3000 '1300 No. 83 1300 828 0 8- 880-1540 Downriver 880-1540 530-1200 (4/4) 0.4 miles (4/4) (4/4) Water Near Gross Beta 4 2.83 No. 87 3.18 3.08 0 Intakes to 36 1.3-4.8 Beaufort 1.8-4.8 1.1-10.0 Water (21/24) Downriver (11/12) (10/12) Treatment Plants 112 miles i- (pci/1) Gamma Isotopic 36 i Be-7 80 (e) NDM NDM NDM 0 Mn-54 15 NDM NDM NDM 0 Fe-59 30 NDM NDM NDfl 0 M

TABLE 3-1 (SHEET 5 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUPMARY VogtlA Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia, Calendar Year 1991 Medium or Type and Lower Lim:1 All Indicator Location with Highest Contiel locations Number of 4 Pathway Sampled Total Number 'of Locations Annual Mean Mean (b) Nonroutine (Unit of of Analyses Detection (a) Mean (b) Name Mean (b) Range Reported Measurement) Performed -(LLD) Range Distance & Range (Fraction) Measurements (Fraction) Direction (Fraction) i C0-58 15 NDM NDM NDN 0 ; Co-60 15 NDM NDM NDH 0 Zn-65 30 NDM NDM NDM 0 : Zr-95 30 NDM NDH NDM 0 ! w i E Nb-95 15 NDM NDH NDM 0 1-131 (f) 15 NDM NDM NDM 0 Cs-134 15 NDH NDM NDM 0 Cs-137 18 NDM NDM NDH 0 Ba-140 60 NDM NDM NDM 0 i La-140 15 NDH NDM NDM 0 Tritium 3000 1630 No. 88 2040 165 0 12 910-3200 Port Went 1600-3200 153-180 (8/8) Downriver (4/4) (2/4) , 122 miles ! i l

i TABLE 3-1 (SKEET 6 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SLM ARY Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 ' Burke County, Georgia, Calenaar Year 1991 Medium or Type and. Lower Limit All Indicator Location with Highest Control Locations Number of Pathway Sampled Total Number- of Locations- Annual Me n Mean (b) Nonroutine (Unit of of Analyses Detection (a) Mean (b) Name Mean (b) Range Reported Measuren,ent) Performed (LLD) . Range Distance & Range (Fraction) Measurements (Fraction) Direction (Fraction) t Finished Water ' Gross Beta 4 1.9 No. 88 2.04 1.53 0 at Water 36 1.1-3.1 Port Went 1.1-3.1 0.9-2.5 t Treatment (24/24) Downriver (12/12) (12/12)

Plants 122 miles l (pCi/1)

Gama Isotopic y 36 l

  ~                                                                                                                                  '

Be-7 80 (e) NDM NDM NDM 0 ! Mn-54 15 NDM NDM NDH 0 Fe-59 30 NDH NDM NDM 0 j 00-58 15 NDH NDM NDM 0 i f Co-60 15 NDM NDM NDM 0 Zn-65 30 I NDM NDM NDM 0 t a Zr-95 30 NDM NDM NDM 0 t Nb-95 15' NDM NDM NDM 0 Cs-134 15 NDM NDM NDM 0 i

TABLE 3-1 (SHEET 7 0F 10). RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUP9tARY 3 Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia, Calendar Year 1991 i Medium or Type and Lower Limit Ali Indicator Location with Highest Control Locations Number. of t Pathway Sampled Total Number of. Locations' Annual Mean Mean (b) Nonroutine (Unit of of Analyses. Detection (a) Mean (b) Name 'Mean (b) Range Reported Measurement) Performed (LLD) Range Distance & . Range (Fraction) Measurements (Fraction) Direction (Fraction) f f Cs-137 18 NDM NDM NDM 0 Ba-140 60 NDM NOM NDH 0 j La-140 15 NDM NDM NDH 0 i y 1-131

- m                                                                         36                   1                       NDM                          NDM                  NDM               0 i                                                                            Tritium              E000                    1470           No. 88        1700                 225               0 12                                           670-3900 '

Port Went 670-3900 180-290 l (8/8) Downriver (4/4) (3/4) {. 322 miles i i Anadromous Fish Gamma Isotopic (pCi/kg wet) 1 Be-7 100 (e) NDM NDM NA 0 Mn-54 130 NDM NDM NA 0 l ! Fe-59 260 NDM NDM WA 0 C0-58 130 NDM NDM NA 0 Co-60 130 NDM NDM NA 0 Zn-65 260 NDH NDM NA 0

TABLE 3-1 (SHEET 8 CF 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUP9tARY Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Ceorgia, Calendar Year 1991 Medium or. Type and Lower Limit All Indicator Location with Highest Control Locations Number of Pathway Sampled Total Number of Locations Annual Mean Mean (b) Nonroutine 2-(Unit of of Analyses Detection (a) Mean (b). Name Mean (b) Range Reported Measurement) Performed, (LLD) Range Distance & Range (Fraction) Measurements I (Fraction) Direction (Fraction) Cs-134 130 NDM NDM NA 0 Cs-137 150 12 No. 81 12 NA 0 12-12 Upriver 12-12 (1/1) 1.7 miles (1/1) y Fish Gamma Isotopic

  * (pCi/kg wet)    11 Be-7               100 (e)       NDH                               NDM                              NDM                                0 Mn-54.             130           NDM                               NDM                              NDM                                0 Fe-59              260           NDM                               NDM                              NDM                                0 l                    Co 58              130           NDM                               NDM                              NDM                                0

Co-60 130 NDM NDM NDM 0 Zn-65 260 NDH NDM NDM 0 Cs-134 130 NDM NDM NDM 0 Cs-137 '150 105 Po. 81 211 211 0 27-33G Upriver 26-890 26-890 0 (5/5) 4.7 miles (6/6) (6/6)

I

TABLE 3-1 (SHEET 9 0F 10) - - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

i Vogtlo Electric Generating Plant, Docket Nos. 50-424 and 50-425 ; i Burke County, Georgia, Calendar Year 1991 i Medium or . Type and Lower Limit All Indicator Location with Highest Control Locations Number of i Pathway Sampled Total Number of Locations Annual Mean Mean (b) Nonroutine (Unit of of Analyses Detection (a). Mear (b) Name Mean (b) Range Reported Measurement) Performed (LLD) Range Distance & Range (Fraction) Meas.c=ments

 ;                                                                                    (Fraction)      Direction    (Fraction)

Sediment Gamma Isotopic (pCi/kg dry) 4 i ! Be-7 300 (e) 026 No. 83 826 427 0 : 4 ' i46-910 Downriver 740-910 400-450 ; (2/2) 0.6 miles (2/2) (2/2) 4 LJ ,

2. Co-60 40 (e) 113 No. 83 113 NDM 0 -

c) 113-113 Downriver 113-113 t (1/2) 0.6 miles (1/2) i Cs-134 150 NDM NDM NDM 0 . Cs-137 180 246 No. 83 246 100 0 -

 ,                                                                                    200-290        Downriver     200-290                82-120                              i

(2/2) 0.6 miles (2/2) (2/2) 2 r

i l e v

TABLE 3 1 (SHEET 10 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50 424 & 50 425 Burke County, Georgia, Calendar Year 1991 TABLE NOTATIONS

a. The LLD is defined in table Notation 3 of Table 4.12-1 of the TS.

Except as noted otherwise, the values listed in the column are those ' fcund in that table. In practice, the LLDs attained are generally much lower than the values listed,

b. Mean and range arc based upon detectable measurements only. Fraction of detectable measurements at specified locations is indicated in parenthesis,

c. No Detectable Measurement (s).

d. Not Applicable.

e. The El has determined that this value may be routinely attained. No value was provided in Table 4.12-1 of the TS.

f. Item 3b of Table 3.12-1 of the TS implies that an 1-131 analysis is not I required to be performed on these samples when the dose calculated from the consumption of water is less than 1 mrem per year.

I I i l 3-11

4.0 DISCUSSION OF RESULTS An interpretation and evaluation, as appropriate, of the laboratory results for each type sample are included in this section. Relevant comparisons were made between the difference in average values for indicator and control stations and the calculated Minimum Detectable Difference (MDD) between these two groups at the 99 percent Confidence Level (CL). The MDD was determined using the atandard Student's t-test. A difference in tne average values which is less than the MDD is considered to be statistically indiscernible. Pertinent results were also compared with past results including those obtained during the period of preoperation. The results were examined to perceive any trends. To provide perspective, a result might also be compared with its Lower Limit of Detection (LLD) and/or Reporting Level (RL) which are nominally provided by Tables 4.12-1 and 3.12-2 of the TS, respectively. Attempts were made to explain any RLs or other high radiological levels-found in the samples. There were no failures in the laboratory analyses of each of the samples in attaining the LLDs required by Table 4.12-1 of the TS for this report period. , t Unless otherwise indicated, any reference made in this tection to the results of a previous period will be the results which have been purged of any obvious extraneous short term impacts. During preoperation these included the' nuclear weapons tests in the fall of 1980, abnormal releases from the Savannah River Site (SRS), and the Chernobyl incident in the spring of 1986. During the part of 1987 after operation commenced, these included abnormal releases from SRS. There were no obvious extraneous short term impacts ' during CY 88, CY 89, and CY 90. Near the end of the fourth quarter of CY 91, approximately 7,500 curies of tritium were released f rom SRS to the Savannah River. (This release 3 discussed fully in the report " Release of 7,500 Curies of Tritium to the Savannah River from the Savannah River Site," Prepared by Georgia Department of Natural Resources. Environrental Protection Division, Environmental Radiation Program, January 1992.) Tritium was detected in drinking water at downstream stations (Nos. 87 and 88). However, the tritium concentrations in drinking water are determined from quarterly composited samples. Therefore, the effect on the annual average of tritium in drinking water was minimal (see Section 4.6). Unless otherwise indicated, any referer.ces to CY 87 will be to the operations portion of 1987. The SRS was previously. called the Savannah River Plant. l The annual land use census required by Section 3/4.12.2 of the TS was I conducted on April .i9 The locations of the nearest milk animal, residence, t and garden of greater than 500 square feet producing broad leaf vegetation in each of the 16 meteorological sectors within a distance of 5 miles are tabulated in Table 4-1. Land within SRS was excluded from the census. Any 4-1

                       - . . _ - - . - . - - -        _ _ _ . - _ ~ . . - . - .

e ir '

consequences of the results of the land- use census upon sample collections are discussed in Sections 4.3 and 4.4. The results of the annual survey conducted downstream of the plant to determine whether water from the Savannah River is being used for drinking or irrigation purposes are presented in Section 4.5. To flag any result which differed from the others in its set by a relatively large amount, the practice of testing all results for conformance to Chauvenet's Criterion' was introduced in CY 90 and continued in CY 91. Identified outliers were investigated to determine reasons for deviating from the norm. If an equipment malfunction or other valid physical reason was found, the anomalous result was deemed non-representative and excluded from the data set.' No datum was excluded for failing Chauvenet's Criterion only.

        ' G. D. Chase and J. L. Rabinowitz, Principles of Radioisotone Methodolooy (Burgess Publishing Comptay, 1962) 87-90.

4-2

TABLE 4-1 LAND USE CENSUS-RESULTS Distance in Mi.es to Nearest locations in Each Sector FECTOR ANIMAL RESIDBC1 GARDEN N 1.6 NNE * * * , NE * *

ENE * *
E ESE- * *
SE
4.3
SSE
4.0
S 4.4
SSW 4.7 SW
3.0 4.9 WSW
1.2 3.1 W 3.7
WNW
2.6
NW 1.6 NNW
None within 5 miles and outside of SRS.

l 4-3

4.1 Airborne As indicated by Tables 2-1 and 2-2, airborne particulates and airborne radiciodine ere collected at 5 indicant stations (Nos. 3,7,10,12,and16) which encircle the site boundary, at a nearby community (No. 35) ard at a control station (No. 36). At these locations, air is continuously drawn through a part culate filter and a charcoal canister in sequence to retain airborne particulates and to adsorb airborne radioiodine, respectively. The filters and canisters are collected weekly. Each of the air particulate filters is counted for gross beta activity. A gamma isotopic analysis is performed quarterly on a composite of the air particulate filters for each

                        -station. Each charcoal canister is analyzed for .-131 by gamma spectroscopy.

Two of the air particulate and two of the airborne radioiodine samples were dernd to-be unacceptable. In CY 90, three of the air particulate and two of the airborne radioiodine samples were found to be unacceptable. The samples collected on Pgust 13, at Station 3 were excluded due to low volume as a consequence or air pump failure due to lightening rtriking a transformer. Both samples failed Chauvenet's Criterion. When collecting the samples at Station 7 on December 3, it was discovered that a fuse had blown, thus stopping pump operation after 29.3 hours, as shown by the system clock. Both samples failed Chauvenet's Criterion and were excluded from the results. As seen in Table 3-1, the average weekly pross beta activity during the year ,- for the indicator stations was 0.1 fCi/m greater than that for the control ; station. However, there is no discernible difference between these stations 3 since the difference is less than the calculated MDD of 2.6 fCi/m . 1 I i l l 4-4 , l

The average weekly gross beta activity in units uf fCi/m 3 for the indicator, community and control stations during CY 91 are compared below with those attained during previous years of operation, with the entire preoperational period (which began in September 1981 for the air monitoring stations) and with the range of annual averages during the calendar years of preoperation. Period Indicator Control [pmmunity CY 91 19.3 19.2 18.6 CY 90 19.6 19.4 18.8 CY 89 19.1 18.2 18.8 CY 88 24.7 23.7 22.8 CY 87 .13.0 23.5 22.3 Preop Overall 22.9 22.1 21.9 Preop Rango 18.1-28.1 18.3-26.5 18.3-26.5 1he average weekly readings for CY 91 are seen to be about one percent less l than -that for CY 90, about 85 percent of that generally found during the previous years of operatior and near the lower end of the range of annual averages for the years of preoperation. No trends were recognized in these data. Like CY 88, CY 89, and CY 90, no positive results for manmade radionuclides were found during CY 91 from the gamma isotopic analyses of the quarterly composites of tne air particulate filters. During CY 87, Cs-137 was found in one indicator composite at a level of 1.7 fCi/m 3. During preoperation, Cs-137 was found in an eighth of the indicator composites and a seventh of the control composites with average levels of 1.7 and 1.0 fCi/m3 , respectively; the required LLD is 60 fCi/m3 . Also, during preoperation Cs-134 was found in about 8 percent of the indicator composites; .the average. level was 1.2 fCi/m3 . The required LLD for Cs-134 is 50 fCi/m3 . l' I l-131 was not detected in any of the charcoal canisters during the year. There were no positive . results du:ing the previous years of operation. During preoperation, positive results were obtained only during the aftermath of the Chernobyl incident when levels as high as 182 fCi/m3 were obtained. 3 The maximum allowed LLD is 70'fCi/m ; however, the LLD usuall{ attained is about 30 percent of this value. The RL for I-131 is 900 fCi/m . L l .' g i. 4-5 L l. I t ._ , , _ - . . , - . . , . _ __ , _ _ , , . , _ , , . . . , _ , , _ . . - , - _ , ,

l i 4.2 Direct Radiation Direct (external) radiation is measured by TLDs. A TLD badge is placed at each station; each badge contains 4 calcium sulfate TLD cards. Hence, each of the TLD badges consists of 4 dosimeters. Two TLD stations are established in each of the 16 meteorological sectors about the plant. The inner ring of statiou (Nos. I through 16) is located near the site boundary, while the outer ri,;g (Nos 17 through 32) is located at a distance of about 5 miles. The 16 stations torming the inner ring are designated as the indicator stations. Each of the 4 control stations (Nos. l 36, 37, 47 and 48) is over 10 miles from the plant. Special interest area:; i consist of a hunting cabin (No. 33), the town of Girard (No. 35), and the GPC i employees' recreational area (No. 43). , Not ir. frequently, TLDs are lost due to theft or vandalism. Near the middle of each quarter, the vast majority (85 percent) of the station: (those

readily accessible) are checked for missing or damaged badges; replacement '

l badges are provided as needed. During the first quarter, it was learned that the . badges at Stations 23 and 31 had been stolen. Replacements were installed and then the rylacement badge at Station 23 was stolen. Last year two badges were lost in the field, one of which was burned in a brush fire. As may be seen from Table 3-1, the average quarterly dose of 16.9 mR acquired a .' indicator stations was 0.2 mR less than that acquired et the control

.+ as; this difference was not discernible since it was less than the .

alculated MDD of 1.9 mR. The quarterly doses -acquired at the outer ring-v M ons ranged from 13.1 to 28.0 mR with an average of 16,7 mR which is 0.2 n, .ess than that found for the inner ring.- This difference is not discernible since it is less than the calculated MDD of 1.0 mR. i Listed below for the indicator, control and outer ring stations, are the ; average levels in units of mR/91 days obtained during each year of-operation and the entire period of preoperation (which began in October 1981, for TLD. stations), and the range of annual averages .obtained during the calendar years of preoperation.- Period Indicator Control Outer Rino CY.91 16.9 17.1 16.7 CY 90 16.9 16.6 16.3 CY 89. 17.9 18.4 17.2 CY 88 16.C 16.1 16.0 CY 87 -17.6 .17.9 16.7 Preop Overall 15.3 16.5 14.7 Preop Range 15.1-16.9 14.1-18.2 12.5-16.2 l 4-6

                                                           = _ -           --                 - -           -       --

Overall, the doses for CY 91 were roughly within 3 percent of those found during previous years of operation and approximately 10 percent greater than those found during preoperation. No trend is recognized in these data. The average levels . in units of mR/91 days ror the special interest areas obtained durIng each year of operation and the entire period of preoperation along with the range of annual averages obtained during the calendar years of preoperation are listed below. Period Station 33 Station 4:i Station 43 CY 91 17.5 19.6 17.0 CY 90 16.8 18.9 16.2 CY 89 21.2 18.7 17.4 CY 88 19.7 18.1 14.8 CY.87 21.3 18,5 15.2 Preop Overall 16.6 15.1 15.3 Prcop Range 13.G-19.9 12.6-17.6 13.9-25.0 The doses. acquired at the special interest areas are seen to be somewhat typical and within the range of those acquired at the other stations. 4-7 .

                                                                                          . . - , . . . - . - . , , _ - . . ~ , . . - - , , -   . . - - .

_ __ =_. _ _ _ _ _ _ . _ . _ _ _ _ _ _ ____ _. .___. __ 4

                                          ' 4.3 Milk As indicated ny Tables 2-1 and 2-2, milk is collected biweekly from two                                            i
                                         ' control stations, Dixon Dairy (No. 98) and the Boyceland Dairy (No. 99).

Gamma isotopic and I-131 analyses were performed 00 each sample. Milk has not been available from an indicator stat on (a location within 5 miles of the plant) since April 1986 when +.he cow from which milk was being obtained went dry and was subsequently removed from the area. As indicated by Table 4-1, no milk animals were found in the land use census. The availability of milk within 5 miles of the plant was meager throughout preoperation. A milk animal is a cow or goat producing milk for human l consumption. ; As utual, the only manmade radionuclide found during CY 91 from the gamma isotopic analysis of the milk samples was Cs-137. Listed below are the , average, minimum and maximum levels in units of pCi/1 along with the fraction l of detectable measurements during each year of operation as well as during l preoperation. Period herJLqq Minimum bximum Fraction CY 91 14.1 14.1 14.1 0.018 CY 90 17.0 17.0 17.0 0.018 CY 89 -7.0 5.8 7.7 0.056 CT 88 6.9 4.9 8.1 0.058 CY 87 10.4 9.9 10.8 0.051 Preop 18.1 9.0 27.0 0.044 Although the fraction of detectable measurements during previous years of operation was about the same as that during preoperation, the average level was only about 60 percent of that during preoperation. The level of the one positive result found (at Dixon Dairy) in CY 91 is seen -to be about 80

                                         - percent of- the average - level found during preoperation.                                  No trend- is recognized from these 'results.                    The LLD and RL for Cs-137 in milk, as reequired by the TS, are 18 and 70 pCi/1, respectively. During preoperation, Cs-134 was also' detected in a sample from an indicator station, and during CY'87, Zn-65 was al;o detected in a sample from Boyceland Dairy.                            ._

l During previous years of operation, I-131 was detected in two milk samples, both in' CY 90. (See Plant Vogtle Annual Raoiological Environmental Surveillance Report for CY 90 for a discussion of the circumstances associated with these results.) During preoperation, positive I-131 results l pre found only during the Chernobyl incident; the levels ranged from 0.53 to 5.07 pCi/1. The LLD and RL required by the TS are 1 and 3 pCi/1, i respectively. l 4-8 l

                                                                    -                 ~_

4.4 Vegetation The TS call for the gamma isotopic locationsanalysis of grass near the site boundary or in leafy vegetation collected monthly from two onsitedifferent Grassmeteorological is collected at sectors (S at about 15 or more miles from the plant (Station 37). each of these locations. No gardens were found in the land use census wh stations at which vegetation is being sampled. 5 As indicated in Table 3-1, Cs-13T was the only t greater than that at the indicator stations. l compare results from the indicator stations with those from i the contro station was not possible due to the fact that only ;ne positive obser { was made at the control location. The conclusion was compared a single observation with the mean of a sample. h that there is no statistical difference at the 99 percent CL betwe indicator and control stations,the control stations than at the indicator s being indiscernible. Except for a short period fc11owing the Chernobyl in vegetation incident, Cs-137 h samples by gamma As a the only manmade radionuclide detected isotopic analysis during both the preoperation and operation periods consequence of the Chernobyl incident: samples collected over a period of several week one of the samples. 4-9

The average level of Cs-137 found in vegetation samples in units of pCi/kg wet ahng . Ith the fraction of detectable measurements at the indicator and control stations is shown below for each year of operation and the period of preoperation. Indicator Stations- Control Stations Period antagg Fraction Averaae fraction CY 91 35.3 0.208 62.4 0.083 CY 90 30.0 0.083 102.0 0.166 CY 89 9.7 0.042 0.0 0.000 CY 88 38.7 0.280 0.0 0.000 CY 87 24.4 0.318 61.5 0.250 Preop 54.6 0.573 4.4 0.193 No trend is recognized in these data. The LLD and RL are respectively 60 and 2000 pCi/kg wet. 4-10

4.5 River Water Surface water is composited '-nm the Savannah River at three 1ccations using 15C0 automatic samplers. Sn.al' quantities of river water are collected at intervals not exceeding a few hours. River water collected by these machines is picked up monthly', quarterly composite are made up from the monthly . collections. The collection points consin of a control station (No. 82) ! which is located about 0.4 miles upriver of the plant intake structure, an indicator station (No. 83) which is located about 0.4 miles downriver of the plant discharge structure and a special station (No. 84) which is located about 1.3 miles downriver. A gamma isotopic analysis was made on each monthly collection. As in all previous years of operation, there were no radionuclides of interest detected in the river water samples during CY 91. A tritium analysis was performed on each quarterly composite. As usual, a positive result was obtained from each analy.is. As indicated in Table 3-1, t- the average level of 1300 pCi/1 found at the indicator station was 472 pCi/1 l greater than that at the control station; this difference is not discernible since it i s 1 cts than the calculated M00 of 626 pCi/1. There was a discernible difference in the tritium levels between these two stations in CY 88 and CY 89. At the special station (No. 84), the result ranged from 620 to 1700 pCi/1 with an average of 1081 pCi/1. The required LLD is 3000 pC1/1 and

                             +he RL is 10 times greater.

Listed below for each year of operation are the average tritium levels found

                           -at the control, indicator, and special stations, the difference between the average values at the indicator and control stations (L, - L,), the MDD between these two stations and the annual liquid releases of tritium from the plant. All of. these values are in units of pCi/1 except for the releases, which are in units of Ci.

Item CY 91 CY 90 CY 89 CY 88 CY 87 Control Station 828- 392 538 427 524 Indicator Station 1300 1142 1293 843 680 dpecial Station 1298 1081- 1268 1430 1411-L-l i e 472 750 755 416 156 MDD 626 - 766 518 271 416 Releases 1094 1172 916 390 321 These data show a generally upward trend for plant releases through CY_90, with CY 91' results decreasing to approximately 60 percent of CY 90 levels. The releases are sufficient to account for the increased levels of tritium at 4-11 s'r . q*+.a e a,-py.cee ,mp - ayw e q-g ,,,,y3g , . _ , , ,, , , , ,_..g m,y,,,_.-_,..r---,,,_,,,.~.7%y--.-_,_m, w., _ , . - . -7,,,__m._,,._,,.m_, _

                                                                                                                                                                                        ,,eie,,w-.,,rq ,,,,.%,

the indicator station. The annual organ dose that the maximum exposed individual (a child) would receive from drinking water with an average tritium concentration of 472 pCi/1 was conservatively calculated to be 0.049 mrem or 1.6 percent of the TS limit. The tritium release from SRS in December, which was mentioned in Section 4.0, entered the Savannah River downstream of the indicator station for river water (No. 83). Therefore, that release had no affect on the river water samples. (See Section 4.6 for a discussion of the effect of that release on drinking water taken from the Savannah River.) On September 24 the annual survey of the Savannah River was conducted downstream of the plant for approximately 106 river miles to identify any parties who may use river water for purposes of drinking or irrigation. The i only parties found to be withdrawing river water for drinking purposes were l the two downriver water treatment plants (Stations 87 and 88) from which l- samples are collected monthly. As in all prcvious surveys, no intakes for irrigation use were observed. The survey results were corroborated by contacting the Environmental Protection Division of the Georgia Department of Natural Resources and the South Carolina Departnient of Health and Environmental Control; it was found that no new surface or drinking water withdrawal permits had been issued in CY 91 for the Savannah River downstream of the plant. l 4-12 i

I i 4.6 Drinking Water i Samples were collected at a (.ontrol station (No. 80), the Augusta Water Treatment Plant in Augusta, Georgia, which is located about 56 miles upriver and at two indicator stations (Nos. 87 and 88), the Beaufort-Jasper County Water Treatment Plant near Beaufort, South Carolina, and the Cherokee Hil' Water Treatment Plant near Port Wentworth, Georgia, which are respectively located about 112 and 122 miles downriver. These upriver and downriver distant es in river miles are the distances from VEGP to the _ point in the river where water is diverted to the intake for each of these water treatment pluts. At each of the water treatment plants, monthly collections were made of river water which was composited near the plant's intake (raw drinking water) and of grab samples of finished drinking water; quarterly composites are made up , from the monthly collections. Gross beta and gamma isotopic analyses were performed on each of the samples collected monthly. Tritium analyses were performed on the quarterly composites. Although an 1-131 analysis is not required to be performed on these samples when the dose calculated from the consumption of water is less than 1 mrem per year (see Item 3b of Table 3.12-1 of the TS), an 1-131 analysis was performed on each of the grab samples of finished water collected monthly since a drinking water pathway exists. As indicated by Table 3-1, the average gross beta xtivity for raw drinking water was 0.25 pCi/l greater for the control station than for the indicator stations. However, this difference is not discernibic since it is less than the calculated MDD of 2.47 pCi/1. For finished drinking water, the average gross beta activity was 0.37-pCi/l greater for the indicator stations than for the control station. This difference is not discernible since it is less than the calculated MDD of 0.47 pCi/1. 4-13

H l Listed below for each year of operation are the average gross beta levels for j raw and finished drinking water in units of pCi/l at the indicator and control stations, and the difference between the average levels at these stations (Li - L,). , 1 Period Indicator Cont ro_1 1Li dl l RAW CY 91 2.83 3.08 -0.25 CY 90 2.53 2.55 -0.02 , CY 89 2.93 3.05 -0.12 l CY 88 2.67- 3.04 -0.37 ' CY 87 2.20 5.50 -3.30 4 FINISHED CY-91 1.90 1.53 0.37 CY ')0 2.08 1.92 0.16 CY 89 2.36 2.38 -0.02 CY 88 2.28 2.35 -0.07 CY 87 2.10 1.80 0.30 With the exception of the high reading for the raw drinking water for the control station for CY 87, the above tabulations show fairly consistent results. The high reading in CY 87 was attributed to sediment being drawn-into a few of the samples. Ignoring this high reading, the overall average gross beta reading for all- years of operation is seen to be 27 percent greater for the raw drinking water than for the finished drinking water; this is expected since the fir' hed water has been filtered. There has not been a . discernible difference between the average gross beta values at the indicator and-control stations during any of the years of operation. As indicated in Table _ 3-1, there were no positive results for the radionuclides of interest frot the gamma isotopic analyses of the monthly collections. Only one positive result has been found since operations began; Be-7 at a level of 68.2 pCi/1 was found in the sample collected for September 1987 at Station 87. 4-14

l l Listed below for each year of operation are the average tritium levels found in _the quarterly composites of raw and finished drinking water in units of pCi/1 collected at the indicator and control stations, the difference between the average levels at these stations (L, - L,) and the MDD. period Indicator Control RA1 _@p_

I RAW CY 91 1630 165 1465 1537 CY 90 1320 266 1054 572 CY 89 2508 259 2249 1000 CY 88 2630 240 2390 580 CY 87 2229 316 1913 793 FINISHED CY 91 1470 225 1245 1082 CY 90 1299 404 895- 1131 CY 89 2236 259 1977 627 CY 88 2900 270 2630 830 CY 87 2406 305 2101 1007 The above tabulations show that in previous years of operation, with the exception of finished drinking water for CY 90, there was always a detectable difference between tritium concentrations at the indicator and control stations. 'A detectable difference exists when the absolute value of (L, - L,) exceeds the MDD.) The tabulations also shca a decided decrease in the tritium levels at the indicator station durin0 Ci 90 and CY 91. During preoperation the. result were_ similar to those for CY 87 through CY 89. As stated in Section-4.0, in late December a liquid release occurred at SRS in which.7,500 Ci of tritium were released to the Savannah River. The actual release. occurred between December 22 and December 25. The release entered the Savannah River via Steel Creek. Due to the approximately 100 river miles distance between the point at which Steel Creek enters the Savannah River and the intake points for the water treatment plants (Stations 87 and 88), the

     -plume containing the released tritium did not reach the intat9s 'for the water
     - treatment plants until about December 29. (See the reportL referenced in                                                             '

Section 4.0.) The early notification of water treatment plant operators by SRS, and the few days delay. between the time _the release occurred and the 4 time the plume reached the water treatment plant intakes, allowed the water

     ~ treatment plant operators - and governmental officials to take action to counter.the potential impact on drinking water supplies. These _ circumstances 4-15 l

ev vi--,,p C1 g r-s -gg r** m y + g-- ir --y --gu y ,--e as- % i+*-tr--r .*$ c +*----y --- ,eycii y-

coupled with the fact that the release occurred so late in the quarter, with tritium concentrations being determiner ' ' analysis of quarterly composite i samples, caused the effect on raw and i J drinking water sample results i (for Station Nos. 87 and 88) to be less might be expected from a relesse i of this magnitude. As shown in the above table, for raw drinking water, the .; annual tritium value for the inlicator stations was not discernible from the control station. For finished drinking water, the tritium value for the indicator stations is discernible from the value for the control station, but I the result is in'the same range as those for previous years of operation. As indicated in Table 3-1, there were no pc<ltive results from the 1-131- ' analysis of the finished drinking water samples; each result was below its Minimum Detectable Activity it!DA) which ranged from 0.13 to 0.56 pCi/1. Similar results were abtained in previous years of operation. The TS call for a LLO and a RL of I and 2 pCi/1, respectively. I i 6 f 4-16 1

                                                                                         "'1rr*'me---we m-awe wy               w es+ie w -a .wy,emy,-w 3,.av 'm v e rwe www wewe s ye e

4.7 Fish The TS call for tne collection of at least one sample of any anadromous species of fish in the vicinity of the plant discharge during the spring spawning season. The TS also call for semiannual collections of any commercially or recreationally important species in the vicinity of the plant discharge area and in areas not influenced by plant discharges. Further, the TS call for a gamma isotopic analysis on the edible portions of each sample collected. About a.five mile-stretch of the river is generally needed to obtain adequate fish samples. For the semiannual collections, the control station (No. 81) extends from approximately 2 to 7 miles upriver of the plant intake structure and the indicator station (No. 85) extends from about 1.4 to 7- miles downriver of the plant discharge structure. For the anadromous species all collection points can be considered as indicator stations.

                               - On March 25, American shad, an anadromous species, was collected at Station

81. The only radionuclide of interest detected was Cs-137 at a level of 12 pCi/kg wet. In CY 88, CY 89 and CY 90, no positive results for the radionuclides of interest were obtained from the gamma isotopic analysis. In CY 87, Cs-137 was found in one of the three shad collected at a barely detectable level of 10 pCi/kg wet. The LLD for Cs-137 in fish as specified by the TS is 150 pCi/kg wet.

On April'15-and October 21, the composition of the catches at the indicator and control stations were as follows: i Qalg Indicator Control April 15 Channel Catfish Channel Catfish Redbreast Sunfish largemouth Bass Redear Sunfish - October 21 Largemouth Bass Channel Catfish Channel Catfish Largemouth Bass Redbreast Sunfish Redear Sunfish As indicated in Table 3-1, Cs-137 was the only radionuclide of interest found in the semiannual collections of commercially or recrestionally important '

                               - species.      Since operation' began, the only other radionuclide of interest-detected was 1-131 which was detected in CY 89'and CY 90 at levels of 18 and 13 pCi/kg wet, respectively.

4-17

                                                    .gm       en ,,-yr-                            r         v- ' r'          vm erwwwdykir-----g-iyNerwi-T--$-ig*te ii=--m--Rr We T

l l As seen in Table 3-1, the average level of 105 pCi/kg wet for Cs-137 at the indicator station is 106 pCi/kg wet less than that at the control station. ! This difference is not discernible, however, since it is less than the calculated MDD of 198 pCi/kg wet. Since operations began, positive values for Cs-137 have been found in all but one of the 47 samples collected. Listed below for each year of operation are the average levels of Cs-137 in units of pCi/kg wet found in fish samples at the indicator and :ontrol l stations. Egrit' Indicator Control CY 91- 105 211 CY 90 103 249 CY 89 117 125 CY 88 66 116 CY 87 337 119 No trend is recognized in this data. ' 4-18

            .._.z.___                                                                                                                                          _ , _ . _ _ _ . _ _ _ - - - . .

d 4.0 Sediment b

iir. t was collected along the shoreline of the Savannah River on April 1

$ tember 30 at Stations 81 and 83. Station 81 is a ontrol station at 2.5 miles upriver of the plant intake structum at RM 153.3

n 83 is ar. indicator station located about 0.6 mih.s downriver of

                   . 2. t charge structure at RM 150.2. The indicator samples for April g            w Se ; ;;er were collected at RM 150.2 while the control samples were g

col' at RM 153.3. A gamma isotopic analysis was performed on each li*ted ' low for each year ci operation are the average levels of 4. reaitnuclides of interest in units of pCi/kg dry found in the regular samples E : c ellected at the indicator and/or control stations along with the frequency of occurrence and the :' Os. Each of there radionuclides is included in the plant's liquid release:.. Period Jndicator Frequency Control Frecuency 3 Be-7, LLO 300 CY 91 3 D. 1.0 427 1.0 CY 90 4C' l.0 545 1.0 CY 89 1300 1.0 415 1.0 e CY 88 9'^ 1.0 810 1.0 CY 87 987 1.0 543 1.0 / Mn-54, LLD-50 CY 89 18 0.5 CY 88 22 0.5 Co-58. LLD=25 CY 90 140 0.5 CY 89 135 1.0 CY 88 190 1.0 Co-6G, LLD-4b CY 91 113 0.5 CY 90 46 0.5 CY 89 46 1.0 CY 88 62 0.5 4-19 1

l l i Period- indicator ff_tqyency Control Frecuency C -137, LLD-180 CY 91 246 1.0 100 1.0 CY 90 155 1.0 140 1.0 CY 89 230 1.0 125 1.0 CY 88 175 1.0 175 1.0 CY 87- 209 1.0 111 1.0 As in all previous years of operation, positive readings in CY 91 for Be-7 and Cs-137 were found in each sample and the readings were on the same order as found previously. For Be-7, the average reading of 826 pCi/kg dry for the indicator station is 399 pCi/kg dry greater than that for the control station;._however, this difference is not dir.cernible since it is less than

    - the calculated MDD of 614 pCi/kg dry. For Cs-137, the average reading o# 246 pCi/kg dry for the indicator station is 146 pCi/kg dry greater than that for the control station; this difference is not discernible, however, since it is less ~ than the' calculated MD0 .of 335 pC1/kg dry.

There has been no

    ' discernible difference between ti' .evels at the indicator and control stations for either Be-7 or Cs-137 during-any past year of operation.

The activation product Co-60 is seen to be present again this year at the indicator station at about twice the level as last year and in only half of the . samples. It is also'noted that Hn-54 (also an activation product) which appeared in' half of the samples from the indicator station in CY 88 and CY 89 was not present in CY-90 or CY 91. Since the Co-60 was only found at the indicator station,- its presence is believed to be due to plant releases. Co-60 was found in sediment samples during preoperations. The radiological impact due to the presence of Co-60 in the shoreline sediment was assessed by calculating the whole body dose due to direct radiation (from the sediment) to an individual using the methodology and parameters of NRC Regulatory Guide 1.109, Revision 1, October 1977, and comparing this dose with that permitted by Sertion 3.11.1.2b of the TS (3 mrem per year). The theoretical dose was conservatively determined to be 4.5 microrem per year or - 0.15 percent of the- TS limit. - This extremely low s potential dose, altho gh calcul able, poses no measurable negative environmental or public health impact. The theoretical doses due to the activation products in CY 88, CY 89, and CY 90, were found-to be 3.6, 2.6, and 2.5 microrem, respectively. l 4-20 o l-i

5.0- INTERLABORATORY COMPARIS0N PROGRAM Section 3.16.3 of the_TS requires that analyses be performed on radioactiv'. materials supplied as part of an Interlaboratory Comparison Program approved by the NRC. The Environmental Protection Agency's (EPA's' Environmental Radioactivity Laboratcry Intercomparis n Studies (Crosst :k) Program conducted by the Environmental Manitoring and Support Laboratory in las Vegas, Nevada, provides such a program. Reported herein, as required by Section 4.16.3 of the TS, are the results of the EL's participation--in the EPA Crosschack Program. The Crosscheck Program was designed for laboratories involved with REMPs; it includes environmental media and a variety of rsdionuclides with activities at or_ near environmental levels. Participation in the program ensures that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed; REMP results can thereby be demonstrated to be reasonably valid. Simulated environmental sample' are distributed regularly to the participants who analyze the samplo and idurn the results to the EP_A for statistical analysis and comparisons with known values and results obtained from othar participating laboratories. The Crosscheck Program provides each participant with documentation of its performance; this can be helpful _ in identifying any instrument or procedural problems. The EL's participation in the program consists of the analyses on the

radioactive materials su,glied by the program that correspond with those required by Table 2-1. Analyses were .aerformed in a normal manner. Each sample. was analyzed in triplicate as required by the program. Results obtained from the gross beta and gamma isotopic analyses of air filters, the gamma isotopic and I-131 analyses of milk samples, and the tritium and gamma isotopic analyses of water samples are summarized in Table 5-1. ' Delineated in Table 5-1 for each -of the environmental media ai-e the type analysis performed, EPA's collection date, the known value and expected precision (one-standard deviation) provided by the EPA, the average result obtained by the EL, the standard deviation of the EL's result, the normalized deviation (from the known result), and the normalized range. The normalized

deviatior and normalized range were also provided by the EPA. 5-1

The normalized deviation ' from the known 'value provides a measure of the central' tendency of the data (accuracy). The normalized range is a measure of the dispersion of the data (precision). An absolute value of three standard deviations was established by the cPA as the control limit. An absolute value of two standard deviations was established as the warning limit. The EL - considers any value greater than the control limit as unacceptable. Investigations are undertaken whenever any value exceeds the warning limit-or whenever a plot of the values indicates a trend. c As may be seen from Table 5-1, the normalized deviation and the normalized range in each- case were within control limits but the warning limit was exceeded for the Cs-137 analysis on air filters on March 29 and August 30. The warning limit was exceeded for the Ru-106 analysis in water on February 8 and October 4. Also the warning -limit was exceeded for the Co-60 analysis in water on October 4. For Cs-l37 on air filters, the investigation into the positive bias for l normalized deviations led to the conclusion that this condition was the l result of differences between the geometry of the calibration standard and the EPA Crosscheck sample. Geometry corrections are being developed. For Ru-106 and Co-60 in water, the investigation led to the conclusion that the positive bias for normalized deviations probably resulted from changes in background count rate following relocation of the detectors. Computer software is being developed to evalcate background data to revise peak background correction values. One sample, collected Jun_e 7, had a normalized range of 2.28 for Ba-133 in Water. The sample analysis results were investigated, found to be correct, and no reason was found for the higher normalized range value. All other nnemalized range values for Ba-133 in water have been within the two standard deviations warning limit. The result was not investigated further since the result was within the three standard deviations control limit and no trend was' indicated._ 5-2

TABLE 5-1 (SHEET 1 0F 2) CR05SCHECK PPFAAM RESULTS Date Known Expected Reported Standard Normalized Normalized Analysis Collected Value Precision Averaae- Deviation Deviation Rance Air Filters (pCi/ filter) Gross Beta 03/29/91 124.0 6.0 122.67 1.53 -0.38 ' 0.3'O 08/30/91 92.0 10.0 92.67 0.58 0.12 0.06 Cs-137 03/29/91 40.0 5.0 46.67 4.51 2.31 1.12 08/30/91 30.0 5.0 36.33 1.15 2.19 0.24 Milk (pCi/1) [ I-131 04/26/91 60.0 6.0 59.67 3.79 -0.10 0.69 09/27/91 108.0 11.0 104.33 3.79 -0.58 0.33 Cs- 137 04/26/91 49.0 5.0 50.67 1.53 0.58 0.35 09/27/91 30.0 5.0 31.67 4.51 0.58 1.12 Water (pCi/1) H-3 02/22/91 4418.0 442.0 4726.67 75.06 1.21 0.17 06/21/91 12480.0 1248.0 13200.00 173.20 1.00 0.12 10/18/91 2454.0 352.0 2713.33 64.29 1.28 0.20 Co-60 02/08/91 40.0 5.9 39.33 3.21 -0.23 0.71 06/07/91 10.0 5.0 13.67 2.52 1.27 0.59 10/04/91 29.0 5.0 35.00 3.46 2.08 0.71 Zn-65 02/08/91 149.0 15.0 152.33 3.21 0.38 0.24 06/07/91 108.0 11.0- 115.67 14.05 1.21 1.96

10/04/91 73.0 7.0 78.33 6.43 1.32 1.02 4

TABLE 5-1 (SHEET 2'0F 2) CROSSCHECK PROGRAr1 RESULTS Date Known Expected Reported Standard Normalized Normalized Analysis Collected Vali. . Precision' Averace- Deviation Deviation Ranae Water (pCi/1)_-(Continued) 3 Ru-106 02/08/91- 186.0- 19.0 -217.00 6.24 2.83 0.37 06/07/91 149.0 15.0 141.00 5.20 -0.92 0.35 10/04/91 199.0 20.0 225.33 8.02 2.28 0.47 Cs-134 02/08/91 8.0 5.0 11.00 0.00 1.04 0.00 04/16/91 24.0 5.0 22.00 4.58 -0.69 1.12 06/07/91 15.0 5.0 17.00 0.00 0.69 0.00 '! u, 10/04/91 10.0 5.0 11.33 1.15 0.46 0.24 Cs-137 02/08/91 8.0 5.0 9.33 0.58 0.46 0.12 l 04/16/91 25.0 5.0 26.67 1.15 0.56 0.24 ;

                        .14.0       5.0        18.67               4.62    1.62       0.95 06/07/91 10/04/91   10.0       5.0        11.67               1.53    C.12       0.35 Ba-133-  02/08/91   75.0       8.0        73.00               2.65  -0.43        0.37 06/07/91   62.0       6.0        63.33               8.62    G.38       2.28 10/04/91   98.0      10.0        98.00               2.65    0.00       0.30 4

r l t E l o i

a 6.0 CONCl.USIONS

 ;                This report confirms the licensee's conformance with Section 3/4.l? of the TS
                - durtr,'; the year. It shows that all data were carefully examined. A summary and a discussion of. the results of the laboratory analyses for each type sample collected are presented.                       All results indicate no measurable adverse radiological impact to the environment as a result of plant discharges to the river or to the atmosphere, i

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6-1 \ . - . - .

y. V GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DCCKEl NOS. 50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 i ANNUAL ENVIRONMENTAL'0PERATING REPORT FOR 1991 (NONRADIOLOGICAL)

V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 AND UNIT 2 ANNUAL ENVIRONMENTAL OPERATING REPORT (NONRADI0 LOGICAL) 1991 i SPECIFICATION In accordance with Section 5.4.1 of the Vngtle Electric Generating Plant Environmental Protection Plar. (Nonradiological), Appendix B to facility Operating License Nos. NFF-68 and NPF-81, this report is submitted describing implementation of the Environmental Protection Plan for the calendar year 1991. REPORTING REQUIREMrNTS A. Summaries ant ialyses of Resnits of the Environmental Monitoring Activities fo- the Report Period

1. Aquatic Monitoring - Liquid effluent monitoring was performed in accordance with National Pollutant Discharge Elimination Sy: tem (NPDES) Permit GA0026786; there was no additional requirement for aquatic monitoring during 1991. Two NPDES noncompliance events were reported to the State of Georgia during 1991,

2. Terrestrial Monitoring - Terrestrial monitoring is not required.

3. Maintenance of Transmission Line Corridors

a. Right-of-way re-clearing was conducted on the VEGP-Thalt,in 500 kV line from June 1, 1991 to August 2, 1991. Work was performed with rotary mowers equipped with low ground pressure tires. In cultural resource areas, clearing was conducted by hand utilizing chain saws and brush axes. _

The herbicide Accord was used on approximately 60 acres of wetland swamps along the VEGP-Thalman 500 kV line between structures 117-118; 119-122; 125-126; 135-137: 419-422; and 424-430 between July and October 1991. The product is registered by the Environmental Protection Agency for this type of application and was applied by licensed applicators in strict compliance with the herbicide label. The VEGP-Goshen 230 kV line between structures 0 and 79 was ) rc cleared between January 1 and January 3, 1991. Low ground pressure tires were util; zed in the two cultural properties along this route; hand clearing along this route is not required by the Final Cultural Resource Management Plan. The VEGP-South Carolina portioc of the corridor was also re-cleared during 1991. There was no transmission line corridor maintenance performed on the VEGP-Scherer 500 kV line in 1991. 1

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b. There were no clearing or maintenance activities conducted within the Ebenezer Creek or Francis Plantation areas during 1991,

c. Routine maintenance activities within the designated cultural properties along transmission line corridors were conducted in accordance with the Final Cultural resources Management Plan.

4. Noise Monitoring - There were no ccmplaints received by Georgio Power Company during 1991 regarding noise along the VEGP-related high voltage transmission lines.

B. Comparison of the 1991 Monitoring Activities with Preoperational Studies, Operational Controls, and Previous Monitoring Reports These comparisons were not. required because no nonradiologicd monitoring programs were conducted during the reporting period beyond those performed in accordance with NPDES Per. nit No. GA0026786 referenced in Section A above. C. An Assessment of the Observed Impacts of Plant Operation on the Environment There was no significant adverse environmental impact associated with plant operation in 1991. D. Environmental Protection Plan (EPP) Noncompliances and Corrective Actions There were no EPP noncompliances during 1991, i E. Changes in Station Design or Operation, Tests, or Experiments Made in Accordance with EPP Subsection 3.1 which involved a Dotentially Significant Unreviewed Environmental Question ! There were no changes in station design'or operation, tests, or I Xpet .:Lents during 1991 which involved a potentially significant unreviewed environmental question. F. Nonroutine Reports Submitted in Accordance with EPP Subsection 5.4.2 ihere were no nonroutine reports submitted in 1991. 2}}

ELV-03678, 1991 Annual Rept Part 2

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