ELV-02749, 1990 Annual Rept-Part 2
| ML20073F176 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/31/1990 |
| From: | Mccoy C GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ELV-02749, ELV-2749, NUDOCS 9105020080 | |
| Download: ML20073F176 (256) | |
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- April 29, 1991 ELV-02749 0941 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
V0GTLE ELECTRIC GENERATING PLANT 1990 ANNt)AL REPORT - PART 2 In accordance with the applicable regulatory requirements, Georgia Power Company hereby submits Part 2 of the 1990 Annual Report of operating informatior.. It includes the remainder of the 1990 reports not previously submitted. Sincer , p C. K, McCoy CKM/JLL/gmb
Enclosure:
Annual Report - Part 2 xc: Georaia Power Company Mr. W. B. Shipman Mr. M. Sheibani Mr. P. D. Rushton NORMS V. S. Nuclear Reattlatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle 9105o20080 901231 ADOCK 050004 &,+ PDR 7 pl R PDM
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- .990 ANNUA:a 33?03"
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l DOCKET NUMBERS 50 - 424/425 LICENSE NUMBERS NPF-68/81 1 - _ _ - -_____-- . _ - - _ - _ .
GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET NOS. 50-424 AND 50-425 FAClllTY OPERATING LICENSE NOS, NPF-68 AND NPF-81 1990 ANNUAL REPORT - PART 2 TABLE OF CONTENTS
- 1. INTRODUCTION
- 11. PLANT MODIFICATIONS AND TESTS OR EXPERIMENTS o PLANT MODIFICATIONS o TESTS OR EXPERIMENTS 111. EMERGENCY CORE COOLING SYSTEMS OUTAGE DATA REPORT IV. ANNllAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT V. ANNUAL ENVIRONMENTAL OPERATING REPORT
1 GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 & 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 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 operation started on May 31, 1987. Unit 1 is operating in its third fuel cycle, Unit 2 received its operating license on February 9,1989, and began commercial operation on May 20, 1989, Unit 2 is operating in its second fuel cycle.
11 GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 AND UNIT 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 PLANT MODIFICATIONS AND TEST OR EXPERIMENTS
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1 1 l l a l' 1990 ANNUAL REPORT - PART 2 1 1 e I 10 CFR50.59(b) 3-I i i 'l PIR(r M OIFICATIONS 1 i 1 l
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l II , 1990 ANNUAL REPORT - PART 2 1 10 CFR50,59(b) REIORT DCP's 87-V1E0031 'Ihis is a " SAFEGUARDS" DCP, the following abbreviated informtion has been declassified for this report. This modiff. cation creates a vital area barrier for the security radio repeater.
- 1. This modification does not involve equipnent important to safety, only the addition of security barriers. 'lherefore, there is no increase in the probability of occurrence or ceasequences of an i accident or m lfunction of equipnent important to i safety previously evaluated in FSAR section 15,
- 2. Only security barriers are added, no equipment ,
inportant to safety is involved. Therefore, there l is no possibility of a different type of accident that needs to be evaluated in the safety analysis report.
- 3. Technical Specifications do not address the plant security system or barrier requirements.
Therefore, this m dification does not reduce the mrgin of safety'as described in the bases for any technical specification. 87-VlE0047 A. This change involved the addition of auto m tic door closers to Aux. Bldg. doors D65, C47,B18, and A17. 'lhe door closers hold these doors open during nonnal operating conditions,-thereby providing a vent path for escaping steam pressure in. the event of a postulated high energy line break (HELB). During a fire, the door closers automatically close the doors, assuring the integrity of the rated fire barriers. For those held-open doors where health / physics access control is a concern, wire msh gates wre installed within the sam door, opening to allow positive control of personnel access. This design solution satisfies-the three. separate requirements of pressure venting, fire rating, and access control. 't F a.x-- ---.....-.:. - . - . . - . . . - . . - . . -
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II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT B ._ Also, Aux Bldg. doors C54 and A09 will be 1 replaced with wire tmsh -gates, and door C73 will be deleted entirely. Aux. Bldg. Service Stair $6 was derated.
- 1. We proposed change does not increase the probability of occurrence or consequences of the mlfunction of any equipmnt or component assunod to function in accidents analyzed in the FSAR. his included a review of FSAR Qiapter 15 (Accident Analyses) 3.4 (Flooding Analysis) and 3.6 (IELBA). !
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- 2. The proposed change does not create the possibility of an accident or equipmnt/cmponent malfunction not described and analyzed in Cha,ter 15 of the FSAR. Bere is no change to the M hazard analysis of 3.4 and 3.6.
- 3. The proposed change does not decrease the margin of safety defined by the bases of the Technical Specifications. This included review of Sections 2, 3/4, and 6.11.
87-VIE 0117 The liquid waste processing system (IMPS) spent resin sluce pump will be downgraded frm AStE Section III to manufacturer's standards and the N-Stamp removed. The design pressure rating of the pump and discharge c mponents will be raised from 150 psig to 240 psig. Appropriate pressure gauges will be changed out to allow monitoring of line-pressures. The nitrogen supply pressure will be increased to 91 psig to efficiently _ transfer the spent resin to-the disposal. facility from the spent resin storage tank.
- 1. The changes involve a non-safety related system.
f Wey do not increase the probability of an accident or malfunction of equipment important to safety. There is no effect on FSAR section 15.7.2 (Radioactive liquid waste system leak or failure). These changes have no effect on the equipmnt- or canponents assumd to function in the accident analyses of FSAR section 15. 2
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II 1990 A!NJAL REIORT - PART 2 10 CFR50.59(b) REPORT
- 2. An increase in spent resin sluice punp discharge pressure and change out of pressure gauges are the only physical changes involved. Since all piping and canponents are designed to acccundate the new pressure system, operation is not othetvise affected. No new possibilities for accident or failures are created.
- 3. The changes as described do not affect the bases as defined in section B 3/4.11.1 of the Technical Specifications.
87-V1E0124 This DCP adda rem te valve operators to the Liquid Waste Processieg Systen (LWPS) (1901) valves located in enclesed pits. 'lhis portion of the LWPS has piping connections betwen different flow paths depanding on liquid volume or operations requirements. The LWPS is designed to control, collect, process, handle, store, and dispose of liquid radioactive waste generated as the result of nonnal operation, including anticipated operational occurrences. Cross connections are available such mat additional flexibility to store and process liquid waste fran one unit can be provided by the liquid waste system of the second unit and to other collection / processing facilities.
- 1. The probability or consequence of the malfunction of any equipment is not changed as a resul' of this modification. The externally added reac's rods will be supplied with torque - limiting canponeta ,.
This system is not assuned to function in ac :idents analyzed in the FSAR. This is based on review of FSAR Chapters 11 and 15,
- 2. This nodification does not create the possibility of any accident or malfunction that is not described and analyzed in the FSAR. The valve reach rods will have their own supporting systan and torque-limiting canponents. This included review of Chapter 11 and 15.
- 3. There is no decrease in the enrgin of safety defined by the Technical Specifications as a result cf this modification. The bases for Tech. Spec 3/4.11.1.3 are not affected by this modification.
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II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPOKr 87-V1E0150 Re-route Waste Gas caupressor suctica lines fran the Recycle Evaporator Vent condensor to eliminate low point drain and check valve ins:alled at low- j point. Install a check valve at a nore preferable location. W e Boron Recycle System is discussed in FSAR section 9.5.4.2. 4
- 1. There is no change to the probability of occurrence-or malfunction consequences a: m aad in FSAR chapter 15, 9.5, Appendix 9A or 9B.
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- 2. W is change does not create a scenario which is not bounded by FSAR chapter 15, 9.5, Appendix 9A or 9B.
- 3. Be Baron Recycle system is not discussed in the l Technical Specifications, herefore, there can '
be no adverse affect on the Technical Specification bases. Bis is based on review of Technical -i Specification 3/4.1 bases. 87-V1E0155 Lis DCP adds drain lines to three systems (1204, 1 1208, and 1212) at the contairment alping penetrations identified below. Eaca system includes containment isolation valves which must receive a local leak rate test (LLRT). We piping's physical routing has low points at or near the genetration which can not be gravity drained for tne LLRT. Eis modification will add drain lines and valves at these locations. Line 1204-148-3/4", Lines 1208-055-3", Line 1212-001-3/8" and Line 1212-033-3/8". ,
- 1. The proposed change does not increase the probability of any accident discussed in the FSAR. The drain valves being added meet all the FSAR cannituents with regard to containnent isolation (6.2.4), leakage testing (6.2.4, 6.2.6), high energy line evaluation (3.6) and system parameters (6.3, 9.3.2, 9.3.4). W e addition of these drain valves does not increase the probability of any accident discussed in Chapter 6 or 15.
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II 1990 N EUAL REPORT - PART 2 10 CFR50.59(b) REPORT
'Ihe probability or consequences of malft'nction of any equipment is not changed as a result of this modification. Addition ot the drain valves does not increase the probability-of any accident discussed in Chapter 6 or 15.
- 2. This change does not create the possibility of any accident or malfunction that is not described and analyzed in the FSAR. Le nodification does not introduce any new stress or high energy line considerations that are not envGoped by the existing Chapter 6 analysis.
Le postulated accidents of Chapter 15 are not affected by this change.
- 3. There is no decrease in the margin of safety defined by the Technical Specifications as a result of this modification. The bases for L Technical Specifications 3/4.6.1, 3/4.6.2, ;
3/4.6.3, 3/4.5.2 and 3/4.5.3 are not affected ' by this modification. 87-VlE0160 Chillers 1-1592-C7-001/002 have flowswitches Seq 1 which alarm and trip on low water flow through the L chiller evaporator. These flowswitches are set 0 410 G N (alarm) and 400 G M (trip). he l proposed change will reduce the setrint of the g l alarm flowswitches to 3R GPM and tae setpoint of t the trip flowswitches to d GPM.
- 1. The change of chiller low-flow setpoints will not increase ~the probability-of malfunction of this equipment or ccxuponents asstned for protection for <
the chiller evaporator to prevent tube freezing which could result frcxn reduced water flow. There L are two additional existing methods of preventing l-evaporator tube freezing which is accomplished by . temperature-sensors for the liquid refrigerant and of the leaving chilled water temperature. The flow-switches will still detect a ccxnplete loss of flow. l' without causing unnecessary chiller trips on reduced l "' chiller loads. This design change does not increase-the probability or consequences of any equipment 5
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.990 1 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT nulfunction related to the coolers serviced by the essential chilled water system. This is based on review of the text and Failure Modes and Effects Analysis (REA) presented in the following FSAR sections:
System 1531 FSAR section 6.4.4 Syst m 1532 FSAR section 9.4.5 System 1539 FSAR section 9.4.1 Systm 1555 FSAR section 9.4.3 System 1561 FSAR section 9.4.. 3 (Ref. FSAR Section 9.2.9.1 and 15.0)
- 2. The lower chilled water flow through the chiller will not cause equipment malfunction as verified with the chiller manufacturer. Unnecessary chiller trips will be avoided by lowering the flow'setpoints to the levels which are possible during chiller operation. The essential chilled water system and coolers serviced by this system do not have any new camponent malfunction created as a result of this design change. This is based on review of FSAR sections including associated REA's for 6.4, 9.2.9, 9.4.1, 9.4.3, 9.4.5, and 15,
- 3. This change does not decrease the taargin of safety defined in the Technical Specification 3/4.7 bases.
87-V1E0lG0 A. Setpoints for flow switches Seq 2 1FSL-1802/1803 and 1FIS-1802/1803 (Chiller Condenser Water Flow) are revised to correspond with the mininun condenser flow provided by throttling valves 1TV-11675 and 11V-11740. The FIS's provided a low-flow alarm and the FSL's provide a low-flow trip for the chillers. The minimum flow provided by the throttling valves is 500 GIM. The alann setpoint is 425 GPM and the trip setpoint is 250 GPM. B. A time delay is added to the chilled water pump auto-start circuit (SI/CRI/ Toxic Gas) to ielay the chiller ccmpressor start past the point of NSCW flow transients which occur following a IDSP. 6
l II 1990 ANNUAL REPORT - PART 2 10 CFR50,59(b) REPORT C. Le chiller recycle inhibit is activated by the empressor pwer circuit breaker instead of the chiller control circuit,
- 1. The proposed changes do not affect the design safety function of the chillers as described in FSAR Section 9.2.9.1.1. The chi.ller condenser design flow rate of 1100 GPM will continue to be available during design heat loads and accident conditions. The added tim delay in establishing chilled water flow to essential HVAC does not inpact design ambient tenperatures (Ref. RER 88-0860). Single failure effects of the added tim delay would only affect one train chiller (other train available - FMEA 9.2.1-2). He tim delay is not activated if a S1/CR1/ toxic gas signal is received while the chiller is in the manual mode of operation.
- 2. L e proposed changes do not affect the safety design function of the chillers. No change is required to the accident analysis of the FSAR based on review of sections 9.2.1, 9.2.9 and 15.0.
- 3. The essential HVAC systems served by the chillers will continue to maintain acceptable ambient air tmperatures. The proposed changes do not affect the margin of safety as defined in Technical Specification sections 2,3 and 4, and their bases.
87-VlE0206 his DCP will revise the diesel fuel oil storage tank low level alarm setpoint from 45" + 3.6" above the centerline of the tank to 56"~+ 3.6" ~ above the centerline of tank.
- 1. This design change gives an advance warning to operations that the Technical Specification limit of 68,000 gallons is being reached and does not increase the pruoability of occurrence or consequence of an accident described in the FSAR.
This includes a review of FSAR chapter 15, 16, and 9.5.4.5. Although the alarm function is not part of the Technical Specifications, it is a good practice to be forewarned of the inpending Technical Specification condition. 7
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. h is design change does not create the possibility' of an accident or equipent/cmaonent m1 function not described and analyzed in tae FSAR. his includes a review of the Failure Modes and Effects Analysis (Table 9.5.4-2) and FSAR section 9.5.4.3.
L e function of the two level switches.(LIS-9022 and ILIS-9023) has not been changed and no safety liinit settings or setpoints are affected.
- 3. This design change does not decrease the trargin of safety defined by the bases of the Technical Specifications. There is no change to the bases-of Technical Specification as listed below:
Section 3/4.3 Section 3/4.7-
-Section 3/4.8.1 87-V1E0209 This change involves replacement of the original.
E hinge pins in valves 1-1305-U4-052 and 060 and 1-1304-U4-002, 005,008, 012, 172, 175, 178 and - 182. The new hinge pins-are of the sane sterial as the originals. Howver the thickness and
-diannter of the head has-been increased and the overall length has been increased. Valves L 1-1305-U4-052 and 060 are located-on the Train A&B-Feedwater Pump discharge lines and serve to protect the pumps fr m reverse flow. The B04 syst m valves i are on drain lines off of the reheater drain tanks (A through D):and are designed to prevent _possible backflow of water into the drain tanks in the event of a feedwater heater tube rupture.
- 1. m is change does not affect systs function or
- operation and therefore does not increase the probability of occurrence or the consequences of any accident described in sections 10.4-or
'15.0.
- 2. This change does not-affect any systs, equipnent, or cmponent's function or operation and based on a review of sections 10.4 and 15.0, would not create-the possibility of an unanalyzed or undescribed g accident, f or equipment /cmponent rnalfunction.
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11 1990 A!NUAL REPORT - PART 2 10 CFR50.59(b) REl%T
- 3. his change does not affect any systun/equipnunt function or operation and therefore does not affect the safety margin defined in Technical Specification section B 3/4.7.
87-VlE0216 Additional INAC is added to the following elevator equipnent rooms to prevent heat buildup: 1) R-502 (Control Bldg) 2) R-217 (Aux Bldg) 3) R-301 (Aux Bldg), he additional INAC consists of a fan / coil unit located in each equipnent room with an associated condensing unit located outside of each roan. Condensing units for rooms 502 and 301 are located outside of the building on the roof. The condensing unit for roan 217 is located on the roof of the elevator equipnent roan which is enclosed in the fuel handling building corridor, Roan 127. There is no safety related equipnent or safety function associated with the elevators.
- 1. The additional IWAC equipment is not associated with any systan assuned to function in the accident analysis of section 15.0 of the FSAR. The additional equipnent operates independently fran other INAC systems for ' he Aux. Bldg. and Control Bldg. referenced in FSAR 9.4.1 and 9.4.3. As identified in revised Table 3.5.1-1, there is no missile that can be postulated s ich will adversely affect any equipment asstned to function in the FSAR accident analysis.
- 2. The tNAC equipment added by this change is non-safety related and does not serve safety related equipment asstned to function in the accident analysis (FSAR section 15.0). We INAC equipmnt serves only the elevator-equipnunt,
- 3. This change does not decrease the nurgin of safety defined in the Technical Specification 3/4.7 bases.
87-VlE0217 h is design change will involve the replacement of low tank level annunciation of the liquid hydrogen storage tank. The present switch will be replaced with a differential pressure switch. The 9
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II 1990 ANNUAL REPORT - PART 2 -l 10 CFR50.59(b) REPORT requirements for hydrogen service are satisfied since the replacumnt switch mets the explosion proof class I, grou) B, C & D specifications. We project class for tae switch is 62J non-safety related, non seismic.
- 1. Le replacemnt of the level switch was perfonmd l to allow for m re reliable annunciation of tank low '
level conditions. The replacamnt switch mets the same standards as the original switch. FSAR section 9.3.5.1.1 states that the Auxiliary Gas System does not arovide any safety function. By using a switch walch mets the intent and design of the original switch, the probability of
. occurrence, or the consequences of an accident or m lfunction of equipment important to safety is not increased.
- 2. he Auxiliary Gas System is not relied upon to provide any safety function as stated in FSAR r section 9.3.5.1.1. As the new switch mets the original design intent and specifications, no new t accident scenarios or malfunctions of any type not ,
previously evahtated in the FSAR are created.- 1 L There is no effect on the chapter 15 analysis. L L 3. Station Technical Specifications 3/4.3 and 3/4.7 do not address the-Auxiliary Gas System or any components within the system. Therefore, the change does not reduce the. margin of safety defined in any Technical Specification bases. 87-V1E02?9 This; design change mdifies the operation of , tornado damper 1-1533-W7-407 located in the Control building freight elevator shaft. The existing design requires the tornado damper to remain in a normally closed position and open upon a signal to vent by the local fire zone indicating panel. Eis
-design change will disable the damper actuator which will allow the damper to be held in a
-normally open position by the damper's springs.
he tornado damper will close only on an excess outflow velocity which would be caused by a tornado. 10
11 1990 ARO\L REPORT - PART 2 10 CFR50.59(b) PH OKr
- 1. We subject tornado damper is required to close when subjected to excess outflow velocity caused by a tornado. This design change has no affect on the damper's intended safety function by changing fran normally closed to nonnally open. We damper actuator being disabled is designed to open the tornado damper during a fire to exhaust st:oke fran the elevator shaft. The danper actuator performs no safety function. This design change does not increase the probability of a malfunctlon to this danper or any equipment assuned to function during accidents. This includes a review of FSAR secticas 9.4, 9.5.2 and 15.0,
- 2. This design change does not prevent the t enado dauper fran closing when required by excess outflow velocity. The elevator shaft will continue to have an exhaust path for smoke renoval provided by the normally open damper configuration. This change does not create any new accidents not described in the FSAR since there is no change to the safet'f function of the tornado damper. 'lhis includes review of FSAR sections 9.4, 9.5.1 and 15.0
- 3. h e nargin of safety as defined in the Technical Specifications bases sections 2, 3 and 4 is not affected by the proposed change. he safety design function of the danper is not affected.
87-VCE0230 he CVCS camon chiller unit A-1208-E6-008 control circuit is designed so that the chiller can be started either fran Unit 1 MCB or fran Unit 2 MCB with proper indication of indication lights on the control handswitches located in both MCB's (Unit 1 and Unit 2). These lights indicate the camen chiller unit status point of starting (fran Unit 1 or fran Unit 2), Presently, the two indicating lights are wired in series with the chiller starting relay coil and because of insufficient voltage drop across the starting relay coil, the chiller can not be started with proper indication. This DCP is to correct the wiring of the catman chiller control circuit so that the chiller can be started with 11
II 1990 ANNUAL RElORT - PAIG' 2 10 CFR50.59(b) REPOPJr proper indication as designed, by rewiring the internal wiring of the control handswitch 1HS-0390 and the control circuit wiring to agree with the approved EFCRB-16899 of the chiller elumntary diagram drawing AX3D-BD-C01E.
- 1. The change inproves the CVCS cctuon chiller operation reliability, and does not increase the probability of equipmnt unlfunction. The CVCS camon chiller is non-safety related and is not assuned to function during an accident. This includes review of FSAR Chapter 15.
- 2. The CVCS ccunon chiller is non-Q. This change inproves chiller operation reliability and does not create the possibility of an accident or equipment nulfunction not described in the FSAR. This includes review of FSAR section 9.3 and 15,
- 3. lhe CVCS ccumon chiller is non-safety related.
This change inproves system reliability. Therefore, the change does not decrease the nargin of safety defined by the bases for Technical Specification. This included review of the Technical Specification bases of 3/4.4 and 3/4.7. 87-VIE 0231 This change involves changing the span of preasure transmitters 1PT-6018A, 1FT-60188, 1PT-6168A and 1FT-6168B. This change requires calibration changes only, there are no hardware che ges required. These instnnmnts are pressure transmitters associated with the reheater steam which perform no safety-related function.
- 1. The change to the turbine generator canponents referenced does not increase the probability or consequences of an accident based on review of FSAR section 10.2 or chapter 15.
- 2. The change does not present a failure mode or impact any ccuponent not previously reviewed in analysis presented in FSAR chapter 15.
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- 3. The Technical Specification bases do not restrict the span of these transmitters in order to maintain L safety margin.
VCE0235 This modification adds-spent resin dewatering capability to the condensate filter deminerali::cr systm (1414). This dewatering system will contain two independent trains for holdin6 and dowatering spent resins frm the condensate tilter domineralizers. Each train will consist of a holding tank, dewatering puup with two cartridge type filters, recirculation pump, filter pump, pressure filter skid-ass mbly, and associated piping, instrumentation and controls. The sment i resin dewatering system will be located in the Turbine Building, level A, primarily in the IE quadrant. It will be a camon- system and will prce m all' Unit I and Unit 2 condensate filter demineralizer spent resins. I 1. This change is to add spent resin dewatering capability to the condensate filter danineralizer system (1414). This change does not increase the-probability of occurrence or the consequences of 1 any equipment important to safety. There is no ] affect on e ipuent or components which are assmed to ction in'an accident as analyzed in i the FSAR, section 15,
- 2. All systems that are impacted by this change (1414, !
1418,- 2401, 2420, 2301, 2419) are non-safety related, and not required to function in the event of an accident. No new accident type is-created by this modification. The FSAR is not _. compromised by this change. This included review of chapter 15 accident analysis.
- 3. This change is asacciated with systems 1414, 1418,-
2401, and 2420. There is no decrease in Technical-Specification safety margins'since these systems have no safety design bases. Reft. Technical ecification bases review, sections 3/4.11.1, Sp/4.11.2,-3/4.11.3, 3 3/4.7, 6.12 and FSAR review, sections 9.3.1,- 9.2.3, 10.4.6, 15.1 and 15.7. 1 L l L 13
l II ! 1990 AIM 1AL REPORT - PART 2 10 CFR50,59(b) REPORT 87-V1E0239 The turbine building smp pmps 1-2412-P4-003, 4, 5 and 6 are being replaced with four submrsible sum pups capable of flowing approxinutely 300 gpm at 100 ft.
- 1. '1he turbine building drain systen is nvt safety-related and is not relied upon to mitigate the consequences of an accident, The increased flow rate will not affect the ability of radiation mnitor RE-848 to aerfonn its intended design function. Thus, tae capability of the turbine building drain systen to direct the pmp discharge to the turbine building drain tanks, in the event of contamination of the effluent, will not be affected by this modification. Therefore, this modification does not increase the probability of occurrence or consequences of the malfunction of any equipment or emponent assumd to function in accidents analyzed in the FSAR. FSAR sections 9.3.3 and 15.0 of the FSAR wre reviewed.
- 2. This system is not safety related. Its canponents cve not required to function in post-accident conditions. Failure of the smp pmps will have no effect on the capability to operate the plant safely or to accmplish safe shutdown. In addition, this change has no effect on the FNEA presented in FSAR section 9.3.3. Therefore, this modification does not create the possibility of an accident or equipment /cmponent unlfunction not described and analyzed in the FSAR.
- 3. This change involves sunp pump and nutor replacement for a system which is not described in the Technical Specification. It does not decrease the margin of safety as described in the bases of the Technical Specifications. Sections 3/4.7, 3/4.11 and 3/4.12 were reviewed.
87-V1E0247 A pressure point saaple connection will be added to lines 1-1574-505-4" and -506-4" downstream of SJAE 'A' and 'B'. The root valve for the 14
11 1990 /MUAL REPORT - PART 2 10 CIR50.59(b) REIOKr connections will be a globe valve and a 3/8" tube
-to-pipe adapter with cap and will be installed on the valve outlet to allow for test instrtment connection.
- 1. The design change does not create a nulfunction mde of any system or emponent not previously analyzed in FSAR chapter 15,
- 2. This design change does not effect plant operations or accident conditions analyzed in FSAR chapter 15,
- 3. 'lhere is no change to the bases of the Technical Specification in section 3/4.7 plant systens.
87-VlE0263 The MFIV air reservoir pressure switches 1PS-5227B, 1PS-5227C, 1PS-5228B, 1PS-5228C, 1PS-5229B, 1PS-5229C, 1PS-5230B and 1PS-5230C are currently Barksdale Fbdel BlT-A12SS.
'Ihese are being replaced with Barksdale Model BlT-lil2SS pressure switches. The vendor (Anchor / Darling Valve) reccmnends that the pressure switches be changed to a model with a smiler differential (deadband). The new pressure switches are qualified to the saae requiretrents as the original pressure switches and will not invalidate the MFIV qualification.
- 1. The operation of the MFIV's will not change as a result of implettenting this DCP. Therefore, the MFIV's can still be assumed to function in the accidents analyzed. There is no change to the F%R analysis of chapter 15 or 6.2. The new pressure switch meets the sme requirements as the original switch. This change will not increase the probability of occurrence or the consequences of an accident or malfunction of equipuent important to safety that has been previously evaluated in the 1%\R.
- 2. The new pressure switches are qualified to the sane requirenents as the original pressure switches. With the only difference being 15
l i l Il 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REP 01E the switch deadband, this will not create the-possibility of an accident-or malfunction of a - .different type not areviously evaluated in the FSAR. Therefore, taere is no change to chapter 6.2, 7.3, 10.4 and 15 analysia.
- 3. By using pressure switches of the same proven quality as the original pressure switches, there will'be no reduction in safety margin as defined in the Technical Specification bases. There is no change in the Technical Specification bases of 3/4.6.3 and 3/4.3.
4 87-V1E0278 Modify.the Steam Jet Air Ejector (SJAE) Radiation Monitor as follows: 1
--Replace existing filter holder with SAIC filter holder that will allow for use of quick disconnect. -Add flex-line between new filter i holder and cooler. 4 I
- Add low point drain addition at discharge of double heat pump.
- Add a flow totalizer for the sample flow stream.
- Modify piping betwen cooler and gas ~ detector.
- Replace filter entrance piping heat trace with
'TZ-10 super flow" set at 215 F.
- Heat trace midstream piping frcin cooler outlet including the gas detector with TZ-5 heat tracing set at 120 F.
- Add water source to existing loop seal.
- 1. 'Ihis change does not affect the performance of any equiptmnt assumed to function in the FSAR accident analysis. 'Ihis-included review of FSAR cha and 15. There is no impact on the monitor'pters s 11
. capability of responding to a postulated steam generator tube rupture described in FSAR section 15.6.3.
-2. 'Ihis change does not create the possibility of any.
malfunction not described and analyzed in FSAR section 15 and 31.5. 16
l II 1990 Nt4UAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3. Based on review of bases 3/4.3.10, " Radioactive gaseous effluent nonitoring "instrmentation", and 3/4.11,2, " Gaseous Effluents , this change does not affect the nargin of safety as defined in the bases reference section of Technical Specification.
87-V1E0310 This design change adds air release valves, vent valves, and associated piping to the water side of the Main Generator Hydrogen Coolers. This piping is part of the Turbine Plant Cooling Water System, Systen 1405. It is non-safety related, project class 626. This system has no safety function. Failure of this systan will not empranise the ability of the plant to accmplish a safe shutdown.
- 1. Equipment affected by this design is not assmed to function in an accident analyzed in section 15 of the FSAR (Accident Analyses). Installation of the proposed change will not cause the malfunction of other equipuent assmed to function. The Flooding and Seismic analysis is not impacted. The new piping and valves are not safety related or seismic category 1, and neither are the existing piping or valves. This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
- 2. This design change does not create the possibility of an accident or equipment / component malfunction not described and analyzed in the FSAR which could affect the health and safety of the public. This was based on a review of FSAR section 15 (Accident Analyses), section 9.2 and 10.2. New cmponents are added to improve the function of the existing sys tans .
- 3. This change is associated with system 1405 and does not decrease Technical Specification safety margins since it has no safety design bases. This !s based on review of Technical Specification bases, including section B 3/4.11.
17
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II 1990 ANNUAL REPORT - PART 2 a 10 CFR50.59(b) REPORT -k 87-V1E0321 Two recorders (1PIUR-41321 & 1PR-41302) will be remved from existing backflush filter panel 1-1224-P5-FBP and be replaced by nore reliable new solid state switches (Bi-stables), located near the backflush filter panel. This new panel will contain the bi-stables necessary to provide contact output for the panel annunciator, and are also in
- coman with the Control Roan annunciator. The-alams processed in this new panel will be Crud Tank high, high-high and filter differential high (for each backflushable filter).
- 1. This change does not affect operation of the
.f -backflushable filter system. The change provides nere direct indication of systm status to the operator. Therefore, this design change will not ;
increase the probability of occurrences or
=
consequences of an accident described in FSAR section 15,
- 2. This change does not affect backflushable filter system operation. This change provides a different nethod of alanning system parameters. Therefore, this change does not create'the possibility of any malfunction not-described and analyzed in FSAR section 15.
- 3. Based on review of Bases for Technical Specifications 3/4.3 and 3/4.11, this change does -
not affect the mrgin of. safety. 87-V1N0341 This modification will provide the addition of
. personnel platform structures and associated
-ladders for enhanced access to essential valves,-
controllers, and other equipment inside the. Unit 1
' Turbine Building.
- 1. The proposed change does not. increase the probability of occurrence or consequencestof any equipment malfunction for equipment assumed to function in FSAR chapter 15 accidents.
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4 II 1990 ANNUAL REIORT - PART' 2
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- 2. L e proposed change does not create the possibility of an-accident or equipment malfunction not described in the FSAR. Bis included review of FSAR sections 3.0 and 15.
- 3. The proposed change does not decrease the unrgin of safety defined in the Technical Specifications.
This included review of bases for Technical Specification 3/4.7 87-VIN 0342 Add float check valve tag number 1-2403-X4-419 to the pneumatic line associated with instrunent 1LI-19192. Add float check valve tag nunber 1-2403-X4-420 to the pneumatic line associated with instrument 1LI-19193.
- 1. The check valves 1-2403-X4-419 and 1-2403-X4-420 are non-safety related valves serving a non-safety related funcKon in Project Class 424 pneuantic tubing. his change will not increase the probability of occurrence or consequences of the j malfunction of any equipment-or e sponent assumed to function in accidents analyzed in FSAR sections 15.0 and 9.5.4. i
- 2. Le check valves 1-2403-X4-419 and 1-2403-X4-420 are non-safety related valves serving a non-safety related function in Project Class 424 aneumatic tubing. Bis change will not create tae possiaility of an accident or equipment /conoonent malfunction not described and analyzed in tTe FSAR sections 15.0 and 9.5.4:.
-3. The proposed change does not decrease the margin l of safety defined by the bases of the Technical l Specification 3/4.8.1. -l
'87-VIN 0351 Install valves 1-1407-U4-207, 1-1407-U4-208, i 1-1407-U4-209, and associated piping on the steam '
generator blowdown system downstream of the existing backflushable filter. Valves 1-1407-U4-207 and 1-1407-U4-208 will be normally closed while valve 1-1407-U4-209 shall-be nonnally open. 19 we
- m rev ye- --- -cgg- *g g-$p+t+_
W II 1990 ANNUAL REIORT - PART 2 10 CFR50,59(b)- REIORT.
- 1. Eis mdification does not affect the probability of occurrence or consequences of any equipnent '
mlfunction or ccmponent assumed to function in an FSAR postulated accident. B is included review of FSAR chapter 15-accidents.
- 2. Although this design does require a new ccxaponent installation, the portion of the Stemn Generator where this ccuponent is to be installed is non-safety related. There is no new component malfunction created by this change that has not been considered in the existing FSAR chapter 15 analysis.
- 3. There is no change to the Technical Specification 1 margin of safety. 'This included review of the
. bases of Technical Specifications 3/4.3.3, 3/4.4.5, 3/4.7.2, and 3/4.11.1.
87-VIN 0358 The Diesel Generator (D-G) building oily waste I sump level switch float displacer suspension cable
-is~too long~because the last dis 31acer is sitting on the bottcxn of the sump when tie water level is at the low level setpoint. The low level switch setting is raised to correct the problea without changing the displacer suspension cable.
- 1. This design change has no affect on accidents postulated in the FSAR because failure of the level switch will not initiate any. accident described in L
=
the FSAR. ' Design elange insures the accurate operation of the D-G building oily waste sump level '- switch low level setpoint. Therefore, this change does not increase the probability of occurrence or-consequences of an accident described in the FSAR.
- 2. No new potential accidents or events-are created as' a result.of this modification. Changing the low level setpoint does not affect other systems to s
create the possibility of an: accident or equipment-malfunction not described and analyzed in the FSAR. 20 i L
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II 1990 N4NUAL REPORT - PART.2 10 CFR50.59(b) REIORT
- 3. There are no Technical Saecification bases defined or inferred for the D-G 'auilding oily waste sup level setpoint, herefore, the setpoint change does not decrease the margin of safety defined by the bases of the Technical Specification.-.
87-V1N0385 This DCP changes the AC power feed of an emergency sealed bean lighting fixture fran 1NLP36-13 to INLP31-4. This fixture is located in Roan 123 of the Unit 1 Control Building and it's laupheads are directed at valve 111V-3019. This 8 hour rated battery fixture provides unergency illunination for local manual operation of valve llW-3019, 11N-3019 is the Steam Generator #2 outlet to the Turbine Driven Auxiliary Feedwater pump. The new. power feed will be field routed fran an adjacent energency fixture that also has it's AC power feed fran 1NI)31-4. Le new feeder will be ALS (aluminun sheathed cable). The existing AC powr.
-feeder is also ALS. It will be abandoned, deteuninated, wires cut and taped.
1, This change does not affect safety systems, setpoints or other equipment or cauponents assumed to function-in accidents' analyzed in the_FSAR (sections 8.3.1, 9.5.3 and 15). This change allows for the proper operation of the emer system (FSAR section 9.5.1, 9A.I.64 7). gency lighting
- 2. The changing of the AC power feed to this emergency light does not create the possibility of an accident or equipment /canponent malfunction not-described and analyzed in the FSAR (sections 8.3.1, 9.5.3.-9A.1, 15).
- 3. The margin of safety as defined by the bases of the=
p Technical Specification (section 3/4.8)-is not affected by this change. 87-V1E0392 This a " SAFEGUARDS" DCP, the following abbreviated information has been declassified for this~ report. Add steel collars to main feedwater lines to eliminate a protected area to vital area security breach. 21
_ . _. . _ - __ ._ -._~ _ _. _ _ . _ . _ _ _ - _ _ _ . _ _ _ _ _ _ . . . i l II 1990 AhWUAL REPORT - PART 2 10 CFR50.59(b) RENRT
- 1. Le addition of steel barriers around the min feedwater lines has no impact on equipent important to safety. H erefore, there is no increase in the probability of occurrence or ,
consequences of an accident or malfunction of equipment important to safety previously evaluated . in the FSAR section 15. l l
- 2. The installation of these steel security barriers does aot increase the possibility of an accident or malfunction of a different'tvp j previouslyevaluatedintheFSAke-thanwas This was I determined by performing a review of all applicable FSAR sections including 3.6 to 3.11 and 11.3. -Other factors considered in the development and final implementation of this DCP included an interference and material consideration for possible seismic issues.
- 3. Technical Specifications do not address the plant security s herefore,ystm or barrier recuirments, this modification coes not reduce the margin of safety as described in the bases for any Technical Specificacion.
87-VlE0408 21s DCP replaces various piston type check valves l with flapper type check valvas in the systems ! identified below. Le sumps identified perfom the ' same generic function of collecting waste effluent l and processing it for recycling or disposal. !
- System 1214 . Containment and Atociliary Bldg.
Drain-System - Radioactive _
- System 1215 - Auxiliary Bldg. Drain System -
Non-Radioactive !
- System 1225 - Control Building Drain System
- System 1227 - Fuel Handling Building Drains
- System 1420 - Waste Water Effluent System
- 1. L e probability or consequence of the malfunction of any_ equipment is not changed as a result of this mdification. Rese systems are not assumed to function in accidents analyzed in the FSAR. This is based on review of section 9.3,11 and 15.
22
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1990 N MIAL REIORT - PART 2
- 10 CatSO.59(b) REIORT
- 2. This nodification does not create the possibility of ar.y accident or taalfunction timt is not described and analyzed in the FSAR. This is based on a review of FSAR section 9.3, 11 and 15.
4
- 3. 1here is no decrease in the m rgin of safety defined by the Technical Specificatico as e result of this txxlification. 1he bases for Technical Specf.fication 3/4 3.3.9, 3/4 7.1.', 3/4 11.1 and 3/412.1 are not affected by this nrxlificatico.
87-V1E0409 This IL'P adds extended rurote operators to valves 1-1215-04-294, 277, 269, 267, 265, 263, and 261 of the Auxiliary Building nmaradioactive drain systun located in the Auxiliary feedwater punp house. The , principle function of the systan is to drain nonnally non-radioactive equiptent and floor liquid waste fran cpen areas of the auxiliary building to the floor drain tank via the auxiliary building or 4 the penettstion roun suup. The auxiliary building non-radioactive drain systun is not required for . the-safe shutdown of the plant. All new catponents are located in roan k-104, b 1. The probability or consequence of the malfunction of .any equipamt is not changed as a result of this rtodification.- The externally added reach rods will be supplied with torcue-limiting carponents. This syntua is not assuret to function in accidents analyzed in'the FSAR. There is no change to the accident analysis of chapter 15 as a result of this nodification.
- 2. This undification does not create the possibility of any accident or m1 function that is not described and analyzed _in the I M . The valve reach rcxis will have their own supporting systein and torque-limiting cauponents. 1here is no clmnge to any catponent malfunction analyzed in the FSAR.
This is based on review of FSAR sections 9.3,10.4 and 15. 'I 23
, _ _ . - .a,_._._-.__.2___________ _ _ _ _ _ _ _ _
t i 11 i
- . 1990 ANNUAL PE10RT - PART 2 10 CFR50.59(b) REPolE 1
) 3. There le no decrease in the trargin of safety defined by the Teclutical Specification as a result of this nodification. The bases for Technical l Specification 3/4 3.3.9, 3.7.1,2 and 3/4 11.1.1 is l not affected by this nouification. 1 l 87-V1N0435 This DCP will rmove IPV-57950 frce t.he control loop for valves IFV-5795A & B. Tubinr will be - installed in place of 1FV-5795C to all,ow 1.1C-5795 9 to directly control IW5795A & B,
- 1. The Auxiliary stercn systen is non-safety related and not assuted to operate in an accident. This .
change does not affect any safety related ! equi} ment or safety systems settings. This is based on a review of FSAR sections 15.1, 15.2, L 9,5 and 10.1.
- 2. This design change will not affect plant conditions.
Also, as a reSJ1t of a review of FSAR sections 15.1, 15.2, 9.5 and 10.1, there is no FSAR accident condition chsage.
- 3. This design change does not affect the Technical Specification bases section B3/4.7.
87-VIN 0452 This DCP adds hour noters to uraitor run tino of each ESF fan supplying flow through the charcoal of 1 the ESF filtration systems. Each noter provides positive indication of fan run tine for the FEF charcoal filter. Thi.s design uses a contact of the_480-volt load center breaker or 480 volt MCC breaker in conjunction with an auxiliary relay to energize the hour noter cicult when ESF fans are operating. These meters are installed in wall mounted panels. Each noter is powered by a non-Class 1E circuit which is separated frce the Class 1E circuit by an-auxiliary breaker position switch or contactor position contact for the fan control circuit. Installation of there neters will not have any detrinental effect
- on the ESF filtration systen operation, t
L 24
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s 4 II 1990 ANNUAL RrJORT - PART 2 10 CML50.59(b) REFORT i
- 1. This design does not change the operation of the
; affected equipmnt described in FSAR sections
! 9.4.1, 9.4l2 and 9.4.5. 1he new equipnent is pwered frun a non-Class IE source scaarated fran the other Class 1E equipmnt. Taerefore, the probability of occurrence or consequences of the m1 function of any equipmnt or emponent i assumed to function in accidents analyzed in the ; FSAR is not increased.
- 2. The new equip ent and associated power sup)1y is I non-Class IE and separate fran the Class 13 l eculpmnt. Failure of this equipment will not atversely inpact the operation of the umrgency filtration syntan and other safety systen equipmnt. Iherefore, the change will not create the possibility of an accident or equipment /
cauponent nalfunction not described and analyzed in the FSAR.
- 3. No chang'e to Technical Specification section 1 3/4 7.6, 3/4 7.7, and 3/4 9.12 is necessary !
as a result of this design change package. This ! design is to enhance m nitoring the usage of the energency filtration systen to ensure Technical Specification surveillance requiremnts are satisfied. Therefore, there is no decrease in the i nargin of safety defined by the bases of the I l Technical Specification, 88-V1N0011 This design package involves installing a resistive voltage divider network in the de-excitation cubicle of the Generrex panels. The outputs of this netwark are to then be routed over to the plant fault recorder located in the control roan in order to provide a ground reference. The inputs to the network will be-connected to the generator field supply with.the circuit rated for 2 KV service as stipulated by General Electric. The canpanents of this network will be assunbled cm an insulated industrial 31 ate munted inside the de-excitation cubicle of t ae unin generator Generrex excitation cabinet.- i l-i 25 ( l b-.,.---.-.--,_,,_.__. __ _ _ _ , _ . . , _ . . . . . _ _ _ _ . . _ _ , . , _ _ , _ _ _ , . , _ _ . _ _ _ _ , , _ __
y. II 1990 A!MlKL RDORT - PART 2 10 CtK50.59(b) RDMT
- 1. The only function this change serves is to relay information after a trip by the generator due to a grid fault - it will not affect any equi ment i or emponent asstnod to function during an FSAR analyzed accident. Review section 15, 8, and 10.2.
- 2. Again, this change provides post turbine trip ground fault detection information. There is no equitrent required to mitigate the effects of any Qiapter 15 accident which will malfunction due to this change.
- 3. The bases of the Technical Specifications do not address the plant fault recorder nor the de-excitation cubicle. Review Technical Specification sections B 3/4 7.1 and B 3/4 8.4.
88-V1N0020 This is a "SAFTEUARDS" DCP, the fo) lowing abbreviated information has been declassified for this report. Add security barriers to penetrations in the auxiliary buildin;;.
- 1. The change adds security barriers over existing openings in the auxiliary building. There is no equil ment inportant to safety involved. Therefore, there is no increase in the probability of occurrence or the consequencea of and accident or im1 function of equipuent inportant to safety as described in FSAR section 15.
- 2. Only security barriers are added, no equilment inportant to safety is involved. Therefore, there is no possibility of a different type of accident that needs to be evaluated i the safety analysis report.
- 3. Technical Specifications do not address the plant security system or barrier requiranents.
Therefore, this modification does n,)t reduce the margin of safety as described in the bases for any Technical Specification. 26
F t. P II 1990 ANNUAL REPORT - PAKf 2 10 CFR50.59(b) kDORT 88-V1N0023 Install reach rods on equipnent drain valves 1-1218-U4-080 (CVCS Centrifugal Charging Puup Pain, 1 > Train "B"),1-1218-U4-081 (CVCS Centrifugal ! Charging Ptup Fem, Train "A"), and 1-1218-U4-082 (CVCS Positive Displacment Puup Roan) and install i chain operators on valves 1-1218-U4-068 (RHR lient i Exchanger Roan, Train "B") and 1-1218-U4-069 (PJIR i lleat Exchanger Roan, Train "A"). Tnese valves will
- still be required to be locked closed.
- 1. We addition of reach rods and/or chain o3erators-m the subject valves will not increase t,e i
probability of occurrence or consequences of-the malfunction of any equipnent or canponent assuned ! to function in accidents analyzed in the FSAR. his included a review of section 9.3.3 and chapter i 15 (Accident Analysis).
- 2. A review of FSAR section 15.7 shows that this
-design change does not create the possibility of an accident or equipnent/cauponent malfunction not described anu analyzed in the FSAR. There is i no change to the negative pressure boundary or flood retaining features of this roan, j' i
his design change does not decrease the margin of 3. safety defined in bases of Technical Specification 3/4.11 or 3/4.3.3. 88-VIN 0026 This is a " SAFEGUARDS" DCP, he followin6 abbreviated infonnation has been declassiried for this report. This nodification adds seven (7) barriers to penetrations and vents that breach vital areas. i
- 1. The addition of steel barriers to secure
. penetrations into vital areas has no impact an l equipment-inportant to safety. Therefore, there is no increase in the probability of occurrence ,
or consequences of an accident or malfunction of equixnent important to safety previously evaluated-l in tae FSAR section 15. L l l-27 i-L
a i a, a 11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. Only security barriers are added, no equipuent inportant to safety is involved. There is no possibility of a different type of accident that needs to be evaluated in the safety analysis report.
i
- 3. Technical Specifications do not address the plant security systan or barrier recuircuents.
Therefore, this mdification coes not reduce the margin of safety as described in the bases for any
- Technical Specification. ,
88-VIN 0035 The proposed chango will pernanently' incorporate the mdifications made by UR's 1-87-269 and 1-87-402, which provfded the changes necestary to drive the i- CRT on the main control board fran the nnergency Response Facilities (ERF) ComputJr. This was acccuplished by running three (3) coax cables fran the D W display on the Shift Supervisor's stand to l the CRT on the QCB and disconnecting the three Proteus cables at the back of the CRT. \
- 1. Since the ERF and Proteus ccuputer systans are not assured to function in an accident, the proposed change does not increase the probability or !
occurrence or consequences of the malfunction of i any equipcent or canponents assured to function in the accidents analyzed in the IG R. 1
- 2. The proposed change does not create the possibility- j of an accident or. equipment /ccunonent malfunction ;
not described and analyzed in tae FSAR. The canputer systems perform nonitoring functions only and are not required in any accident scenarios.- ,
- 3. The proposed change involves the ERP and Proteus L ccuouter systems, which are not addressed in the Technical Specifications. Therefore, this change does not decrease the nargin of safety defined by the bases of the Technical Specificattons.
28 ~._ _ -.- _._.,.....___.;_.________
11 1990 M4NUAL REIOitT - P/Kr 2 10 CFR50.59(b) REIOKr 88-VCN0056 311s change adds an overhead bridge crane of 40 ton capacity to the existing Alternate Radwaste Building (ARB). This crane replaces a hydraulic crane of lower capacity (Ref. ICP 88-VCE0061-0-1) . The nodification enccopasses the folicving work activities:
- 1) Core bare for anchor bolts
- 2) Crane support structure design / installation details
- 3) Electrical wiring and conduit installation for a) Pcw r b) Kurote control panel c) Video cuaera systen
- 1. Addition or the bridge crane does not change the consequences or probability of malfunction of any equi cent asstmed to function for accidents analyzed in tle FSAR. This is based on review of sections 9.1.5 and 15,
- 2. No new equipnent m1 function or accident is created as a result of this change. 'Ihis is based on review of the heavy load analysis of 9.1.5 and chapter 15.
- 3. Diere is no change to the mrgin of safety as defined in the bases of Technical Specifications 3/4 11 and 6.14, 88-V1N0070 This DCR inserts a RFIAY DRIVFR board and a hETT MODULE board into each diesel generator (DG) load sequencer logic cabinet (1-1821-U3-001 for train A and 1-1821-U3-002 for train B) . 'lhis modification will allow the IG secuencer to be reset by an operator should an unt,ervoltage start sequence occur followed by a restoration of offsite power during the DG loading sequence.
- 1. This design chant;e reduces the chance of a malfunction associated with the diesel generator logic sequencer since the ability to reset the 29
4 3 ] II
, 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT ; sequencer after restoration of pcwr, follcuing i a loss-of-offsite pcwr, has been en1 anced and the conse reduced. quences of a nn1 function have been
- 2. The change of the logic cabinets does not alter the sequence
- systun operation as described in the FSAR. 'ine nanual reset function is inhibited for 60 seconds during the undervoltage start and l
loading of the DG sequencer. Therefore, the new design does not create an accident or a un1 function not described in the FSAR. , I
- 3. This change effects the nanner of restoring the DG sequencer lo61 c clock to the nonnal node when pcur )
has been restored, follcuing a loss-of-power, and ' does not change the nnrgin of safety as defined in sections 3/4.8 and B 3/4.8 of the Technical Specification. 88-VIN 0079 The Unit 1 Control Building Control Rocxn (CBCR) ESF Nuclear Filtration Units (I-1531-N7-001/002) were nodified to produce a lead / lag fan control logic to prevent nure than one fan frcin running when a Control Rottu Isolation (CRI) signal is present. The associated outside air dmpers will be changed ! frmi tbrun11y Closed-(NC) to locked Open (LD). The CBCR INAC isolation dmpers (11N-12146, 47, 48, 49, 2tN-12146, 47, 48, 49) will close inuediately upon recei)t of a CRI signal frcin either Unit 1 or Unitb.
- 1. This design change does not change the consequences or probability of any equipnent malfunction, for j- equipnent assmed to function in an FSAR postulated accident. This design change does not adversely ,
affect the capability of- the control rocxn energency INAC systan frcxn neeting the design requirments of FSAR section 6.4.
- 2. This change does not create the possibility of an accident or equiptent malfunction that is not described and analyzed in the FSAR. There is no single failure that prevents the control- rcun mergency INAC systan frcxn meeting the design regulrunents of FSAR section 6.4.
30
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. 1990 A MUAL REPORT - FART 2 l 10 CFR50.59(b) REIURT
- 3. '1he bases of control rcxui !!VAC Technical Specification 3/4.7.6 and ESTAS instrtrentation Technical Specification 3/4.3.2 do not change.
After this design change, the Technical Specifications retain the stre criteria. The criteria of 10CFR50, GDC19 are net with the design. Therefore, there is no decrease in the mrgin of safety as defined in the Technical Specifications.
'lhere is a change to one of the tindng parameters, FSAR Table 16.3-2, referenced in bases of Technical Specification 3/4.3.2. This does not affect the urgin of safety, since with t.he new tine parmeters, the safety analysis still neets GDC 19 requirments.
88-VIN 0088 The presently med URAL board and the AC Volts / URAL Cate boards in panel 1-1328-F5-GRC will be replaced with redesigned boards. *1he exchange of boards involves adding tw wirca to the backplate of the regulator control section, installing jtnpers on both boards, adjusting pots on the URAL board (as outlined in G.E. change notice ECN G314-028 and Instruction Book IX4AA01-275) and a perfonittnce check with the new boards in place (also per ECN G314-028). Alac, this change provides doctmentation to mke pennanent a Synchronizer III board which is presently installed by tenporary mdification 1-88-046.
- 1. The proposed change does not increase the probability of occurrence or consequences of the nulfunction of any equipnent or ecoponent asstned to ftriction in accidents analyzed in the FSAR.
This included a review of section 10.2 and chapter 15 of the I W ,
- 2. The proposed change does not create the possibility of an accident or equi} ment /ccuponent m1 function not described and analyzed in the I M . This change does not alter operability of the system and malfunction of the equilment/cmponents will not change due to this axilfication. This includes a review of the followin!; FSAR sections 15.0.8, 15.1, and 15.2.
1 31
Il 1990 A!M!AL IWlORT - PART 2 10 CtK50.59(b) IT.lViG
- 3. 'lhis proposed change dcvs not decrease the ntrgin and r,afety defined by the bases of the Tecimical Specifications. 'Ihis includes a review of the follwing Technical Specification sections 3/4,7, 3/4.8.
88-VIN 0094 'lhis doewent allws the nitrogen dme pressure witches on : MSIV actuator to be changed fran Whitman Ge;.- to P.arksdale. To accatplish this, nxlificetiona to the tubing and conduit are required. *lhe function and wiring do not charu;e, r - L'ng tub f 4 cations required to accmplish
~ .ii.t ige have been analyzed seisntically and are
- t. rep W e to prevent loss of nitrogen during a
.1 MI LC
/er.( ,
- 1. Since the witch was qualified for its intended functic,n and configuratitu by test and calculations, the probability of occurrence or consequences of t.he wilfunction of t.he MSIV's is not inpacted, including reviw of PSAR chapter 6.2, 10, and 15. In addition, there is no change to the WJ% of Table 10.3.3-1,
- 2. Since the switch was qualified for its intended function by test and calculations, there is no new possibility for an accident or equipnent/
cmponent wilfunction, including, reviw of FSAR chapters 6.2, 10 and 15,
- 3. *Ihere is no decrease in the nargin of safety defined by the bases of the Technical Specifications, 'lhis is based on review of the bases for Technical Specification 3/4.'/.1, 88-V1N0096 '1his change adds six lightning protection dwn conductors to the contaisment building, 'Ihese dwn conductors will be coursed down the side of the contalment building and thrcugh the tendon gallery access shaft. Core drills through the tendon gallery access shaft basennt will provide penetrati.ons through which ground rods (6) will be driven to serve as ground tenntnals, 'Ihese ground 32
Il 1990 AtM AL RE10RT - PART 2 10 CFR50.59(b) REIOKr
- tenninals will be tied together with a
; " counterpoise" conductor ecursed through the tendon gallery which setTes to equalize inbalance in discharge currents to ground. This conductor will further cerve to band the fonning steel of the tendon gallery roof, the building / site grid, and raiscellaneous bodies of conductance and inductance to the lightning grid and ground rods.
- 1. During an electrical discharge (lightr/ng) the counterpoise serves to equalize the discharge currents to ground as w11 as equalize ground potentials betwen building / eculpnent grounds and the lightning protection grounts such that any ground potential rise rmote to the contaiment Luilding ground tenninals will be nearly equal.
This serves to decrease the probability of equipuent dicmre or m1 function due to current surges and/or side flashes betwen grounds of different potentials and thereby increases the reliability of the equip:ent described in FSAR section 15.0 during electrical stonm.
- 2. Since lightning protection serves only to shunt lightning discharg;e currents to ground and is Non-1E. t.he proposed design change does not create a potential for accidents or equiptent malfunctions as described in FSAR section 15.0.
Additionally, this proposed change is an addition to an existing grid system which betters its operation. 'lhere is no change to the ground water consideration of FSAR section 2.
- 3. 3he proposed design improves the lightning protection and thereby equipmnt reliability.
There is no change to the margin of safety as described in Technical Specification bases 3/4.8. 88-V1N0097 Modification of various floor penetration seals and containnent electrical penetration lxn:es to provide water tight / leak proof seals. Scope of work to be perfonaed and location of penetrations are specifically addressed in NPFSG Letter 2591, 33 1 . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ -
II 1990 ANNUAL REIORT - PARf 2 10 CFR50.59(b) REIOKr fran R. L. George to S. H. Chesnut. These nodified seals will nret all hazards and fire protection design criteria. The nodifications to the penetration seals were perforced per procedure 00432-C (Penetration Seal Control).
- 1. This change does not involve any equipnent or ccuponent. The hazard analysis is not affected by this change. The proposed change provides increased protection of safe shutdown and water sensitive safety related ccraponents. There is no change to the fire hazard analysis of chapter 9 or accident ar'alysis of chapter 15.
- 2. This change does not create the possibility for any accident or equipnent malfunction not previously described and analyzed in the FSAR, The unterial used does not adversely affect the fire protection requirenents of FSAR section 9.
- 3. This change neets the nurgin of safety defined by bases for the Technical S3ecifications. This included a review of the 3ases of Technical Specification 3/4.7.6.
88-VCE0102 The InBS (Boron Thermal Regenerative System) operation is unitized with independent water cailler and puups in the chilled water loop with the provisicr of a camon chiller to serve either Unit 1 or Unit 2 BTRS operation in the event any of the unitized chillers is down for maintenance or repair. The existing design only provides for the return of the camon chiller oil cooler water to the Unit 1 chilled water loop. This design change calls for addition of a one inch by3 ass line with an isolation valve to enable tie ecnnon chiller oil cooler water to be returned back to the Unit 2 chilled water loop when the camon chiller is used to support Unit 2 B7ES operation.
- 1. The proposed change added a 1" bypass line with an isolation valve to enable the backup camon chiller to support Unit 2 BTRs operation without 34
II l 1990 ANNUAL REPORT - PART 2 ! 10 CFR50.59(b) REPORT the opening of the 6" cconon valve located at the pump suction. his proposed change does not increase the probability of occurrence or consequences of the malfunction of any eculpmnt or emponent assunud to function in accicents analyzed in the FSAR. This is based on review l of FSAR sections 9.3.4 and 15. )
- 2. This change does not create the probability of an !
accident or equipmnt/ccuponent malfunction not . described and analyzed in the FSAR. This is i based on review of section 9.3.4 and 15 of the' I FSAR. I
- 3. E is change does not decrease the margin of safety the bases of the Technical Specification defined sections 3 by/4.4, 3/4.7 and 3/4.9.
88-V1N0106 h is design adds a conduit seal assembly to the Main Steam Isolation Valves (1-IN-3006A&B, 1-HV-3016A&B , 1 -1N- 3026A&B , 1 -IN- 3036A&B) . The seal assenbly wi'l be placed between the valve junction box and the full open and full closed limit switches. To allow for this change, the-existing vendor supplied junction box will be replaced with a box which will be mounted approx. 8 inches off of the valve yoke, he design enecmpasses the support structure required to maintain the box intact during a seismic event. Wiring changes are also implanented to ensure that failure of non-qualified devices do not impact PAMS indication,
- 1. Since this indication is a category 2 device, it is not assuned to function in accidents, but rather, it emi fail and there are sufficient back-up devices to provide similar information.
Also, this change will enhance the probability that this device functions during an accident. Werefore, there is no-change to the accident analysis or PAMS consideration of FSAR sections 7.5.2 and 15. 35
II 1990 A!M1AL REPORT - PART 2 10 CFR50.59(b) RD UKr
- 2. h is change involves only MSIV position indication and only one train of redundant electrical equiptent per MSIV. h erefore, this change does not create the possibility of an accident or equiptent/
camponent malfunction that has not been previously unalyzed. This is based on review of FFAR section 7.5 and 15.
- 3. There is no Technical Specification applicable to the ccuponents involved in the tradification (other than the MSIV Technical Saccification 3.7.1.5 which is not inpacted by this caange) . Therefore, the margin of safety of any Technical Specification bases is not affected.
89-V2E0011 This is a "SAFFDJARDS" DCP. The follcving abbreviated information has been declassified for this report. Allow access to the electrical tutneis through the level "B" control building doors. Weld rebar barriers into the electrical chase in the diesel generator building.
- 1. Relocation of the vital area boundary is a massive physical modification to the facility, whica enhances access to the tunnels and conforms with FSAR section 3.6. Le proposed change does not affect the enviroment in the area, nor does it affect any equipment inportant to safety as described in FSAR section 15.
- 2. Relocation of the vital area boundary is a passive physical tradification to the facility which enhances access to the tunnels and conforms to FSAR section 13.6. The nrdification does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
- 3. The security system and security barriers are not discussed in the Technical Specifications. We security plan is discussed in the Technical Specifications, section 6, in relation to plant audits. This change does not affect the margin of safety as defined in the bases for any Technical Specificaticm.
36
~ 11 1990 A! RIAL REPORT - PART 2 10 CFR50.59(b) REPORT
\
89-V1N0016 his design change relocates the existing filler / i breather cap on the MSIV actuator hydraulic l I reservoir and replaces it with a desiccant type filler / breather assunbly. This design affects control valve tag numbers llN-3006A&B, llN-3026A&B, j and llN-3036A&B. l
- 1. This design change does not modify the design function or the operation of the MSI\"s herefore, there is no increase in the probability l of occurrence or consequences of the mlfunction of 1 any equipment or canponent assumed to function in I accidents analysed in the FSAR. This included a review of FSAR table 10.3.3.1 (Main Stean Systen -
Failure Modes and Effects Analyses) and FSAR 1 section 15.0 (Accident Analyses).
- 2. A review of the results of the FMEA calculation MX4 CPS.0075.259 indicate that the failure of the desiccant type filter / breather cap will not impact
' the operation and qualification of the MSIV's. 'Ihis was based on a review of FSAR section 6.2 (Essential- Containment Systans), 7.3.8 (ESFAS - Main Stean and Feedwater Isolation), and FSAR table 10.3.3.1 (Main Stean Systen - Failure Modes and Effects Analyses).
- 3. This design change does not-decrease the u rgin of ,
safety defined by the bases of the Technical Specifications. This included a review of Technical Specification 3/4.7.1.5 and its bases. 89-V1N0017 Bis desi:;n change remves the existing MSIV actuator aydraulic manifold isolation valve inserts and replaces than with a " balanced" mnifold. isolation valve insert. B is design clmnge affects control valve tag nunbers llN-3006A&B,11N-3026A&B,- and llN-3036A&B.
- 1. This design change does not increase the probability of occurrence or consequences of the m1 function of:
any equipment or canponent assumed to function in accidents analysed in the FSAR. his included a - review of FSAR table 10.3.3.1 (Main Stean Systan - Failure Modes and Effects Analyses) and FSAR section 15.0 (Accident Analyses). 37
i i I 11 . 1990 ANNUAL REIORT - PART 2 i 10 CIK50.59(b) RE10RT 4
- 2. This design change does not create the possibility of an accident or equiptent/catponent malfunction '
not described and analyzed in the PSAR. This was based on a review of FSAR sections 6.2 (Essential - Contalment Systems), 7.3.8 (ESFS - Main Steam and Feedwater Isolation), and FSAR table 10.3.3.1 (Main Stean System - Failure 1kxles and Effects Analyses).
- 3. This design change does not decrease the nargin of safety defined by the bases of the Technical Specifications. This included a review of Technical Specification 3/4.7.1.5 and its bases.
89-VlN0027 The MFIV fast closing operation schure contains a redundant control relay. This control relay contact is wired to initiate the de-energization of the MFIV pilot solenoid valve when the control relay coil is de-energized. De-energizing the pilot bolenoid
- valve will cause the ITIV to close in the fast close 4 node. During nontal operation the redundant control
! relay coil is continuously energized. If the relay coil fails due to either a short or open circuit, the pilot solenoid valve will de-energize and. initiate the undesired fast closing operation of the MFIV. This change involves rewiring the if1V control circuitry to disconnect the redundant relay , coil' and contact fran the MFIV fast closing control scheme,
- 1. The control wiring change does not change the ifIV operation and no new cauponent is added to the existing control circuit to make this change.
Therefore, the change does not increase the probability of occurrence or consequences of the , unifunction of any equiprent or canponent asstamd to function in accidents analyzed in the FEAR. o 2.- This minor wiring change does not chenge the original system operation and no new cmponent is added to the existing control circuit to make this change. Therefore,:the change does not create the possibility of an accident or equiptent/canonent malfunction not described and analyzed in tae FSAR. 38
a + adi d s.,4+1as__M=. A h44a.r-5.5h Ae JL..4 Am%2J.-4 t 4 a4 4.42m-4 i II 1990 N NUAL R120RT - PART 2 10 GK50.59(b) Rf20RT
- 3. The change is designed to neet the intent of the eriginal system design bases while increasing the electrical control circuit design reliability. Therefore, the change does not decrease the margin of safety defined by the bases of the Technical Specification, particularly sections 3/4.4.5 (Stean Generator) and 3/4.6.3 (Contattront Isolation Valves).
89-V1N0028 An orifice with a bore diarmter of 0.109 inch is instelled in the "P" port of the fast actity; 1 solenoid valve for each of the Main Turbine Control 1 Valves IXV-6005, 6006, 6007, and 6008.
- 1. The fo11cving conclusion is based on review of FSAR sections 10.2 and 15, ihe noin turbine control valves are non-safety related. The turbine overspeed protection feature of the control valves, so as to prevent potentially damging missiles fran striking safety related structures, is unimpaired. !
- 2. The design change affects only the internal flow /
pressure characteristics of 12S fluid to the control valves and is canpletely enveloped by existing descriptions and analyses in FSAR sections 10.2 and 15. No new accident possibilities or malfunctions are created. 1
- 3. The design change does not affect the operability or overspeed protection functions of the control D valves as discussed in Technical Specifications bases Section 3/4.3.4.
89-VIN 0029 Orifice plate 11V-5552 in line 505-2" is being renoved. Gate valve 1-1326-U4-527 is being- . replaced with a globe valve that can be used for
- throttling. The valve is also in line 505-2" on l
the Stator Cooling Water Skid. General Electric-I has recanmnded this clange for both units. This , change is to a non-safety, non-seismic system located in the 1brbine Building. 39 _ .,__.2, _ . _ _ _ .- _ _ . _., . . _ . . - -._ ... _ _ . _ . _ __ _ _ _ _ _ _ __ _ _ _ _
j 11 1990 AMUAL REIORT - PART 2 ' 10 C1K50.59(b) Rr10RT I
- 1. The Stator Cooling Water Systs is not safety-
! related and is not assmed to function in any accident analyzed in ITAR including section 15.
- 2. This change does not affect plant conditicus. There is no change to section 15 or section 11 and 12.
- 3. The Stator Cooling Water Systm serves no safety function and has no safety design bases. Bases _ ,
to Technical Specifications are not affected. L 89-V2E0030 This DCP adds additional lightning aroteccion bonding for the Unit 2 contaiment milding INAC ventilatico duct, and the adjaeont ladder. The bonding is to be installed where the dcne ring i conductors pass under the netal emponents at the i lower end of the vent duct, and ladder runs. Bonding to the existing dcun conductors will be required.
- 1. The design change provides additional bonding tor the contaiment building exterior INAC vent duct, and ladder, which decreases the
probability of malfunction for this equipmnt, and increases the reliability of the existing e mponents as described in the FSAR section 15.0, during' an electrical stom. Therefore, decreasing the probability of occurrence or consequences of an 1 accident or malfunction of equiptent inportant to safety. , 2. This design change adds non-class IE bonding connections to the existing contaiment building lightning protection syst s and does not create-any potential for accident or equiatwne malfunction to tae plant grounding syst s , FSA3 section 8.0, or to any equipment assumed to function as described in FSAR section 15,
- 3. The design change inproves the lightning protection and thereby irproves the equipnent reliability. There is no change to the margin of safety as defined in bases of the Technical Specification 3/4.8.
40 _.a_.______._.. _ _ . _ _ . . - - - . . _ _ . . _ . _ . . _ -
II 1990 AN!M REIVRT - PART 2 10 Cat 50.59(b) RElORT 89-V2E0031 'Ihis design change adds a heat insulator betwen the process connection and Ficher controller for the following inctnoents: TAG 10. DESCRIFf10N 21EL-4361 FEUr A 21fL-4362 tEDT B 21EL-4371 RUr A ZlfL-4373 RDT A 21LL-4372 RUr B 2tfil-4374 RDT B 2LEL-4522 MSDT C ZlEL-4523 MSUr D 21EL-4532 RUr C 21f11-4534 RIrr C 21EH-4533 RUr D 21EH-4535 kl>T D
- 1. This change does not adversely affect the perfontance of any curponent assuted to function in the accident analysis of chapter 6 or 15. There is no increase in the probability or consequences of any equip:ent assured to function in accidents analyzed in the PSAR.
- 2. This change does not create the possibility of any accident or emponent tralfunction not presently analyzed in the PSAR. This included a review of FSAR'section 7, 10 and 15,
- 3. 1here is no change to the nurgin of safety defined by the bases of the Technical Specification. This included review of Technical Specification 3/4.3, 3/4.7 and 3/4.11.
89-VlN0032 This design chant;e adds a heat insulator between the process connection and the Fisher controller for the following instnnents: TAG NO. DESCRIFr10N 11rL-4361 MSUI' A 11LL-4362 MSUf B 41
. ..._. _ .._. _ _ _ _ _ _ _ _.. _ . . _ _ _ _ _ . - _ . . _ _ . _ _ _ _ . ~ .
i 11 1990 NRIAL RDVRT - PART 2 10 CITL50.59(b) RDORT
- I W 4371 RUr A 11EL-4373 RDT A 11EL-4372 RUr B 11DI-4374 RDT B IIEL-4522 MSUr C 11fL-4523 MSUr D lifL-4532 RUr C 11DI-4534 RDT C i IIIL-4533 RUr D 11DI-4535 RUr D In addition, the air regulators are to be remved fran the follcuing level controllers and munted rumtely.
TAG NO. DESCRII'rION lifL-4361 MSUr A Iici-4363 MSUI A I W 4362 MSUr B 11411-4364 MSUr B 11rL-4522 MSUr C 11 Ell-4524 MSUr C 11rL-4523 MSUr D l 11f11-4525 MSUI D
- 11LL-4371 RTD A 11ril-4373 RDT A 1LCL-4372 RUT B 11Lil-4374 RUr B 11rL-4532 RUr C 11Lil-4534 RUr C ILCL-4533 RUr D t RUr D 11D!-4535
'Ihe function of each of the above instrumnts is to control the level of its associated drain tank (misture separator or reheater drain tanks). All affected instru mnts are non-seismic, non-safety related class 62J instruments.
1 42
II 1990 ANNUAL R13 ORT - PART 2 10 CFR50.59(b) lEIOKf
- 1. This change dces not adversely affect the perfontance of any emponent assmed to function in the accident analysis of chapters 6 or 15, or any other section of the FSAR. There is no increase in the probability or consecuences of any equirent asstred to function in accidents analyzed in t le FSAR. All instrunents and e this change are non-seismic non-safety / qui} rent related. affected by
- 2. This change does not create the possibility of any accident or cwponent malfunction not presently analyzed in the FSAR. 'lhis included a review of FSAR section 7, 10, and 15.
- 3. This design change does not decrease the targin of safety defined by Technical Specifications B 3/4.3, B 3/4,7 and D 3/4.11 and their bases.
89-VIN 0034 1he proposed change will nodify the existing Control Rod Drive Mechanism (CRI14) Unit 1 Seismic Support Lugs. This nodification will involve the elongation of a 3" dimeter nctninal hole in the radial and torsional support lugs. The elongation of the subject lugs will enhance the unde cimnges durn.g plant operation and reduce refuelint; duracion by allowin{; the seismic tie rods to be rmoved without returning the reactor vessel to operating taTperature,
- 1. This design change does not increase the probability of occurrence or consequences of the malfunction of any equi;xTent or couponent anstood to function in accidents analysed in the FSAR.
This included a review of FSAR sections 3 and 15. The proposed design package only nodifies the Seismic Support Lugs of the CRI14 and does not affect the function of the CRL14 during an earthquake.
- 2. This design etange does not create the possibility of an accident or equi txient/cmnonent nulfunction not described and analyzed in tile FSAR. This was based on a review of PSAR sections 3 and 15, 1
43
_. ~ . . _ . . _ __ _ ___._._ m. _ __ . . .._ _ _._____ '1 - 11 1990 ANtRIAL REIORT - PART 2 10 CFR50.59(b) REPORT , 3. This design change does not change the nargin of safety defined by the bases of the Technical Specifications. This included a review of j Technical Specification 3/4,4.10 and its bases. 89-VIN 0036 A. *1his DCP involves replacing the Lovejoy notor l
) controller and its associated canponents on i the Unit 1 Sigra Refueling Machine (1-2101-R6-003) with a Veearc Super 7000 Controller, a joystick powr supply and associated cmponents. Also included in this change is the addition of a fault relay. The unin difference betwen the i controllers is that the notor speed for the !
Veearc can be set digitally, whereas the l Invejoy is set in an analog unnner. l Westinghouse safety evaluation checklists 1 SECL-89-1096-C and SECL-89-1187 address this change. B. In addition, to prevent the refueling unchine powr cable fran snagging on bolts protruding fran the stailw11 wall, a pipe / miler assently will be added to oct as a sw. doff, preventing danage to the cable.
- 1. a. This change involves the replacenent of the notor controller on the Unit 1 Sigin Refueling Machine. DC-1010, Rev. 5 indicates the refueling nachine as a non-safety related, Seismic Category 2 nochanical systun. 1he refueling nachine does not constitute part of the protection system and it is not specified as electrical Class IE. FSAR section 15.7.4 addresses accidents involving the refueling machirn,
- b. 1he pipe / roller assembly added to act as a standoff preventing dmage to the refueling machine powr cable has i>een designed to Seismic Category 1 criteria precluding any Seimnic 2 over one interactions.
44
II 1990 ANNUAL PIPORT - PART 2 [ 10 CIK50.59(b) PIIORT Therefore, this change does not increase the probability of occurrence or consequences of a malfunction of any equ[ rent assmed to function in an accident.
- 2. a. This change involves replacing the Unit 1 i Sigma Refueling Machine m tor controller to improve reliability and maintainability of the machine. The new controller is-
; functionally similar to the original one and !
it does not degrade the perfornvnce of the m ehine. Also, it does not inpact the fuel - handling accident analyses or the operation ' of safety-related equipmnt as described in the FSAR. 3
- b. The pipe / roller assembly added to act as a standoff preventing damne to the refueling machine pcuer cable has seen designed to Seismic Category _1 criteria precluding any Seismic 2 over one interactions.
Therefore, this change does not create the possibility of an accident or ecuiprent/_ ccinponent malfunction not alreacy described in the FSAR. Sections 9.1.4 and 15.0 were reviewed.
- 3. The replacment of the mtor controller on the refueling mehine and the addition of the pipe /
roller assanbly has no effect on the urgin of safety defined by the bases of the Technical , Specifications, section 3/4.9.6. 89-V1N0053 This change is the replacenent of--the existing Crane nrxlel GVH-20k Unit 1 Reactor Coolant Drain Tank Pumps (1-1901-P6-001 and 1-1901-P6-002) with Crane nodel 4301 canned meer pumps, and the replacment of their associated 100A contaiment penetration backup breakers with 125A breakers. The Reactor Coolant Drain Tank pmps are part of the Liquid Waste Processing Syscun (1901), which i is addressed by FSAR section 11.'. i It is i non-safety related,_ project class 427. This systan l has no safety function. Failure of this system will not compranise the ability of the plant to 45
l II i 1990 AMIAL R13 ORT - PART 2 10 CFR50.59(b) REIDIT acccuplish a safe shutdown. Contahnent menetration protection is provided by the backup areaker.
- 1. Equiptent affected by this design is not asstned to function in an accident analyced in chapter 15 of the FSAR (Accident Analyses). Insta11atico of the proposed change will not cause the tulfunction of other equip:ent asstried to function. 'lhere is no increase in the probability of occurrence or the consequences of an accident or tu1 function of equiptent inportant to safety. *1he flooding and ceismic analysis in not inpacted. The new pttp and pipiru category 1; and are not safety neither are therelated existingor seismic piping or valves. Containnent penetration backup protection is provided in accordance with IX'-1823 and FSAR section 8.3.1.1.12,
- 2. '1his design change does not create the possibility of an accident or equipnent/cctnxnent malfunction not described and analyced in tae FSAR khich could affect the health and safety of the public, 'Ihi s was based on a review of PSAR chapter 15 (Accident Analyses) and section 11.2. The new ptops are being added to inprove the function of the existing systen. Contalment penetration backup protection is provided per IX'-1823 and FSAR section 8.3.1.1.12.
- 3. This change is associated with systun 1901 and decs not decrease Technical Specification safety nargins since it has no safety design bases. '1his is based on review of the Technical Sp'ecification bases, including Section B 3/4.11, Radioactive Effluents",
and also section B 3/4.4 " Reactor Coolant Sys tan" . Contain:ent penetration backup protection is arovided and the cargin of safety as described in Tecanical Specification bases 3/4.8.4 is not decreased. 46
- f k
j 4 11 1990 M RIAL REPORT - PART 2 10 CFR50.59(b) REPORT 89-VIE 0057 For heater drain ptops 1-1304-P4-001 & 002:
- 1) Perform seal wld of ptmp 1cwr aligning ring
- to ptmp shell.
- 2) Change ty>e gasket used betwen xmp shell and ptrin disc Targe head fran red rubaer to a Garlock gastet.
< 3) Drill hole to accmodate 1/4" thread top thru one of the thirty-six bolts holding the discharge head to the shell. Tap the drill hole to accmodate a 1/4" mehine scro.
- 4) Remve grout at punp tirxinting flann such that top of grout is even with bottczn of ptstp munting base flange.
- 1. The changes do not inpact any equiptent or ccuponent analynod in FSAR sections 15.1 or 15.2.
The changes are to non-safety related, non-1E ccx:ponents. Failure of realacment gasket or integrity of seal wld or solt could Icad to condensate / steam leaks at the foundation area but wuld not inpact safe secondary side operation. There is no change to the Fl u of Table 10.4.7-1,
- 2. The changes do not affect the reliability of the heater drain ptrrps. The changes are mde to reduce the possibility of condensate Icakage frcxn the non-safety related heater drain systun. 1his is ,
based on a review of FSAR sections 15.1 and 15.2. There is no change to Fin Table 10.4.7-1,
- 3. Based on review of the Bases for Technical Specifications B 3/4.7 the changes have no effect of mrgin of safety.
89-VIN 0059 This design changes (a) Replaces the wided end caps of spare penetrations (15, 55 and 90 located in the containtent building) with bolted blind flanges, (b) Adds flange
" fixtures" during Mode 5 and/or Mode 6 for outage related work activities which require containnent penetration. The flan 6"
" fixtures" will replace the blind tlanges for the insertion of eddy current, sludge lancing f
4 47
' - - + , , ,.-ry-y .-.,-yy. - ., , .ym.,* ,,,,,p-m, y,,_,,,_,-m,,_._,._,.m..m.,_,,, . _ , - . . , . , - , , - _ , , . . _ , e--,.-
i II 1990 ANNUAL REIORT - PART 2 10 CFR50.59(b) RE10RT
& IS1 ec outagefixture"
'luipnent during of is mde-up a Unit a flange 1 outage.1he plate with threaded sleeve ports. These sleeve ports allw access for cables and connection points for lead-in hoses during outage related work,
- 1. This change (a) does not increase c% probability of occurrence or consequences of the a lfunction of any equipnent or camponents asstred to function in the accident analyzed in the FSAR including those of Sections 3, 6, 9 or 15 or FSAR. The design criteria, analytical nethods and construction procedures for these penetrations are the sicae as those used for other nochanical penetrations. Thus, there is no inpact on the integrity of the contahnent liner plate fission product barrier.
This design chan;;e (b) neets the intent of FSAR chapters 6.2.4.1 (Contaurent Isolation), 15.7.4 (Fuel iktndlinr, Accident) and GL 88-17 (loss of Decay llent kwval) in that the outage
" fixtures" prevent the direct ccnuunication of the contairmunt enviroruent with the outside envirotuunt under the postulated conditions of a refueling accident. Therefore, this change does not increase the probability of occurrence or consequences of the mlfuncthu of any equi; ment or ccuponents asstned to function in the accident analyzed in the FSAR including those in sections 3, 6, 9 or 15 of the FSAR. The " fixture" being installed is considered tenporary, and will only be installed when the plant is in Mode 5 (cold shutdwn) or 6 (Refueling). It will only have to perform its design funciton under the wrst case conditions of a refueling accident and GL 88-17, which does net postulate contabount pressurization. 1his " fixture" does not have to be designed to function for other postulated design basis Accidents.
Therefore, this fixture does not have to be designed, fabricated, or installed to the ASME Section III, Class 2, requirenents of the penetration. A seisntic evaluation has been 1 48
J t 4 11
' 1990 ANNML PZIORT - PART 2 i 10 CFR50.59(b) REIORT
- perfomed to insure that the structural integrity of the penetration (and " fixture") is
! naintained during a Design Basis Earthquake. Thus, i there is no impact on the integrity of the contaiment liner plate fission product barrier.
- 2. This design change for the nodified condition of the pencerations naintains a boundary that will prevent the exchange of the contalment enviroment with the outside enviroment under postulated refueling accident conditions. The
^ nodified penetrations neets the intent of all applicable Technical Specifications under nornni and postulated refueling accident . conditions as defined in the FSAR. This change does not impact any systun, equipnent or ecxnponent's function or operation and btsed on a review of WAR sections 3,6, 9 and 15 would not create the possibility of an unanalyzed or undescribed accident or equipnent/ccuponent i malfunction.
- 3. After testing of the penetrations in accordance with the 11RT procedures, the new configuration will neet the currently existing basis for contalment leakage. This design change naintains the contaiment penetration in a condition that will prevent the direct ecununication of the -
contaiment enviroment with the outside enviroment in accordance with the bases for Technical Specifications 3/4.9.4 for normal and postulated refueling accident conditions during core alterations. Therefore, these changes inpact do not the bases affect safety of Technical unrgin defined Specification 3/4.6.1by and 3/4.6.2. 89-VIN 0070 This nodification strengthens the Icft bank intercooler plenums for both e & B train
- diesel generators. These plenums, which supply o
ccrubustion air to the dicac1 engines, are nodified by attaching extemal ribbing arranged to ecxup1 ment 49 L-
l l l 11 1990 N MIAL REPORT - PART 2 10 CFR50,59(b) RE10RT l l the existin stiffening schmes and b improving I the interna vane attaciment to the s tell. In i addition, a 1/2" coupling is added to facilitate inspection in accordance with Delaval instnictions in SIM-365, Rev. 2. L 1. This nodification only involves the diesel generator intercooler plenuns and does not degrade the function of the diesel generator or any other equiprent or cauxment. Therefore, it will not increase the pro > ability of occurrence or consequences of the txilfunction of any eculpuent ' or component assuned to fur.:: tion in accicents ; i analyzed in the W AR section 15. I 2. The proposed nodification reinforces exir: ting-intercooler plenuns by attaching extemal ribbing and improves the internal air vane attactment to the shell. It does not affect the operation of-diesel ! generators in any way. Therefore, it can not create- l l the possibility of an accident or equipment / couponent tnalfunction not described and analyzed in the FSAR. I
- 3. The proposed nodification does not decrease the 1 nargin of safety defined by the bases of the Technical Specifications, sections 3.0/4.0.
89-V1N0072 Remove the undervoltage trip function of the CVQ L relays for the non-1E 13.8KV and 4.161N switchgear. I Motors fed fran the non-1E 13.81N switchgear tuses 1NAA and 1 NAB and 4.16KV switchgear buses 1NA01 and INA04 will no longer be tripped for undervoltage-l ' . conditions. Annunication of undervoltage conditions will be retained. The renoval of the trip functions will,be accmplished by disconnecting jmper wires in each affected feeder breaker. L l 1. The rmoval of undervoltage and negative sequence trips for non-1E switchgear will not increase the probability of occurrence or consequences of an accident described in the FSAR including chapters 8 and 15. l l 50
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 1he undervoltage relay trips being disconnected are part of the non-lE ous protective relaying and are separate frm the reactor coolant putp notors' lE breaker undervoltage and underfrequency relays as discussed in the FSAR section 15.3. Rmoval of the trips will enhance the function of the RCP's undervoltage relays by eliminating nuisance trips due to miscoordination between the non-lE and lE undervoltage relays.
- 2. The undervoltage trips, which are to be renoved, served two functions. The first was to load strip notors in the event of a dead bus. The second was to protect notors frm damging undervoltage conditions.
The transfonmra feeding the buses affected by this DCP (lNAA, ItMB, INA01 & INA04) are not capable of starting all notor loads connected to the bus at the sa:m tine. Therefore, in the event of a dead bus, the operator will be required to strip the notor loads before the bus is re-energized. The probability of the station service voltage being degraded to 90% or lower and remaining at that level long enough to damage notors is very low. The low voltage condition would probably be caused by a failure of a prctective device to clear a fault. Backup protection should clear the fault before notors are damaged by undervoltage. Tripping of critical notors (RCP) during this period could result in an unnecessary unit trip. The lE switchgear buses are not affected by this change.
- 3. The margin of safety as described by the bases for Technical Specification 3/4.8 is not decreased by this char.ge.
51
11 1990 ANNUAL PEPORT - PAR 1' 2 l 10 CF850.59(b) PIPOPl1' i 89-V1N0074 The breathing air systun supplies breathing l nir to contairment workers during refueling outages. It is class 626 except at the containnent penetration which is class 212. This DCP will accanplish the following:
- 1. Extend and enlarge breathing air header supp/2 11 .igpiping. All header Renove filters and existing pipe to be bottle supply systen. Add electrical pcwr supply. Portable canpressor to be located outside at the east end of the Auxiliary Building at Main Stean Tunnel 1T1,
- 2. Delete r itrol roan annunciation.
- 3. Containm:nt isolation valve, 1-2401-U4-2L1 will be replaced with a 1 1/2" equivalent valve.
- 2. The Breathing Air System does not interface with any equipnunt assuned to function in accidents analyzed in the PSAR and is not essential for the response to any accidents analyzed in the FSAR.
(Chapters 9.3.1 and 15.0 reviewed)
- 3. This change does not create the possibility of an accident which could jeopardize the reactor coolant pressure % dary or the health and safety of the public. Only mrxlerate energy lines are being added to the Aux./ Cont. Bldgs. Pipe hangers will be seismically analyzed and the piping will be supported such that it cannot fall and danage any equipnent or conponents assuned to function in accidents. The rotary conpressor (a possible missile source) is located outside the Auxiliary Building and cannot fail in such a way as to dantage safety related equipannt. Contairment isolation will not be changed by this proposed change.
52
11
.1990 ANNUAL REPORT - PART 2 i 1: CFR50.59(b) REPORT
- 3. W e Breathing Air System is not di.ccussed in the Technical Specification. However, section 3.6.3 states that contairunent isolation valves shall be operable. Therefore, valve 1-2401-U4-211 will not be changed until after plant shutdcun, and the original design, QA, and operability requirements will be met by the new larger valve.
89-VCN0086 This DCP involves replacing the lovejoy motor controller and its associated camponents on the the emnon fuel handling machine (A-2203-R6-002) with a Veearc Super 7000 controller and associated canponents. The main difference betwen the controllers is that the notor speed for the Veearc can be set digitc11y, whereas the lovejoy is set in an analog manner. Also included in this , change is the replacement and addition of relays associated with the new controller.
- 1. This change involves the replacement of the notor controller on the fuel handling machine. DC-1010, Rev. 5, identifies the fuel handling machine as a safety-related, seismic category 1 mechanical system, ikwver, the- fuel handling machine does not constitute part of the protection system and it is not specified as electrical class 1E. FSAR section 15./.4 addresses accidents involving the fuel handling machine. This change will necur during nornal reactor operation during which time the fu21 handling machine is not in use.
W erefore, this change does not increase the-probability of occurrence of malfunction of any equipment assmed to function in an accident.
- 2. This change involves replacing the fuel handling machine notor controller to improve reliability and maintainability of the machine. Be new controller is functionally similar to the original ~
one and it does not degrade the perfonnance of the machine. Also, it does not impact the fuel handling accident analyses or the operation of safety-related equipment as described in the FSAR. 53
II 1990 ANNUAL REPORT - PART 2-30 CFR50.59(b) REP 0iG' Therefore, this change does not create the ; possibility of an accident not already described i in the FSAR. Sections 9.1.4 and 15.0 wre reviewd.
- 3. The replacement of the m tor controller on the fuel handling machine has no effect on the margin of safety defined by the bases of the Technical Specifications.
89-VIN 0089 The proposed change will replace the bolted blind flange on the 11RT Test Connections with a 3/8" globe valve and a threaded cap. The test I connections are used to perfom local leak rate testing.of the containment isolation valves associated with penetrations 2' 67A, and 67B. The design is in accordance wi.i the requirem nts of General Note 13.h of the Piping Material Classification (Doc. # AX4DR001, Rev. 19). The isolation provisions for each test connection will consist of tw globe valves in series versus the current design of a globe valve and a blind flange. Fourofthetestconnectionsareprojectclass2f2 and are part of the Reactor Coolant Systs. The rmaining six are project class 424 and are part of the Nuclear Sampling Systs. Several of these valves interface with the Post Accident Samling Systm.
- 1. Replacing the 11RT test connection blind flange
-with a second valve is in accordance with the a3 proved design given in document AX4DR001. The cvinge does not affect any equipment or camponent assumed to function in accidents analyzed in the FSAR. Criteria for leak testing in 10CFR50 App. A, GDC 54, is satisfied.. Operation of the Reactor Coolant System, Post-Accident Sampling System and the Nuclear Sampling Systs is not affected by the change. Furthermore, the ability to aerform the LLRT is enhanced. Therefore, the proaability of occurrence or consequences of the malfunction of any equipment or cmponent assumed to function in accidents analyzed in the FSAR is not increased.
54
i 11 1990 AhWUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. The change is in accordance with existing approved design requirements. Implementation of the modification will be in accordance with existing !
plant approved specifications and procedures. The worst case accident would be a breach of the Reactor Coolant System pressure boundary which is analyzed in the FSAR section 15.6.2 and bounds this change. The modifications required by this DCP do not create the possibility of an accident or equipment /com3onent mlfunction not described and analyzed in tae PJAR.
- 3. The modifications required by this DCP are to be 1 implemented in accordance with plant approved specifications and procedures. The modificatione required by this DCP will be empleted in accordance with the Design Criteria and the Codes and Standards applicable to VEGP. In addition, applicable Technical Specifications (section 3/4.6.1.2 and 3/4.4.6.2), including the associated basis, have been reviewd and determined not to be affected by this change. Therefore, the modifications required by this DCP do not decrease the u rgin of safety defined by the bases in the 1 Technical Specification.
89-V2E0091 The main turbine high exhaust hood temperature trip feature is to be converted to a high-high tmperature alam. This will be accomplished by deleting internal wiring to the DiC cabinet connecting the exhaust hood temperature sensors (TSH-6417, 6420, 6423) to the 125 VDC trip bus. The temperature sensor inputs to the alC cabiriet will be re-connected directly to the high high temperature alarm and camputer output of the DiC cabinet. The annunciator window reading " Exhaust hood hi-hi temperature turbine trip" will be changed to read " Exhaust hood hi-hi temperature". The computer alam (Proteus) will be similarly revised. l- { l l ' 55
. - . - - - - . .- - - - . - . - ..-..-.~ - - - . . ~ . . - . - . ----
11 1990 ANNUAL REPOKr - PART 2 10 CFR50.59(b) REPORT
- 1. The design change does not increase the probability of occurrence or consequences of the malfunction of any equisnent (safety analyzed or otherwise) since the caange is associated with a non-safety systan which is not relied upon in any accident analysis or accident condition. There are several layers of protection for a high-high exhaust hood tmperature condition. (1) Hood sprays (2) High tauperature alarm (200 degrees fahrenheit) (3) Hood temperature indication on the main control board (4) Local hood tmpera:.ure gauges (5) Im vacuum trip (2/3 logic). The addition of a high alann (200 degrees fahrenheit) will give the operator sufficient final warning for a manual trip if necessary.
- 2. This design change does not create the possibility -
of an accident or malfunction of safety equipent-or system not already described and analyzed in the FSAR. A back-up of turbine hood-spray is provided. FSAR turbine missile analysis is based on low and high speeds of 'the turbine which is unrelated to exhaust hood temperature transients.
- 3. The proposed change will not affect the margin of safety defined in-the Technical Specifications since the function is not addressed in sections 3/4.3.4 and 3/4.7.1 of the Technical Specification
.and is not relied upon in any accident analyses in
~c hapter.15.
VIN 0098' This change removes-the protective ring from the Reactor Internals Lifting Device (1-2206-R6-001). Westinghouse Refueling Equipment Engineering Group was advised of this-change request. Although they do not reemmend the remval of the protective ring they do not prohibit its-removal and non-use by the-operating companies.
- 1. The protective ring is not for fuel protection and
-is not assumed to function during a refueling accident discussed in the FSAR section 15.7.4.
56
! II 1990 ANNUAL PIPORT - P/JtT 2 10 CFR50.59(b) REPORT
- 2. This does not create the possibility of an accident damging the 0-Rings because the possibility of damge has always 5een present. Inspection of the 0-Ring surface for defects is done prior to vessel assembly per procedure 93240-C, " Reactor Vessel Assembly / Disassembly Instructions", section 4.21,
" Head and 0-Ring Installations".
- 3. This change does not decrease the margin of safety as defined in Technical Specification bases. This includes a review of section 3/4.9.
89-V1N0102 The following changes are being made in the Turbine ElC Panel (1-1615-QS-DIC) . The Valve Position Driver Boards which control the four Turbine Control Valves (1XV-6005, 6, 7, & 8), three Intercept Valves (1XV-6009, 10, & 11), and one Stop Valve (1XV-6002) are being modified. The modification consists of adding a jumper on the backplate of the board frm test point TPS to test point TP11 (ground). This jumper will disable the closing bias circuit. The purpose of the closing bias circuit was to cause the valves to go slowly closed if the control signal is lost,
- 1. The proposed change does not increase the probability of occurrence or consequences of the malfunction of any equipnent or component assumed to function in accidents analyzed in the FSAR.
This includes a review of the following sections: 10.2, 15.1 and 15.2
- 2. The proposed change does not create the possibility of an accident or equipment /cmponent malfunction not described and analyzed in the FSAR. This includes a review of the following sections:
13.0.8, 15.1, and 15.2.
- 3. The proposed change does not decrease the margin of safety defined by the bases of the Technical Specifications. This include a review of the following sections: 2.2.1, 3/4.3.4.
57
11 1990 ANNUAL REPORT PART 2 10 CFR50.59(b) RD ORT 89-VCN0108 h is DCP provides for transfer of the power supply source to lighting transfonmr ANPA16X and distribution panel ANLP82, frm HCC ANM to MCC ANBS. Also, the power supply to the solenoid valves for River Make-Up pumps 003 & 004 and one Battery Rom exhaust fan in the River Intake Structure Building are transfered from 120 VAC distribution panel ANYAl to ANLP82 with the inplementation of this DCP. This change requires lighting transfonmr ANBA16X to be re-numbered to ANBS03X and one new cable to be installed (frm panel ANLP82 to JB6257) in existing raceway. All other cables are existing and are to be disconnected from their present power sources, re-numbered and reconnected at the new power source. One new conduit is also to be installed from panel ANLP82 to an existing tray.
- 1. W e changes provided in the DCP are to be made to non-class lE equipmnt/systs. The operation and function of the affected emponents are not changed. Bere are no seismic or enviromental concerns involved, h e new cable used was procured as class lE. L e capability of MCC ANBS to feed L the 45 kva lighting transformer ANBS03X has been evaluated'and found acceptable, hese changes will-not introduce any new interfaces, or alter setpoints to any safety related equipmnt or system.. h erefore, these changes do not increase the probability or occurrence or consequences of j' .
the malfunction of any equipment or component
. assumed to function in accident analyzed in the FSAR.
- 2. These changes enhance the availability of the river water make-up system which has no. safety, environmental, or seismic concerns. These changes
[ do not affect the function / operation of-the equipment / system involved, nor will they affect u the operation of any safety related equipmnt. Therefore, these changes do not create the l possibility of an accident or. equipment /comaonent malfunction not described and analyzed =in the FSAR. I 58 L
11 1990 ANNUAL REPORT - PAP l1' 2 10 CFR50.59(b) REPORT
- 3. The DCP provides for changes to non-safety related equiprent/ system. There is no reduction to the mrgin of safety as defined in the bases for any Technical Specification involved.
89-V2E0lll 'Ihis change adds a new floor drain to Auxiliary Building Roan A99. The new drain piping will be routed to un existing drain located in Auxiliary Building Roan B135. The drain discharges to the floor drain tcak for eventual processing in the radwaste systan.
- 1. The addition of the floor drain to Auxiliary Building Roar A99 has no effect on the equipient or systems assuned to function in FSAR section 15.0 accident analyses. Therefore, the change will not increase the probability of occurrence or consequences of the malfunction of any equipcent or cauponent assured to function in accident analyzed in the FSAR,
- 2. 'Ihe addition of the floor drain to Auxiliary Building Roan A99 is enveloped by existing FSAR analyses, and therefore, will not create an accident or equipment /camponent malfunction not descrd' sed and analyzed in the FSAR. Sections 9.3.3, 9A.2, 13.5 and 15.0 were reviewed.
- 3. The addition of the floor drain to Auxiliary Building Roan A99 has no effect on the margin of safee y as defined by the bases of the Technical Specifications.
89-V2E0113 The proposed change is the additicn of a Gaitronics telephone /page wall mounted handset to Auxiliary Building Roan #A103 and one wall munted handset with speraker in the Fuel Handling Building Roan lA01. 'Ihe handset to be located in Roan A101 will be powered fran a handset on the floor below which will require a core drill and penetration seal. 59
11 1990 ANNUAL REIWP - PART 2 10 CFR50.59(b) RElWF The handset to the added in Roan A103 will be powered from an existing emplifier located within the roan. Based on the size of the roan, no speaker will be required for the new handset since there is a speaker connected to the existing amplifier.
- 1. The proposed change to plant camunications does not affect any eculpnent or camarant which is assumed to function curing an accident analyzed in FSAR section 15.
- 2. Since the camunication system is non-safety related, the proposed change does not create a potential for accidents or equipment malfunctions as described in FSAR section 15.0 based on a review of FSAR sections 9.5.2, 9B-69 and 15,
- 3. The addition of telephone /page handsets will not affect the margin of safety defined by Section B3/4.9.5 which addresses having proper cannunications during re-fueling. 7his requirenent is satisfied by the soundpowerecl phones as referenced in FSAR 9B-69.
89-VCN0ll5 1his design change package provides pennanent electrical power feeders for use during the Unit 1 outages. These power feeders are all h80 volt, 3 phase feeders and the scurce and rating are as follows. One 100 amp service will be provided fran load center INB02 to the alley between the Turbine and Control Buildings. Second, a 400 amp feeder fran load center 1NB12 will service Invel 3 (Turbine Deck) of the Turbine Building. Third, a 600 aup supply will be fed fran load center ANB06A to an area outside the Demineralizer Building on the south west corner. Because this load center has a tie-through breaker to ANB06B, interlocks have been provided to prevent sinultaneous operation of the outage feeder and the tie-through breaker, thus overload of either switchgear transformer is prevented. Last , two i 60
II 1990 ANNTAL REPORT - PART 2 10 CFR50.59(b) REPORT separate 230 amp services will be provided on Level 1 of the Containnent Building. Iko services wre required since the Electric Penetration Assemblies (E.P. A) wre only rated at 355 amps for the Im Voltage Penetrations. Three of the existing 500 Fm feedthrough conductors are being upgraded fran 258 anps to 355 arps to supply the continuous rated current required. All of these powr supplies are non-1E and are fed fran non-1E sources.
- 1. The Contaiment Building feeders are not required during reactor operation. The tenninations to the EPA will be perfonned using prequalified tenninations and insulating materials in accordance with FSAR sectton 8,3.1.1,12. There is no addition of aluninum. Raceways to be added in the Containment Building will be rigid steel and will be constructed in accordance with construction specification X3AR01 E.8 to provide the proper train separations. Also, these raceways and disconnects will be supported per applicable seismic category 1 criteria. The design of these feeders is such that series redundant overcurrent protection is provided by 480 volt breakers 1NB0902. Although three of the E.P.A. Ieedthrough conductor ampacity ratings were increased from 258 amps to 355 amps, calculations have been performed to analyze this configuration which have been performed, reviewed and approved by Conax Corp. (vendor). The E.P.A.s have been previously qualified to IEEE 317 - 1976 and Specification, X3AB03. Thus, because the original design criteria is being met by this proposed design change, there is no increase in the probability of occurrence or consequences of the malfunction of any equipment or emponent assumed to function in accidents analyzed in the FSAR. This includes a review of FSAR sections 6.4, 8.3, and 15, i
61
11 1990 ANNUAL REPORT - PART 2 10 CFK50.59(b) REPORT
- 2. Based on a review of FSAR chapter 15, the proposed change does not create the possibility of an accident or equipuunt/canponent malfunction not described and analyzed in the FSAR. This included review of section 15A.2.1, " Accident Release Pathways", for which the Electric Penetration Assunblies are not listed. Le contairnent feeders are not required during reactor operation and the E.P.A.s are qualified per 1EEE 317, 1976 and Specification X3AB03.
- 3. Revision is made to FSAR section 16.3 Technical Specification improvenent program which will include the additional load center 1NB09 breakers into Table 16.3-5, and thereby assure periodic testing of these contairment penetration conductor overcurrent devices. B erefore there is no decrease in the margin of safety defined by the bases of the Technical Specifications, including the bases to Technical Specification section 3/4.8.4.
89-V1N0117 he proposed change is an addition of a Gaitronics telephone /page wall nounted handset to Auxiliary Building Roan #C106 near the steam generator bloulown panel, h is is an extension of the existing system such that the additional handset will be powered from an existing speaker amplifier (A05) located in the adjacent room #C105. he additional telephone /page equi;xuent and raceway are seismic category 2 equipnrnt which will be installed in a seismic category 2/1 area and supported to seismic category 1 criteria. Per FSAR section 9.5.2.1, the camunications systan is non-safety related and serves no safety function,
- 1. Ec change to plant camunications does not affect any equipurnt or component which is assumed to function during an accident analyzed in FSAR section 15. Ikwver, only seismic category 1 support design will be used in the support of raceways during implenentation of this design.
h erefore, there are no seismic category 2/1 Concerns. 62
II 1990 MHJAL REPORT - PART 2-10 CFR50.59(b) REPORT
- 2. Since the comunications system-is non-safety related, and this change is seismically supportec.
and electrically separated per FSAR 8.3, and passes.through a penetration which is not required to be sealed, the proposed change does i not create a potential for accidents or equipment malfunctions not already described in FSAR section j 15.0, based on a review of the FSAR including sections 8.3, 9.5.2, 9B-69 and 15,
- 3. We addition of telephone /page handsets will not affect the nargin of safety defined by Technical Specification section B3/4.9.5, which addresses having proper comunications during re-fueling.
Eis requiramnt is satisfied by the soundpowered i phones as referenced in FSAR 9B-69, i 89-V2N0119 The change will replace the existing distribution ) class arresters with intenmdiate class arresters l
- for non-class 1E transfoumrs. Surge arresters provide electrical insulation protection due to switching impulses generated in the operation of the electrical system. W e lower switching surge rating of the replacement arresters will provide an increasedn. nrgin of protection against switching surges and consequently increase the reliability of the transfonmrs.
- 1. - Le change does not increase the probability of occurrence or consequences of the malfunction of any equipment or cauponent assumd to function in accidents analyzed in the FSAR.- The proposed change will involve'only non-1E transformers. The transfonmrs and the loads supplied fran these-transformers are not required to function for accident mitigation or for safe shutdown. he loss of any non-1E transformer is bounded by the loss of
.Non-emergency AC power to-the Plant Auxiliaries analysis. This includes a review of FSAR' sections 8.3.1 and 15.0, specifically sections 15.2.6 and 15.0.8.
63
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPOKr
-2. h e change does not create the possibility of an accident or equiment or cmponent malfunction not descriaed and analyzed in the FSAR.
The change involves changes to non-1E transformers, h e consequences of failure of these i transforners is bounded by the loss of Non-energency AC power to the Plant Auxiliaries analysis. This includes a review of FSAR sections 8.3 and 15.0 1 specifically section 15.2.6 and 15.0.8. I l
- 3. We change does not decrease the margin of i safety defined by the bases of the Technical J Specifications. We change involves only _non-1E transforners. These transforners do not supply safety related equipment required for safe shutdown or mitigation and control of accident conditions. This includes a review of Technical Specification bases 3/4.8.
89-V1N0220 This DCP corrects a wiring error so that the local Suppression Indicating Panel 1-2301-Q3-F013 can transadt a trouble signal to the fire protection console in the control roan. h is will be accanplished by- relocating conductors at panel 1-2301-Q3-F013 fran TB1-23 to TB1-2l: and fran TB1-22 to TB1-23. 1, The change is inside a non-1E panel and does not affect the probability of occurrence or consequences of a malfunction of any safety related equipnent or component assumed to fulction in accidents analyzed in the FSAR. The wiring change reflects the design as described in FSAR sections 9.5.1.2.2, Table 9.5.1-9 (NFPA 72D-1979), and 9B.C.6.C(2), and agreer with vendor design document 1X4AX03-5041,
- 2. Relocating of the existing conductors at the terminal block inside local Suppression Indicating Panel to agree with design intent does not create the possibility of an accident or equiptent/
camponent malfunction not described and~ analyzed in the FSAR, f 64
11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3. This change will maintain the safety margin defined in the Technical Specification bases for systems and ccuponents associated with the affected fire zones because the nodification does not change the operability or design of these systems and canponents.
89-V2N0287 1his DCP involves replacing the lovejoy notor controller and its associated ccuponents on the Unit 2 Signa Refueling Machine (2-2101-R6-003) with a Veearc Super 7000 controller, a joystick power supply and associated cauponents. Also included in this change is the addition of a fault relay. The min difference between the controllers is that the m tor speed for the Veearc can be set digitally, whereas the lovejoy is set in an analog mnner.
- 1. This change involves the replacement of the mtor controller on the Unit 2 Sigma Refueling Itichine.
DC-1010, Rev. 5 identi.fies the refueling machine as a non-safety related, seismic category 2 mechanical systan. The refueling m chine does not constitute part of the protection systen and it is not specified as electrical class IE. FSAR section 15.7.4 addresses accidents involving the refueling l machine.- This change will occur during the first refueling outage after core loading verification. Therefore, this change does not increase the probability of occurrence or consequences of-a malfunction of any equipment assumed to function in an accident.
- 2. This change involves replacing the Unit .2 Sigma Refueling Machine mtor controller to improve reliability and maintainability of the machine.
The new controller is functionally similar to the original one and it does not degrade the perfonnance of the machine. Also, it does not impact the fuel: l handling accident analyses or the operation of L safety-related equipment as described in the FSAR. Therefore, this change does not create the possibility of an accident or equipment /ccmponent i- u lfunction not already described in the FSAR. Sections 9.1.4 and 15.0 were reviewed. 65 (
l 11 1990 AtWJAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3. %e replacement of the notor controller on the Refueling Machine has no effect on the nnrgin of safety defined by the bases of the Technical Specifications, section 3/4.9.6.
89-V1N0290 his DCP contains two basic nodifications to the Extraction Stean check valves. We first change is to add packing leakoff drains. Rese leakoff drains will be piped to the existing, normally closed, drain lines for the Fxtraction Steam drain acts. Le second nodification is the welding of tne disc nut to the disc stm. Valves which will be undified are W-4112, W-4113, W-4114, W -4115, W-4112, W-4113, W-4132, W-4133, and W-4134. Rese valves are located in the Extraction Steam supply lines to the number 6A & 6B, 5A & 5B, 4A & 4B, and 3A, 3B, & 3C feedwater heaters. The Extraction Steam Systan (1303) is described in the FSAR section 10.1 and 10.2, and has no safety design basis. Valves are project class 424, Seismic Category 2.
- 1. h is change does not affect the probability of occurrence or consequences of a malfunction of any equipmnt assumed to function in the accident analyses in FSAC Chapter 15. W is change is to a non-safety related system (1303), whose failure will not impact any safety related equipnent, or caupranise the ability of the facility to effect a safe shutdown. B is is based on a review of FSAR sections 10.1, 10.2, and 15.0.
- 2. This change does not affect the operation or function of any canponent. It will not, based on review of FSAR sections 10.1, 10.2, and 15.0, create the possibility of an accident or canpanent malfunction not described and analyzed in the FSAR.
- 3. This change does not affect the operation of the systen or any canponent of th( system. E is change 1s to system 1303 which, based on review of the Technical Specifications, FSAR sections 10.1, and DC-1303, has no safety design basis and is not required or assumed to function in the event of an accident. Here is, thercfore, no decrease in the margin of safety and the Technical Specifications do not require any change, 66
11 1990 ANNUAL REPORT - PART 2 = 10 CFR50.59(b) REPORT 89-VIE 0302 he following cable tenninations are to be lifted and taped in the control roan annunciator cabinets: INCPWEPVA, ANKUAVW, -1NCBDi2AlVA, INCBD12A1VB, INCBD12A1VC, INCBH41A1VT, and ANCPREPVA. Annunciator window tiles ALB54F01, ALB60C04,
.ALB60C05, ALB60C06, ALB60D06, ALB60E06, ALB63B06, and ALB63C06 will be replaced with blank tiles.
The cables and annunciator windows are related to the radwaste facility and the recycle and waste evaporators. We annunciator system is safety class 61J.
- 1. The change does not increase the probability-of occurrence or consequences of the malfunction of '
any equipment or cauponent assuned to function in accidents analyzed in the FSAR since the affected systems are not in operation. Eis includes a review of chapter 15, Section 11.2.
- 2. he change does not create the possibility-of an accident or equipment /can3onent malfunction not described and analyzed in tae FSAR since the affected systems are not in operation. Eis includes a review of chapter 15, Section 11.2.
.3. The change does not decrease the margin of safet./, defined by the bases.of the Technical Specifications. Bis includes a review of the the bases Technical Specification to the-following bases sections: including 3/4.11.1, 3 /4.11.3, ,
6.12, 6.14. 89-V1N0306 This design change revises the existing hydraulic e pump and-thermal relief valves (1-PSV-3006AH & BH,_ 1-PSV-3016AH & BH, 1-PSV-3026AH & BH and 1-PSV-3036AH & B). h e thermal relief valves setpoint will be increased to 4,350 psig, and the. pneumatic pressure regulator will be set to correspond to a hydraulic fluid _ pressure of 3,650 psig. E is will eliminate the overlap of the 67
11 1990 ANNUAL REIORT - PART 2 10 CFR50.59(b) RE10KC hydraulic air paup stall pressure of 3,960 psi and the relief valve reseat pressure of 3,511 psi. The new setpoints will achieve a hydraulic air pump stall pressure (plus gauge accuracy) of 3,700 psig and a relief valve rescat pressure of 3.726 psi.
- 1. Revising the setpoint of the thermal relief valve and the hydraulic pump on the MSIV actuators will not increase the probability of occurrence or consequences of the m1funciton of any equipaent or camponent assuned to function in accidents analyzed in the FSAR, because the fast closure ability of the MS1V's will not be adversely affected. This included a review of FSAR section 10.3, FSAR table 10.3.3.1 (Main Steam System - Failure Modes and Effects Analyses), and FSAR section 15 (Accident Analyses).
- 2. Revising these set points will not create the possibility of an accident or equipment /camponent malfuncion not described and analyzed in the FSAR and does not affect the health or safety to the public. This was based on a review of the FSAR, including sections 6.2 (Essential - Containnent Systems), 7.3.8 (ESFS - Main Steam and Feedwater 1 solation), and FSAR table 10.3.3.1, and FSAR chapter 15,
- 3. Revision of these set points does not adversely affect the MSIV's, or any other system function, operation or response. Therefore, there is no decrease in the margin of safety defined by clw bases of the Technical Specifications. This included a review of Technical Specification 3/4.7.1.5 and its bases.
89-V1NO307 The change is the addition of a Gaitronics telephone /page wall munted handset to Auxiliary Building Roan JA09 and one wall mounted handset with speaker in the Fuel Handling Roan (A10. The hanaset to be located in Rcan A10 will be powered fran a handset on the floor below, which will require a core drill and penetration seal, i l 68
l I II 1990 AtNUAL PIPORT - PART 2 10 CFR50.59(b) REPORT he existing speaker amplifiers in Roan A09 will be replaced will a wall nounted handset which will be relocated to an area ncar the Reactor Coolant Pump (R.C.P.) scal injection valves. These additional telephone /page handsets are category 2 equipment which will be supported per seismic category 1 criteria, along with all associated raceways. Per FSAR section 9.5.2.1, the camunications system is related and serves no safety function.
- 1. The change to plant camunications does not affect any equipment or cauponent which is assumed to function during an accident analyzed in FSAR section 15. Both the camunications equipment and associated raceway will be installed to seismic category 1 criteria such that there are no seismic category 2 over 1 concerns. Also, per construction specification X3AR01- E.8.7.A, the raceway supports shall be designed to seismic category 1 criteria.
- 2. Since the camunications system is non-safety related, the ch a ge does not create a potential for accidents or equipment malfunctions as described in FSAR section 15.0, based on a review of FSAR sections 9.5.2, 9B-69 & 15.
- 3. The addition of telephone /page handsets will not affect the margin of safety defined by section B3/4.9.5, which addresses having proper camunications during re-fueling. This requirenent is satisfied by the soundpowered phones as referenced in FSAR 9B-69.
89-V2NO308 This DCP removes the undervoltage trip function of the CVQ relays for the non-1E 13.8KV and 4.16KV switchgear. Motors fed directly fran the non-1E 13.8KV switchgear buses 1NAA and 1 NAB and 4.16KV switchgear buses 1NA01 and INA04 will no longer be tripped for undervoltage conditions. Annunciation of undervoltage conditions will be retained. 69
1 II 1990 ANNUAL REPORT - PART 2
-10 CFR50.59(b) REPORP Disable the undervoltage trips for noters fed directly frcxn non-1E 480V switchgear buses except buses 2NB01 and 2NB10 which are supplied frcin 1E- ,
buses, i
- 1. This change has no effect on any accident described in the FSAR. (Based on a review of FSAR chapters 8 and 15.) The notors affected by this change, with the exception of the RCP notors which have 1E j undervoltage protection, are non-1E and are not 1 included in tie analysis, i
- 2. The undervoltage trips, which are to be renoved, served two functions. We first was to strip notors !
'in the event of a dead bus. We second was to protect notors frcxn damaging undervoltage conditions. The transforuers feeding the 13.8KV, ,
4.16KV and 480V buses affected by this DCP are not i capable of starting all notor loads connected to the l bus at the same tine. Therefore, in the event of a dead bus, the operator will be required to strip the notor ic> ads before the bus is (re) energized. Scxm of the 480V buses which only feed one or two notors may not require load stripping before (re)energization, h e probability of the station service voltage being' degraded to 90% or lower and remaining at that IcVel long enough to damage notors is very low, h e low voltage condition would probably be caused by a failure of a protective device to clear a fault. Backup Protection should clear the fault before notors are damaged by undervoltage.- Tripping of critical _ notors (RCPs) during this' period could result _in an unnecessary unit trip, he 1E switchgear buses are not affected by this clunige.
- 3. The margin of safety as described by the bases for Techrdcal Specification 3/4.8 is not decreased by this change. No changes to safety related equipment or the function of safety related equipment is made by this DCP. In addition, the bases for the containment penetration conductor overcurrent protection devices. discussed'in 3/4.8.4.1 and safety related notor operated valves thermal overload protection discussed in 3/4.8.4.2 are both unaffected by these changes.
l 70
i II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 89-V2N00310 Lis DCP will renove the existing 4" 90 degree elbow on the river water m ke-up lines to the NSCW Towers, upstream of valves 21402U4013 and 21402U4015 for Trains A and B, respectively, and replace them with a 4" tee. A 4" gate valve will be welded to the tee with a flange on the downstream side of the valve and a 2" reducing flange with a threaded cap on the 2" nipple. W e project class for the piping system is 626. Le chemical injection will be provided by the plant using portable equipment. Le chmicals will be existing chemicals presently used by the plant.
- 1. We river water mke-up pipe is non-safety related and no failure of the ma<e-up pipe caused by this nodification would result in a malfunction of any equipe nt or camponent assumed to function in any accident analyzed in the FSAR. This includes a review of sections 2.4.11, 9.2.1, 9,2.5, 10.4.5 and 15,
- 2. The river water make-up pipe is non-safety related and this modification does not create the possibility of an accident or equipment /cmponent malfunction not described and analyzed in the FSAR. B is includes a review of FSAR sections 2.4.11, 9.2.1, 9.2.5, 10.4.5 and 15,
- 3. B is change is to provide a ch mical injection point and flush point to the make-up piping frm the river water to the NSCW System and does not decrease the margin of safety defined by the bases of the Technical Specifications. This includes a review of Technical Specifications sections B 3/4.7.4 and B 3/4.7.5.
89-V1E0315 Lis DCP deteminates leads K and K1 from switches 1ZS-17801, -17802, -17803, -17804 at junction boxes 1NPJB4831 and 1NIUB4835. Rese leads conu frm position switches on the doors of the RHR Train A & B pump rows and the containmut spray Train A & B pump roms. The purpose of the switches is to alert the control room operator of the status of the watertight doors via annunciator window ALB61F06, " Level D leak detection", and the position status lights on the misc. system equipment panel ZLB-12. 71
Il 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REIORT Leads K and K1 are for the input to the annunciator window. Lifting and taping back the K and K1 leads will eliminate the door inputs to the annunciator wu1dow and retain the status lights on the misc. systems equipannt panel 1-1605-QS-PCF (QPCP). These switches and the annunciator system are not safety related.
- 1. Renoval of the door inputs to the annunciator window -
does not affect the accident mitigating equipnent i assumed to function in FSAR sections 3, 5, 6, 9, and
- 15. his included review of Appendix 3F. The RHR and containment spray system operation is not affected and the pump room doors will continue to be maintained closed.
- 2. Removal of the door inputs to the annunicator window will not create the possibility o? an accident or malfunction not described and analyzed in the FSAR including sections 3, 5, 6, 9, and 15 as discussed i above. !
- 3. % e door inputs to the annunciator window has ne I affect on the margin of safety defined by the Technical Specifications bases, including bases B 3/4.4.6, B 3/4.5, and B 3/4.6.2. he-doors will be maintained closed, so the requirements of fire protection,- flooding, post-loca recirculation filter system and redundancy will be maintained.
89-V2E0316- This DCP deteminates leads K and K1 from switches 2ZS-17801, -17802, -17803, and -17804-at . junction boxes 2NFJB4831 and 2NPJB4835. These leads come from position switches on the doors of the RHR Train A & B pump rooms and the containment spray Train A & B pump rooms. The purpose of the switches is to alert the control rocxn operator of the status of the watertight doors via annunciator window ALB61F06, "Leval D leak detection", and the position status lights on-the misc. systems equipaent aanel 2LB-12. Leads K and K1 are for the
-input to tTe annunciator window. Lifting and- 4 taping back the K and K1 leads will eliminate the door inputs to the annunciator window and retain i
72
. x ~- . - _ : . . - . - . . - a...
i 11 1990 ANNUAL REPORT - PART 2 10 CIV50.59(b) REPORT the status lights on the miscellaneous systems equipent panel 2-1605-QS-PCP (QPCP). These switches and the annunciator system are not safety related.
- 1. Renoval of the door inputs to the annunicator window does not affect any accident mitigating equipnent assuaed to function in FSAR sections ;
3, 5, 6, 9, and 15. Eis included review cf Appendix 3F. he PJR and contaiment sprey system operation is not affected and the pump rom doors will continue to be maintained closed.
- 2. Renoval of the door inputs to the annunciator window will not create'the possibility of an accident or malfunction not described and analyzed in the'FSAR including sections 3, 5, 6. 9 and 15 as discussed above.
- 3. The door inxits to the annunciator window has no .!
affect on the margin of safety defined by .Se i Technical Specifications bases, including bases B 3/4._4.6, B 3/4.5, and B 3/4.6.2. W e doors will be maintained closed, so the requirements of fire protection, flooding, post-loca recirculation filter system and redundancy will be maintained. 89 "'N0322 E is design change modifies the containment spray flow indication circuit fran a square log function to a linear function. Bis design change requires the modification of existing hardware currently shown on 2XGAU01-451, Rev. 8. his DCP routes the analog input signal to the:RPU associated with the PSMS computer through the new NLP1 card instead-of the existing NQP1 card. However, no change is made to the signal, i.e.i both cards are 61J and the 1-5 VDC signal is still proportional to NaOH-delta-p. Bere is no change to the PSMS. We PSMS is a safety-related post-accident infonnation processing system. his change will correct IED 1148, identified during the control roan design review, which addresses a problem with readitg the contaiment spray additive flow on existing 73
11 1990 ANNUAL PETOKr - PAKr 2 10 CFK50.59(b) PEPORT indicator 2F1-930. Linearizing the flow sig,nal and replacing the indicator with a 0-10 VDC linear scale indicator will provide a more accurate and discernible indicator reading for verifying the proper Na0li flow.
- 1. We square root function card (1601) and the powr supply card (N121) are 61J cards in the non-cafety related cabinet 2-1604-QS-PC3. This change canplies with IEEE 279 which requires that this indicator nodification for the spray additive tank eductor flow not affect the function or operation of any process instrununtation protection system function.
We installation of the circuit cards will be per the physical separation crit m a of IEEE 279 and 384 herefore, this design change does not increase the occurrence or consequences of the malfunction of any equipnent/canponents assuned to function in a ac^idents analyzed in the FSAR. This loop is utilized for indication only and has no control function.
- 2. This change linearizes the contaiment spray additive tank eductor flow signal. We signal is used for control roan indication only and does not have any control function which could interact with equipuent/canponents or cause any accident. The integrity of the containnent spray indication function is not degraded by this nodification. The qualification of the control roan panel and the 7300 panels is not adversely affected. Electrical separation and seismic mounting are maintained.
Therefore, this proposed change does not create the possibility of an accident or equipment /canpanent malfunction not described and analyzed in the FSAR.
- 3. Technical Specification bases, particularly B 3/4.6.2, are not affected by this change to the containment spray additive tank eductor flow signal. his signal is used for indication only and does not alter the flow rates, system performance criteria or other assumptions made in the radiological assessments presented in the FSAR.
74
11 1990 AtNUAL RFJORT - PART 2 10 CFR50.59(b) REl W T 89-VIN 0323 Increase the setpoint on hydraulic puup start pressure switch (41A) frctn 1500 + 25 psi to 2000
+ 25 psi on valves 1-PV-3000, 30TO, 3020 and
'3030. Rese valves are in the main stean line atmspheric relief valves (ARV's) which are project class 212 and required for safe shutdown fran the hot standby ntxle . The pressure relief setpoint of the valve is not being changed, only the hydraulic punp start pressure.
- 1. This setpoint change will change the ARV accunulator pressure range fran 1500-2500 psi to 2000-2500 psi. Le smaller delta-p has a posit %
effect on the fatigue life of the accuaulator, out has no effect on the maxinun stress. We hydraulic punp and notor my cycle more frequently and therefore experience increased war. Howver, the mxinun pressure is unchanged. Vendor reconnendation on hydraulic pum cycling is not changed. It therefore does not increase the probability of occurrence or consequences of the mlfunction of any equipment or canponent assumd to function in accidents analyzed in the FSAR including chapters 7.4,10.3 and 15.
- 2. This design raises the low pressure serpoint in the ARV actuator. h is does not adversely affect any other cauponent in the valve and any other equipmnt with the possible exception of the hydraulic aunp and nutor due to increased cycling.
However, the vendor recannendation on hydraulic punp cycling is not changed. It, therefore, does not increase the possibility of an accident or equipmnt/ component malfunction not described and analyzed in the FSAR.
- 3. This el ange does not decrease the margin of safety as defined in the Technical Specification bases since the function and operation of the valve is not changed. his includes a review of the bases in sections 3/4.3 and 3/4.7.
75
i 11 l 1990 ANNUAL. REPORI' - PART 2 10 CFR50.59(b) REPORT 89-V2N0324 h is DCP increases setpoint on hydraulic pump start 1 pressure switch (41A) fran 1500 +25 psi to 2000
+ 25 psi on valves 2-PV-3000, 30TO, 3020 and
'J030. Rese valves are the min steam line atmspheric relief valves (ARV's) which are project class 212 and required for safe shutdown fran the hot standby node, h e pressure relief setpoint of the valve is not being changed, only the hydraulic pump start pressure.
- 1. 'Ihis proposed setpoint change will change the ARV accunulator pressure range from 1500-2500 psi to l 2000-2500 psi. The sunller delta-p has a positive -
effect on the fatigue life of the accunulator, but i has no effect on the maxinun stress, he hydraulic panp and mtor nay cycle mre frequently and therefore experience increased wear. However, the maxinun pressure is unchanged. Vendor recanmndation on hy,draulic pump cycling is not It, theretore, doea not increase the changed. probability of occurrence or consequences of the malfunction of any equipnent or cauponent assumed to function in accidents analyzed in the FSAR including chapters 7.4, 10.3 and 15.
- 2. h is design raises the low pressure setpoint in the ARV actuator. E is does not adversely affect any other couponent in the valve or any other equipaent with the possible exception of the hydraulic pump and notor due to increased cycling. However, the L
vendor recamendation on _ hydraulic pump cycling i is not changed. It, therefore, does not create the possibility of' an accident or equipment /cauaonent malfunction not described and analyzed in the FSAR.
- 3. Eis change does not decrease the margin of safety as defined in the Technical Specification bases since the function and operation of the valve is not changed. E is includes a review of the bases, sections 3/4.3 and 3/4.7.
89_-VIN 0326 -This change involves connecting and remving lighting fixtures to/fran the Normal or Essential lighting system. The Normal and Essential lighting system are both safety class 6 Seismic category 2,
-per table 3.2.2-1 in the FSAR and Design Criteria
-DC-1808, p 76 l
11 1990 ANNUAL PIIGT - PART 2 10 CFR50.59(b) IU ORT 1, h is change does not_ affect safety systems, setpoints or other equip e nt or components assuaed i to function in accidents analyzed in the PSAR, including sections 8.3, 9.5.3 and 15,
- 2. h is change does not create the possibility of an accident or equi;nent/camponent malfunction not described and analvzed in the FSAR, including sections 8.3.1, 9.l.3, 15 and the Plant Security Plan. L e fixtures are not m unted over any equipmunt or safety related conduit or cables.
- 3. The margin of safety as defined by the bases of the !
Technical Specification section 3/4.8 is not affected by this change. 89-V2N0327 Rese changes involve connecting lighting fixtures to the Normal lighting system. %e Nornal lighting system is safety class 6, Seismic category 2, per ; table 3.2.2-1 in the FSAR and Design Criteria i I DC-1808.
- 1. B is Cange does not affect safety systems, setpoints or other equipent or carponents assumed to function in accidents analyzed in the FSAR, including sections 8.3, 9.5.3 cad 15,
- 2. This' change does-not create the possibility of an accident or equipment /cutponent umlfunction not described and analyzed in the FSAR, including sections 8.3.1, 9.5.3, 15 and the Plant Security Plan. The fixtures are not mounted over any equipment or safety related conduit or cables.
- 3. The margin of safety as defined by the bases of the Technical Specification section 3/4.8 is not affected by this change.
V1N0015 This modification affects the electro-hydraulic actuators for the atmospheric relief valves (1PV-3000/3010/3020/3030). L e modification involves the installation of a separate drain line fran the valve port of the pilot-to-open check
-valve (285*) to the oil reservoir to eliminate pressure build-up downstream of the check valve.
77
. . . _ . _ _ _ . . . _ _ _ . - ~ - _ _ _ , _ _ . - _ _ . _ _ . . ~ _ _ _ _ _ .
l l a II-1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT To accomplish this, the existing e M 3 5
- s. .ne pilot-to-open check valve (item 'e, " te # 4 18*
is removed, and new tubing, from OP - - 9 01 1' reservoir of the hydraulic powr unEs n .; _ be installed. Connection for valve 18* is to be i plugged. * - Vendor drawing item number 1 ~. This modification improves the reliability of the R ARV by eliminating the back pressure concern at the ! pilot port. All changes are made internal to the hydraulic actuator and no other safety related equipment or system is affected. Tubing added is-safety-related. There is no change in the quanity_ of combustibles. All material / equipment used in this nodification will meet original design requirements. Therefore, the change does not affect the probability of occurrence or consequences of the malfunction of any eculpcent or ccmponent assumed to function in accicents analyzed in the FSAR.
- 2. This modification involves tubing internal to the ARV hydraulic actuator to improve the availability and reliability of the ARV. No other equipment or system will be involved in the proposed change. Folicwing installation, testing will be perfonned to ensure operability of the valves after nultiple stroking. Therefore, the proposed change will not create the possibility of an accident or equipment / component malfunction not described and analyzed in the FSAR.
- 3. The change improves the-reliability of the ARVs..
Therefore, the margin of safety defined in Technical Specifications will not be decreased by the change. 90-VIN 0033 A. It is proposed to delete the active valve status associated with valves 1LV-459 and ILV-460._ 1hese va: .as are project class III. 78
l 11 1990 NNUAL REIOKr - PART 2 10 CFR50.59(b) RElVRT
'Ihe position of these valvas is detennined by their pilot solenoid valves project class 11J.
This vill be a paper change to change the solenoid valves project class fran 11J to 62J, and will not affect the shell pressure boundary function of the valves and vill not necessitate changes to plant hardware. B. Utis design change proposed to rove the speed control needle valves for valves 1LV-459 and 3LV-460. It also proposes to install check valves around the speed control valves to allow for quick opening of the process valves. The process valves (1LV-459 and 460) are project class 111. The solenoid valves are currently project class 11J but are being reclassified to project class 62J oince itun (A) above has determined that the solenoid valves are not required to perfonn a safety-related function. The sensing line was originally installed as project class 212 but is being reclassified as project class 424 based on the solenoid valve reclassification.
- 1. a. Historically, the IDCA-related accident analyses have asstned that the letdown isolation valves close, as designed. A failure of the letdown isolation valves to close was postulated and reviewed. This review included the ,otential for long-tenn loss of pressurizer aenter function (due to uncovering the heaters) and also mass and energy reicases fran the Reactor Coolant Systan. This review concluded that these valves do not need to be assuned to close in an accident situation. This review is doctrrented in the Revision 1 response to RER 88-0861 and Westinghouse letter GP-14369, dated t'. arch 31, 1989. Therefore, this change does not increase the probability of occurrence or consequences of the malfunction of any equip:ent er camonent assuned to function in accident analyzed in the FSAR.
79
b 11 1990 M MIAL RE10RT - PART 2 10 ClK50.59(b) REIVRT I I
- b. This design change will inprove the reliability l of the letdom isolation valves and therefore will not increase the probability of l malfunction of these valves. The consequences '
of malfunction are covered in (a) above. , 2. a. During certain accident scenarios, the Pressurizer Relief Tank (PKr) rupture disks my open due to f1cu frctn the thennal relief valve on the letdom line inside Containment, should the letdcun isolation valves not close. This has been reviewed and failure of the PRT ruptc e disks under these conditions is bounded by the existing plant safety analyses. Therefore, this change does not create the possibility of an accident or equipmnt/ ccamonent mlfunction not described and analyzedintheFSAR.
- b. If a check valve fails in the open position, the quick closure of the associated letdcun isolation valve could cause water hanner or flashing to occur in the regenerative heat exchanger. The effects would be localized
! to'the heat exchanger. Repeated water hanmer or flashing in the heat exchanger my cause damage to the tubes. Mcwver, damage to the shell of heat exchanger is unlikely duc its design for high pressure service. If a breat were to occur at the heat exchanger, the consequences have already been analyzed as. part of the high energy line break analyses and are acceatable. Since reactor coolant f1cus on bota the shell side and tube side of this heat exchanger, a heat exchanger tube rupture would not be a breach of the reactor coolant pressure bcundary or would radioisotopes carry over to other systems. 80
11 1990 N MIAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3. The letdwn isolation valves provide a neans of purification and securing inventorythe nonm1f1wpath.
control clean-up,These valves are not operated by the Reactor Trip or Engineered Safety Features Actuation Systmis (3/4.3.1 and 3/4.3.2). They are not required for high energy line break isolation (3/4.3.3.11), or contaiment isolation (3/4.6.3). The bases for establishing the RCS leakage requirments of section 3/4.4.6.2 are not impacted by the changes being tmde to these valves since the operators can detennine any RCS inventory letdwn through this f1wpath. Therefore, the unrgin of safety defined for the bases of the - Tecanical Specifications listed above is not decreased. 90-VIN 0036 With the present design, if a leak or rupture were to occur in the turbine building instrument air supply, the turbine air header would isolate when ! the low instrument air pressure is sensed. This isolation action would assure that the prinary plant had sufficient instrunent air pressure to operate. This design change rmoves the autanatic isolation valve function by rmoving actuator air supply, disconnecting power to the valve pilot solenoid, and eliminating one setpoint output fran aressure switch IPLS-19414. The instrument air 1 ender isolation vo've will no longer isolate the turbine building st.,. ply header. However, the lw ! air pressure alann will runain operable to alert the operator.
- 1. The design change renoves the autunatic signal required to isolate the instrunent air supply header. Valve function remins the sme for the manual action. The instrunent air systun provides supply air to the safet
'in FSAR Table 9.3.1-2. As y related valves specified identified in FSAR i
Section 9.3.1.3, pneumatically operated valves which are essential for safe shutdown and accident mitigation are designed to assunu a fail safe I l l 81 l
II l 1990 ANNUAL REPORT - PART 2 l 10 Cm50.59(b) REPOKr l l position upon loss of air pressure. Perfonunce of the nodification has no affect on the asstmed failure position of these valves. The valve will continue to fail as designed during a aostulated accident follwing inplenentation of t tis i i nodification. Therefore, this change does not increase the probability of occurrence or consequences of the m1 function of any equiptent or ccuponent asstuned to function in accidents analyzed , in the FSAR. l l 2. No new potential accidents are created as a result j of this nodification since no new failure nodes I are introduced. This change eliminates the l possibility of spurious valve operation leading ! to unit tri s. In addition, the worst scenario associated ith this nodification would result in failure to isolate the turbine building portion I of the instnnent air system following a line l l break downstrean of the instnnent air header i isolation valve. This occurrence would result in reduced air supply to the safety-related pneunatic valves. Howver, as described in FSAR section 9.3.1.3, these valves fail in their fall safe position. Therefore, this nodification does not create the possibility of an accident or equipment emponent malfunction not described and analyzed in the FSAR.
- 3. This nudification will enhance the availability of the turbine building instnment air header and will l provide the control roan operator with annunciation should low air pressure exist within the instnnent air header. Therefore, this nodification does not reduce the margin of safety as defined in the bases of Technical Specifications.
90-V2N0037 If a leak or rupture occurred in the turbine building instnnent air supply, the turbine air header is isolated when the low instrunent air i pressure is sensed. This isolation action assured l that the prinnry plant had sufficient instnoent l air pressure to operate. This design change t renoves the autanatic isolation valve function l 82
! 11 1990 AWUAL RF.lOKr - 1%1tT 2 10 CitSO.59(b) RElOKr by ret ving actuator air supply, disconnecting power to the valve pilot solenoid, and eliminating one set point output fran pressure switch 2PSL-19414. The instrumnt air header isolation valve will no longer isolate the turbine building supply header, lkuever, the lw air pressure alann will rennin operable.
- 1. The design chance renoves the autarntic sigrml required to isoiate the irmtrunent air supply header. Valve function rainins the sano for the nanual action. The instrunent air systen provides supply air to the safety-related valves identified in FSAR Table 9.3.1-2. As specified in FSAR section 9.3.1.3, pneuratically operated valves which are essential for safe shutckun and accident mitigation are designed to assure a fail safe position upon loss of air pressure. Perfoninnee of the nodification has no affect on the asemed failure position of these valves. The valves will continue to fail as designed during postulated accidents following inplenentation of this nodification. Therefore, this change does not increase the probability of occurrence or consequences of the unlfunction of any equiptent or emponent assured to function in accidents analyzed in the FEAR.
- 2. No new potential accidents are created as a result of this nodification since no new failure nodes are introduced. This change eliminates the possibility of spurious valve operation leading to unit trips. The wrat scenario associated with this nixlification would result in failure to isolate the turbine building portion of the instnnmnt air system follwing a line break domstream of the instruient air header isolation valve. This occurrence would result in reduced air supply to the safety related pneuratic valves.
Ikuever, as described in FSAR section 9.3.1.3, these valves fail in their fail safe position. Therefore, this modification does not create the possibility of an accident or equiptent ccuponent ma11 unction not described and analyzed in the FSAR. i 83
II 1990 ANNUAL PEPORT - PART 2 . 10 CFR50.59(b) REPORT i 3. This rmdification will enhance the availability of the turbine building instnrmnt air header and will j provide the control roan operator with annunciation should lw air pressure exist within the instrumnt air header. Werefore, this nodification does not reduce the nargin of safety as defined in the bases of Technical Specifications. 90-V2N0068 h is DCP involves rumval of the internal emponents of check valves 2-1592-U4-186 and 2-1592-U4-187 associated with the Unit 1 essential
- chilled water system (Train A&B) as shwn on P&ID 2X4DB221. The check valves and associated piping are project class 313.
- 1. Rmoval of the check valve internal cwponents eliminates the original concern of possible damage to the valve cwponents caused by disc oscillations in the present configuration. No new cwponents are added to the check valves which could restrict the chilled water flw path. Since the chilled water flw is not inpacted and backflow prevention is not required, there will be no increase in the probability of a malfunction of the essential chilled water systun or any other equipment or cmponents assumed to fuiction in accidents analyzed in the FSAR. mis includes a review of the FSAR, including sections 7.3, 9.2.9, 9.4 and 15, i 2. The safety function and nornal operation of the essential chilled water systen is not inpacted by climinating backflow prevention. No new failures and/oraccidentsarecreatedbytheproposedvalve nodification. This is based on a review of the FSAR, including sections 7.3, 9.2.9, 9.4 and 15.
- 3. ' h e essential chilled water system will continue to perform this function since rmoval of the check
- valve's internal cmponents will have a negligible effect on the systen flwrate and results in no. change to the system operation. Therefore, this change does not decrease the nnrgin of safety defined by the bases of the Technical Specifications, including section 3/4.7.11. , 84
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II 1990 A!4NllAL RElUKr - PAKr 2 10 Clt50.59(b) RE10Kr 90-VltiOO71 This DCP provides the an extended narrw range level tap on the loop 4 steam generator (1-1201-B6-004). The follcving design and construction wrk will be perfonted by Westinghouse in accordance with Field Cimnge tiotices (FClis) CEO-40524, 40525 and 40525A (log !4trnbers 1XfM19-20, 004 and 20, 005) . Thic wrk will involve penetrating the secondary side of the stean generator belcw the transition cone. A new nozzle and root valve will be added and tubing will connect t.his tap to the lw pressure side of a new differential pressure transtuitter (1Ur-30548). The high pressure side of 1Ur-20548 will be connected to the drain valve on the high aressure sensing line of 1Ur-548. IUr-10548 will x nounted on the stee floor stand with 1Ur-548. IUr-10548 will not be penranently wired.
- 1. 'Ihe new level tap, additional piaing and tubing connected to stean generator nteber 4 has been designed in accordance with the requira:ents of the /GE Txailer and Pressure Vessel Code, Section III, subsection NB,14C, and NF. The tmrgin of safety inherent in the use of this code is sufficient to ensure that the structural integrity of the stean generator pressure boundary is not reduced, ccinpared to that of the original design.
The existing steen generator instrtmentation rennins operable. She operation of the plant is not being changed by this ICP. All the existing setpoints, alanns, and trips will still be operable and are not affected by this change, with the exceation of the PKr level alann which will be disaaled during pwer ascension. Annunciator Response Procedure 17012-1 states that no initial operator actions are required for this alann. Since the PKr level will be indicated on the nain control board, operation with the alann disabled will have no adverse affect on plant safety. Instrtment 1Ur-10548 is connected to the high pressure sensing line of ILT-548. 85
-.- - . . . . - . _-- . -.~..
I 11 1990 AN! CAL lu30RT - PART 2 10 CFit50.59(b) lu30RT ILT-548 is used to mnitor stearn generator narrow range level in conjunction with the reactor control, reactor protection, and uterg,ency safety featurer actuation systens. 1LT-1054 3 is scimnically qualified to the sano requirments as 1LT-546. The interconnecting tubing installation confoms to the ASME Section III, Class 2 code, , which is the smo class and code that the ILT-548 installation conforms to. 1here is no electrical l connection shared between the two devices. The accident toalyses docunented in imAR sections 3.6 6.2, 10.3 (Table 10.3.3-1), and 15 are not adversely inpacted by this design change.
- 2. The data acquisition configuration changes the location of pressurizer relief tank (PRT) hvel indication on the main control board fran IL1470 to IL1-462. Also the PRT high/ low alann will be out of nervice. The operators will need to une a tenporary data logger if RCS leak rate trending is required during the data acquisition period. Data will be collected during power ascension following the 1R2 refueling autage. The data collection is estimated to take less than 72 hours and the PKf level loop L-470 will be restored to its normal configuration as quickly as possible. The data acquisition configuration has been evaluated by
- operations personnel for impact. Based on these evaluations, this change will not create the possibility of an unanalyzed accident or equiptent failure.
- 3. The bases for the Technical Specifications, including sections 2.2, 3/4.3, 3/4.4.5, 3/4.4.10 and 4.0.5, are not adversely inpacted by this change. No safety limits, lbniting safety systun settings, reactor trip instrunentation, or essential instrumentation is adversely affected by
< this change and data acquisition process.
~86
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l 11 1990 /&%\L R130RT - PART 2 10 cm50.59(b) RI:lORT 90-V2N0076 This DCP involves the deletico of the unin SEQ 2 and auxiliary fecckater tutperature nonitoring systun by deleting all systan to:perature indication, differential tarperature indication, and differential tutperature alanm fran the noin control board. The rumining systuu thernuwells, tutperature clu:ents and cabling up to and including those boards (In the QBCP) associated with the turperature indications, will rurain in place. All cards associated with differential tarperature indications and the differential tenperature alanns will be retuved. If necessary, the tutperature indication cutputs can be utilized to derive backup plant calorinetric tutperature data. Currently, the unin and auxiliary feeckater tu:perature nonitoring systen senses the line tutperature at two separate points of both the t.nin feedwater and auxiliary feedwater line upstream of the steam generators. 'Ihese tumerature sensors serve to detect the back leakage of steam and/or hot water frau the steam generator into the feedwater lines. This tutperature infonintion is used to minimize the possibility of water hanner due to the introduction of cold feedwater should back leakage occur.
- 1. The Vogtle unin and auxiliary feedwater systun configuration (separate nain and auxiliary feedwater lines, nultiple in-line check valves, and pipe routing) and the S/G design minimize the potential for steam backleakage and therefore the potential for feedwater line water luumer. In addition, to culpensate for the deletion the nain and auxiliary feedwater temperature nonitoring system, Westinghouse has recumended the inplenentation of additional feedwater systun unintenance and 87
1 i
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I II 1990 /M4UAL REIVIC - PART 2 10 Cn60.59(b) RI'IVIC startup procedural changes. 'Ihese naintenance and procedural changes, in conjunction with the feedwater systen design, will serve to i further preclude feedwater systun water hsnner conditions innediately upstrenn of the stean generators. Deletion of the tm perature indicators, differential tutperature indicators, and differential tmperature alarms fran the main control board will have no effect on either unin or auxiliary feedwater systan operation. Rese temperature readouts wre not mployed in the perfonnance of nonini unin or auxiliary feedwater systun actions as other systun readouts are available. In
-addition, these tmperature readouts and alanns wre not relied on to assess or perfonn auxiliary feedwater system operations during an accident. Westinghouse reemnended operating procedural changes call for increased unintenance of those valves which prevent systun backflow, and maintaining closer control of feedwater f1w rates during those low power operations under which water hamer nay occur. Therefore, their runoval will not affect either the norcal or mergency feedwater systen operation.
Therefore, the deletion of the unin and auxiliary feedwater tmperature monitoring system will not increase the probability of occurrence or consequences of the nn1 function of any equipnent or emponent assuned to function in accidents analyzed in the FSta.
- 2. We feedwater tmperature monitoring systan is non-safety related and is utilized only during low power operations'. Its deletion will therefore have no offect on the safety function of the auxiliary feedwater systan.
l 88
'l 11 1990 ANNLIAL REPORT - PART 2 10 CHL50.59(b) REPORT Although the potential for feedwater systm water hamur has not been entirely eliminated, the possible existence of conditions in the feedwater piping necessary for water hm mer
- have been significantly reduced by a combination of feedwater piping configuration, steam generator design, and procedural changes. Therefore, the mnitoring of the feedwater temperature adjacent to the steam generator is no longer required and the main and auxiliary feedwater temperature mnitoring system can be deleted.
With or without the feedwater temperature 4 m nitoring system, the worst caso design bases accident (loss of feedwater due to a IIne break) r eains unchanged. Therefore, the proposed deletion of the main and auxiliary feedwater temperarure m nitoring system is enveloped by existing systen accident analyses. The main control board 01CB) has been seismically qualified by Westinghouse. Ratoval of the tenperature indicators and alanns fran the MCB, along with their replacement covers, does not adversely affect the dynamic responses of the MCB and therefore has no impact on its seismic qualification.
-Based on the above, the proposed change does not create the possibility of an accident or equipment /cauponent malfunction not described and analyzed in the FSAR.
- 3. The plant Technical Specifications do not contain any specific references to w in or auxiliary feedwater system start-up differential temaerature limits at the steam generators. Despite tae deletion of-the main and auxiliary feedwater-tauperature nonitoring systens, this change will not alter or reduce the. ability of the auxiliary feedwater system to mitigate the consequences of the design basis loss of feedwater accident, 89
- II
! 1990 AtMIAL REPORT - PART 2 10 CFR50.59(b) REIOiG Relevant sections of the Technical Specifications (3/4.3 and 3/4.7) have been reviewed and there is no decrease in the mrgin of safety defined by the bases of the Technical Specifications. 90-Vlt40077 1his design change nodifies the Unit 114atural Draft Cooling Tower by: (a) adding pipe extensions at the perinuter of the tower, (b) adding additional nozzles (256 locations) at the perineter of the towr, (c) replacing nozzle essenblies at the flune bottans with french sprayers, (d) replacing all reruining single-s alashplate assunblies with sonic-welded douale-splash plate nozzle assemblies (R-C type), (e) rearranging nozzle sizes to provide nore water at the outer portions of the tower, reduced water loading in the center, (f) adding diverters (stainless steel angles) to imrove entering flow to the distribution laterals in t ae first section of the fitnes at the second and third risers, and (g) adding PVC fill at the perineter and center of the tower.
- 1. 7he Circulating Water Cooling Towers are not assmed to function in an accident described in FSAR chapter 15 (Accident Analysis). Installation of the proposed change will not cause the mlfunction of other equipment asstned to function.
- 2. This design change does not create the possibility of an accident or equipnent/canaonent malfunction not described and analyzed in tae FSAR which could affect the health and safety of the public. The Circulating Water Cooling Towers are not assuned to function during the limiting cases described in FSAR chapter 15 (Accident Analysis).
- 3. This design change is associated with systen 1401 and does not decrease Technical Specification safety mrgins since it has no safety design bases. This is based on review of the Technical Specification bases, including section B 3/4.7.
90
Il 1990 N M RL REPORT - PART 2 10 CFR50.59(b) REPORT 90-V1E0082 Replace all bearings of heater drain pmps 1-1304-P4-001, 1-1304-P4-002 and the spare pmp, with the nw tri-land desigi bearings to eliminate vibration problans. Furtlier, replace pmp's discharge head /nounting flange present gasket with a new flexitallic gasket to prevent leakage prob 1 mis. In addition, run a high pressure flush line fran 2nd stage to bottan bearing of the pmps to prevent cavitation problun. Le heater drain punps are located in the turbine building and are project class 424,
- 1. Le nodifications are internal to the heater drain pmps. Le pmps are non-safety related, and are not assaned to function during postulated accidents. No safety-related carponents will be canpranised as a result of inplanenting the design change. 21s included a review of FSAR section 10.4.7, and chapter 15.
- 2. Le nodifications to the heater drain pmps will be performed in accordance with the design, material and cuality standards applicable to the pwp. The nodfications are internal to the pmp and will not create the possibility of an accident or equipment /camponent malfunction not described and analyzed in the FSAR.
- 3. Based on review of Technical Specificat'on 3/4.7 and B 3/4.7, the changes have no effect ut the margin of safety.
90-V2E0083 Replace all bearings of heater drain pmps 2-1304-P4-001and1-1304-P4-002with the new tri-land design bearings to eliminate the vibration aroblans. Furthernore, replace cumps's discharge ,ead/nounting flange present gasket with a new flexitallic gasket to prevent leakage probluns. In addition, run a high pressure flush line fran 2nd stage to the bottan bearings to improve lubrication and extend the life of the bearings. Le heater drain pumps are located in the turbine building and are project class 424 91
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II 1990 N W AL REPORT - PART 2 10 CFR50.59(b) RDWT
- 1. Le nodifications are internal to the heater drain pumps. The pwps are non-safety related, and are not assmed to function during postulated accidents.
No safety-related ccuponents will be ccupranised as a result of inplemnting the design change. his I included a review of FSAR section 10.4.7, and I chapter 15.
- 2. Le nodifications to the heater drain punps will be perforned in accordance with the design, unterial and cuality standards applicable to the pmp. The nocifications are internal to the punp and will not create the possibility of an accident or equipment /canponent mlfunction not described and analyzed in the FSAR.
- 3. Based on review of Technical Specification 3/4.7 -
and B 3/4.7, the changes have no effect on the unrgin of safety. 90-VCN0088 h is DCP provides a routing for unscheduled, non-safety related " fiber optics" camunication cable in existing underground duct run (outside the pcuer block) betwen the Service Building and the Production Warehouse.
- 1. his nodification routes unscheduled, non-safety related " fiber optics" camunication cable in an underground conduit adjacent to other conduits
; carrying non-1E cables betwen the Service Building and the Production Warehouse. No equipaent or canponent assuned to function in accident analysis are affected directly or indirectly as a result of this change. In addition, this modification does not affect fire protection / safe shutdown analysis. Hence, the pro >osed change does not increase the
-probability of occurrence or consequences of the
- nn1 function of any equipcent or couponent asmzned to function in accidents analyzed in the IMR.
92
11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REP 01T 2, No new potential accidents or events are created as a result of this nodification since no new failure nodes are introduced. The proposed routing of " fiber optics" carmunication cable is non-safety related. No other safety system will be affected by this change. Therefore, this change does not create the possibility of an accident or equipu m t/ catponent mlfunction not described i.nd analyzed in the TSAR.
- 3. This change does not affect the Technical Specification bases for any portion of the Technical Specification since there are no bases defined or inferred for installing non-cafety related " fiber optics" cable in an underground duct.
90-V1N0094 The design change is to replace the existing two coil Main Turbine Electric Trip Solenoid Valve (ETSV) with a new, one coil ETSV. The previous four wire, tw coil circuit consisted of tw cables in one conduit for each of the tw coils that are witvd in parallel. The four wires are grou>ed together at the receptacle of the two coil TrSV. The rcalacemnt single coil ETSV will reuse the i t two caales in each of the two conduits and a loss l of continuity in the wiring in one conduit will not ! " de-energize the single coil ETSV and trip ! the Turaine. Therefore, the redundancy of the , external wiring fran the ETSV will remain. The 1 redundancy of two coil ETSV is replaced with a one coil design that is more reliable overall. The FISV is a non-seismic class 62J device and is part of non-safety related system 1613.
- 1. This change should decrease the probability of occurrence of the malfunction of the ETSV that is assum d to function during a Turbine trip that is analyzed in the FSAR, sections 15.2.2, 15.2.3, 15.2.5 and 15.2.6.- The consequences of the malfunction of the ErSV are unchanged since the-present tw coil solenoid and replacemnt one coil-i solenoid are both fail-safe devices in that their L
failure results in a Turbine trip. l I e 93 p
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11 1990 AtNUAE REIORT - PART 2 10 CFR50.59(b) REIORT
- 2. Le ETSV is a fail-safe device in that, upon its
- failure, a Turbine trip will occur. Berefore, i the change does not create the possibility of an accident or the malfunction of equiment/
canponents not described and analyzed in TSAR section 15.2.2, 15.2.3, 15.2.5 and 15.2.6. This , change does not affect the response time on turbine trip - oil pressure that is listed in Table 7.2.1-3.
- 3. We EIW retrofit will decrease the probability of Turbine trips and the ETSV will remin a fail-safe device in that its failure will achieve a turbine trip. Berefore, the change does !
L not affect the margin of safety defined by bases 2.2.1, 3/4.1, 3/4.3.2, and 3/4.7.1 in the Technical Specifications for Wrbine Trip and Reactor Trip Systun Interlocks. 90-V2N0095- The design change is to replace the existing , two coil Main Turbine Electric Trip Solenoid Valve (ETSV) with a new one coil ETSV. The previous four wire. two coil circuit consists of two cables in i one conduit for each of the two coils that are
- i. wired in parallel. We four wires are grouped together at the receptacle of the two coil ETSV.
The re>1acamnt single coil ETSV will reuse the s two caales in each of the two conduits and a loss of continuity in the wiring in one conduit will not de-energize the single coil EISV and
. trip. Werefore, the redundancy of the external wiring fran the ETSV will remain. The redundancy i of two coil ETSV is replaced with a one coil design ,
that is m re reliable overall. The ETSV is a non-seismic class 62J device and is part of non-safety related systan 1613,
- 1. - his change should decrease the probability of occurrence of the malfunction of the ETSV that is-assumed to function during a Turbine tri analyzedintheFSAR, sections 15.2.2,1$.thatis 2.3, 15.2.5 and 15.2.6. .The consequences of the "
malfunction of the E7SV are unchanged since the ~ present two coil solenoid and replacamnt one coil L . solenoid are both fail-safe devices in that their failure results in a Turbine trip. L- 94 ; _.u.___ __ . . _ _ _ _ _ _ _ _ _ . _ . - . . _ _._. .. _ . . _ . . _ _ _ _ _ _ _
i II 1990 NMIAL RElVRT - PARP 2 10 CFR50.59(b) ICIORT
- 2. The IIISV is a fail-safe device in that, upon its failure, a 'Ilirbine trip will occur. Therefore, the change does not create the possibility of an accident or the unlfunction of equip:ent/
cmponents not described and analyzed in ITAR section 15.2.2, 15.2.3, 15.2.5 and 15.2.6. This change does not affect the response tire on turbine trip - oil pressure that is listed in Table 7.2.1-3,
- 3. The EISV retrofit will decrease the probability of Turbine trias and the EISV will ra:nin a fail-safe device in L ut its failure will achieve a turbine trip. Therefore, the change will not affect the nargin of safety defined by bases 2.2.1, 3/4.3.1, 3/4.3.2, and 3/4.7.1 in the Technical Specifications for Turbine Trip and Reactor Trip system Interlocks.
90-VlliOO98 1his change unkes a section of piping on lines 1-1301-384-1" and 1-1314-159-2" renovabic by adding a flanged section of piping to each line. These lines are surdl bore piping, preject class 424.
*1his change also nodifies the pipe support design for supports V1-1301-38441631, V1-1305-060-H068 and V1-1314-159-il612 by unking than renovable. The supports are seismic category 2, M4SI B31.1, designed to neet 2 over 1 design criteria. The system function or support integrity is not effected by the,' changes as substantiated by calculations, as oferenced in the ICP calculation record. Supports and piping are only to be renoved during an outage and nust be reinstalled before systen operation.
- 1. This piping and supports in this DCP were reviewd for impact to interconnecting equip:ent and emponents. This review, and calculations perforned for this change, denonstrate the adequacy of the nodifications. Therefore, based on a review of the FSAR, including section 15, this change does not affect any equipcent or emponent function.
95
II 1990 AtM1AL REPORT - PART 2 10 Cnl50.59(b) REPORT
- 2. W e piping and supports in this DCP are substantiated by calculations that shw the change is within the design criteria and code allowables, The change is for piping, flanges and supports, and does not affect any equixent/ccuponent, function or operation and, taerefore, does not create any possibility of accident or malfunction.
- 3. The calculations performed for this DCP shw the piping and support stresses to be in accordance with the applicable design criteria, codes and standards as identified in the Design Insut Record of this DCP. These criteria estaalish the design bases for piping and sup3 ort stresses.
Inherent in these design bases is tae same mrgin of safety as the original design. Since the. calculations are within desi p criteria, the margin of safety as defined in the'aases of the Technical Specifications including the bases of Technical Specification 3/4.7, is not-decreased. 90-VIN 0109 This change is the addition o_f two bypass lines around Heater Drain Punp discharge control valves LV-4331 6 LV-4332. These lines, (1-1304-570-8", 1-1304-571-8") will be 8", sch. 40, ASIM A-105 piping with a single globe valve (1-1304-U4-760 and 1-1304-U4-761) in each line. These lines will be in the 1304 (Heater Drain) system which is non-safety related and project class 424.
- 1. 'Ihe change does not increase the probability of occurrence or consequences of the malfunction of any equipment or emponent which is assued to -
function in the FSAR accident analysis. FSAR section 15 does not identify LV-4331 and LV-4332 as ccuponents which are assumed to function in the event of an accident, and the failure or rupture. of a bypass line will-not cause the malfunction of any equip. Tent assue d to function in accidents analyzed in the FSAR. 96
11 l 1990 AMUAL REPORT - IV,RT 2 , 10 CIISO.59(b) REPORT
- 2. The piping and supports in this DCP are substantiated by calculations that shcv the change is within the design criteria and code allcuables.
The change is for piping, flanges and supports, and does not affect any equirent/ccinponent, function or operation and, taerefore, does not create any possibility of accident or malfunction.
- 3. The calculations perforned for this DCP shcw the piping and support stresses to be in accordance with the applicable design criteria, codes and standards as identified in the Design In3ut Record of this DCP. These criteria estaalish the design bases for pi ing and supaart stresses.
Inherent in these desi bases is t1e stum nurgin of safety as the orig 1 design. Since the , calculations are within design criteria, the unrgin of safety as defined in the bases of the Technical Specifications, including the bases of Technical Specification 3/4.7, is not decreased. 90-VIN 0109 This change is the addition of two bypass lines around licater Drain Puap discharge control valves LV-4331 & LV-4332. These lines, (1-1304-570-8", 1-1304-571-8") will be 8", sch. 40. ASIM A-105 piping with a single globe valve (1-1304-U4-760 and 1-1304-U4-761) in each line. These lines will be - in the 1304 (Heater Drain) system which is non-safety related and project class 424,
- 1. The change does not. increase the probability of occurrence or consequences of the malfunction of ,
any equipment or couponent which is asained to function in the FSAR accident analysis. FSAR section 15 does not identify LV-4331 and LV-4332 as ccxuponents which are assumd to function in the event of an accident, and the failure or rupture of a bypass line will not cause the malfunction--of any equipment assuned to function in accidents analyzed in the PSAR.
-96 !
II 1990 AN!UAL REIVRT - PART 2 10 CFR50.59(b) REPORT
- 2. This change does not create the possibility of an accident or the tm11 function of equiptent or
_ canponents not described or analyzed in the FSAR. This is based on a review of the FSAR, including section 15 (Accident Analysis). 1his piping nodification will nect original design requirments of the Heater Drain System.
- 3. Based on a review of Technical Specification b;ses, including sections 3/4.4 and 3/4.7, this change does not decrease the nargin of safety.
90-V2N0110 This change is the addition of two bypass lines around Heater Drain lirup diccharge control valves-LV-4331 and LV-4332. These lines, (2-1304-570-8", 2-1304-571-8") will be 8", Schedule 40, AS7M A-106 , piping with a single globe valve (2-1304-U4-760 and 2-1304-U4-761) in each line. These lines will be in the 1304 (Heater Drain) systan which is non-safety related and project class 424,
- 1. The change does not increase the probability of occurrence or consequences of the unlfunction of any equipnent or canponent which is assuned to function in the FSAR accident analysis. FSAR section 15 does not identify LV-4331 and LV-4332 as canpanents which are assuned to function in the event of an accident, and the failure or rupture of a bypass line will not cause the malfunction of any equipnent assuned to function in accidents analyzed in the FSAR.
- 2. This change does not create the possibility of an -
accident or the nulfunction on equipaent or canpanents not described or analyzed in the FSAR. This is based on a review of the FSAR, including section 15-(Accident Analysis). This piping i modification will neet original design require ents of the Heater Drain Systen. l 97
l 11 1990 N M1AL REIORT - PAIC 2 10 CIT 30.59(b) R13 ORT
- 3. Based on a review of Technical Specification basca, including sections 3/4.4 and 3/4.7, this change does not decrease the rmrgin of safety.
90-V2N0111 This design change adds flange " fixtures" to the spare penetrations (IS, 55 and 90) located in the contaiment building, during Mode 5 and/or Mode 6 for outage related work activities, Wich require contaiment penetration. The flange " fixtures" will replace the blind flanges for the insertion of eddy current, sludge lancing & ISI equiptent during a Unit 2 outage. Le outage " fixture" is imde-up of a flange plate with threaded sleeve ports. Rese sleeve ports allow access for cables and connection points for lead-in hoses during outage related wrk.
- 1. This change does not increase the probability of occurrence or consequences of the imlfunction of any equiptent or emponents assmed to function in the accidents analyzed in the FSAR including those in sections 3, 6, 9 or 15 of the PSAR. We
" fixture" being installed will only be installed when the plant is in Mode 5 (cold shutdown) or 6 (Refueling) . It will only have to perform its design function under the wrat case conditions of a refueling accident and GL 88-17, which does not postulate contaiment pressurization. This
" fixture" does not have to be designed to function for other postulated Design Basis Accidents.
Berefore, this fixture does not have to be designed, fabricated, or installed to the ASME Section III, Class 2, requirmento of the penetration. A seismic evaluation has been perfonwd to insure that the structural integrity of the penetration (and " fixture") is maintained during a Design Basis Earthquake, h us, there is no inpact on the integrity of the contalment liner plate fission product barrier. 98
i II 1990 A!ML PEOKf - PAKf 2 10 CFK50.59(b) lu2VKr
- 2. This design change for the nodified condition of the penetrations unintains a boundary that will prevent the exchange of the contaiment enviroment with the outside enviroment under postulated refueling accident conditions. 1he tulified penetrations tects the intent of all applicable Technical Specifications under nornal and postulated refueling accident conditions as defined in the FSAR. Based on a review of FSAR sections 3, 6, 9 and 15, this design change wxld not create the possibility of an unanalyzed or undescribed accident or equi ttent /camponent an1 function.
- 3. This design change unintains the containnent penetration in a condition that will prevent the direct camunication of the containnent enviroment with the outside enviroment in accordance with the bases for Technical Specifications 3/4.9.4 for nornal and postuluted refueling accident conditions during core alterations. Prior to ascending to Mode 4, the outage " fixture" will be replaced with the bolted blind flanges. After testing of the penetrations in accordance with the llRr procedures, the new configuration will neet the currently existing basis for containnent leakage. Therefore, this change does not affect the systen, equi'nent function or operation, and does not affect t,e safety nargin defined by the bases of Technical Specification 3/4.6.1, and 9.6.1.2.
90-VIE 0112 This design change installs flange joints downstreign of the 3/4" drain valves to the stemn generator blowdown heat exchangers 1-1407-E6-001, 2, 3, 4, 5, 6, 7, 8. This portion of the SG blcwdown systun is non-safety related, project class 424. The new flanges will be located in the Auxiliary Building P,oan tio. C-108,
- 1. The SG blowdown heat exchangers are non-safety related and no failure of the systun caused by this nodification would result in a un1 function of any equipnent or canponent assuned to function in any accident analyzed in the FSAR. This includes a review of FSAR sections 10.4 and 15, 99
i i 11 1990 M4NUAL RElORT - PART 2 10 CFR50.59(b) REIORT
- 2. The SG blodwn heat exchangers are non-safety related and no failure of the system caused by this nodification would create the possibility of an accident or ecultuent/cmponent un1 function not described anc. analyzed in the FSAR. Original design criteria continues to be net for the line after nodification. This includes a review of FSAR sections 10.4 and 15,
- 3. This design change does not decrease the nargin of safety defined by the Technical Specification bases because neither the heat exchanger nor their functions are discussed in Technical Specifications, and all the original design criteria continue to be net by this change. This included a review of the bases of Technical Specification sections 3/4.4.5 and 3/4.7.
90-V2E0114 This design change addresses the installation of 4 piping, valves, and instrunentation which will allcw the Turbine Plant Cooling Water Systan to supply seal and cooling water to the Unit 2 Circulating Water pumps and notors as -a backup supply in the event Utility Water is not available. Sequence 1 of the DCP will allow the installation of piping fccxn the discharge of TPCW pump 2-1405-P4-501 to the first isolation globe valve on the'new line (2-1405-L4-594). It will also cover nodifications to Utility Water line 2-2419-L4-543 , and the portion of the new line 2-2439-L4-543 and the portion of the new line (2-1405-L4-594) which , taps into line 543 up to the first isolation globe < valve. These lines will be installed during a alant , outage and will be capped off. . Sequence 2 of t1e DCP will rmove the caps and join the two new lines to form one new line. The new piping is located at the Circulating Water pump structure. 1.- Equipnent affected by this design change is not assunod to function in.an accident analyzed in chapter 15 (Accident Analyses) of the FSAR. 100
11 1990 AtMAL REIVRT - PAKP 2 10 ClR50.59(b) REIORT Installation of the changes proposed in the DCP will not cause the nalfunction of other equituent asstraed to function. The nw piping is not safety related or seisudcally qualified. This is consistent with the existing criteria for these systcxns. The piping is not located in a seismic 1, or 2 over 1, area.
- 2. This design change does not create the possibilit,f of an accident or equipment /ccnnonent m1 function not described and analyzed in t1e FSAR which could affect the health and safety of the public. This was based on a reviw of the FSAR, including section 15 (Accident Ar.alyses). i
- 3. The m difications described in this design change do not decrease Technical Specifications safety mrgins since the systems affected, systans 2419 and 1405, have no safety design bases. This is based on a review of Technical Specification bases, including section B 3/4.11, 90-V2E0116 This design change will allow the use of split or cartridge type acchanical seal or the existing packing on Turbine Plant Cooling Water Pumps (TlW) 2-1405-P4-501 and 2-1405-P4-502. The change will also route utility water to the TPCW ptops for cooling and flushing of the mechanical seal, and add a blind flange to line 2-1401-L4-610 upstream of valve 2-1401-U4-545. The TPCW ptmps are located at the circulating water ptmp bat,in pad and are project class 626,
- 1. The TPCW Punps are non-safety related and no failure of the ptops caused by th4s mdification would result in a nalfunction of any equi lment or l
ccuponent assumed to function in any accident analyzed in the FSAR. This includes a review of sections 9.2 and 15 of the FSAR. l 2. The TPCW Ptmps are non-safety related and this m dification does not create the possibility of an accident or ecul nent/ccuponent t un1 function not described anc, analyzed in the FSAR. This includes a review of FSAR sections 9.2 and 15. 101
l 11 1990 NME REIORT - PART 2 10 CFK50.59(b) REPORT
- 3. This change is to allow the use of a split or cartridge type nochanical seal or packing on TIG hnps 2-1405-P4-501 and 2-1405-P4-502 and does not decrease the nurgin of safety defined in the bases of the Technical Specifications. This includes a review of Technical Specifications section B 3/4.7.
90-V1H0122 This design change will increase the length of the check valve disc back stop on check valves 1-1306-U4-004 & 006 (MPP Ibrbine Driver Im Pressure Check Valves). The length of the disc sto, will be increased by approxinategy 1-inch, watch will limit thediscogeniggtoa60 angle, instead of the present 80 -85 angle. This change will have un insignificant effect on the flow capacity of the check valves. These check valves are a part of systen 1306 (Turbine Drive Steam Systun), which is non-safety related, project class 424,
- 1. Modification of the opening angle of these check valves is being done to increase their reliability by decreasing the inpact of flw turbulence on the check valve disc. This change will not have any adverse impact on the function of system 1306, or any other system. This design change will not increase the probability of occurrence or consequences of the tralfunction of any ecuiprent or cmponent assteed to function in accicents analyzed in the FSAR. Check valves 1-1304-U4-004 and 006 are not identified ao equirent/cmponents which are assmed to function in the event of an accident. These check valves are installed to prevent high pressure stean from back-ficuing to a lower pressure side during unin turbine start-up.
This included a review of FSNI sections 10.2, 10.4.7, and FSAR chapter 15.
- 2. This design nodification only revises the full open angle of these check valves. There will be no changes to system eneration or response to systan 1306, or any other system, as a result of this cha':ge. This change does not create the possibility cf an accident or equipaent/ccuponent malfunction not described and analyzed in the PSAR.
This included a review of FSAR sections 10.2, 10.4.7, and FSAR chapter 15. 102
11 1990 AN!A1AL REIORT - IMT 2 10 CFR50.59(b) REKET
- 3. Decreasing the full open angle of these check valves will not decrease the nargin of safety as defined in the Technical Specification bar,es, including the bases to sections 3/4./.1 or 3/4.7.2, as the flw resistance through the valve will not be appreciably increased and the valves should be nere stable.
90-V2N0126 Inpluientation of this design change will result in the following nodifications to the Unit 2 Condensers "A", "B", and "C", equip:ent tag nunber 2-1305-E4-005, 006, and 007, respectively.
- 1) The 20" dianeter, Schedule 5, lienter Drain Tank "B" liigh Irvel Dmp sparger (Condenser A, Connection 19) will be replaced with 20" diuneter, Schedule 20 pipe, with pipe cap, and drilled per the original design. The nw sparger hole pattern varies slightly fran the original design due to additional field welds.
7hese welds are required, since the replacenent sparger will be installed in short r.ections due to limited access into the hotvell. The minor variances frar the original design have been discussed with, and approved by, the original supplier (Ecolaire). Additional drain holes are included to expedite drainage and minimize the potential for water / steam ha:mer in the event of rapid drain valve cycling. Also, a detail will be added to the Condenser "A" vendor drawing allowing the field to install stiffener plates in existing sparger support nebers, if required.
- 2) The new sparger (with design subnitted by Ecolaire L'ondenser) will be added to connection 61 on Condenser "B" (Main Steam Drain Pot MaLnifold). Presently, the internal design for this connection consists of a baffle plate.
New support (s) will be required. 103
i II i 1990 NEAL REPORT - PART 2 10 CFR50.59(b) RElWI l 1
- 3) A note will be added to the approp"riate condenser drawings for Cwicnsers A", "B", j
! and "C", which will allcv the field to install the condensate suction scret.'.s in the hot-well in sections, instead of one p.%ce. ! 4) The lateral structural suppo-ts for the Main ! Steam Dunp Spargers (connection 9, Main i Turbine By-:> ass) will be replaced with heavier !
. mnbers . Tacse are the two lower min steam
- dtmp spargers that turn down and are routed
!- to the hot-well (2 on each condenser). Also, a pad /bm per will bo installed on the sparger where it contacts the lateral support surface to distribute inpact loads more evenly int.o the pipe.
- 1. . Based on a review of FSAh scctions 15.1.3, 15.2.2, i 15.2.3, 15.2.5, and 15.6.2 and the fact that the nain condensers and the Tutbine By-Pass System are i non-safety related ccxnponenu, there will be no increase in the probability of occurrence or consequences of the m1 function of any eculpnent or caponent asstred to function in accicents analyzed in the FSAR.
- 2. Based on the review of FSAR sections 15.1.3, 15.2.2, 15.2.3,-15.2.5 and 15.6.3, and because this design change will be accordance with the original design criteria, and the fact that the main condensers and the Turbine By-Pass System are non-safety related, this design change will not create the postibility of an accident or equipment /
ccuponent m1 function not described and analyzed in the FSAR.
- 3. The changes described in question 1 (Description of proposed change, test, or experinent) are internal to the main condensers. The main condensers are o not addressed in the Technical Specifications or its associated bases. This is based on a review of Technical Specifications,-including section 3/4.7.1 Plant System - Turbine Cycle) and its assc>ciated bases.
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i . 11 1990 /mtAL REPORT - PART 2 10 CFR50.59(b) PIIORT 90-V2N0131 The 500KV breaker current transfon: ors (CI's) which are located at PCB 161520 and PCB 161620 in the switchyard, for the Unit 2 main transfon:cr differential relay (587U1), will be changed fran the 3000/5A ratio tap to the 2000/5A ratio tap.
- 1. The change of the tap settings fran 3000/5 to 2000/5 on the current transfoniers has no effect on the systen design or operation of any equipent or canponent asstred to function in accidents analyzed in the FSAR, section 15. The offsite pcuer systen is not safety-related and is not taken credit for in the accident analyses. Therefore, this change does not increase the probability of occurrence or consequences of the m1 function of any equiptent assa ed to function in accidents analyzed in the FSAR.
- 2. The proposed change of the tap setting does not add any new sources of accidents. Therefore, it does not create the possibility of an accident not described and nnalyzed in the FSAR. This includes a review of FSAR sections 15 and 8. This change should prevent the miseperation of the breaker and still provide the correct balance of sensitivity with security.
- 3. This change will result in a nore stable electrical switchyard and correct the problen which caused the Unit 2 trip on 3-20-90, this change in the protective relaying for the main transfonier does not decrease the margin of safety as defined in the bases to the Technical Specifications, including the bases to section 3/4.8, 5 or 6.
90-V1N0132 A. Rencve RTD Bypass loop 1 1 solation valve 1-1201-U4-007 and replace it with a spool piece. Both the pipe and valve are project class 111. This design change is required because this valve will no longer function, i i 105
11 1990 Ath'UAL REPOKr - PART 2 10 CFR50.59(b) REPOKr B. The addition of tmporary shielding uny be required around valve 007 while it is being cut out to reduce the radiation exposure to personnel in the area. Up to 200 pounds of shielding nny be placed on the valve while alant is in Modes 5 or 6. The shielding nust ae rmoved before proceeding to Mode 4,
- 1. a. The KrD's on this bypass line are asstned to function in accidents analyzed in the FSAR, however, this design change will not adversely affect their operation or system calibration.
This change does not adversely affect the HELBA results including pipe whip, jet inpingment } or enviromental parancters,
- b. %e addition of the tenporary shielding will c
not affect the operation of this line. The f - KrDs are not required to be operable in Mode 5 or 6 and the shielding and pipe removal process ' will be controlled to prevent any seismic 2 over 1 concerns. This will be empleted by [' securing the snielding to the valve, and supporting the pipe negnur .ith chain falls during its rmoval. Therefore, neither the eliminatien of the valve, nor the addition of the teaporary shielding will increase the probability of occurrence or consequences of an accident described in the FSAR, including those in sections 3.6, 6.2 or 15.
- 2. a. Le RTD Bypass line will neet the original RCS piping design criteria. We valve is used only as a backup during maintenance to isolate the RCS loop 1 cold leg,
- b. L e addition of temporary shielding will not alter the system operation or permanent design margin ar.d will not create a seismic 2 over 1 concern. Berefore, this design change does not create the possibility of an accident or equipnent/ component malfunction not described and analyzed in the FSAR, including sections 3.6, 6,2 and 15, 106
11 = 1990 ANNUAL REFORT - PART 2 10 CFR50.59(b) REPORT
- 3. h e system function, operation and calibration is not being changed. The original piping _
criteria is being met by this change. Berefore, the margin of safety defined in the bases of the Technical Sae:ifications is not being changed, including tae bases to sections 2, 3/4.1, 3/4.2, 3/4.3, 3/4.4 ar.d 3/4.9. 90-V1N0133 Plant Vogtle uses two safety related diesel generators per unit to provide A.C. power to the essential loads when the offsite sources are not available. he diesel generators are project class 015 and provide power to the project class llE 4.16-KV switchgear. We controls for the diesel generators are project class 11J, 61J, 62J, or 015 depending'on their function and location. h is DCP proposes to disable several diesel engine autmatic trips which provide engine protection for operation during a loss of offsite powr (IDSP) event. ,
- 1. Bypassing the trips on LOSP start will have no effect on the probability of malfunction of the diesel generator or its cmponents, h is design change will require the operator to mnitor the alam status 'of the diesel generators during IDSP initiated operation. If the operator does not take appropriate action quickly enough and the diesel is damaged, the system response will be no different that if the diesel had tripped.- Failure mde effect analysis shown on FSAR Table 8.3.1-3 (items 4 and 41) describes the consequences of failure of a diesel generator to start or run, his table shows that there is no adverse impact on-equipment assumed to function in the accident analyzed in chapter 15. The consequences of a catastrophic failure (missile generation) have been evaluated as part of the hazard evaluations for the diesel generators and the results are not 107
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'l l
11 1990 ANNUAL REPORT -'PART 2 10 CFR50.59(b) REFORT I affected; Adding contacts parallel to safety 1 injection celay contacts -K609,15-16 has no affect ! on the perfomance of the Solid State Protection l System (SSPS). The diesel generator control l panels, the essential load sequencers, and the SSPS - 1 are all safety-related and train separated. This i design change-does not create any new train interconnections.
- 2. During MSP initiated diesel generator operation, the operator in the control room will now have to determine whether or not to trio the diesel if it develops an alam condition. Currently, the -
operator only has to perform this oversight during safety injection or mergency mnual initiated operation. The effects of failure of the-diesel generators to run are documented in-FSAR Table 8.3.1-3 (items 4 and 41). The circuit mdifications only involve mving the IDSP start contacts to the emergency start circuit on the diesel generator. The correct operation of this circuit is tested during surveillances required-
-by the Technical Specifications. No new single-failures are created. Therefore, this design change can not lead to an accident or malfunction -
not described or analyzed in the FSAR. a
- 3. This design change will not decrease the margin of safety as defined by the bases of the Technical-Specifications, incluting the bases for 3/4.4, 3/4,5,3/4.8,3/4.9, or the bases for Surveillance Requirement 3/4.8.1. 'Ibe mart,in of safety of 1 Engineered Safety Feature sys; ems and refueling operation is related to the availability of the site electrical powr system to provide sufficient power to the safety-related equipment required for the safe shutdown of the facility and the mitigation and control of accident conditions.
within the facility. . A loss of offsite powr is a 108
--II -
1990 ANNUAL PZPORT - PART 2 10 CFR50.59(b) REPOR1' contributing event during other postulated l accidents. The ability of the diesel generator to i start-and continue to run during events which H contribute to accidents is an increase in the u rgin of safety. A faulted diesel generator would now continue to run instead of autmatically tripping upon receipt of one of the bypassed trip signals. A single faulty sensor will not erroneously trip the diesel engine. The operability of the onsite A.C. power sources will not be degraded by inplementing this design change. 90-V1N0135 This DCP adds a carbon steel plate to the .I Integrated Head Package CRDM Cooling Shroud assembly to seal an accessway which was cut during i investigation of a canopy weld seal leak. The reactor vessel, reactor vessel head and CRDMs are l project class 111. However, the CRDM cooling { shroud does not perfom a safety related function, j 1.- The Integrated Head Package CRIN Cooling Shroud helps to remvc heat frmi the CRDMs. However, the ! shroud does not perfom a safety function and is ! not required to function during an accident. The ! structural integrity and mounting of the shroud is not adversely inpacted by this repair. ! Therefore, the shroud will not fall and damage other equipent during a seismic event. There will ; be no change to syst m operation, and the repair materials and coatings will be compatible with the original material. Therefore, this repair will not increase the probability of occurrence or consequences of malfunction of any equipent or component assumed to function in accidents l analyzed in the FSAR, including those in chapters 3, 4, or 15. t 2. Restoration of the Integrated Head Package CRDM
- i. Cooling Shroud creates no possibility of an-1 l \
)
109 1 l
11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) RElWl' accident or equipuent/canponent malfunction not addressed in the FSAR, including sections 3.9.4, 4.5.1, 4.6.1, or 9.4.6. No change in system operation is being made.
- 3. This DCP does not decrease the margin of safety defined by the bases of Technical Specifications, including the bases to 3/4.1.3 and 3/4.10, 90-V1N0136 Reactor Vessel Level Instrunentation System (RVLIS) reactor vessel head sensors will be inverted (top to bottom) to accanmodate calibration inaccuracies due to air in-leakage to the sealed portion of the sensor during refueling disassembly and maintenance.
Westinghouse safety evaluation checklist SEL-891127-C addresses this change.
- 1. The RVLIS provides caerators with infonnation to monitor and assess tie reactor coolant inventory following an accident to ensure t. hat adequate core cooling is available. During implenentation of this modification the plant will be in a cold shutdown condition. As described in Technical Specification 3/4.3.3.6, the RVLIS is not required to be operable in this node of operation. By reorienting the transmitter und minimizing the effects of air inleakage which affect instrument calibration, the safety related function of the system will be enhanced. Ranounting of the transmitter will be perfonned in accordance with the original design details so as not to affect the seismic qualification of the equipment. In addition, the anomalies with the RVLIS have been addressed by Westinghouse in NSD-TB-89-09, which provides a basis for the proposed changes.
Therefore, the proposed change does not increase the probability of occurrence of the malfunction of any equipment or couponent assuned to function in accidents analyced in the FSAR. 110
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II 1990 ANNUAL REPORT - PART:2 10 CFR50.59(b) REPORT-
- 2. During'implenentation of the modification, the plant will be in a cold shutdown condition.- In this node l of operation, the RVLIS is not-required to be operable. No new potential accidents are created as a result of this nodification since no new failure nodes are introduced. The function of the' transmitter will be enhanced by this nodification and will neet the original design requirenents. Therefore, this nodification does -
not create the possibility of an accident or equipnent emnonent malfunction not described and analyzed in t3e FSAR.
- 3. The Technical _ Specification bases for accident 1 nonitoring instrumentation indicates that the RVLIS is provided to ensure that sufficient information is available to nonitor and assess selected plant variables following an accident. This nodification will enhance-the operability of the RVLIS and will. I provide; the operator with nore reliable and accurate information following an accident. Therefore, this I modification does not reduce the margin of safety as defined in the bases of Technical Specifications.
90-V2N0137 Plant Vogtle uses two safety related diesel generators per unit to provide A.C. power to the essential-loads when the offsite sources are not
-available. The diesel generators are project class 015 and provide power to the project class <
11E 4.16-kV switchgear. The controis-for the diesel generators _are aroject class 11J, 61J, 62J, or 015, depending on tieir function and location. This DCP proposes to disable several-diesel engire - automatic trips which provide engine protection for operation during a loss of offsite power (IDSP) event.
- 1. _ Bypassing the trias on LDSP start will have no effect on the proaability-of malfunction of the diesel generator or its c mponents. This design change will require the operator to nonitor the alarm status of the diesel generators during IDSP initiated operation. If the operator does not take appropriate action quickly enough and the diesel is damaged, the system response will be no different l
111
i 11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) RITORf that if the diesel had tripped. Failure mde effect analysis shown on FSAR Table 8.3.1-3 (items 4 and 41) describes the consequences of failure of a diesel generator to start or run, his table shows that there is no adverse impact on equipnent assumd to function in the accident analyzed in chapter 15. We consequences of a catastrophic failure (missile generation) have been evaluated as part of the hazard evaluations for the diesel generators and the results are not affected. Adding contacts parallel to safety injection relay contacts K609,15-16 has no affect on the perforwince of the Solid State Protection System (SSPS). h e diesel generator control panels, the essential load sequencers, and the SSPS are all safety-related and train separated. This design change does not create any new train interconnections.
- 2. During IDSP initiated diesel gene ~ or operation, the operator in the control roan vn.1 now have to detennine whether or not to trip the diesel if it develops an alarm condition. Currently, the operator only has to perform this oversight during safety injection or umrgency manual initiated operation. The effects of failure of the diesel generators to run are docu m nted in FSAR Table 8.3.1-3 (items 4 and 41). The circuit modifications only involve moving the 1DSP start contacts to the emrgency start circuit on the diesel generator. The correct operation of this circuit is tested during surveillances required by the Technical Specifications. No new single failures are created, herefore, this design change can not lead to an accident or malfunction not described or analyzed in the FSAR.
- 3. This design change will not decrease the margin of safety as defined by the bases of the Technical Specifications, including the bases for 3/4.4, 3/4.5, 3/4.8, 3/4.9, or the bases for Surveillance Requirement 3/4.8.1. The margin of safety of the reactor coolant systan, emergency core cooling system and refueling operation is related to the 112
11 1990 ANNUAL REPORT --PART 2
-10 CFR50.59(b) REPORT availability of the site electrical power system to provide sufficient power to the safety-related equipnent required for the safe shutdown of the facility and the mitigation and control of ac ident conditions within the facility. A loss of offsite power is a contributing event during other postulated accidents. W e ability of the diesel generator to start and continue to run during <
events which contribute to accidents is an increase in the margin of safety. A faulted diesel generator would now continue to run instead of automatically tripping upon receipt of one of the bypassed trip signals. A single faulty sensor will not erroneously trip the diesel engine. The operability of the onsite A.C. power sources will not be degraded by implementing this design clwige. 90-V1N0138 'lhis proposed change is to allow the emergency diesel generators lA and 1B, project class llE, to be started by an arcrgency signal and have the high jacket water tria bypassed, in addition to the other trips whica are already bypassed by an emergency start signal (Safety Injection (SI), Ioss of Offsite Power (LOSP), or Energency Manual Start). This change will add isolation valves in the instrurient tubing between the DG high jacket water temperature elements and the local DG control panel. The valves will nomally be closed, but the valves may be opened to allow additional engine protection when perfonning a non-emergency manual start or surveillance of the diesels.
- 1. The valves being added are seismically and environmentally qualified for their environment and will be mounted in such a way as to not overstress the instrument tubing. Isolating this trip will not cause a malfunction of the diesel engine or any other equipment or ccxuponents assumed to function in accidents analyzed in the FSAR. The valves will be located away from the engine, near a support, to prevent any significant vibration effects. Proper valve position will L
be ensured by administrative control. Incorrect l positioning of the valves would only make the trip active, as it is currently. 113
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- 2. W ere-is no change to the DG trip logic, other than the blocking of the trip. A separate high jacket water temerature alam is still available locally and in the control room. The safety benefit gained by the increased reliability of the DG when this trip is bypassed is judged to outwigh the effect of an additional operator decision to trip the DG if a high temperature alam is received. B is change does not create the possibility of an accident or equipent/canponent malfunction not described or analyzed in the FSAR, including the failure of the DG, As docmonted in the REA in FSAR Table 8.3.1-3, if one DG fails, the other is available. In addition, the operator can manually stop the DG if a high tauperature alarm is received.
- 3. E is design change will not decrease the margin of safety as defined by the bases of the Technical Specifications, including the bases of 3/4.4, 3/4.5, 3/4.8, 3/4.9, or the bases for the surveillance requirement in 3/4.8.1. The margin of safety of the Engineered Safety Feature systems and refueling operation is related to the availability of the site electrical powr system to provide sufficient power to the safety-related cauipment required for the safe shutdown of the facility and the mitigation and control of accident conditions within the facility. A loss of offsite powr is a contributing event during other postualated accidents. L e ability of the DG to start and continue to run during events which contribute to accidents'is an increase in the margin of safety.
Removal of this automatic trip would allow the DG to continue to operate in the event of a false high-jacket water tauperature. Le operability of the onsite A.C. power sources will not be degraded by implementing this design change, and will be enhanced in certain conditions. 90-V1N0139 This DCP mdifies the (A'NS) Anticipated Transient
- Without Scram (MSAC) Mitigating System Actuation Circuitry panel (1-1626-QS-#E), he mdification is within the panel. This DCP 114 l
l 4 i ,I l l 11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT upgrades the diagnostic software by replacing the existing Erasable Progrannable Read Only Ksory (EPROMs) with new EPROMs and also e difies the N EAC circuitry by relocating a wire fr m the ANA REF COM to 15V COM bus.
- 1. The change provided in the DCP is to be mde to non-safety related equipmnt/ system. This mdification does not affect the isolation devices provided to buffer the AMSAC outputs frm the safety-related final actuation device circuits.
The system is considered as being a control system that is not required for plant safety as verified by Westinghouse safety evaluation SECL 90-119,
'Iherefore, this change does not increase the probability of occurrence or consequences of the un1 function of any equipmnt or emponent assumed to function in accidents analyzed in the FSAR.
- 2. No new potential accidents or events are created as a result of this modification since no new mdes of failure are introduced. The change covered in the DCP is to a non-safety related equipment / system. This change does not affect the function / operation of the equianent/ system involved, nor will it affect the operation of any safety related equipent. Modification within the AMSAC panel will not affect seismic qualification of the panel. Therefore, this change does not create the possibility of an accident or equipnent/cmponent m1 function not described and analyzed in the FSAR.
- 3. The DCP provides for change to a non-safety related equipment /syst m which is not addressed in the Technical Specifications. In addition, this m dification does not affect the isolation devices provided to buffer the AMSAC outputs fr a the enfety-related final actuation device circuits.
Therefore, there is no reduction in the nargin of safety defined-by the bases of the Technical Specification. 115
l 11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) RFf0RT 90-V2N0142 This DCP allows 30 amp breakers 2AY1A-05 and 2BY1B-05 and 20 amp breakers 2CY1A-05 and 2DY1B-05, in the Vital 120V AC Distribution Panels 2-1807-Q3-VII, VI2, VI3 and VI4 to be replaced _with 35 and 30 amp breakers respectively. Similarly 15 amp breakers 2AD12-03, 2AD12-08 and 2BD12-03 in the 125V DC Distribution Panels 2-1806-Q3-DA2 and DB2 will be replaced with 20 amp breakers.
- 1. h is increase in breaker sizes have no effect on the downstream loads because there is no increase of downstream loads on the breakers. The new breakers are same in form, fit and function as the existing breakers. Bis modification will eliminate the inadvertent tripping of the Breakers during a-Station Black Out condition due to loss of ventilation to the Vital AC and DC SWGR rooms.
Berefore, the change does not increase the probability of occurrence or consequences of malfunction of any equipment or ccxuponent assumed ! to function in accidents analyzed in the FSAR.
- 2. No new potential accidents are created as a result of this modification since no new failure modes are intrcxluced. The increase in breaker ratings, breaker coordination and the associated existing cables have been analyzed in calculation number MX3CT08, Rev. A1, and found to be adequate and safe. The emparable Unit 1 panels are addressed in FSAR table 8.3.2-5. Failure of these panels have been evaluated and determined to be acceptable. Berefore, the proposed change does not create the possibility of an accident or equipment /com3cnent malfunction not described and analyzed in the FSAR.
- 3. The panels in which these breakers are located are listed in Technical Specification section 3.8.3.1 h e increase in breaker rating to eliminate the potential of inadvertent tripping of the breakers during a Station Blackout condition due to loss of ventilation to the Vital AC and DC SWGR rooms does not affect the limiting condition for operation of the 120V AC/125V DC Vital busses. This modification will enhance the availability of the Class 1E Vital 116
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REE W F 120V AC/125V DC power sugply fran the Distribution Panels 2-1807-Q3-V11, VIc, VI3 and VI4 and 2-1806-Q3-DA2 and DB2 subsequent to a Station Blackout. Therefore, the proposed change does not decrease the u rgin of safety defined by the bases of the Technical Specification. 90-V2N0144 he Private Autmatic Branch Exchange (PABX) services the plant telephones and interfaces with Southern Bell. The PABX system switching unit is located in the Service Building. The Merlin units allow handsets to have multiple line appearances (mitiple extensions which are switchable) to allow several handsets to have the sam extension at one time. The Merlin units are connected via trunk lines to the PABX switching unit. This design change will add an uninterruptible power supply (UPS) unit for the Unit 2 PABX system Merlin equipment only. The UPS unit will be located in roan R-A31. The UPS unit is sized to provide a 2 hour (min.) backup for the Merlin equipment only. Any additional loads added to the unit could decrease the backup time below the time required for PABX equipment in DC-1702.
- 1. The proposed change does not increase the probability of occurrence or consequences of the m lfunction of any equipe nt or component described in chapter 15 or sections 3F and 9.5.2 of the FSAR.
The PABX equipment is not safety related and is not 1 required for safe shutdown of the plant. Separation criteria is maintained by this design.
- 2. The proposed change does not create the possibility of an accident or equipment /carponent malfunction not described and analyzed'in sections 3F, 9.5.2, or chapter 15 of the FSAR. The batteries in these units are the sealed maintenance-free type and produce no gases when charging. 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 i
117 1
II 1990 ANtPJAI. REPORT - PART 2 10 CFR50.59(b) REPORT has been not between the non-lE lighting panels and the safety related load center which feed thun, he heat load generated by the UPS unit is 147 B'N/hr. The heat load was evaluated and determined to be negligible.
- 3. Le proposed change does not decrease the unrgin of safety defined by the bases of the Technical Specifications, including the bases to sections 3/4.8.2 and 3/4.9.5. Separation criteria is not violated by the addition of the UPS units. The additional UPS units will ensure that the PABX equipment has battery back-up power in the ovent that both the nortnal power source and the diesel-backed source are de-energized.
90-V2N0150 Le MFIV fast closing operation schane contains a redundant control relay (AX6 for Train A, BX6 for Train B). E is control relay contact is wired to initiate the de-energization of the MFIV pilot solenoid valve den the control relay coil is de-energized. De-energizing the pilot solenoid valve will cause the MFIV to close in the fast close node. During nontul operation the AX6 (BX6) control relay coil is continuously energized. If the relay coil fails due to either a short or open circuit, the pilot solenoid valve will de-energize and initiate the undesired fast closing operation ' of the MFIV. E is change involves rewiring the MFIV control circuitry to disconnet the AX6 (BX6) relay coil and contact fran the MFIV fast closing control scheme.
- 1. The control wiring change does not change the MFIV operation and no new component is added to the existing control circuit to make this change.
Berefore, the change does not increase the probability of occurrence or consequences of the malfunction in accidents analyzed in the FSAR.
- 2. This atinor wiring change does not change the original system operation and no new component is added to the existing control circuit to make this change. H erefore, the change does not create the possibility of an accident or equipnunt/
canponent Inalfunction not described and analyzed in the FSAR. 118 l _ - - - - _ -
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II l l 1990 ANNUAL REPORT - PART 2 I 10 CFR50.59(b) REPORT
- 3. Le proposed change is designed to meet the intent of the original system design bases while increasing the electrical control circuit design reliability. Therefore, the change does not decrease the margin of safety defined by the bases
-of the Technical Specification, particularly sections 3/4.4.5'(Steam Generator) and 3/4.6.3 (Containment Isolation Valves).
90-V2N0151 W is design change will increase the length of the check valve disc back stop on check valves 2-1306- l U4-004 and 006 (MFP Turbine Driver Low Pressure l Check Valves). The length of the disc stop will be i
- ingreasedtolimitthediscopeningtogpproximately l 60 angle, instead of the present 80 -85 angle, j m is change will have an insignificant effect on the ;
flow capacity of the check valves. Rese check i valves are a part of system 1306 (Turbine Drive l Steam System), which is non-safety related, project class 424,
- 1. Modification of the opening angle of these check valves is being done to increase their reliability by decreasing the impact of flow turbulence on the check valve disc. This change will not have any adverse impact on the function of syst m 1306, or any other systs. E is design change will not increase the probability of occ.trrence or consequences of the malfunctiott of any ecuipment or ccmponent assumed to functi .n in accic ants analyzed in the FSAR. Check valves 2-1306-U4-004 and 006 are not-identified as equixnent/ccmponents which are assumed to function in tne event of-an e accident. These check valves are installed to prevent high pressure steam frcm back-flowing to a lowr pressure side during main turbine start-up.
B is included a review of FSAR sections 10.2,
-10.4.7, and FSAR chapter 15.
- 2. This design modification only revises the full open angle of these check valves. There will be no changes to system operation or response to system 119
l II 1990 ANNUAL REPORT - PART 2 10 CFR50,59(b) REPORT 1306, or any other system, as a result of this change. This change does not create the possibilit,f of an accident or equipmnt/canonent mlfunction not described and analyzed in tne FSAR. This included a review of FSAR sections 10.2, 10.4.7 and FSAR chapter 15,
- 3. Decreasing the full open angle of tPcse check valves will not decrease the margin of safety as defined in the Technical Specification bat,es, including the bases to sections 3/4.7.1 or 3/4.7.2, as the flow resistance through the valve will not be appreciably increased and the valves should be more stable.
90-V2N0154 This desi p change will determinate ZSC-0459 and ZSC-0460 CIDSED contacts and retenninate ZSO-0459 and 250-0460 "NOT OPEN" contacts in the non-Q (62E) control circuits or relays 2LCV-0459X and 2LCV-0460X, which control letdown orifice isolation valves 21W-8149A, 8149B, and 8149C in the Chemical and Volune Control System. The ZSC & ZSO limit switches are project cl. ass 11J and valves 2HV-8149A, B, and C are project class 212.
- 1. The proposed change does not add any new tmchmiisms of failure or change the function of the letdown orifice isolation valves 21W-8149A, B, & Ci it only affects the timing of their operation. Therefore, this change does not increase the probability of occurrence or consequences of malfunction of any equipment or component assuned to function in accident analyzed in the FSAR. This includes a review of FSAR sections 6.2, 6.3, 7.6, 7.7, 15.4 and 15.6.
- 2. This change does not create the possibility of an accident or equipment /ccaponent malfunction not described and analyzed in the FSAR, based on a review including FSAR sections 6.2, 6.3, 7.6, 7.7, 9.3 and chapter 15. No new components are being utilized and only the timing of valves 2HV-8149A, B and C is being changed. Letdown isolation, i
120
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-1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Contaiment isolation, and HELB isolation on this-line is not affected, his change will help prevent damage to the Regenerative Heat Exchanger.
- 3. This change does noe decrease the margin of cafety as defined in the bases to the Technical ecifications, including the bases to sections Sp/4.1.2 3 and 3.3.3.6. Technical Specifications section 3/4.1.2. requires a boron-injection flow path via a charging pump to the Reactor Coolant System, and section 3.3.3.6 requires the operability of Pressurizer level instrumentaiton loops 459 and 460. h is change has no affect on the bases of either of these sections.
90-VAN 0156 This is a " SAFEGUARDS" DCP, the following abbreviated information-is declassified for this report. This nodification involves changes to non-class 1E, seismic category 2 equipaent, he , change replaces the existing distribution class arresters with intermediate class arresters for non-class 1E transforner ANBilX located in the PESB. Surge arresters provide electrical insulation protection due to switching impulses generated in the operation of the electrical system. The lower switching surge rating of the replacement arresters will provide an increased margin of protection against switching surges and
. consequently increase the reliability of the transtonners . The electrical distribution system is described in'section 8 of the FSAR and section 3/4.8 of the Technical Specification.
- 1. Le change does not increase the probability of occurrence or consequences of the malfunction of of any equipment or component assumed to function in accidents analyzed in the FSAR. The change involves -only one non-1E transfonmr. Be transformer and the loads- supplied fran this transformer are not required to function for accident mitigation or for safe shutdown. The loss of any non-1E transformer is bounded by the Loss of Nonemergency AC Power to the Plant Auxiliary analysis. Bis includes a review of FSAR sections 8.3.1 and-15.0, specifically sections 15.2.6 and i.
15.0.8. 121 l L- _- --
II 1990 ANNUAL PIPOKf - PAP 3' 2 10 CFR50.59(b) REPORT
- 2. The change does not create the possibility of an accident or equipnent or emponent malfunction not described and analyzed in the FSAR. The nodification involves changes to a non-1E transforner. The consequences of failure of this transformer is bounded by the Ioss of Non-energency AC Power to the Plant Auxiliaries analysis. This includes a review of FSAR sections 8.3 and 15.0, specifically sections 15.2.6 and 15.0.8.
- 3. The change does not decrease the margin of safety defined by the bases of the Technical Specifications. The change involves only one non-1E transforcer. The transformer does not supply safety related equipment required for safe shutdown or mitigation and control of accident conditions. This includes a review of Technical Specification bases 3/4.8.
90-V1E0157 This design change allows the use of the Diesel Generator Fuel Oil Transfer Punps (tag numbers 1-2403-P4-001, 002, 003 and 004) with mininun of of 2 packing rings in the stuffing box. Vendor concurrence has been gianted to exclude these punps fran the 6 packing ring requirenent that is currently shown in the punp unnual. The removed packing rings are to be replaced with lantern rings. This design change will also allow Plant Maintenance to use shims between the diesel fuel oil tank flange and fuel oil transfer punp discharge head cs necessary to establish vertical aligunent, replacing the previously applied tolerance with a vertical clearance of less than or equal to 1/8" between the pump discharge flange and the mating flange. Vendor concurrence has also been granted for this deviation fran the current punp instruction manual.
- 1. This change will allow a reduction in the nunber of packing rings in the stuffing box of the Diesel Generator Fuel Oil Transfer Punps, which are part of the Emergency Diesel Generator System (2403).
Per ASME Section III, Subsection ND 3413.2, 1977 Edition, and Section XI, Subsection Ih -7400, 1983 122
II 1990 AtNJAL REPORT - PART 2 10 CFR50.59(b) PIPoic Edition, the stuffing bcx is exmpt fran Section III and Section XI code requirments and is not considered to be within the pressure boundary of the pump. Therefore, this change will not degrade any equipnent or cauponents which are assmed to function in an accident as described in the FSAR, including sections 8.3, 9.5.4 and 15.
- 2. '1he proposed change reduces the number of packing rings in the pmp stuffing box. A failure of the packing rings or lantern rings would not cause a complete failure of the pm p. The punp could still operate and pump fuel oil even though the stuffing box was leaking due to a packing failure. In addition, even if the punp flowrate is reduced causing an insufficient mkeup to the day tank or a low discharge pressure, the second transfer pump would auto start. Each punp is designed to supply fuel oil at a rate of approximately 3 times the demand at full load of the diesel. It will not introduce nozzle loadings nor significantly affect the pipe stress calculation for the punp to be installed within the nunufacturer's prescribed vertical tolerance.
Therefore, this change does not create the possibility of an accident or equiptmnt/cccponent malfunction not described and analyzed in the FSAR.
- 3. A failure in the packing would not lead to a failure in the ability of the pwp to operate and deliver fuel oil to the day tanks, even if the packing leaks. Also, the diesel fuel oil system is designed such that if the flowrate to the day tatis is not sufficient to supply enough makeup or if the pump discharge pressure is low the second transfer pump would auto start. Each punp is also designed to supply fuel oil at a rate of approximately 3 tines the demand at full load of the diesel.
I 123
II 1990 NMJAL REPORT - PART 2 10 CFR50.59(b) REIM 1herefore, a failure of the packing will not decrease Technical Specification safety unrgins defined by the bases of the Technical Specification. This is based on a review of Technical Specification basis, including the bases of section 3/4.8.1, 90-V2E0158 1his design change allows the use of the Diesel Generator Fuel Oil Transfer Punps (tag nunbers 2-2403-P4-001, 002, 003 and 004) with a mininun of 2 packing rings in the stuffing box. Vendor concurrence has been granted to exclude these pumps from the 6 packing ring requi.rment that is currently shown in the punp nunual. The renoved packing rings are to be replaced with lantern rings. This design change will also allow Plant Maintenance to use shims between the diesel fuel oil tank flange and fuel oil transfer punp discharge head as necessary to verify that the punp is vertical, replacing the previously applied tolerance with a mininun vertical clearance of 1/8" between the pump discharge flange and the unting flange. Vendor concurrence has also been granted for this deviation from the current pump instruction manual.
- 1. This change will allow a reduction in the nunber of packing rings in the stuffing box of the Diesel Generator Fuel Oil Transfer Pumps, which are part of the Energency Diesel Generator System (2403) .
Per ASME Section III, Subsection ND 3413.2, 1977 Edition, and Section XI, Subsection IWA-7400, 1983 Edition, the stuffing box is exmpt from Section III and Section XI code requirements and is not considered to be within the pressure boundary of the pump. Therefore, this change will not degrade any equipnent or emponents which are assumd to function in an accident as described in the FSAR, including sections 8.3, 9.5.4 and 15. 124
l l 1 II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. The proposed change reduces the nunber of packing r "s in the puun stuffing box. A failure of the pac rings or lantern rings would not cause a cmplete failure of the punp. The pump could still operate and pump fuel oil even though the stuffing box was leaking due to a packing failure. In addition, even if the pump flowrate is reduced causing an insufficient makeup to the day tank or a low discharge pressure, the second transfer pump would auto start. Each punp is designed to supply fuel oil at a rate of aaproximately 3 tines the demand at full load of the diesel. No unanalyzed pipe stresses or nozzle loadings will result if the pump is installed within the manufacturer's prescribed vertical tolerance. Therefore, this change does not create the possibility of an accident or equipment /ccuponent malfunction not described and i analyzed in the FSAR.
- 3. A failure in the packing would not lead to a failure in the ability of the pump to operate and deliver fuel oil to the day tanks, even if the-packing Icaks. Also, the diesel fuel oil system is designed such that if the flowrate to the day tanks is not sufficient to supply enough makeup, or if the pump discharge pressure is low, the second transfer pump would auto start. Each punp is designed to supply fuel oil at a rate of I aaproximately 3 times the demand at full load of tae diesel. Therefore, a failure of the packing will not decrease Technical Specification safety margins defined by the basis of the Technical ecification basis, including the bases to section Sp/4.
3 8.1. ; 90-V2N0161 This desi change package nodifies the return line (2-1213-0 7-3") frcm the Spent Fuel Pool Cooling and Purification System (SFPCPS) to the refueling cavity. The modification consists of the addition of renovable piping seggents interconnected by quick disconnect fittings. These renovable piping segrrents and their supports will be installed during refueling outages with the reactor in Fbde 5 or 6 and will be renoved following refueling for decontamination and storage outside of containment. 125
II 1990 ANNUAL REPORT - PART 2 10 C1150.59(b) REFOKr
- 1. Failure of the piping during installation or while
[ the system is in operation during refueling outages will have no adverse effect on seismic category 1 structures, systems, and camponents in the vicinity of the piping. Extending the return line below the norunl refueling cavity water level creates the possibility of gravity draining the cavity in the event of reverse system flow. To prevent this ( occurrence, an anti-syphon hole has been provided ) in the return line. Thus, the ability to unintain the minimum cavity water level and the decay heat renoval of the spent fuel pool / refueling cavity will not be comprcxnised.
- 2. Installation of non-seismic piping within the contairment has been addressed and deened acceptable I
since failure of the piping will not adversely affect the seismic category 1 structures, system, or emponents in the vicinity of the piping (Ref. FSAR Table 3.2.2-1, Note 2) .
- 3. The SFPCPS is not specifically addressed in the Technical Specifications. Technical Specifications 3/4.9.10 and 3/4.9,11 and the associated bases address the water level in the spent fuel pool and the refueling cavity during refueling. The addition of the anti-syphon hole will ensure that any reverse flow in the SFPCPS will not drain the water level below the minimum Technical Specification y requirenents . Therefore, the margin of safety as detined in the Technical Specifications will not be reduced.
90-V2N0166 This proposed change is to allow the mergency diesel generators 2A and 2B, project class llE, to be started by an emergency sigtml and have the high jacket water trip bypassed, in addition to the other trips which are already bypassed by an emergency start signal (Safety Injection (SI), Loss of Offsite Power (IDSP), or Emergency Manual Start) . 126
II i 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT This change will add isolation valves in the instnment tubing betwen the DG high jacket water tmperature elwents and the local DG control panel. The valves will normally be closed, but the valves uny be opened to allcv additional engine protection when performing a non-emergency unnual start or surveillance of the diesels.
- 1. The valves being added are seismically and environmentally qualified for their enviroment and will be nounted in such a way as to not overstress the instrument tubing. Isolating this trip will not cause a malfunction of the diesel engine or any other equipment or cmponents assumed to function in accident analyzed in the FSAR. The valves will be located away fran the engine, near a support , to prevent any significant vibration effects. Proper valve position will be ensured by administrative control. Incorrect positioning of the valves would only unke the trip active, as it is currently.
- 2. There is no change to the DG trip logic, other than the blocking of the trip. A separate high jacket water tmperature alann is still available locally and in the control romt. The safety benefit gained by the increased reliability of the DG when this trip is bypassed is judged to outweigh the effect of an additional operator decision to trip the DG if a high telperature alann is received. This change does not create the possibility of an accident or equipannt/
cmponent malfunction not described or analyzed in the FSAR, including the failure of the DG, As docununted in the RIEA in FSAR Tables 3.3.1-3, if one DG fails, the other is available. In addition, the operator can nunually stop the DC if a high temperature alann is received.
- 3. This design change wilt not decrease the margin of safety as defined by the bases of the Technical Specifications, including the bases for 3/4.4, 3/4.5, 3/4.8, 3/4.9, or the bases for the surveillance requirenunt in 3/4.8.1. The unrgin of safety of the reactor coolant system, emergency 127
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT core cooling system and refueling operation is related to the availability of the site electrical power system to provide sufficient power to the safety-related equipnent recuired for the safe shutdown of the facility anc the mitigation and control of accident conditions within the _' facility. A loss of offsite power is a contributing event during other postulated accidents. The ability of the DG to start and continue to run during events which contribute to cccidents is an increase in the margin of safety. .I Removal of this autanatic trip would allow the -DG to continue to operate in the event of a falso high-jacket water tuuperature. 'Ihe operability of the onsite A.C. power sources will not be degraded by implenenting this design change, and will be enhanced in certain conditions. 90-V2N0173 l Thischangemakespipesup'portsV2-1304-061-H057and N058 on line 2-1304-061-6 (feedwater heater drain) ! 1 and pipe supp' ort V2-1305-080-H060 on line 2-1305-080-6 (condensate and feedwater system) I removable. The supports are seismic category 2 ,
-ANSI B31.1 designed to meet 2 over I design I criteria. The system function or support integrity is not effected by these changes as substantiated by calculations as referenced in the DCP calculation record. Supports are only to be removed during an outage and nust be reinstalled before system operation.
- 1. The piping and supports in this DCP were reviewed for impact to interconnecting equipment and components. This review and calculations performed for this change demonstrate the adequacy of the modifications. Therefore, based on a review of the FSAR, including section 15, this change does not affect any equipment or cauponent function.
- 2. The supports in this DCP are substantiated by calculations that show the change is within the design criteria and code allowables. The change
- l. is for pipe supporta and does not affect any o equipment /cauponent function or operation and therefore does not create any possibility of accident or malfunction.
I 128 i.
11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPOKr
- 3. 'Ihe calcult.tions performed for this DCP show the suppart stresses to be in accordance with the applicable design criteria, codes and standards as identified in the Design input Record of this DCP. These criteria establish the design bases for piping and support stresses. Inherent in these design bases is the sane mrgin of safety ao the original design. Since the calculations are within design criteria, the margin of safety as defined in the bases of the Technical Specificatiens, including the bases to Technical Specification 3/4.7, is not decreased.
90-V1N0179 Electrical protection of the safety related, seismic category 1, diesel generators lA and 1B ouring normal and loss of offsite power (iDSP) operation includes loss of Field, Voltage Controlled Phase Overcurrent (OC), Ground Overcurrent (GOC), and other protection. The GOC trips for diesel lA and 1B will be removed by disconnecting the ground fault auxiliary relay (151NX) contact frcxn the diesel generator breaker lock-out relay 186B. This is acca911shed by rmuving two jumpers and adding a third jumper in the generator control panels for diesel generators lA and 1B.
- 1. Rmoving the diesel generator breaker ground overcurrent trip will have no effect on the probability of malfunction of the diesel generator or its camponents. This design change will require the operator to monitor the alarm status of the diesel generators during LOSP and nonnal initiated operation as he currently does for SI initiated operation. If the diesel is dannged due to a second ground fault with one of the faults occurring in the generator, the system response will be no different than if the diesel had tripped. FSAR Table 8.3.1-3 (items 4 and 41) describes the conscquences of failure of a diesel generator to start or run. This table shows that there is no adverse impact on equipnent assuned to function in the accidents analyzed in FSAR chapter 15.
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l II l 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. During lOSP and nonml initiated diesel generator operation, the operator in the control room will now have to determine whether or not to trip the diesel if it develops a ground fault. Currently, the operator only has to perform this oversight during safety injection initiated operation. The effects of failure of the diesel generators to run are documented in FSAR Table 8.3.1-3 (items 4 and 41). The circuit nodifications involve disconnecting the diesel generator ground -)
overcurrent trip. For postulated fire initiating i events, an additional single active failure (i.e. a I second ground fault with one or both occurring in the generator) is not required to be postulated (ref. FSAR 9.5,1.3). No new single failures are created. Therefore, this design change can not lead *:o an accident or malfunction not described or analyzed in the FSAR.
- 3. This design change will not decrease the margin of safety as defined by the bases of the Technical Specifications including the bases for 3/4.4, 3/4,5, 3/4.8, 3/4.9 or the bases for Surveillance Requirement 3/4.8.1. The margin of safety of the reactor coolant-system, emergency core cooling system and refueling operation is related to the availability of the site electrical power system to provide sufficient power to the safety-related equipnent required for the safe shutdown of the facility and the mitigation and control of accident conditions within the facility, A loss of offsite power is a contributing event during other postulated accidents. The ability of the diesel generator to start and continue to run during events which contribute to accidents is an increase in the margin of safety. A diesel generator would now continue.to run with a ground fault existing instead of automatically tripping on ground overcurrent. The operability of the onsite A,C.
power sources will not be degraded by implanenting this design change. 1 130 i
- - - - _ __ . - ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 90-V2N0180 Electrical protection of the safety related, seismic category 1, diesel generators 2A and 2B during normal and loss of offsite power (10SP) operation includes loss of Field, Voltage Controlled Phase Overcurrent (OC), Ground Overcurrent (GOC), and other protection. Each of these conditions is detected by a protective relay which actuates an auxiliary relay. The atmiliary relays for OC, GOC, and loss of field actuate lock-out relay 186B which trips the diesel generator breaker. The GOC trias for diesel generators 2A and 2B will be rmaved by disconnecting the ground fault auxiliary relay (151NX) contact frcxn the diesel generator breaker ~~ lock-out relay 186B. This is acccxuplished by rmoving two jumpers and adding a third jumper in the generator control panels for diesel generators 2A and 2B,
- 1. Rmoving the diesel generator breaker ground overcurrent trip will have no effect on the probability of nn1 function of the diesel generator or its cmponents. This design change will require the operator to nonitor the alarm status of the diesel generators during WSP and nonnal initiated oparation as he currently does for SI initiated operation. If the diesel is damaged due to a second ground fault with cne of the faults occurring in the generator, the system response will be no different than if the diesel had tripped. FSAR table 8.3.1-3 (items 4 and 41) describes the consequences of failure of a diesel generator to start or run. This table shows that there is no adverse impact on equipnent assumed to function in the accidents analyzed in chapter 15.
- 2. During WSP and nonnal initiated diesel generator operation, the operator in the control rocxn will now have to detennine whether or not to trip the diesel if it develops a ground fault. Currently, the operatm only has to perfonn this cuersight during safety injection initiated operation. The effects of failure of the diesel generators to run are documented in FSAR table 8.3.1-3 (items 4 and i
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-II 1990 ANNUAL REPORT - PART 2 l
10 CFR50.59(b) REPORT 41). The circuit modifications involve 1 disconnecting the diesel generator ground _ overcurrent trip. For postulated fire initiating events, an additional single active failure (i.e. a second ground fault with one or both occurring in ' l the generator) is not required to be postulated (ref. FSAR 9.5.1.3). No new single failures are created.. Therefore, this design change can not j lead to an accident or malfunction not described or : analyzed in the FSAR.
- 3. This design change will not decrease the margin of safety as defined by the bases of the Technical-ecifications including the bases for 3/4.4,3/4.5, sp/4.8, 3 3/4.9 or the bases for Surveillance Requirement 3/4.8.1, The margin of safety of the reactor coolant system, . emergency core cochng .
system and refueling operation is related to the availability of the site electrical p w er system > to provide sufficient power to the safety related equipment required for the safe shutdown of the facility and the mitigation and control of accident conditions within the facility. 'A loss of offsite power is a contributing event during other postulated accidents. The ability of the diesel generator to start and continue to run during events which contribute to accidents is an increase in the margin of safety. A diesel generator would now continue to run with a ground fault existing-instead of automatically tripping on ground overcurrent. The operability of the onsite A.C. power sources will not be _ degraded by implenenting
~
this design change. 4 90-V2N0187 This change involves adding two supports on the Turbine Generator Stator Cooling System. One support is for two, 1/2" diameter flow lines for flow element 2FE-6859 and the other is for flow
-switch 2FS-6832. The hanger supportin 1/2" flow lines is attached to the 6"diameter g the two,-
alpe they originate from. The flow switch is araced off of the Stator Cooling Water Storage Tank. The Turbine Generator Stator Cooling System 132 ,
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REEVRT is located in the Turbine Building and is a non-safety related system. The 1/2" flow lines, 6" pipe, flow switch, and storage tank are all Project Class 424 items.
- 1. By adding the supports required under this proposed change, the resulting reduction in vibrations of the subject flow devices will improve the reliability of the Turbine Generator Stator Cooling System and improve the reliability of the Turbine Generator.
However, both systems serve no safety function and have no safety design basis and, therefore, are not assumed to function in accidents analyzed in the FSAR.
- 2. The purpose of the supports is to reduce the vibration in the flow devices so as to reduce the possibility of an unnecessary turbine trip.
Therefore, the aroposed change has been analyzed so as to improve tae operation of the system and does not create the possibility of an accident or an equipment /ccxnponent tx11 function not described and analyzed in the FSAR.
- 3. The proposed change improves the reliability of the Turbine Generator Stator Cooling System and improves the reliability of the Turbine Generator by reducing the vibrations in two flow devices which eliminates a source of a possible unnecessary turbine trip.
Therefore, the margin of safety as defined by the bases of the Technical Specifications, including the basis to section 3/4.7, Plant Systems, is not decreased. 90-V1N0189 This design change adds a test connection between CVCS letdown containment isolation valve HV-8152 and containment penetration 48. The connection includes one, 3/4" manual globe valve, one flange connection and blind flange, and 3/4", Sch. 40S pipe (Class 212, HGO).
- 1. This portion of the CVCS is not assumed to function during an accident. No new pipe break locations are created and no hazard analyses are affected by the addition of the 3/4" test connection. This includes a review of FSAR sections 15 and 9.3.
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- 2. The accident which is described in the FSAR for the I
nornal letdown is a full pipe break and is not affected by this change. The operation of the c letdwn isolation valves and the contaiment ] isolation valves are unaffected by this change. i This includes a reviw of FSAR sections 3, 6, and 15.
- 3. Technical Specification bases 3/4.6.3 describes
- requirments for containnent isolation valves. The valve added by this design change will be nontally closed and blind flanged which gives double
, isolation. Contalment isolation and the function of the CVCS is unaffected. Therefore, this design change does not decrease the nargin of safety as defined by the Technical Specification bases, including the bases to sections 3/4.6.3, 3/4.1 and , 3/4.4. 90-V2N0194 This design change provides the following: , A. For each Main Steam Isolation Valve (MSIV), add qualified 1 try fuses in the Haskell Pmp air supply solenoid circuit. 'lhe fuses will be located in spare fuse blocks nounted in Auxiliary Relay Panels. Designate the circuit frctn the fuses to the solenoid valves as an associated circuit. W e MSIVs are project Class 212 and considered safe shutdown ccuponents. B. At each MSIV, insert a condulet in the conduit - run between the vendor installed splice box 3 and the hydraulic fluid reservoir. 7he condulet will facilitate splicing to the fluid level witch pigtails. The switches are project classification llE.
- 1. a. The design change does not alter any existing equipment or ccuponent assumed to function in accidents analyzed in the FSAR. The addition
-of fuses as described in responses-la and 4a does not constitute an alteration because the fuse blocks utiliced
, 134 e- .v e e-r-r~wmu-w.,-+-,-ym, nwv v m a r,,, r._ ,-,._,_w-
- . , . ~ , . , , , , - - , .
11 1990 NEAL REIOlE - l'AIG 2 10 CFK50.59(b) REIOla are spares and are already located in the mtxiliary relay panels. The Haskel punp solenoid is not environnentally qualified and is not required to operste under accident conditions. The changes proposed will decrease the probability of occurrence and consequences of an accillent by ensuring the MSIV control and indication circuit is not affected by the potential failure of the unqualified solenoid valve.
- b. The additional condulet will provide a point of access to the llydraulic Fluid Level Switch pigtails, thereby allowing substitution of a qualified splice in place of the existing unqualified splice. This will enhance the overall circuit relit.bility and decrease the probability and consequences of any accident ohich could alter the MSIV area environment.
Reviewd FSAR sectiona 7.3.8, 8.3.1.4, 10.3, and chapters 8 and 15.
- 2. None of the proposed changes create the possibility of an accident or equip ent/ca ponent t nlfunc. tion.
The changes are enluaicerrents which inprove reliability of the MSIV control and indicating circuits. 1he MSIV and its associated circuits and devices are functionally unaffected. 'lhe chary;es are only to bring the installed circuits into ccu1pliance with cannitnents trade in the FSAR with regard to independence of Class 1E circuits and the use of qualified splices, as discussed in responses 4a and 4b. Reviewed sections 7.3.8, 8.3.1.4, 10.3 and chapter 8 and 15 of the PSAR.
- 3. Independence of Class lE circuits and cabic splicing are not addressed in the Tech;.. cal Specification. It is assuned Class it and associated circuits are adequately separated or isolation frcol non-Clas,1E circuits and cctponents, or justified by analysis, and that qualified splices are used where approprute. hse irrplied 1
135
11 1990 AhWUAL PIPORT PART 2 i 10 CFR50.59(b) PIPORT I asstuptions are still valid. None of the Riin Steam System features or bases are affected in as nuch as there is no change to the MSIV control or indicating function characteristics. 91-VIE 0016 L is change re, laces the existing non-safety related (limit' switches for MOV IHV-7244 and llN7245) plastic limit switch wonn gear, limit switch wheel and add-on pak gear set (parts) with notal parts for valve tag nos. IN-7244 and 1-IN-7245, which serve as - circulating water pump discharge valves within-the Circulating Water systun (Systun No.1401). ! he limit switches provide valve position and pilot light indication. In addition, the lintit switch is interlocked with the circulating water putp-start electrical circuit. Ecse valves are within the circulating water systen which does not perfonn a safety related function as indicated
- in design criteria DC-1401, Para. 3.1.
- 1. Modifications associated with this DCP affect the Circulating Water System which has-no
. safety design aasis and is not assumd to function in any accident analysis described in the FSAR. The replacaent of the plastic limit switch parts with tmtal parts is in accordance with the valve actuator manufacturer's-(PUIORK) plant spproved instruction manual. herefore, there would ' be no-adverse impact to the intended function-of the valve limit switches. ! Implanentation of the modifications permitted by this DCP, in accordance with existing plant approved specifications and procedures, will not increase the arobability of occurrence or consequences of tae malfunction of any equipment or component asstned to function in accidents analyzed in the FSAR. i l 136
11 1990 AIM 1AL RL10RT - PART 2 10 CHL50.59(b) REIVRT
- 2. There are no new potential accidents or events created, nor are there any new f ailure nodes introduced, lbdification's associated with this DCP affect non-safety related systen/
canponents . B ere are no safety related systans affected by this DCP. Irplarentation of the nodification permitted by this DCP will be in accordance with existing plant approved specifications and procedures.
%e nodifications permitted by this DCP do not create the possibility of an accident or equipnent/cocponent nalfunction not described and analyzed in the FSAR.
- 3. he Technical Specifications do not address the Circulating Water Systun purp discharge valves or the Circulating Water System, licwver, the rnodifications required by this DCP are to be inplarented in accordance with plant approved procedures and valve nanufacturer instructions which will result in a design which is consistent with the original plcnt design requirenunts. The nodifications required by this DCP will be ccupleted in accordance with the Design Criteria applicable to VEGP (and as identified in the Design Input Record). Therefore, the nodifications required by this DCP do not decrease the nargin of safety defined by the bases in the Technical Specification, 91-V2N0032 Safety Injection Accu 1ulator level system bellows for level transmitters (2LT-950, 2LT-951, 2LT-952, 2LT-953, 2LT-954, 2LT-955, 2LT-956 and 2LT-957) will be rotated 180 degrees (Calibration Swaglock fitting vill be in vertical position pointed up).
- 1. Bis trodificatien enhances perfornunce of existing non-safety related Accunulator IcVel system.
137
Il 1990 NNUAL RElWf - PART 2 10 CFR50.59(b) REl W r lhe acetriulator tanks are safety related equipaent required to initigate the consequences of a loss of Coolant Accident (Section 15.6.5). The non-safety related instnnentation affected by this nodification is not relied upon to actuate any safety-related equi lment. Furtheniore, follwing the nodification, tae instnnentation lines will be leak checked to ensure the pressure boundary is naintained. Therefore, this change does not increase the probability of occurrence or consequences of the nnlfunction of any equi tment or canpanent asstmed to function in accidents analyzed in the FSAR.
- 2. No new potential accidents or events are created as a result of this nodification since no new nodes of failure are introduced. 'lhe change covered in the DCP for the level transmitters will not affect the seismic qualification of the transtnitters. The pressure boundary of the instnoentation lines have not b~en adversely affected as a result of this nodification and will be Icak checked follwing rotation of the bellws. In addition, rotation of the bellws will enhance the instnznentation's perfoninnce. Failure of the instnnent following the nodification will be no worse than those failures postulated prior to rotating the bellows.
Therefore, this change does not create the possibility of an accident or equipnent/cainonent malfunction not described and analyzed in tae FSAR.
- 3. This DCP provides a change to the non-safety related level indicating systen of the Safety Injection Acetuulator tank. This nodification enhances the system by indicating nore accurate reading.
Therefore, there is no reduction in the unrgin of safety as defined by the bases of the Technical Specification 3/4.5.1. 138
11 1990 NM1AL KDORT - PART 2 10 CFR50,59(b) RD ORT 92-V2N0015 This design change adds flexible connections to the Electro-Hydaulic Control (DlC) tubing at the Wtin Turbine Control and Stop Valves. The DiC lines consists of the mergency trip supply (1 rS), fast acting supply (FAS), and fluid cooler drain (FCD). Also, supports will be added to the existing, unsupported tubing clmps at the flexible connection locations. The Main Turbine DIC Systen is non-safety related and seim.ic category 2 (Project Class 424), 1, The Wtin Turbine DiC System performs no safety relatedfunctionandisProjectClass424. The nodifications required by this design change will neet the design, naterial, installation, non-destructive exantination, testing and quality requirulents of the existing system, Therefore, the prcoosed change vill not increase the probabikityofoccurrenceorconsequencesofthe malfunction of any equi ment l or ccuponent austmed to function in accidents analyzed in the FSAR,
- 2. The Wtin Ibrbine DiC System perfarno no safety related function and is Project Class 424. The nodifications required by this design change will tret the design, material, installation, non-destructive exanination, testing and quality requiraients of the existing systun. The proposed change does not introduce a new accident, failure node, or luu.ard to the plant. Inplurentation of the proposed chnnge will be in accordance with existing plant approved specifications and procedures.
Therefore, the proposed clutnge will not create the possibility of an accident or equilment/cmoonent malfunction not described and analyzed in Lac FSAR. 3, 1he nodifications required by this design change are to be inplurented while the plant is in cold shutdown, and will be in accordance with plant approved specifications and procedures. The i 139
II 1990 NEAL RDORT - PART 2 10 Cnt50.59(b) RDORT nodifications are to be canpleted in accordance with the Design Criteria and the Codes and Standards applicable to VEGP, which will result in a design Waich is in compliance with the existing plant desip requirements. Technical Specifications 3/4.3.1, 3/4.3.4 and 3/4.7.1 have been reviewd-and it has been determined that they are not affected by this design chtnge. Therefore, the proposed change will not decrease the margin of 1 safety defined by the bases of the Technical ! Specification. l E i i r-l- I I 140
, . ~ . . . . . , - ~ . _ - _ _ . . _ _ . _ _ _ _ _ . . _ . . . _ . . . . _ _ . _ _ _ _ _ . _ . . - . . . . _ _ , . _ _ _ _ _ __ ..___ -.-
] Il 1990 NNUAL REIORT - PART 2 0 10 CFR50.59(b) RDORT MDD's 4 89-V2M026 The addition of differential aressure gaubes to the existing instrunentation on the polishing filter section of the Main Turbine and Stean Generator i Feed l\stp Turbine lube Oil Conditioners.
- 1. The addition of the referenced instrunentation of the labe 011 conditioners, which are non-safety related, would not cause a failure or malfunction i
+ of the systuns they suppe-t, not previously evaluated in the PSAR. litrehernore, the instrunentation, or its failtus, can not cause any nn1 function of any safety rWated systen or cmponent not previously evaluMed in the FSAR.
- 2. Failure of the oil conditioners or re?crenced instrtmentation can not create an accident condition not previously reviewed in the FSAR.
- 3. The failure of the oil conditioners or the referenced instrunentation can not effect any I safety related emponent and, therefore, does not effect any bases for any Technical Specification. )
i 89-V1M030 Change tim dial setting for the RCP Tine ; Overcurrent Relays fran 960 cycles at 500% of
.. TAP to 1080 cycles at 500% of TAP _and change 1 High Dropout Unit Setpoint for the RCP Tine Overcurrent Relays fran tap 7 (480 anps) to tap 7 (560 aups). ,
4
- 1. The change does not increase the probability of occurrence or consequences of an accident as described in FSAR chapter 15. The new see mints for the RCP overcurrent relays (560A) are aelow-the design limit of the containnunt penetrations ,
(636.6), Ref. Fig. 8.3.1-7, sheet 12 of 19, 141
I 11 1990 A!M1AL RDORT - PART 2 10 CIK50.59(b) RDORT
- 2. The new setpoints for the RQ' Tine Overcurrent Relays simply allw all the RCP'c to start reliably without nuisance tripping while still providing adequate locked rotor protection for the RCP's and adequate protection for the contaiment penetrations.
- 3. The change does not decrease the nnrgin of safety defined by the bases for the Technical Specification (See section 3/4.8.4). The new setpoints are still below the design limits of the contaiment penetrations - Ref. FSAR Fig. 8.3.1-7, r.heet 12 of 19, 89-VlM057 in DGlA and DG1B engine control panels, add a tee connection in lines E10-A, E10-B and E10-C before the lines exit the panel to connect lw pressure lube oil cwitches on the en;;ine. Install a 3/8" line fran each tee with a 0.014" orifico in-line and connect to the pneumatic control air supply.
- 1. This MDD decreases the probability of occurrence or consequence of a malfunction of safety related equipnent previously evaluated in the FSAR by inoroving the reliability of the low pressure lube oil trip systun.
- 2. This MDD does not create the possibility for an accident or canponent previously evaluated, in the FSAR. Effect of losing one diesel already been evaluated in FSAR I W A Table 8.3.1-3.
- 3. This MDD does not decrease the narg;in of safety defined by Technical Specification bases 3/4.0.
89-VlM063 Change the orientation of the condensate pumps seal water piping to allow for the installation of nochanical seals in the condensate puT- Design of the pu:p allows for the option of a stuffing box 142
i II 1990 ANh'UAL RElVRT - PART 2 10 CFR50.59(b) REPORT arrangment or nechanical seal in these pmps. Seal injection piping was originally designed to supply water to either pmp sealing systen, with only minor piping orientation changes required.
- 1. The orientation change of the non-safety related seal water piping and the replacaent of the nechanical seal in the non-safety related condensate pmps was done in accordance with existing plant arocedures and specifications. The condensate pwps nave no safety design basis and are not assmed to function in any accident described in the FSAR, and therefore, will not increase the probability or consequences of an accident ortolfunction of equiptent inportant to safety, previously evaluated in te FSAR.
- 2. The change of the seal water orientation and replacment of a nechanical seal in the condensate pm ps does not introduce any new possible failure nodes and will not create the possibility of an accident or equipnent/ component malfunction not described in the FSAR.
- 3. The referenced change will not effect any safety related ccuponent, and therefore, will not decrease the nurgin of safety design bases in the Technical Specifications.
89-V1M064 This MDD renoves internals from check valve 1-1322-U4-509.
- 1. The change effects only system that are not safety related or required to mitagate the effec'ts of any accident. This is based on a review of FSAR sections 3, 9, 10 and 15.
- 2. The change does not create the possibility of an accident not evaluated in the FSAR due to the systan having no effect on safety related system or having effects on any systun accident evaluation done in FSAR sections 3, 9, 10 and 15,
- 3. The change does not effect the bases of Technical Specification as reviewed in section 3/4-7 (Plant Systems).
143
II 1990 AtM1A1. REIVRT - PART 2 10 CFR50.59(b) REPORI' 89-V1M070 Change pressure switch PS-50B setpoint fran 45 psi (rising) to 40 psi (rising). Revise drawing 1X4AR01-52 to reflect the new setpoint and recalibrate pressure switches 1PS-4736 for Train A DG and IPS-4846 for Train B DG. (Reference ODR T-2-88-073)
- 1. This MDD iawolves a setpoint change which will decrease the probability of occurrence or consequence of a cauponent unlfunction.
- 2. This MDD involves a setpoint change only and does not create the possibility for an accident or canpanent un1 function of a different type than previously evaluated in FSAR section 15,
- 3. This MDD does not decrease the uurgin of safety defined by the bases for Technical Specifications.
89-VlM071 'Ihis MDD adds a 1/2" drain valve on each end of both intake mnifolds for Unit 1 Train A and B Diesel Generators. Existing drains are 1/4" tubes which are open to the atmsphere.
- 1. The failure of these drain valves will have no adverse effect on or increase the consequences of a m1 function on the Diesel Generators.
- 2. '1he failure of these drain valves will not increase the possibility of canponent malfunction or Diesel Generator failure of a different type than previously evaluated.
- 3. The addition of these drain valves does not decrease the margin of safety as described in Technical Specification section 3/4.8.1.1.2.
i l I l 144
i Il 1990 ANNUAL REIORT - PART 2 10 CFR50.59(b) REIORT 89-V1M075 This MDD adds vent and drain valves to Steam Generator Feed Pump (SGFP) Seal Injection Duplex Filters to replace the existing threaded plugs used during maintenance of the filters,
- 1. The SGFP Seal Injection system has no safety design basis and is not assumed to function in any accident described in the FSAR. The vent and drain valves are only used during maintenance of the duplex filters. Implenentation of the change will be in accordance with existing plant approved saccifications and procedures, and will not increase tae probability of occurrence or consequences of the malfunction of any equipment or cauponent assumed to function in accidents analyzed in the FSAR.
- 2. The change adds non-safety related vent and drain valves to the non-safety related SGFP Seal Injection system. It does not introduce m1y new aossible failure modes. Implenuntation of the change will be in accordance with the existing plant approved saecifications and procedures, and will not create tae possibility of an accident or equipmnt/
canponent malfunction not described in the FSAR.
- 3. The Technical Specifications do not specifically address the SGFP Seal Injection system. However, the change will be inp1mented in accordance with existing plant approved procedures and with the design criteria applicable to VEGP (and as identified in the Design input Record). Therefore, the change will not decrease the margin of safety defined by the bases of the Technical Specifications.
89-V2M077 This MDD adds vent and drain valves to Steam Generator Feed Punp (SGFP) Seal Injection Duplex Filters to replace the existing threaded plugs used during maintenance of the filters. 145
1 II 1990 NNUAL lEPORT - PART 2 10 CFR50.59(b) IEIORT
- 1. The SGFP Seal Injection systan has no safety design basis and is not assmed to function in any accident described in the FSAR. The vent and drain valves are only used during unintenance of the duplex filters. Inplenuntation of the chnage will be in accordance with existing plant approved specifications and procedures, and will not increase the probability of occurrence or consequences of the malfunction of any equipment or emponent assumed to function in accidents analyzed in the FSAR.
- 2. Le change adds non-safety related vent and drain valves to the non-safety related SGFP Seal Injection system. It does not introduce any new possible failure modes. Inplutentation of the change will be in accordance with the existing plant approved soccifications and procedures, and will not create the possibility of an accident or equipnent/
couponent malfunction not described in the FSAR.
- 3. The Technical Specifications do not specifically address the SGFP Seal Injection systen. However, the change will be inp1mented in accordance with existing plant approved procedures and with the design criteria ap311 cable to VEGP (and as identified in the 3csign Input Record). h erefore, the change will not decrease the margin of safety defined by the bases of the Technical Specifications.
89-V1M078 This MDD renoves existing aftercooler safety valves unde by Runkle Valve Cmpany, Fkxlel # 6010D1 or 6000D1, 1/2", setpoint 275 psig, and installs Crosby valves, Model i 41A, Bronze, 1/2" x 3/4", setpoint 275 psig, service fluid / air, relieving capacity: 606 SCR4 and changes drawing 1X4AK01-54 to reflect the new valves to be installed. 146
Il , 1990 At@GL REPORT - PART 2 10 CFR50.59(b) REPOIC
- 1. his MDD decreases the probability of occurrence or consequence of a m1 function of safety related canponents by installing a nore reliable brand of aftercooler safety valves.
- 2. Bis MDD does not create the possibility for an
- accident or cauponent m1 function of a different I type other than previously evaluated.
- 3. This MDD does not decrease the urgin of safety defined by Technical Specification bases.
89-V2M081 This change replaces the existing 2 mgaword (FM) , fi.xed head dnan storage devices presently used in the Proteus Cacputer System with solid state static mmary devices (Megaram). A tape backup device (Megastream) will se installed on one of the two redundant Megarams, and controller cards interfacing the Megarams with the system will require revision. Bis change is a hardware change only and does not affect the functionality of the systen software.
- 1. E is change represents a change to the Proteus Computer . systems' nxmory hardware and does not degrade the software controlled functions. Le software controlled functions provided by the erations camputer system include rod bank important to plant op/ deviation position indication (Technical Specification 3/4.1.3),_ axial flux difference indication (Technical Specification 3/4.2.1), and the calorim tric calculation. Since the system software is unaffected by this change and these software functions are not included as functions available for mitigation of accident effects listed in FSAR section 15.0.8, then this change does not increase the probability of
--occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
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11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. This change to the Proteus Ccuputer affects systun hardware only. The cauputer systen does provide indication of plant parmuters that would be helpful in the event of an accident. Hcrever, the systan is not required for any safety related functions. Therefore, this change does not create the possibility of an accident or nalfunction of a different type than previously evaluated in the FSAR.
- 3. This change does not affect the canputer systun sof tware or the I/O hardware used to nuet the Technical Specification recuirerrnts of sections 3/4.1.3 and 3/4.2.1. In acdition, if the systan is inoperable, procedures are in place to perfonn these functions nnnually on un increased surveillance cycle. Therefore, this change does not reduce the margin of safety as defined in the bases for the indicated Technical Specification sections.
89-V1M082 This MDD neves root valve 1-1305-X4-954 and IPC-4446 process sensing connection downstream of 1PV-4446 to an area of laminar flow.
- 1. This design change does not affect safety related equipment or canpanents. Therefore, it does not increase the probability of a un1 function of safety related equip' ment or canponents evaluated in FSAR section 15, Accident Analyses
- 2. No new equipment or cauponent unlfunctions are created by this rodification. This included review of the FSAR section 10.4.7 and section 15,
" Accident Analyses".
- 3. There is no change to the Technical Specification nargin of safety by this nodification. This included review of bases of Technical Specification 3/4.7.
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Il 1990 NMht REPORT - PART 2 10 CFR50.59(b) REIWf 89-VQ4095 This FDD will lengthen the " depth of bury" for fire hydrant C-2301-U4-928 in order to raise the hydrant base to grade elevation.
- 1. The fire hydrant will operate exactly as the current configuration, it will only be nore accessible.
There is no increase in the probability or consequences of any accident or un1 function of any safety related equilment not already addressed in the FSAR.
- 2. This change only extends a fire hydrant bury depth.
No new type of taalfunction is possible.
- 3. VEGP Technical Specifications do not address fire hydrants. Therefore, the Technical Specification bases are not affected 3/4.7,0.
89-VQ4096 This FDD revises the setpointa for tine-delay relays (62-devices) for fans 1-1556-B7-007 (Control Building Piping Restraint Fan) and A-1531-B7-008 (CR lLitchen and Toilet Exh. Fan) and will also revise tl'e setmoints for tenperature switch ITISH-22513, waich starts fan 1-1540-B7-006 (East-West Nonnal Electrical Tunnel Exhaust Fan) on high tunnel tuuperature.
- 1. The proposed changes do not affect any safety related equipmnt, instrumnt, or other canponents and theretore will not increase the probability of occurrence or consequences of a malfunction of any safccy related equipuent or canponents previously evaluated in the FSAR.
- 2. The proposed changes will enable the affected equipmnt to operate as designed and as described by or inplied in the FSAR. 1his equipmnt has no safety design function. Therefore, the changes will not create the possibility for an accident or equinxmnt/ component un1 function other than those whic'a have been previously addressed in chapter 15 of the FSAR.
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- 3. The equiparnt affected by the proposed changes is not addressed by the Technical Specifications.
Therefore, there will be no decrease in the turgin of safety defined by the bases for any of the Technical Specifications. 89-V2M098 This !OD deletes Annunciator A1306D01, "illlfr liter!%L STRATIFICATI0ti", and associated cables.
- 1. The piping tarperatures are obtained on a regular basis instead of a response to an annunciator.
Trends are obtained prior to annunciator alann.
- 2. 'lhe nonitoring of tenperatures will detennine if thental stratification is occuring. A Technical Support Departnent progran nonitors critical piping to:peratures using TJR-27734.
- 3. At the present tine, the annunciator ALB06001 is not being used due to the constant tonitoring progran and as such, the annunciator is a nuisance alant. There are no Technical Specificaticas or bases associated with the thenm1 stratification progran.
89-V1M105 This MDD increases the " Fast overload Shutdown" setpoint of Inverters 1AD1111 & IBD1I12 frun 165% to 195% of inverter rating.
- 1. 'Ihe proposed change will not increase the probability of occurrence or consequences of a un1 function of safety related equipnent or cmponent previously evaluated in FSAR chapter 15.
Increasing the fast overload shutdown to 195% of inverter rating does not affect nonnal operation of the inverter.
- 2. The proposed change does not create the possibility for an accident or equipnent carponent un1 function of a different type other timn evaluated previously in the FSAR chapter 15. Increasing the fast overload shutdown to 195% of inverter rating does not affect nonnal operation of the inverter.
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- 3. The proposed change does not decrease the turgin of safety defined by the bases for the Technical Specifications. (Ref. chapter 16 bases for Technical Specifications.)
89-V2M105 This MDD increases the " Fast Overload Shutdcun" setpoint of inverters 2AD1111 and 2BD1112 fran 165% to 195% of inverter rating.
- 1. The proposed change will not increase the probability of occurrences consequences of a uent or malfunction of safety-related emponent previously evaluated in equi}FSAR chapter 15.
Increasing the fast overload shutdown to 195% of interter rating does not affect nomal operation of the inverter.
- 2. The proposed change does not create the possibility for an accident or equi ment t ccuponent tmlfunction of a different type other than evaluated previously in the FSAll chapter 15. Increasing the fast overload shutdcum to 195% of Inverter rating does not affect nonnal operation of the inverter.
- 3. 'Ihe proposed change does not decrease the margin of safety defined by the bases for the Technical Specification. (Ref, chapter 16 bases for Technical Specifications.)
89-V2M106 A. This MDD removes (ficw orifice) ID-5560 trcxn Steam Generator Feed Pmp (SGFP) "A" lube oil drain to conditioner and renoval of 10-5557 fran SGFP "B". B. Rotates valves 2-1307-U4-503 and 505 frcm a vertical to a horizontal stem position. Valve 505 will be noved to allcw this rotation.
- 1. FSAR Table 10.4.7-1 addresses the loss of a steam generator feed pm p. This change will not increase the probability or consequences of such a cmponent failure. Ccoplete loss of the lube oil conditioning system will not fail the feed punp.
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- l 11 1990 AtNUAL REPORT - PART 2 10 CFR50.59(b) REIVRT l- 2. The worst case failure of the lube oil conditioning
- system will not increase the probability or consequences of an accident described in FSAR chapter 15.0,
- 3. Steam Generator feed pum,ps are not included in the bases for Technical Specifications steam -
waterlevel(lossoffeedwater), Tabic 2.Sencrator -1 #13, 89-V2M107 This MDD nodifies the Tendon access shaft covers.
- hese covers are located at Contairnent Buttress No. 2 and 3. Unit 12.
- 1. This change will not reduce the structural integrity of the cover. The access cover cannot create the malfunction of any safety related equipr:ent, d
- 2. The cover will nect the saae design criteria as original construction. The nodification will not create an additional failure node, h e change will not create the possibility for an accident other than previously evaluated in the FSAR, chapter 15. ,
- 3. This change will have no affect of the bases for the Technical Specification. The nodification will ensure the-covers can be renoved safely. - he change will neet the codes and specifications used for plant construction, 89-V1M108- his MDD. revises the Anertap system by rmaval of the ball circulation nonitor and its associated functions.
- 1. Bere is no safety related equipment involved with this change. Per FSAR sections 2.2, 4, and 1.3, there are no safety design bases for the main condenser.
- 2. his change will only renove a nonitor that-serves no function due to its aaplication. It cannot in any way create the possibnity of an equiprent malfunction not discuand in FSAR.
- 3. Condenser is not safety related and has no safet,f design bases in Technical Specifications.
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i 11 1990 AIM \L REIORT - PAllT 2 10 C1R50.59(b) id3 ORT 89-Vl!G09 This l@D installs a top uvunted handkeel type up (adiustable) travel stoa on tenperature control valve '1V-6800. This will ne accmplished by rmoving existing upper diaphram casing and installing a new diaphram casing which has a handwheel / stun travel stop incorporated into the casing.
- 1. The change does not effect any safety related systen as described in the FSM1, nor is stator cooling water required for safe shutdown of the plant (Reference FSAR Systen definition - section 10.2).
- 2. The change to stator cooling does not nor can not cause an accident or m1 function of a caponent that is different fran those described in FSAR sections 10.0, 15.1, and 15.2.
- 3. The change does not effect the nargin of safety as defhied in Technical Specifications section 3/4.7.
89-V2M110 This MDD installs a top nounted handwheel type up (adjustable) travel ston on tmperature control valve TV-6800. This will b acccuplished by renoving existing upper diaphran casing and installing a new diaphran casing which has a handwheel / sten travel stop incorporated into the casing.
- 1. The change does not effect any safety related systun as described in the 1741, nor is stator cooling water required for safe shutdown of the plant reference FSAR systan definination - section 10.2.
- 2. The change to stator cooling does not nor can not cuase an accident or malfunction of a canponent that is different fran those described in FSAR sections 10.0, 15.1 and 15.2.
- 3. The change does not effect the unrgin of safety as defined in Technical Specifications section 3/4.7.
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I I II l 1990 MM1AL REIORT - PART 2 1 10 CFK50.59(b) RDVRT , 89-V1M113 1his MDD involves ratoval of sight glasses frctn feedwater heaters, toisture separator drain tanks, heater drain tanks, and steau line drcin collection points. . 1. Ratoval of these sight glasses are not associated in any way with any safety telated equiptent as referenced in IM sections 10.0 and 10.1, on steam ' and power conversion and definitions.
- 2. Rmoval of sight glasses does not create the possibility of any accident or equipnent un1 function not previously evaluated in IM sections 15.1 and 15.2.
- 3. Rmoval of sight glasses does not effect any itan discussed in Technical Specifications. Reference i section 3/4.7.
l 89-V2H114 This MDD involves rmaval of sight glasses fran feedwater heaters, noisture separator drain tanks, heater drain tanks, and steau line drain co11cetion points.
- 1. Renoval of these sight glasses are not associated in any way with any safety related equiptent as referenced in FSAR secticos 10.0 and 10.1, on steam and pcuer conversion and definitions.
- 2. Rmoval of sight glasses does not create the possibility of any accident or equipnunt malfunction not previously evaluated in FSAR sections 15.1 and 15.2.
- 3. Ranval of sight glasses does not effect any itmi discussed in Technical Specifications. Reference section 3/4.7.
89-VD1115 This MDD deletes the autunatic Steam Generator Feed Pum Turbine trip on high vibration. The high , vibration alann will runnin active to ensure corrective actions can be taken in the event a high vibration condition actually exists. L 4 l 154
11 1990 N M1AL RDVRT - PART 2 10 CR60.59(b) ImRr
- 1. The SGRT's have been considered in failure analyses both on Table 10.4.7.1 and chapter 15.2.
These deal with a loss / trip of the pw p. This change will decrease the probability of a pw p trip and the subsequent safety related actuations.
- 2. A SGHT trip is evaluated in FSAR chapters 10.4.7 and 15.2.7. This change will decrease the probability of a pw p trip and the associated safety related actuations.
- 3. The SGFPT trip on hign vibration is not addressed in the Technical Specifications bases.
89-V2M117 This MDD deletes the autantic Steam Generator Feed Pmp Turbine (SGFP) trin on high vibration. The high vibration alam will rennin active to ensure corrective actions can be taken in the event a high vibration condition actually exists.
- 1. The SGFIT's have been considered in failure analyses both on Table 10.4.7.1 and chapter 15.2. These deal with loss / trio of the pw p. This change will decrease the probaallity of a pwp trip and the subsequent safety related actuations.
- 2. A SGnT trip is evaluated in FSAR chapters 10.4.7 and 15.2.7. This nodification will decrease the probability of a pmp trip an the subsequent safety related actuations.
- 3. The SGFIT trip on high vibration is not addressed in the Technical Specifications bases.
89-VlM118 This MDD replaces the original noter relay for 1 TIS-6810 (stator cooling) with a different (nodel/ manufacturer) type noter relay and changes terminal wiring at the noter relay to keep the circuit operation / logic the saae. 155
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- 1. 1he noter relay type chmiges does not affect the safety related systuu as described in the ITAR, nor is stator cooling water required for safe shutdwn of the plant, as referenced in FNR section 10.2.
- 2. The noter relay type change to stator ecoling does not, nor can not, cause an accident or un1/ unction of a ccuponent that is different fran those described and evaluated in FSAR sections 10.0, 15.1, and 15.2.
- 3. The change of txter relay type does not effect the nargin of safety as defined in Technical Specifications section 3/4.7.
89-V1M119 'Ihe air supply regulator for the actuators of heater drain tank high 1cvel dtnp valves 1LV-4333 and 1LV-4334 is being increased fran 60 paig to 80 asig. This nny also require the actuator spring to ae changed.
- 1. The proposed change is not to a safety related systen and has no effect on any safety related equipnent evaluated in chapter 15 of the FSNI.
- 2. The increased air supply pressure to the actuator ck>es not charge the valve's function. Failure of these valves is bounded by the Turbine Trip Analysis.
- 3. These valves are not addressed in Technical Specification 3/4.7 and they are not a part of any Technical Specification bases.
89-V2M120 The air sup31y regulator for the actuator of llenter Drain Tank aigh level dinp valves 11N-4333 and 1LV-4334 is being increased fran 60 psig to 80 psig. This tray also require the actuator spring to be changed. 156
i I II 1990 Ati! RIAL PFtDRT - PART 2 10 Cal 50.59(b) REIORT
- 1. The proposed change is not to a safcty related systen and has no effect on any safety related equipnent evaluated in chapter 15 of the FSAR.
- 2. The increased air supply pressure to the actuator does not change the valve's function. Failure of these valves is bcunded by the Turbine Trip Analysis.
- 3. These valves are not addressed in Technical Specification 3/4.7 and they are not a part of any Technical Specification bases.
89-V1M121 This change adds six flanges in the sprinkler piping for systun 541 (Unit 1 T.B.) and replaces welded su; ort steel nuubers with bolted support steel ud era.
- 1. This chanE;e involves non-safety related sprinkler piping and supports located in the 1\irbine Building. There is no increase in the probability of occurrence or consequences of a malfunction of safety related equi}uent described in the FSAR.
- 2. 'Ihis change does not alter the sprinkler systan's function in any way. It only adds the capability for casier re: oval when needed. There is no possibility for any accident, failure or nulfunction of any canponent or systen that is not already evaluated in the FSAR.
- 3. V mP Technical Specifications do not address 11trbine Building sprinkler systans. This change only affects Turbine Building sprinkler systan 541 und associated supports. Therefore, this chmige does not affect the Technical Specification unrgin of safety in any way.
89-V1M150 This MDD will add a aressure snubber to the camon sensing line of the deater Drain I\rnp (HDP) seal injection controller and pressure indicator IPC-4379, IPI-4385 for lirup A,1N:-4380, IPI-4386 for lirnp B. 157
i Il 1990 M MIAL REIORT - PART 2 10 CFR50.59(b) REIVRT
- 1. This clunge does not involve safety related equipwnt but it should be noted that even ecuplete loss of the affected line (condensate systun seal injection supply) will not cause a llDP trip. On loss cf seal pressure, a back-tn dunineralized seal water supply will be provided LTrough a nornully closed check valve.
- 2. loss of feedaater is evaluated in FSAR section 15.27. Toere is also no increased danger of flooding per review of section 3.4 '%2ter level (FloocU Design".
- 3. There is no decrease in the margin of safety defined or inplied by Technical Specifications.
93-VCM089 This l@D isolates individual heating elumnts which are causing ground indication and alanns on 480 volt switchgear MG13. It clarifies nimilar isolations previously done on A-1541-A7-002-Il001/ H02 per ABN 88-V1000A167.
- 1. The Fuel llandling air supply units A-1541-A7-001/002 are class 626 and the respective feeder breakers NG13 and ANB14 are also non-Q. Therefore, this change cannot increase the probability of occurrence or consequences of an accident or malfunction of equiprent inportant to safety as evaluated in FSAR section 15.
- 2. As all final design parmeters are still being net, this change cannot affeet the llVAC systun function. This change cannot cause any type of accident.
- 3. The fuel handling nonnal system IIVAC is not included in the Technical Specification bases. The nonnal cooling function which could affect sme of the rotat listed in table 3.7-3 (Fuel B008) is not affected.
90-VlM091 This MDD replaces IFSHL-1928 and 1FSHL-1929 with a f1cu switch that ir.corporates a five second tine delay. This is in accordance with the reemnendations made in REA VG-9638.
- 1. Addition of a short tino delay to the alann in no way affects the ability of safety related equi; ment to fulfill its safety function.
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- 2. Addition of a five second tim delay on a non-Q alann function has no imaet on safety related equ1} ment . Therefore, this change does not create the possibility of an accident or malfunction different than previously analyzed in the FSAR.
- 3. This change does not decrease the mrg; ins of safety defined in the bases for any Technical Specifications, including 3/4.7.3.
90-V2M092 his MDD replaces 2FSIL-1928 and 2FSIL-1929 with a flow switch that incorporates a five second tire delay. This is in accordance with the recamendations mde in REA VG-9633,
- 1. Addition of a short tine delay to tb alann in no way affects the ability of safety relcted equip:ent to fulfill its safety function.
- 2. Addition of a five second tine delay on a non-Q alonn function has no innact on safety related equirnent. Therefore, tais change does not create the mossibility of an accident or m1 function different tun previously analyzed in the FSAR,
- 3. his change does not decrease the mrgins of safety as defined in the bases for any Technical Specifications, including 3/4.7.3.
90-V1M099 This MDD raroves heat tracing fran the electronic portion of Radiation Monitor 1RE-12444C, Update load chart drawing lyJAF05-479 to reficct 41 feet of heat tracing for ekt C1-11 and a new low alann setpoint of 90 F.
- 1. Remving 39' of heat tracing fran Radiation Fbnitor 1RE-12444C and reducing the low alann setpoint to 90 F for the heat tracing will not increase the probability of occurrence or consequences of a m1 function of safety related equipnent or canpanent previously evaluated in FSAR chapter 15.
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- 2. The proposed change will not create the pooibility for an accident or equi tent /ca.ponent m1 function ot a different type other than any evaluated previously in FSAR chapter 15.
- 3. The proposed change will not decrease the mrgin of safety defined by the bases for the Technical Specifications . Reference Technical Specification bases 3/4.11, 90-V1M106 This MDD provides temporary rigging for lifting the Unit 1 Containnent Equipnent Hatch, which weiglw approximately 32,000 lbs, with the polar crane. This tmporary rigging will be used in lieu of the existing permnent hoists located above the equiptent hatch. Add stiffeners and liftirg lugs to the structural steel at elevation 269'-0 and 244'-7" to support srmtch blocks and slings for the tmporary rigging.
- 1. The tmporary rigging will only be used during Mode 5 and 6. Failure of the tmporary rigging my cause the dro) of Equipient Hatch over on the structural steel aernns and checker plate directly below the equip:ent batch opening. As described in FSAR section 9.1.5.3.1.1, this will not inpair decay heat rmoval or the maintenance of cold shutdown due to the physical separation of the RHR systen and its power supplies. In addition, since the load path for lifting the equip;ent hatch cover using the polcr crane is identical to the load path when using the installed hoists, the consequences of a failure would be no different than those associated with the failure of the perm nently installed hoists and rigging. Therefore, the proposed tecporary rigging will not increase the probability of occurrence or consequences of the m1 function of any equipnent or emponent assured to function in accidents armlyzed in the FSAR. ,
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- 2. Failure of the tenporary rigging nay create the possibility of the equipuent hatch cover falling on the structural steel beams and checker plate located directly below. The possibility of an equiptent hatch cover load drop is not specifically deacribed in the FSAR. Hcwver, based on a review of FSAR section 9.1.5.3.1.1, this failure will not cause damage to any safety related equiptent required during Mode 5 and 6, and will not areclude decay heat raroval or maintenance of cold shutdown.
Also, the consequences of this accident will be the sano as the failure of existing hoists during the lifting of the equiptent hatch cover. Utilizing the polar crane to lift the equiptent hatch cover will not adversely affect the polar crane since the polar crane will be subjected to a straight vertical pull significantly less than its rated capacity. When in une, the polar crane 41 be under adninistrative controls to ensure t.. no lateral loads are applied to the crane. Therefore, utilization of teuporary rigging does not create the possibility of an accident or equip:ent/ nulfunction not inplied or analyzed in the FSAR.
- 3. This undification will have no effect on the safety enrgins provided by Technical Specifications sections 3/4.6.1 and 3/4.9.4, or associated bases.
90-VlM107 '1his MDD changes position transmitter 1-2T-9450 (on cooling tower blowdown valve) fran Fisher nodel 3552 to Fisher model 4211 and reduces the power supply voltage fran 40 VIE to 24 VDC.
- 1. Ib malfunction of the cooling tower blowdown valve could result in an event nore severe than the loss of circulating water, which is already analyzed in FSAR section 15.2.5. No safety related equiprent is effected by circulating water.
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- 2. As' discussed above, any malfunction is still bounded by the analysis in FSAR section 15.2.5.
- 3. h e circulating water system is not discussed in the
' basis for any Technical Specification based on a !
review of the bases for 3/4.5 and 3/4.7. l 90-V1M108 he position transmitter on turbine lube oil l temperature control valve l'IV7116 is being changed i frm a Fisher mdel 3552 to a mdel 4211. The power supply voltage is being reduced fran 40 VDC to 24 VDC.
- 1. Any failure is bounded by a turbine trip, which is analyzed in FSAR section 15.2.3. No safety related equipment is effected by turbine lube oil or TPCW.
Since the new part is similar to the old part, the failure probability is unchanged.
- 2. Any failure will, at worst, result in a turbine trip, which is analyzed in FSAR section 15.2.3.
- 3. Turbine lube oil and TPCW are not involved with the bases for any Tecletical Specification, including 3/4.5 and 3/4.7.
90-V1M109 This MDD provides an alternate method for installing the turbocharger exhaust flange bolting on the Emergency Diesel Generator (EDG). Bis method will be used when existing field conditions prevent installation of the bolts per the existing-design drawings.
- 1. The alternate bolting nethod will not increase the probability of occurrence or consequences of a malfunction.of safety related equipment. The alternate bolting has no more probability of failure than the original bolting material. This material and method of installation were recomiended by the vendor. W erefore, it will-not increase the probability of an EDG failure.
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- 2. The alternate bolting nethod can not create an accident or equipuent/ccxuponent malfunction of a different type other than any evaluated in the FSAR.
This change will have no affect on the single failure criteria assu:ed in the FSAR. The bolting noterial is recatmended by the vendor. Therefore, it cannot change the FSAR analysis. 3, This change has no effect on the bases of the Technical Saecification. This change has no effect on tae operation of the EDG. This MDD will not require a cha e to Technical Specification bases 3/4.8.1 or 3 4.8.2. 90-V1M110 In the hotwell of the "A" nmin condenser (1-1305-E4-005), a broken baffle plate is being repaired and reinforced by adding various utinor structural steel plate and angle naubers. The condenser will be returned to near the original design condition by repairing damage observed during 1R2 outage, which was nest likely caused by other dannge which was found and repaired during 1R1.
- 1. The condenser is not safety related, and is not assumed to function in accident conditions. No safety related equipnent is affected by the condenser. FSAR 10.4.1.3 states that the condenser has no safety design basis.
- 2. Any malfunction of the condenser internals will at worst result in a loss of condenser vacuan (turbine trip) or loss of unin feedwater, both of which are already analyzed in the FSAR Section 15.2.
- 3. The condensers are not discussed in the basis for any Technical Specification. This is based on a review of the bases for sections 2 and 3/4.7.
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i l l 11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 90-VQ4112 . The setpoint drawing and the logic diagrams associated with "Q" tmperature switches 1/2-TISH-12051 and 1/2-TISH-12054 require revision to show the "as-built" condition in the field. These switches provide the auto start. runction for the second essential fan in the Diesel Generator building when the diesel irrunning and the taperature rises above 807. Revise Unit 1 switches to show the two stage action for the United Electric switch. Revise Unit I and Unit 2 reset values to be consistent. Revise logic diagrams to show a separate signal for the stop function, not just a termination of the start signal.
- 1. The affected temperature switches are not considered directly in the failure modes analysis included in FSAR section 9.4.7. As the switches are not being physically modified, no new chance for failure other than previously evaluated is anticipated.
- 2. This' change is a drawing change to make setpoint and logic diagrare agree with the higher level-elementaries and wiring diagrans. The setpoint change had previously been evaluated and approved for Unit 2 during the construction phase.
- 3. The margins-of safety for Specification 3/4.7.13 are unaffected as the system will continue to function as shown on the elementaries. The establishment of the "stop"'setpoints will ensure that the fan is allowed to run long enough to meet its design bases and prevent _ unnecessary cycling of the essential tan.
90-V1M113 This MDD remaves type H6B and 4AS anchor bolts as shown on minor departure sketch drawings.
- 1. The removal of these bolts cannot cause the malfunction of any safety related equipent. The removal of these bolts is for personnel safety only.
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- 2. As stated above, the rmoval of these bolts cannot create a cmponent tmlfunction. The rmoval of these bolts will provide personnel safety.
- 3. The anchor bolts are not covered in the bases of the Technical Specification. The renoval of these bolts will Imve no effect on the bases of the Technical Specification, 90-V2M114 This MDD adds a personal emputer (PC) interface to the Unit 2 Proteus caputer systen. The equipment added is a MID card to the switchable progranmers bus, a cable (XWO39) fran the MID card to the PC, and a PC, powered fran a 120V distribution panel internal to Proteus. The necessary software change associated with this change will be controlled and documented per procedure 50015-C, CSCP CR-90-0020.
The systan response and/or operation of the cmputer systan is not affected by this change. This PC will perfonn as a batch terminal to the systan. This will allow for editing and sane progrmning changes on-line.
- 1. The addition of the PC interface will not affect the probability of canponent failure. The Proteus Cmputer is not a safety related systen and it does not directly interface with safety related emponents . The change therefore, does not increase the probability of occurrence of a malfunction of a safety related emponent in a nunner described in the FSAR. It is not referenced in the accident analysis in section 15 of the FSAR..
- 2. The accident analysis described in section 15 of the FSAR was reviewed to determine that no possibility for an accident or equipment / component malfunction of a different type than described in the FSAR would be caused by the addition of the PC interface.
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- 3. We Proteus Cmputer's function of muitoring for Axial Flux Difference and Rod Position Deviation are not changed or emprcaised in any way by this change. Therefore, the mrgins of safety defined by the bases for Technical Specifications 3/4.2.1 and 3/4.1.3.2 are not decreased by the addition of the PC interface.
90-VlM115 This MDD eliminates the internal flow switch frmi the filter unit heater for the Post-loca Purge Exhaust, 1-1508-N7-001-H01.
- 1. Le heater, the filter housing and all ecuponents downstream of the filter housing are non-safety related ccraponents and their failure has not been analyzed in the FSAR.
- 2. The heater nulfunction is satisfactorily prevented by the redundant thermal protection provided and by the administrative guidelines found in procedure 13130-1. Le accident analysis of FSAR section 15 are unaffected by this change.
- 3. B is heater was designed to reduce the relative htraidity of the inecming air to 70% R.H. where mre efficient iodine adsorption can occur. However, the included flow protection was improperly designed for the extrm ely low velocity pressures seen at tue heater coil. This change will allow the heater to operate as designed without undo risks. Therefore, the margin of safety for Technical Specification 3.11.2.4. is not reduced.
90-VCM120 This MDD rewires internal wiring on the Blowdcun Sulfonator panel. Rese changes will provide nunual operation of the Sulfonator skid.
- 1. Inplementing this change sinply makes operable the dechlorination system which is determined in the FSAR not to be a threat to aersonnel or equipment required for a safe shutdown. Reference section 2.2.3 of the FSAR. Rus, safety equipnent malfunction probabilities and consequences are unchanged.
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11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 2. Inplemnting this change sinply makes operable the dechlorination systan which is determined in the FSAR not to be a threat to equipw nt or personnel required for a safe shutdown (per FSAR section 2.2.3). 'Ihus, no new accidents or failures are created.
- 3. Neither the circulating water dechlorination system, cooling tower bloulown nor the chanicals used in these systems are a part of the Technical ecifications. A special review of section Sp/4.11 3 was made. Thus, the Technical Specification margin of safety is unaffected.
90-V2M124 This MDD adds one support and the stiffening of another existing support on line 2-1326-505-2" in the stator cooling piping system area of flow switch 2FS-6832.
- 1. loss of stator cooling water can not effect any safety related system as described in FSAR accident analysis sections 15.1, and 15.2, nor is stator cooling required for safe shutdown as described in FSAR section 10.2.
- 2. loss of the stator cooling water system does not, nor can it cause, an accident or malfunction of a safety related cauponent that has not been previously described in FSAR sections 10.0, 10.2, 15.1 and 15.2.
- 3. The change or loss of stator cooling water does not effect the margin of safety in Technical Specification because stator cooling can not erfect any safety related cauponent required for safe shutdown of the urat.
90-V2M125 'Ihis change replaces welded steel supports with bolted steel supports located in Unit 2 Turbine Building. 167
4 11 1990 ANNUAL PLIVRT - PART 2 10 CFR50.59(b) REPOKr
- 1. There is no increase in the consequences of an accident previously evaluated in the FSAR.
- 2. This change involves non-safety related supports located in the Unit 2 Turbine Building. Therefore, there is no increase in unlfunction oI equipment important to safety.
- 3. The support steel piping supports are located in turbine building Unit 2 and are non-safety related, so there is no increase of consequences of nulfunctioning equipment inportant to safety previously evaluated in the Technical Specifications.
90-V2M133 This MDD adds a reducing pipe tee and root valve arrangenent in the TPCCW cooling water outlet line of air canpressor #3 aftercooler.
- 1. The failure of the canpressed air system or TPCCW can not, thru their design or function, effect any safety related couponent.
- 2. The failure of the caupressed air and TPCCW have been accounted for in the FSAR accident analysis section 15.0, with decrease / increase heat removal by secondary plant systems.
- 3. The bases for safe sintdown of the plant is not affected by cauplete failure of the caupressed air system or TPCCW based on a review of Technical Specifications bases 3.0 and 4.0.
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A 1 I i 1990 MHJAL PIPORI' - PART 2 10 CFR50.59(b) TEST OR EXPERIMENI'S ~ _ _ - _ _ - - _ _ _ _ _ _ _
ni II 1990 ANNUAL PEPORT - PART 2 10 CFR50.59(b) REPORT TENG'S TENG-90-03 Lis procedure obtains flow and pressure data for design engineering to detetmine the ACCW-system setpoint change for valves 1/2HV-2041. ;
- 1. This procedure operates the ACCW system within all design limits. Valve 2HV-2041-is blocked open areventing a " loss of coolant to the thennal aarrier" as described in FSAPisection 9.2.8.
herefore, this procedure will not increase the
-occurrence or consequences of-a malfunction as analyzed in the FSAR and Technical' Specification.
2, The systan is operated within FSAR and Teclutical Specification limits. The addicion of pressure and flow gauges-does not increase the chances-of safety equipment malfunctions. The pressure gauges attached to.the safety related piping will be rated for RCS pressures.
- 3. The ACCW system.is operated within limits set by .
the FSAR and Technical Specification. . Berefore, there is no decrease in the margin of safety. T-ENG-90 he Turbine Torsional Vibration Test consists of TER.t 90-004 for.line-up and performing the test, and temporary modification #1-90-009, 10, and 12 to facilitate data collection and create conditions to perfonn the test. - (Reference to TIL-1012-2, T-ENG-90-05, T-OPER-90-02 and 03 for detail)
- 1. -- his change does not' increase the, probability of malfunction of any equipment or s mponent of safety related systems as described in FSAR section 15.0.8. The turbine generator is not affected as described in sections.10.2.4 and 10.2.1,
- 2. This test will not create accident other than '
evaluated in FSAR 10.2.4. The test is conducted at low power and is within'the design limits of canponents. . Any malfunction in--turbine or 4 generator is analyzed in sections 15.1.3 or 15.2.3. 169
II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPOIC
- 3. The test does not alter any response twe.
Therefore, the nnrgin of safety is not empranised. The generator serves no safety function and has no safety design base. T-ENG-90-07 1his Tenporary Engineering Procedure denonstrates the proper bypass of the Diesel Generator autanatic electrical trips upon receipt of an Emergency Start Signal during S1 with concurrent LDSP.
- 1. This test procedure represents no physical or operational nethodology cbange to the plant. For tais reason, the probability and consequences of an accident or un1 function previously analyzed in the FSAR rennins unchanged by this test.
- 2. Since this Technical Specification required testing represents no physical change to the plant, this test does not in any way affect the possibility of an unanalyzed accident. The function tested by this procedure is detailed in FSAR 8.3.1.1.3.c.
- 3. The non-physical nothodology change nature of this test procedure ensures no inpact to the nnrgin of safety defined in the bases of the Technical Specifications. This testing is required by Technical Specification Surveillance Requirenrnt 4.8.1.1.2.6.c.
T-ENG-90-09 This temporary procedure perfonns the following on both Train A and B Diesel Generators:
- 1) Runs the D/G at the 2 hour rating for about 10 minutes and nonitors perfonnance paraneters,
- 2) Perfonns a largest single load rejection while
- 3) nonitoring/pertonnance paraneters, Runs the D G at the continuous ra 8 hours and nonitors perfonnance paraneters,
- 4) Perfonas a 100% load rejection while nonitoring perfonnance parannters.
In addition to these, the B train section (5.2) loads the SIP, RHRP, CSP, and the ESF Chiller onto the 1E Bus at the approxinute tine intervals as would the sequencer. 170
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II 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 1. This temporary procedure performs testing which sup31ments test requirments called out in Tecanical Specification 4.8.1.1.2. Therefore, this temporary procedure actually decreases the
- probability of a nelfunction of equipnent important to safety by testing beyond the requirments of Technical Specifications. This i
I temporary procedure has no affect on the probability of occurrence or consequences of an accident.
- 2. This temporary procedure operates each train DC independently and separately. No nodifications affecting operating characteristics of any cmponent are unde under this 3rocedure. Therefore, no new type of malfunction has been introduced by this l procedure. This procedure has no affect on the potential for accidents or initiating events.
- 3. Since each DG train is operated independently, we are assured of rmiaining within the limiting conditions for operation as called for in Technical Specifications. Therefore, the margin of safety as defined in the bases of Technical Specifications 3/4.8.1 and 3/4.8.2 is not reduced.
T-ENG-90 This procedure provides detailed instructions to obtain contact operating times for auxiliary contacts used in the bus transfer schene for switchgear 1NAA and 1 NAB. This procedure provides instructions for obtaining ?;reaker operating times which are also required to evaluate the operation of the bus transfer schme. _ This procedure was a necessary to facilitate. the adjustment of the 1 breaker auxiliary contacts and provide further l information in the investigation of the 1NAA fast l transfer failure. l
- 1. This procedure only provided instructions for obtaining contact operating. times. Therefore, since this procedure does not change the plant as {
described in the FSAR (section 8.3.1) or change the design configuration of the plant in any way, it does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. 171
i 11 ! 1990 A!UUAL REPORT - PART 2 10 CFR50.59(b) REPORT l
- 2. The procedure will be perfonned during outage conditions when the switchgear or equipment supplied by the switchgear is not required.
Therefore, performnce of this procedure will not affect the plant or create the possibility of an accident or m1 function of a different type than previously evaluated in the FSAR.
- 3. The procedure only obtains contact timing infonnation and does not reduce the margin of safety as defined in the bases for Technical Specifications section 3/4.8.
T-ENG-90-11 This procedure sinulates, as close as possible, the conditions and sequence of events for Unit one A-Train loss of Offsite Power incident which occurred March 20, 1990. Data collected during this test will be used to detennine the root cause of the DG/ Sequencer anomlies experienced during the March 20,199010SP event. This test alone is not intended to prove DG operability.
- 1. This temporary procedure will not increase the probability of occurrence or consequences of failure of any camponent evaluated in the FSAR.
The procedure is being performed to identify the cause of the DG trips during the March 20, 1900 LOSP event. Corrective actions taken as a result of this informtion will provide the assurance that the DG is prepared to respond to an IDSP as described in section 8.3 of the FSAR. Testing will be perfonmd as described in 3/4.8.1.2.h of the Technical Specifications and FSAR section 8.3.
- 2. This temporary procedure will not create the possibility for an accident or equiptmnt canponent malfunction not previously evaluated in the FSAR.
The structure and methodology of this procedure is similar to that of the ESFAS System testing ecifically required by Technical Specification sp/4.8.1.1.2.h 3 and described in FSAR section 8.3. This testing is being performed to ensure that the Diesel Generator and Sequencer is prepared to perform its safety related function in response to an IDSP. 1 172
Il 1990 ANNUAL P2 PORT - PART 2 10 CFR50.59(b) REPORT
- 3. Le twthodology and structure of this procedure is similar to that of the ESFAS System Tests, 54055-1 and 54065-1. Le relative safety of this procedure is the same as that of the ESFAS procedures which are specifically required by Technical Specification 3/4.8.1.1.2.h and FSAR section 8.3.
T-ENG-90-12 This temporary procedure aligns the plant and simulates a loss of Offsite Power. The plant's response is recorded, documenting the proper operation of the Diesel Generator, the Sequencer and various safety related, sequenced loads. The test equipment is removed at the end of the test, and the plant is restored under the control of the Unit 1 Shift Supervisor. L e test structure and methodology is similar to that of the recently ccupleted ESFAS surveillance procedures, 54055-1, and 54065-1.
- 1. This tmporary procedure will not increase the probability of occurrence or consequences of failure of any ccruponent evaluated in the FSAR.
Le procedure is being performed to identify the cause of the DGlA trips during the March 20, 1990 10SP event. Corrective actions taken as a result of this information will provide the assurance that the DG is prepared to respond to an IDSP as described in section 8.3 of the FSAR. Testing will be performed as described in 3/4.8.1.1.2.h of the Technical Specifications. 2. The structure and methodology of this procedure is similar to that of the ESFAS System testing specifically required by Technical Specification 3/4.8.1.1.2.h and described in FSAR section 8.3. Wis testing is being perfonned to ensure that the Diesel Operator and Sequencer are prepared to perfonn their safety related function in response to an IDSP. 173
II 1990 AhWUAL REPORT - PAKr 2 10 CFR50.59(b) REPORT
- 3. The methodology and structure of this procedure is similar to that of the ESFAS System tests, 54055-1 and 54065-1. The relative safety of this procedure is the same as that of the ESFAS procedures which are specifically required by Technical Specification 3/4.8.1.1.2.h and FSAR section 8.3.
T-ENG-90-13 This test procedure performs testing of safety features sequencer circuitry. Testing is based on vendor reccenundations but has not been reflected into a pernunent plant procedure,
- 1. This testing has no negative affect on probability of occurrence or severity of consequences of an accident or malfunction evaluation in the FSAR.
This testing is composed of normal operation of the Sequencer Test Panel as directed by the vendor in the supplied vendor manual. Any receipt of a valid W or SI condition has been previously proven to place sequencer immdiately out of the test unde and into the nonnal operating mode.
- 2. For the same reasons identified in the inmediately preceding answer, this test has not created the possibility for any accident of a type not yet addressed by the FSAR.
- 3. Since this test is essentially operating the sequencer as designed and tested, the sequencer will innediately terminate an ongoing test upon SI or W receipt. This test represents no challenge to the margin of safety defined in the bases for any Technical Specification (Reference Technical Specification 3/4.3, 3/4.8).
T-ENG-90-14 Test of the function of the Sequencer Reset Module added by DCP, and all attended circuitry and camponents. 174
II 1990 N MJAL REPORT - PAPf 2 10 CFR50.59(b) REPORT
- 1. Since this test was perfonmd at a tine when the associated components were not required for safe operation of the plant, the probability of occurrence and severity of consequences of an accident remain unaffected. This test is required to ensure the Safety Features Sequencer System can be reset in the' event of a DG Breaker trip during stepping or a failure to close (Reference Technical Specification 3/4.3, 3/4.8).
- 2. Since this test is essentially design operation of the Sequencer as supplied frcxn the vendor und nodified by DCP (Separate SE was perfonwd for DCP), this test does not place the Sequencer in any configuration other t.han that to which it normally might be expected to respond. The powering down of the 4160 V buss is acceptable since all equipment affected is not required for plant support at the time of the test.
- 3. Due to the conditions and controls identified in the two previous blocks, the margin of safety for the plant as defined in the bases of the Technical Specifications 3/4.3, and 3/4.8 is maintained within pre-established limits. The powering down of the 4160 V buss is a relatively small challenge and is acceptable in the plant condition the test was perfortned in.
T-ENG-90-15 Test of the function of the Sequencer Reset Module added by DCP, and all affected cicruitry and components.
- 1. Since this test was perfonned at a time when the associated ccxnponents were not required for safe operation of the plant, the probability of occurrence and severity of consequences of an accident remain unaffected. This test is required to ensure the Safety Features Sequencer System can be reset in the event of a DG Breaker trip during stepping or a failure to close (Reference Technical Specification 3/4.3, 3/4.8),
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II 1990 ANNUAL REPORT - PART 2 10 CFK50.59(b) REPORT
- 2. Since this test is essentially design operation of the sequencer as supplied from the vendor and nodified by DCP (separate SE as performed for DCP),
this test does not place the Sequencer in any configuration other than that to which it normally might be expected to respond. The powering down of the 4160 V buss is acceptable since all equipment affected is not required for plant support at the tinu of the test.
- 3. Due to the conditions and controls identified in the previous blocks, the margin of safety for the plant as defined in the bases of the Technical Specifications 3/4.3 and 3/4.8 is maintained within pre-established limits. The powering down of the 4160 V buss is a relatively saall challenge and is acceptable in the plant condition the test was perforned in.
T-ENG-90-16 This test required removal of 1-TSH-19111 and installation of an RTD which would be connected to a chart recorder. 1TSH19111 is one three tmperature sensors located in the Diesel Generator Jacket Water return line to the standpipe. This tradification was reconnended by Cooper Energy Services to obtain a temperature profile at the sensor during standby, engine start ard operation unloaded.
- 1. The consequences of a Diesel Generator failure are as previously analyzed in the FSAR Table 8.3.1-3.
The probability of an accident is not increased because there are two remaining high jacket water sensors in service which will trip in any mode. The high jacket water alarm was still in service and nonitoring of the temperature is possible at the local panel. The local operator and engineering personnel will be present for the test.
- 2. Installation of an RTD in place of one temperature sensing device cannot create any new accident or malfunction not already analyzed in the FSAR. High Jacket Water tripping remains in effect on a two out of two basis and all alams are still active.
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11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3. The Diesel Generator system or subsystem are not made mre unsafe by the modification. Two out of two liigh Jacket Water Trips are in service including alarming and monitoring capability. Reliabilities of the diesel is not decreased. (Technical Specification 3/4.8 bases)
T-ENG-90-17 DCP 90-V1N0133, Rev 0, disabled several diesel engine automatic trips such as " crank case pressure hi turbo and oil pressure low, Jacket Water pressure low high vibration engine bearing temperature hgh, lube oil tenp hi, and loss of field or over-current (186 Block Out)", which provided engine trip during loss of Offsite Powr (LDSP) event. This change was perfoured by mving LDSP contact fran auto start circuit to energency circuit, by connecting parallel to SI contact. T-ENG-90-17 will verify that Design Change as properly inplemented by injecting 00SP signal to start the engine end operato in emergency mode, each with all the normal engine trips (bypass) bypassed and engine trips fran normal trip signal (bypassable) after reset of II)CA.
- 1. The test is to verify the result of the design change which was approved by NRC. The test is merely proving that diesel engine will not trip during loss of Off Site Power (LDSP) fran a just like during SI. All the bypassable juupers install trip 'ed are independently verified during installation and removal. All the operationa are within the bounds of FSAR and chapter 15 of the FSAR accident analysis and controlled under procedural restrictions.
Therefore, this test will not increase the probability of cecurrence or consequence of an accident or malfunction of equipment or component important to safety previously evaluated in the safety analysis report. i 177
I 11 1990 NNUAL lEIWP - PART 2 10 CfK50.59(b) IEPORT
- 2. his test is the functional test to prove this inplemntation of the design change DCP 90-V1N0133, his would be similar to a mnthly surveillance test, except the signal is initiated by connecting a junper in the sequencer panel.
These jumpers are removed after initiating the signal. Both installation and removal are independently verified. The test will be conducted in a controlled manner and could be termtinated at any tim without any adverse effect to systems or emponent. Operation of the Diesel Generator is similar in function to other test conducted to prove operability. Therefore, this test does not create the possibility of an accident or malfunction of a different type than previously evaluated.
- 3. This test is performed on a Diesel Generator that is taken to inoperable status to perform the functional test for DCP 90-V1N0133. This testing will prove the Diesel Generator would be available during IDSP condition when there is an actual or spurious trip frcm bypassable trip signals. Margin of safety is enhanced by increasing the availability of the Diesel Generator, here are alarms in both the control roam and local panel to alert operators in an event any byaassed trip parameters exceed the alarm setpoint waich are below the trip point. herefore, this test wi.11 not reduce the margin of safety as defined in the bases of Teclutical Specifications.
T-ENG-90-18 DCP 90-V1N0137, Rev 0, disabled several diesel engine automatic trips such as " crank case pressure hi turbo and oil pressure low, Jacket Water pressure low high vibration engine bearing teaperature hgh lube oli tenp hi, and loss of field, or over-current (186 Block Out)", which provided engine trip during Loss of Offsite Power (IDSP) event, i 178
. 11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) RF20RT This. change was perfonmd by rnoving IDSP ocntact from auto start circuit to emergency circuit, by connecting parallel to S1 contact T-ENG-90-18 will verify that Design Change was properly inplemented by injecting 1DSP signal to start the engine and operate in emergency mde, each with all the nonnal engine trips (bypass) bypassed and engine trips frcxn nonnal trip signal (bypassable) after reset of IDCA.
- 1. The test is to verify the result of the design change which was approved by NRC. The test is merely proving that diesel engine will not trip during loss of Off Site Power (IDSP) frcxn a bypassable trip just like during Sl. All the junpers installed are independently verified during installation and removal. All the o>erations are within the bounds of FSAR and cnapter 15 of the FSAR accident analysis and controlled under procedural restrictions.
Therefore, this test will not increase the probability of occurrence or consequence of an accident or m lfunction of equipe nt or ccxnponent important to safety previously evaluated in the safety analysis report.
- 2. This test is the functional test to prove this implementation of the design change DCP 90-VIN 0137.
This would be similar to-a m nthly surveillance test, except the signal is initiated by connecting a jumper in the sequencer manel. These jumpers are removed after initiating tTe signal. Both installation and removal are independently verified. The test will be conducted in a controlled m nner and could be terminated at any time without any adverse effect to systems or cmponent. Operation of the Diesel Generator is similiar in function to other tests conducted to prove operability. Therefore, this test does not create the possibility of an' accident or mlfunction of a different type than previously evaluated, i 179
11 1990 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3. This test is performed on a Diesel Generator that is taken to inoperable status to perfonn the functional test for DCP 90-VIN 0137. This testing will prove the Diesel Generator would be available during LOSP condition when there is an actual or spurious trip frm bypassable trip signals. Margin of safety is enhanced by increasing the availability of the Diesel Generator. There are alarms in both the control room and local panel to alert operators in an event any bypassed trip parameters exceed the alann setpoint which below the trip point. Therefore, this test will not reduce the margin of safety as defined in the bases of Technical Specifications.
T-ENG-90-20 This procedure provides instructions to treasure the response tine of the Control Roan Ventilation Systun in the Energency Fbde. The response timing will start when the slave relay K616 (Safety Injection) is energized. When this relay is energized, this will de-energize the Diesel Generator Unit / Parallel sequencer contact. This contact takes the Diesel Generator out of the Parallel node and puts the Diesel Generator into the the Unit node. The Unit node is the nonini node during an utergency situation. While in the Unit node, the Diesel should not be nanually started and/or loaded to the grid for surveillance testing. If an LOSP or a valid S1 signal is present, the perfonnance of this procedure will not prohibit the normal emergency response of the system. In this procedure, the ESF fans are performing their design function when a differential aressure of 1/8" to the adjoining areas is reacaed. A review of the data collected from procedure 54054-1 shows that a differential pressure of 0.25" to atnes,here will meet the 1/8" differential pressure to tae surrounding areas. 180
i q 11 1 l 1990 ANNUAL RENRT - PART 2 10 CFR50.59(b) RENRT
- 1. After reviewing FSAR sections 9.4.1 and 6.4.2, it j is detennined that this procedure does not increase j the probability of occurrence or consequences of a j malfunction of safety-related equipnent previously i evaluated in the FSAR. Le Diesel Generator should :l
-not be mnually started and/or loaded to the grid ;
for surveillance testing when this procedure is- j being performed, j
/
- 2. The procedure is written with the intent to operate - _i the ESF Control Room Ventilation well within its !
-design limitations.- After reviewing section 9.4.1 .I and 6.4.2, it is determined that.this procedure
-does not create the possibility for an accident or equipment malfunction of a different type other than i any evaluated previously in the FSAR.
- 3. After reviewing section 3/4.7.6 of the Technical i Specification bases, it is determined that the
^
procedure doe _s not decrease the margin of safety j defined by the Technical Specification bases. 1 T-ENG-90-023 mis procedure will test the modified Annunciator-A1309 (DCP-89-V2N0095) to ensure proper operation. L e Self-Test mode as well as individual windows. will be shown to:be operational. The First Out function will be shown to be per design. Le Design Objective of DCP-95 (Prevent loss of first out data) will-be demonstrated. A1310 will also be tested-because-it shares a logic Chassis with A1309 that is changed out.
- 1. The annunciator systen is electrically isolated
- frcm class lE equipment, herefore, the system cannot increase the probability _of occurrence or-a malfunction in equipment important to_ safety.
II 1990 ANNUAL REPORT - PART 2 10 CIR50.59(b) RENRT information my lead to operator error. However, this test does not change the annunciator inputs (except for jumpers placed and subsequently remved) . Therefore, window displays are not changed and the possibility of operator error due to incorrect window presentation is avoided.
- 3. '1he annunciator system is not used for the basis of any Technical Specification. The margin of safety is not affected by this test.
T-ENG-90-024 This is a special test of the response tino of the sequencer auto /mnual block function in the auto start circuit of the Control Rocm Energency Filtration System (CREFS) fan. It is intended that this test satisfy the requirenents of Technical Specification 4.3.2.2, for Unit 1, Train "A". Die requirement to perform this or a similar test are delineated in Technical Specification 4.3.2.2 which was reviewed and approved by "the NRC. 'Ihe bases for this specification state ... sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance consistent with maintaining an appropriate level of reliability..". D us, the probability of occurrence does not increase. FSAR sections 15.4.6 and 15.5.1 analyze the potential boron dilution and baron injection accidents associated with inadvertent operation of the CVCS and SIS respectively. Both of these potential accidents are analyzed for a spectrum of initial power conditions. Thus the consequences of an accident are not increased.
- 1. The electrical isolation of equipment (SSPS and DG) or placement in an operating mode which will not change as a consequence of this test (fans and pumps) results in a configuration consistent i
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l 11 1990 AN!UAL REPORT - PART 2 10 CFR50.59(b) RER Rf which the design bases of the equipment. The test itself will sinulate the occurrence of a single failure of the lead-lag logic for which the systan is designed. The out-of-service time to perform this test will be less than the limitations of the Technical Specifications as described above. Thus the probability of occurrence of a malfunction is not increased.
- 2. The switches temporarily installed to perform this test are rated consistent with the design of the ESFAS, and will be renoved at the canpletion of the test. No equiptent will be required to operate outside of its design paraneters. The isolation of the SSPS and DG is equivalent to the single failure of the train, should normal power be lost, for which the system is also designed. /c described in FSAR 15.6.5, three trains running has been analyzed for potential impact on Post-li)CA control roam doses and found to be acceptable.
Thus, no new failure nodes will be introduced.
- 3. The test will be performed within the tinu limitation of the AcrION statements for 3/4.3.2 and 3/4.8.1. Furthenmre, the bases of 3/4.3.2 recognize the requirements to perfonn periodic testing of the ESFAS and state that sufficient redundancy is maintained to pennit a channel to be out-of-service for this testing. Therefore, the margin of safety defined by the bases of the Technical Specifications will not be decreased.
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111 GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 AND UNIT 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 EMERGENCY CORE COOLING SYSTEMS OUTAGE DATA REPORT l
III VOGTLE ELECTRIC GENERATING PLANT - UNITS 1 &2 1990 ANNUAL REPORT - PART 2 EMERGENCY CORE COOLING BYSTEM OUTAGE DATA REPORT This report contains: a) outage dates and duration of outages b) ECCS systems or components involved h the outage c) cause of the outage, and d) corrective actions taken
- - UNIT 1 -
Unit 1 Emergency Core Cooling System components were out of service a total of 232 hours and 16 minutes in 1990.
- 1. a) 1-1-90 1 hour and 31 minutes b) & c) Train B RHR removed from service for inservice testing.
d) Testing completed and Train B restored to service.
- 2. a) 1-1-90 53 minutes b) & c) Train A RHR removed from service for inservice testing.
d) Testing completed and Train A restored to service.
- 3. a) 1-7-90 24 minutes b) & c) Safety Injection Train B removed from service for valve stroke testing, d) Testing completed and Train B restored to service.
- 4. a) 1-18-90 24 minutes b) & c) Train A Centrifugal Charging Pump removed from service for inservice testing.
d) Testing completed and pump restored to service.
- 5. a) 1-18-90 1 minute b) & c) Train B Centrifugal Charging Pump removed from service for inservice testing, d) Testing completed and pump restored to service.
- 6. a) 2-2-90 20 minutes b) & c) Train A RHR removed from service for inservice testing, d) Testing completed and Train A restored to service.
- 7. a) 2-2-90 53 minutes b) & c) Train B RHR removed from service for inservice testing.
d) Testing completed and Train B restored to service. )
III 1990 ANNUAL REPORT - PART 2 ECC8 OUTAGE DATA REPORT UNIT 1 (CONTINUED)
- 8. a) 2-15-90 35 hours and 42 minutec b) & c) Train A RHR pump removed from service to install design change.
d) Design change installed and pump returned to service.
- 9. a) 2-21-90 17 hours and 43 minutes b) & c) Train B RHR pump removed from service to install design change.
d) Deuign change installed and pump returned to service.
- 10. a) 3-20-90 40 minutes b) & c) Train A RHR rendered inoperable due to loss of power.
d) Power restored and Train A returned to service.
- 11. a) 4-14-90 13 minutes b) & c) Train A Safety Injection removed from service for surveillance testing, d) Testing complete? and Train A restored to service.
.2. a) 4-22-90 3 hours and 6 minutes b) & c) RHR Train B removed from service far preventive maintenance on a miniflow switch.
d) PM completed and Train B returned to service.
- 13. a) 5-13-90 14 minutes b) & c) RHR Train A pump placed in " Pull-to-Lock" for inservice testing.
d) Testing completed and Train A restored to service
- 14. a) 5-22-90 11 hours and 17 minutes b) & c) RHR Train B removed from service for pump preventive maintenance.
d) PM completed and Train B returned to service.
- 15. a) 5-25-90 2 hours and 37 minutes b) & c) RHR Train B removed from service for inservice and response time testing.
d) Testing completed and Train B restored to service
- 16. a) 6-1-90 44 hours and 11 minutes b) & c) Train A RHR, Safety Injection, and Centrifugal Charging Pumps removed from service while NSCW Train A out of service for valve repair.
d) Valve repaired and systems restored to service. _2
1 l. III 1990 ANNUAL REPORT - PART 2 ECC8 OUTAGE DATA REPORT UNIT 1 (CONTINUED)
- 17. a) 6-21-90 1 minute b) & c) Centrifugal Charging Pump Train B removed from service for troubleshooting valvo problem.
d) Troubleshooting completed and Train 3 restored to service.
- 18. a) 6-22-90 16 minutes b) & c) RHR Train B removed from service for surveillance testing.
d) Testing completed and Train B restored to service.
- 19. a) 7-3-90 2 hours and 10 minutes l b) & c) Train A RHR miniflow valve removed from service for preventive maintenance.
d) PM completed and valve returned to service.
- 20. a) 7-11-90 40 hours and 2 minutes b) & c) Train A RHR, Safety Injection and Centrifugal Charging pumps removed from service while NSCW Train A out of eervice for snubber work.
d) Work completed and systems restored to service.
- 21. a) 7-22-90 54 minutes b) & c) Refueling Water Storage Tank suction valve removed from service to tighten packing.
d) Work completed and valve returned to service.
- 22. a) 8-9-90 18 hours and 41 minutes b) & c) Train A RHR removed from service to rework pipe support struts, d) Work completed and Train A roscored to service.
- 23. a) 8-18-90 6 hours and 19 minutes b) & c) Train A ECCS Subsystem removed from service for channel calibration and valve stroking.
d) Calibration and valve stroking completed and subsystem returned to service.
- 24. a) 9-14-90 6 minutes b) & c) Train B RHR pump taken to " Pull-to-Lock" for inservice testing.
d) Testing completed and pump returned to service. l
III 1990 ANNUAL REPORT - PART 2 ECCS OUTAGE DATE REPORT UNIT 1 (CONTINUED)
- 25. a) 9-16-90 1 hour and 29 minutes b) & c) Train B Containment spray system removed from service for valve stroking, d) Valve stroking completed and system restored to service.
- 26. a) 11-7-90 1 hour and 26 minutes b) & c) Train A RHR removed from service for valvo leak testing, d) Testing completed and Train A returned to service.
- 27. a) 11-7-90 2 hours and 18 minutes b) & c) Train B RHR removed from service for valvo leak testing, d) Testing completed and Train B returned to service.
- 28. a) 11-14-90 40 hours and 49 minutes b) & c) Train B RHR, Safety Injection and Centrifugal Charging pumps removed from service while NSCW Train B out of service for valve repair.
d) Valve repaired and systems restored to service.
- 29. a) 12-7-90 39 minutes b) & c) Train B RHR removed from service for valvo leak testing, d) Testing completed and Train R restored to service.
III 1990 AWNUAL REPORT - PART 2 ECCS OUTAGE DATA RElQBT
- UNIT 2 -
Unit 2 Emergency Core Cooling System components worn out of service a total of 61 hours and 15 minutos in 1990.
- 1. a) 1-12-90 37 minutes b) & c) Train A Contrifugal Charging Pump removod from service for inservice testing.
d) Testing completed and pump restored to service.
- 2. a) 1-12-90 3 minutos b) & c) Train A contrifugal Charging Pump removed from service for incorvice testing, d) Teating completed and pump rostored to service.
- 3. a) 2-2-90 2 hours and 56 minutos b) & c) Train A ECCS equipment taken out of service whilo cable tray support brackets loosened.
d) Brackots retightened and torquod and Train A restored to service.
- 4. a) 2-27-90 3 hours and 48 minutos b) & c) Train A RHR removed from service for preventivo maintenance on heat enchangor outlet valvo.
d) PM completed and Train A restored to servico.
- 5. a) 3-18-90 3 minutos b) & c) Contrifugal Charging Pump Train A removed from service for valvo stroking.
d) Valvo stroking completed and Train A rostored to service.
- 6. a) 3-30-90 6 minutes b) & c) Trhin B RHR pump placed in " Pull-to-Lock" for inservice testing, d) Testing completed and pump returned to servico.
- 7. a) 5-3-90 26 minutes b) & c) Accumulator /4 removed from servico due to low pressure.
d) Prossure restored, accumulator returned to service.
- 8. a) 5-13-90 28 minutos b) & c) Train A RHR pump placed in " Pull-to-Lock" for inservice valve testing.
d) Testing completed and pump returned to service. l
1
)
III 1990 ANNUAL REPORT - PART 2 ECC5_ OUTAGE DATA. REPORT 1 i UNIT 2 (CONTINUED) i-
- 9. a) 5-14-90 2 minutes b) & c) Train A RHR removed from service for stroking of heat exchanger bypass valve following packing adjustment. j d) Valve stroking completed and Train A restored to i service. i
- 10. a) 6-19-90 4-hours and 8 minutes b) & c) Train B RHR removed _from service for stroking of heat exchanger outlet isolation valve following I preventive maintenance.
i d) Valve stroking completed and Train B restored to-service.
- 11. a) 6-25-90 3 hours and 51 minutes b) & c) Train B RHR removed from service for preventive maintenance on the containment sump suction isolation valve breaker, d) PM completed and Train _B restored to service.
- 12. a) 7-12 17 hours and 58 minutes
- b) & c) Train A Safety Injection Pump removed from service }
for preventive maintenance to supply breaker.
-d) PM completed and pump returned to-service.
- 13. a) 5-90 19. minutes b) & c) Train A RHR pump placed in " Pull-to-Lock" for valve stroking.
d) Valve'stroki.g completed and pump returned to service.
.14. a) 8-16-90 1 hour and 7 minutes
, b) & c) Train B RHR; removed _from service for inservice testing. d) Testing completed andLTrain B restored to service.
- 15. a). 8-16-90 40 minutes .
Train A RHR removed from service for inservice L b)_ & c)
' testing.
n d) -Valve stroking completed and Train A restored to service. l-
- 16. a) 9-5-90 5 hours and 33 minutes b) & c) Train A RHR removed from scrylce for valve stroke testing.
L d) Testing completed and Train A restored to service. ; P ' u.._...__.. . . _ . _ _ ____ _ . _ _ _ . _ _ _ _ _ . _ _ , __ _._._
e , III 1990 ANNUAL REPORT - PART 2 ECCS OUTAGE DATA _. REPORT UNIT 2 (CONTINUED)
- 17. a) 9-15-90 21 minutos b) & c) Train B RHR pump removed from service to test pump i motor current indication. l d) Testing completed and Train B rostored to service.
- 18. a) 9-29-90 35 minutos '
b) & c) Train A Safety Injection pump removed from service for testing, d) Testing completed and pump returned to service.
- 19. a) 11-6-90 2 hours and 1 minuto b) & c) Train A RHR removed from service for valvo leak I testing. _
l d) Testing completed and Train A returned to ' service.
- 20. a) 11-6-90 1 hour and 3 minutos b) & c) Train B RHR removed from service for valvo leak testing.
d) Testing completed and Train B returned to servico.
- 21. a) 11-7-90 11 minutos b) & c) Train A Safety Injection Pump removed from service for valve leak testing.
d) Testing completed and pump returned to service.
- 22. a) 11-7-90 21 minutos b) & c) Train B Safety Injection Pump removed from service for valvo leak testing, d) Testing completed and pump returned to rervice.
- 23. a) 11-7-90 8 minutos b) & c) Train A RHR removed from servico for valvo leak testing.
d) Testing completed and Train A returned to servico. l 24. a) _11-7-90 6 minutos b) & c) Train B RHR removed from servico for valvo leak testing. d) Testing completed and Train B returned to service. l
e i III 1990 ANNUAL REPORT - PART 2 4 ECC8 OUTAGE DATA REP 02T UNIT 2 (CONTINUED) i
- 25. a) 11-8-90 4 minutes )
b) & c) Train B Centrifugal charging pump removed from service for response time testing. 1' d) Testing completed and pump restored to service.
- 26. a) 11-8-90 24 minutes b)__ & c) Train A Safety Injection' Pump removed from servico i
for valve leak testing. d) Testing completed and pump restored to servico., ,
- 27. a) 11-9-90 1 hour and 9 minutes ,
b)-& c) Centrifugal Charging Pump discharge throttio -l injection needle valve unlocked for maintenance. d) Maintenance completed and valve relocked.
- 28. a) 11-25-90 1 hour and 34 minuteo b) & c) Safety Injection flowpath valves removed from service for stroke time testing.
d) Testing completed and valves rostored to servico. ;
- 29. a) 11-30-90 4 hours and 46 minutes b).& c) Train A RHR removed from service for flow switch calibration, j d) Calibration completed and Train A returned to I service. .i F 30. a) 12-5-90 8 minutes b) & c) Accumulator removed from service due to low pressure.
d). Pressure restored and accumulator returned to
; service.
i 31. a)- 12-7-90 10 minutes , b) . & c) Train B RHR pump taken to " Pull-to-Lock" for valvo i' testing. d) Testing _ completed, pump restored to service. 3 2. - a) 12-9-90 33 minutes b)' & c) LTrain B Safety Injection removed from service for valve testing. p d)- Testing completed and Train B rostored-to servido. 4
IV GEORGIA POWER COMPANY , V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 AND UNIT 2 NRC DOCKET NOS. 50-424 AND 50-425 l FACILITY OPERATING LICENSE N05. NPT-68 AND NPF-81 1 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT l CALENDAR YEAR 1990 l i
~~
! V0GTLE ELECTRIC GENERATING PLANT RADIOLOGICAL ENVIRONMENTAL SVRVilLLANCE REPORT TABLE OF CONTENTS SECTION TITLE EAEL
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . , . . .11 2.0
SUMMARY
DESCRIPTION ,, ............... 21 3.0 RESULTS
SUMMARY
. . . . . . . . . . . . . . . . . . . . 3-1 4.0 DISCUSS 10R OF RESULTS ................. 41 4.1 Airborne . . . . . . . . . . . . . . . . . . . . . . 4-4 4.2 Direct Radiation .................. 46 4.3 Milk ........................ 48 4.4 Vegetation .....................410 4.5 River Water . . . . . . . . . . . . . . . . . . . . . 4 11 4.6 Orinking Water . . . . . . . . . . . . . . . . . . . . 4-13 4.7 Fish . . . . . . . . . . . . . . . . . . . . . . . . 4-15 4.8 Sediment ........ .............417 5.0 INTERLABORATORY COMPARISON PROGRAM . . . . . . . . . . . 51
6.0 CONCLUSION
S ...................... 61 i
I LIST Of TABLES IMLL lilLE f%nE 21
SUMMARY
DESCRIPTION Of RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 22 22 RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS 2-7 31 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
3-2 41 LAND USE CENSUS RESVLTS 43 51 CROSSCHECK PROGRAM RESULTS 5'c 11 l _ - _ . . . . . _ _ _ . . _ . . _ . _ . _ _ _ _ _ _ _ _ . _ _ . _ _ _ . _ . _ _ - . _ _ . - , _ _ _ _ , ~ _ . . . . _ .
( LIST Of flGURES 11GliRE IllLL PAGI 2-1 TERRESTRIAL STATIONS NEAR SITE BOUNDARY 2 10 2-2 TERRESTRIAL STATIONS BEYOND SITE BOUNDARY OUT TO APPROXIMATELY SIX MILES AND RIVER STATIONS 2 11 2-3 TERRESTRIAL STATIONS BEYOND SIX MILES 2-12 2-4 DRINKING WATER STATIONS 2 13 iii
ACRONYMS CL Confidence level CY Calendar Year El Environmental Laboratory EPA Environmental Protection Agency GPC Georgia Power Company LLD Lower Limit of Detection MDD Minimum Detectable Difference MDA Minimum Detectable Activity 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 RM River Mile SRS Savannah River Site TLD Thermoluminescent Dosimeter . TS Technical Specifications VEGP Alvin W. Vogtle Electric Generating Plant iv
m . . . _ -- i V0GTLE ELECTRIC GENERATING PLANT RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT
1.0 INTRODUCTION
This is the fourth 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) 1990. Hence all dates in this report are for 1990 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 concentrations 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 comparison 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 impicmented. 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 distance 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 sampics 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
2.0
SUMMARY
DESCRIPTION A summary description of the REMP is provided in Table 21. This table portrays the program in the manner by which it is being regularly carried out; it is essentially a copy of Table 3.121 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 through 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 (0DCM). It is stated in Footnote (1) of Table 3.121 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 particular sample type in Section 4. For CY 90, all the laboratory analyses except for the reading of the - thermeluminescent dosimeters (TLDs) were performed by Georgia Power Company's (GPC's) Environmental Laboratory (EL) in Smyrna, Georgia. The reading of the TLDs was provided by Teledyne isotopes Midwest Laboratory in Northbrook, Illinois. 2-1
TABLE 2-1 (SHEET I 0F 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING AND TYPE AND FREQUENCY EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES OF ANALYSIS AND SAMPLE LOCATIONS COLLECTION FREQUENCY AND/0R SAMPLE Quarterly Gamma dose quarterly
- 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 stations, one in each meteorological sector in the 6 mile range from the site; s, A, and Special interest areas such as population centers, nearby residences, schools and control stations. I
TABLE 2-1 (SHEET 2 0F 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE AND SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS
- 2. Airborne Radioiodine and h oples from seven locations Continuous sampler oper- Radioiodine Cannister:
Particulates ation with sample collec- I-131 analysis weekly tion weekly, or more five locations close to the frequently if required by site boundary in different dust loading Particulate Sampler: sectors; Gross beta analysis (1) i following filter change and m gamma isotopic analysis (2) i L, of composite (by location) ! A community having the highest ! calculated annual average ground-j level D/Q; and
- A control location in the vicinity
, of a population center at a distance of about 15 miles.
TABLE 2-1 (SHEET 3 0F S)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING AND TYPE AND FREQUENCY EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES OF ANALYSIS AND SAMPLE LOCATIONS COLLECTION FREQUENCY AND/0R SAMPLE
- 3. Waterborne Composite sample over Gama isotopic analysis (2)
- a. Surface (3) One sample upriver monthly. Composite for 1-month period (4) tritium analysis quarterly Two samples downriver Composite sample of I-131 analysis on each
- b. Drinking Two samples at each of the two sample when the dose nearest water treatment plants river water near the that could be affected by plant intake at each water calculated for the treatment plant over consumption of the water is discharges 2-week period (4) when greater than 1 mrem per I-13I analysis is year (5). Composite for Two samples at a control gross beta and gamma m
location required to be performed i on each sample, monthly isotopic analyses (2) on raw c0mposite otherwise; and water monthly. Gross beta, grab sample of finished gama isotopic and I-13I water at each water analyses on grab sample of treatment plant every 2 finished water monthly. weeks or monthly, as Composite for tritium appropriate analysis on raw and finished water quarterly Semiannually Gamma isotopic analysis (2)
- c. Sediment from One sample from downriver semiannually Shoreline area with existing or potential recreational value One sample from upriver area with existing or potential recreational value
1. TABLE 2-1 (SHEET 4 0F 5) StPEARY DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i EXPOSURE PATHWAY. NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE AND SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS
- 4. Ingestion
- a. Milk Two samples from milking animals (6) Biweekly at control locations at a distance Gama isotop)ic analysis (2,7 biweekly of about 10 miles or more
- b. Fish At least one sample of any co:nmer- Semiannually Gama isotopic analysis (2) cially or recreationally important on edible portions species in vicinity of plant semiannually discharge area
- At least one sample of any a comercially or recreationally l important species in an area not
- influenced by plant discharge i
At least one sample of any Curing spring spawning Gama isotopic analysis (2) anadromous species in vicinity of season on edible portions annually plant discharge
- c. Grass or Leafy One sample from two onsite locations Monthly during growing Vegetation near the site boundary in different season Gama isotop)ic analysis (2,7 monthly l
sectors One sample from a control location at about 18 miles distance I i I
TABLE 21 (SHEET 5 Of 5) i
SUMMARY
DESCRIPTION Of RADIOLOGICAL ENVIRONMENIAL MON 110 RING 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 particulatt samples is greater than 10 times the yearly metn of control samples, gamma isotopic analysis shall be performed on the individual samples. (2) Ganna 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 downriver 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 00CM. (6) A milking animal is a cow or goat pro 6ucing milk for human consumption. (7) If gamma isotopic analysis is not sensitive enough to meet the lower Limit of Detection (LLD) for 1 131, a separate analysis for 1 131 will be performed. 2-6 i 1
TABLE 2 2 (SHEET 1 Of 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LCCATIONS Station Station Descriptive Direction (2) Distance (2) Sample Number lypp_(11 Location- (miles) Iyp1_QJ _1 1 Hancock Landing Road N 1.1 D 2 1 River Bank NNE 0.8 0 3 1 Discharge Area NE 0.6 A 3 I River Bank NE 0.7 0 4 1 River Bank ENE 0.B D 5 1 River Bank E 1.0 D 6 I Plant Wilson ESE 1.1 0 7 I Simulator Building SE 1.7 D,V,A 8 1 River Road SSE 1.2 D 9 I River Road S 1.1 D 10 1 Met Tower SSW 0.8 A 10 1 River Road SSW l.1 D 11 1 River Road SW l.2 0 12 I River Road WSW l.2 D,A 13 I River Road W l.3 0 14 I River Road WNW l.8 0 15 i Hancock Landing Road NW l.5 0,V 16 I Hancock Landing Road NNW l.4 D,A 17 0 Savannah River Site (SRS) River Road N 5.5 0 18 0 SR$ D Area NNE 5.1 0 19 0 SRS Road A.13 NE 4.7 D 20 0 SRS Road A.13.1 ENE 4.8 D 21 0 SRS Road A.17 E 5.6 0 22 0 River Bank Downstream of Buxton Landing ESE 5.2 0 23 0 River Road SE 4.6 0 24 0 Chance Road SSE 4.9 0 25 0 Chance Road near Highway 23 S 5.2 0 26 0 Highway 23, Mile 15.5 SSW 4.6 0 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 0 30 0 Nathaniel Howard Road WNW 5.0 0 31 0 River Road at Allen's Church Fork NW 5.0 0 32 0 River Bank NNW 4.8 0 33 0 Hunting Cabin SE 3.3 0 27
TABLE 2 2 (SHEET 2 0F 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station Descriptive Direction (2) Distance (2) Sample Number I_voe (11 Location (miles) lype (3) 35 0 Girard SSE 6.6 D,A 36 C Waynesboro WSW 14.9 D,A 37 C Substation (Waynesboro) WSW 17.5 0,V 43 0 Employees Recreation Area SW 2.2 0 47 C Oak Grove Church SE 10.4 D 48 C McBean Cemetery 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 1 Savannah River RM150.4) ENE 0.8 R,5(6) 84 0 Savannah River RM 149.5) ESE 1.6 R 85 i Savannah River ESE 4.3 f(5) 87 1 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 H 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 0 - Direct Radiation F Fish M - Milk R - River Water S - River Shoreline Sediment W - Drinking Water V - Vegetation 2-8
TABLE 2 2 (SHEET 3 Of 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCA110NS 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 Mile (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 become 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-Jas)er County Water Treatment Plant is located at the end of a canal w11ch 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. l 2-9 i
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l l l 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 designated indicator and control stations are presented in Table 31 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 control stations or in addition to those stipulated by l Table 21 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. 31 i
TABLE 3-1 (SHEET I 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Vogtle Electric Generating Plant, Docket Nes. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1990 Location with Highest Controi Locations Number of Medium or Type and Lower Limit All Indicator Mean (b) Nonroutine of Locations Annual Mean Pathway Sampled Total Number Name Mr.n (b) Range Reported (Unit of of Analyses Detection (a) Wean (b) Range (Fraction) Measurements Performed (LLD) Range Distance & Measurement) (Fraction) Direction (Fraction) 20.3 19.4 0 Gross Beta 10 19.5 No. 12 Airborne 7-43 River Road 8-38 7-41 Particulates 316 1.2 miles (52/52) (53/53) (263/263) (fCi 'm3) WSW Gamma Isotopic 24 NDM NDM 0 Cs-134 50 NDM (c) w NDM 0 NDM E Cs-137 60 NDM hDM t.I)M 0 I-131 70 NDM Airborne Radiciodine 316 (fCi/m3) 21.0 16.6 0 Gamma Dose NA (d) 16.9 No. 3 Direct 13-23 River Bank 19-23 11-22 Radiation 79 0.7 miles (4/4) (15/15) (mR/91 days) (64/64) NE l
TABLE 3-1 (SHEET 2 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGPJJi ANNUAL
SUMMARY
Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1990 Location with Highest Control Locations Number of Medium or Type and Lower Limit All Indicator Nonroutine of Locations Annual Mean Mean (b) Pathway Sampled Total Number Name Mean (b) Range Reported (Unit of of Analyses Detection (a) Mean (b) Measurements Performed (LLD) Range Distance & Range (Fraction) Measurement) Direction (Fraction 1 (Fraction) Milk Gamma Isotopic (pCi/1) 56 NDM NDM 0 Cs-134 15 NA 17.0 17.0 0 Cs-137 18 NA No. 99 Boyceland 17-17 17-17 24.5 miles (1/28) (I/56) WNW
- ys w NDM NDM 0 Ba-140 60 NA NDM NDM 0 La-140 15 NA I.82 1.25 0 I-131 1 NA 1.8-1.8 0.7-1.8 56 (1/28) (2/56) l Grass Gamma Isotopic (pCi/kg wet) 36 NDM NDM 0 I-131 60 NDM NDM NDM 0 Cs-134 60 NDM
C
~
TABLE 3-1 (SHEET 3.OF 10) ji RADIOLOGICAL ENVIRONMENTAL-MONITORING PROGRAM ANNUAL SUPMARY i Vogtle Electric Generating Plant, Docket Nos.' 50-424 & 50-425 j Burke County, Geortjia, Calendar Year 1990
' 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 l Measurement) Performed (LLD) Range Distance & Range (Fraction) Measurements (Fraction) Direction (Fraction) l Cs-137. 80 -30 No. 37 102 102 0 25 Substation 44-160 44-160 (2/24) -17.5 miles (2/12) (2/12)
W5W River Water Gamma Isotopic (pCi/1) 36
- j. y Be-7 80-(e) NDM NDM NDM 0 i*
l Mn-54 15 NDM NDM NDM 0 '.l Fe-59 30 NDM NDM NDM 0 Co-58 15 NDM NCM NDM 0 l Co-60 15 NDM NDM NDM 0 l
'Zn-65 30 NDM NDM NDM 0 i
Zr-95 30 NDM NDM NDM 0 i
- Nb-95 15 NDM NDM NDM 0 t
- j. I-131 15 NDM NDM NDM 0 L
4
TABLE 3-I (SHEET 4 0F.10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1990 Location with Highest Control Locations Number of Medium or Type and Lower Limit All Indicator Mean (b) Nonroutine of Locations Annual Mean Pathway Sampled Total Number Name Mean (b) Range Reported of Analyses Detection (a) Mean (b) (Unit of Range Distance & Range (Fraction) Measurements Measurement) Performed (LLD) (Fraction) (Fraction) Direction NDM NDM 0 Cs-134 15 NDM NDM NDM 0 Cs-137 18 NDM NDM NDM 0 Ba-140 60 NDM NDM NDM 0 La-140 15 NDM l 1142 392 0 Tritium 3000 1142 No. 83 y' 610-1550 Downriver 610-1550 284-578 cn 8 (4/4) (4/4) 0.4 miles (4/4) 2.57 2.55 0 Gross Beta 4 2.53 No. 87 Water Near Beaufort 1.2-3.8 1.2-3.6 Intakes to Water 36 1.2-4.4 (24/24) Downriver (12/12) (10/12) Treatment Plants 112 miles (pCi/1) Gamma Isotopic 36 NDM ADM 0 Be-7 80 (e) NDM NDM NDn 0 Mn-54 15 NDM NDM NDM 0 Fe-59 30 NDM
- TABLE 3-1 (SHEET 5 0F 10)-
RADIOLOGICAL ENVIRONMENTAL MONITORING.' PROGRAM ANNUAL SLMiARY Vogtle Electric Generating Plant, Docket'Nos.. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1990 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 . Measurement). Performed - (1.LD) Range _ . Distance & Range (Fraction) Measurements i (Fraction)- -Direction (Fraction)
'Co-58 15 NDM NDM NDM. 0' Co-60 15 NDM NDM NDM 0' I Zn-65 30 NDM NDM NDM 0 ;
Zr-95 30 NDM NDM NDM 0-Nb-95 15 NDM NDM NDM 0
'l-131.(f) 15 NDM NDM . NDM 0 ,
'Cs-134 15 NDM NDM NDM 0 ;
i Cs-137 18 NDM NDM NDM 0 - Ba-140 60 NDM NDM NDM 0
^
'La-140 15 NDM NDM NDM 0 i
Tritium- 3000 1320 No. 88 1423 266 0 12- 803-1910 Port Went 1130-1910 156-378 (8/8) Downriver (4/4) (4/4) 122 miles i
TABLE 3-1 (SHEET 6 0F 10): RADIOLOGICAL: ENVIRONMENTAL MONITORING PROGRAM ANNUAL:SIMIARY Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar. Year 1990 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 ,
- ('Jnit of. of.. Analyses Detection (a) .Mean (b) Name. Mean (b) Range Reported i Measurement) Performed (LLD) Range- Distance & Range (Fraction) Measurements i i (Fraction) Direction (Fraction)
Finished Water at Gross Beta 4- 2.08 No. 87 2.18 1.92 0 Water Treatment 36 1.2-3.6 Beaufort 1.5-3.6' O.9-3.1 Plants (24/24) Downriver (12/12) (11/12) (pCi/1) '112 miles i - Gamma Isotopic t' 36 jy Be-7 80 (e) NDM NDM NDM 0
- N Mn-54 15 NDM NDM NDM 0 Fe-59 30 NDM NDM NDM 0 t Co-58 15 NDM NDM NDM 0 i 60 15 NDM NDM NDM 0. [
r Zn-65 30 NDM NDM NDM 0 [ l Zr-95 30 NDM NDM NDM 0 Nb-95 15 0 ( NDM NDM NDM ; i l Cs-134 15 NDM NDM NDM 0 [
y . _ . TABLE 3-1'(SHEET 7 0F 10)' RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1990 Medium or Type and Lower Limit All Indicator Location with Highest- Control Locations Number of Pathway Sampled Total Number of Lccations Annual Mean Mean-(b) Nonroutine (Uisit 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 18 NDM NDM NDM 0 Ba-140 69 NDM NDM NDM 0 La-140 15 NDM NDM NDM 0 I-131 36 1 NDM NDM NDM 0 y Tritium 2000 1299 No. 88 1460 404 0 oo 12 689-2330 Port Went 979-2330 236-653 (7/8) Downriver (3/4) (3/4) 122 miles Anadromous Fish Gamma Isotopic (pCi/kg wet) 1 Be-7 100 (e) NDM NDM NA 0 Mn-54 130 NDM NDM NA 0 Fe-59 260 NDM NDM NA 0 Co-58 130 NDM NDM NA 0 Co-60 130 NDM NDM NA 0 Zn-65 260 NDM NDM NA 0
TABLE 3-1 (SHEET 8 0F 10) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1990 Location with Highest Control Locations Number of Medium or Type and Lower Limit All Indicator Mean (b) Nonroutine of Locations Annual Mean Pathway Sampled Total Number Name Mean (b) Range Reported of Analyses Detection (a) Mean (b) (Unit of Range Distance & Range (Fraction) Measurements Measurement) Performed (LLD) (Fraction) (Fraction) Direction NDM NA 0 Cs-134 130 NDM NDM NA 0 Cs-137 150 NDM Fish Gamma Isotopic (pCi/kg wet) 11 NDM NDM 0 Be-7 100 (e) NDM
't' NDM NDM 0 ") Mn-54 130 NDM NDM NDM 0 Fe-59 260 NDM NDM NDM 0 Co-58 130 NDM NDM NDM 0 Co-60 130 NDM NDM NDM 0 Zn-65 260 NDM 12 0 I-131 15 (e) 13 No. 85 13 13-13 Downriver 13-13 12-12 (1/5) 4 miles (1/5) (1/6)
NDM NDM 0 Cs-134 130 NDM 249 249 0 Cs-137 150 103 No. 81 26-370 Upriver 28-1300 28-1300 (5/5) 4.7 miles (6/6) (6/6) f
iTABLE'3-1 (SHEET.9 0F'10) i: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUPMARY Vogtle Electric Generating , Plant, Docket'Nos. 50-424 & 50-425 - Burke County, Georgia, Calendar Year'1990 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-Measurement) Performed (LLD) .. Range Distance & . Range- (Fraction) Measurements I (Fraction) Direction (Fraction) l Sediment Gamma Isotopic (pCi/kg dry) 4 ' l Be 300.(e) 465 - No. 81 545 545 . 0
.210-720 Upriver 300-790 300-790 (2/2) - 2.5 miles- (2/2) (2/2) 00-58 25 (e) 140 No. 83 140 NDM 0 140-140 Downriver 140-140 is .(1/2) 1.3 miles (1/2) l Co-60 40 (e) 46 No. 83 46 NDM 0 46-46 Downriver 46-46 (1/2) 1.3 miles (1/2)
'Cs-134 15v NDM 'NDM NDM 0 Cs-137 180 155 No. 83 155 140 0 110-200 Downriver 110-200 120-160 (2/2) 1.3 miles (2/2) (2/2) i t
l l
._m. .. , = _ _ - . - __m = . = = = , = - . . . - - = > _mm m m m.m _,_m.m_m--;_m..___. 2._m.
_-..,,,,.wj.,_mm
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 1990 TABLE NOTATIONS
- a. Thc LLO 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 found in that table. In practice, the LLDs attained are generally much lower than the values listed,
- b. Mean and range are 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 required to be performed on these semples when the dose calculated from the consumption of water is less than 1 mrem per year.
3-11
_ _ - _ _ _ _ _ - _ _ ___..m - _ . _ _ _ _ _ _ _ ..__.-___m. 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 standard Student's t-test. A difference-in the 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 l
. 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 ex) lain any RLs or other high radiological levels found in the samples. Tiere 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. Unless otherwise ihdicated, any references made in this section to the results of a previous period will be 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, Also unless otherwise indicated, any references to CY 87 will be to the operations portion of 1987. The SRS was.previously called the Savannah River Plant. The annual land use census required by-Section 3/4.-12.2 of the TS was conducted on April 16. The locations of the nearest milk animal, residence 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 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. .
As reported in CY 89, samples of-an aquatic vegetation named Eaeria densa but commonly called water weed were collected in the river upstream 'and downstream of the plant on- a trial basis to determine its
- suitability as an environmental sample to monitor any-radiological impact due to liquid releases. This vine-like densely foliaged plant
-grows underwater at a depth of 3 meters or less and acts somewhat like a 4-1
filter. The results of the gamma isotopic analysis performed on the-water weed samples indicated that it indeed would be a good radiological environmental monitoring sample. Water weed samples were not available - for analysis in CY 90. 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 1 was introduced this year. 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.
- 1. G. D. Chase and J. L. Rabinowitz, Princiole of Radioisotoog Methodoloav (Burgess Publishing Company, 1962) 87-90.
4-2
TABLE 4-1 LAND USE CENSUS RESVLlS l Distance in Miles to Nearest locations in Each Sector l I SICIQB MILK RESIDENCE LEAFY ANIMAL GARDEN l N
- 1.6
- NNE NE * *
- ENE E
ESE SE
- 4.3
- SSE
- 4.0 S
4.3
- SSW
- 4.2
- SW
- 2.8 4.9 WSW
- 1.2 4.6 W 1.9 WNW 1.8 NW
- 2.4 4.5 NNW 1.6
- None within 5 miles and outside of SRS.
4-3
l l 4.1 Airborne As indicated by Tables 2-1 and 2-2, airborne particulates and airborne radiciodine are collected at 5 indicator stations (Nos. 3, 7, 10, 12, and 16) which encircle the site boundary, at a nearby community (No. 35) and at a control station (No. 36). At these locations, air is continuously drawn through a particulate filter and a charcoal canister in sequence to retain airborne particulates and to adsorb airborne radiciodine, 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 1 131 by gamma spectroscopy. Three of the air particulate and two of the airborne radiciodine samples were deemed to be unacceptable. Last year, six of the air particulate and four of the airborne radioiodine samples were found to be unacceptable. The samples collected on January 30, at Station 16 were excluded due to low volume as a consequence of a pump failure; the gross beta result for the air particulate filter failed Chauvenet's Criterion. When collecting the samples at Station 12 on July 30, it was discovered that a storm had blown the cabinet off its base and broken glass had cut the tubing to the rotometer but the pump was still running; there was no way to determine the volume or flow rate. The gross beta result for the air particulate filter collected at Station 35 on November 6, failed Chauvenet's Criterion; an examination of the filter showed an unusual pattern of the deposit which indicated leakage around the filter due to defects in the filter; no reason was found to exclude the I-131 result. As seen in Table 3-1, the average weekly gross beta activity during the year for the indicator stations was 0.2 fC1/m3 greater than that for the control station. However, this difference is not discernable since it is less than the MDD which was calculated as 2.5 fCi/m3, The average weekly gross beta activity in units of fCi/m3 for the indicator, community and control stations during CY 90 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 Communi.ty 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 23.0 23.5 22.3 Preop Overall 22.9 22.1 21.9 Preop Range 18.1-28.1 18.3-26.5 18.3-26.5 4-4
i l The average weekly readings for CY 90 are seen to be a few percentage points greater than that for CY 89 and about 85 percent of that generally found during the previous years of operation 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 and CY 89, no positive results for manmade radionuclides were found during CY 90 from the gamma isotopic analyses of the quarterly composites of the air particulate filters. During CY 87 found in one indicator composite at a level of 1.7 fCi/m$.Cs-137 During was 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 is 50 fCi/m3 1-131 was not detected in any of the charcoal canisters during the year. There were no positive results during 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. The maximum allowed LLD is 70 fCi/m3; however, the LLD usually attained are about 30 percent of this value. The RL is 900 fCi/m3, 4-5
l 4.2 Direct ~ Radiation Direct (external) radiation is measured by TLDs. A TLD badge is placed at :ach 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 stations (Nos. I through 16) is located near the site boundary, while the outer ring (Nos.17 through
- 32) is located at a distance of about 5 miles. The 16. stations forming the inner ring are designated as the indicator stations. Each of the 4 control stations (Nos. 36, 37, 47, and 48) are over 10 miles from the plant. After being used on a trial basis _ for 2 quarters, Stations 47 and 48 were added at the beginning of the year to enhance the statistical base for the control stations. Special interest areas consist of a hunting cabin (No. 33), the Town of Girard (No. 35), and the GPC employees' recreational area (No. 43).
~,1ot infrequently, TLDs are lost due to theft or vandalism. Near-the middle of each quarter, the vast majority (85 percent) of the stations (those readily accessible) are checked for missing or damaged badges; replacement' badges are provided as needed. At the end of the first quarter, it was learned that the badge at Station 28 had been stolen and the badge at Station-48 had been burned in a brush fire. Last year a -
total of 5 badges were lost in the field and a sixth was lost in its shipment to the contract laboratory. As may be seen from Table 3-1, the average quarterly dose of 16.9 mR acquired =at the indicator stations was 0.3 mR greater than that acquired at the control stations; this difference was not discernable, however, since it was less than the MDD of 1.6 mR. The quarterly doses acquired at the outer ring stations ranged from 12,7 to 24.0 mR with an average of 16.3 m_R which-is 0.6 mR'less than that found for the inner ring. There was no discernable difference between the averages for the inner
.and outer rings since-this difference was less than the MDD of 0.9 mR.
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 the TLD stations), and the range of annual averages obtained during the calendar years of preoperation. 4 Period Indicator Control Outer Rina
-CY 90 16.9 16.6 16.3 CY 89 17.9 18.4 17.2 CY 88 16.8 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 4-6
l
.. i Overall,- the. doses for CY 90 were roughly 4 percent less.than those -I found during previous years of operation and nearly-10 percent greater than-those-found during preoperation. No trend is recognized in~these data.
The average levels in units of mR/91 days for-the special interest areas obtained during each year of operation and the entire )eriod of preoperation along with the range of annual averages catained during the ' calendar years of preoperation are listed below. Period Station 33 Station 35 Station 43 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 Preop Range 13.6-19.9 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. . It is noticed that the average level at Station 35 is steady and 20 to 25 percent greater than its average icvel during preoperation. 4-7
4.3 Milk As indicated by Tables 2-1 and 2-2, milk is collected biweekly from two control stations, Dixon Dairy (No. 98) and the Boyceland Dairy (No. 99). Gamma isotopic and 1-131 analyses were performed on each sample. Milk has not been available from an indicator station (a location within 5 miles of the plant) since April 1986 when the 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 cansus. 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 consumption. As usual, the only manmade radionuclide found during CY 90 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/l along with the fraction of detectable measurements during each year of operation as well as during preoperation. Period Averace Minimum Maximum Fraction CY 90 17.0 17.0 17.0 0.018 CY 89 7.0 5.8 7.7 0.056 CY 88 6.9 4.9 t.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 30 percent greater than that during preoperation, the average level was only about 40 percent of that during preoperation. The level of the one positive result found (at Boyceland Dairy) this year is seen to be close to the average level found during preoperation. No trend is recognized from these results. The LLD and RL for Cs-137 in milk, as r9 quired 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 also detected in a sample from Boyceland Dairy. Positive 1-131 results of 1,82 and 0.67 pCi/l were found in the March 13 milk sample from Boyceland Dairy and the October 2 sample from Dixon Dairy, respectively, Positive activity for each of these samples was confirmed by recounting; the Boyceland sample was recounted 5 times while the Dixon sample was recounted only once, in each case, the half life estimation could not confirm or rule out I-131 (8.05 day half life) due to large counting uncertainties. 4-8
< Using a regression analysis, the best fit of the data for the Boyceland sample provides an apparent half life of around 16 days. However, within the 95 percent CL, the half life extended from 7.3 days to infinity. The estimated half life at the 95 percent CL for the Dixon sample based upon the two counts was 3 9 days. Laboratory contamination was ruled out as a cause of the positive 1-131 results when the Boyceland sample was reanalyzed and showed positive results after washing all glassware. In the future, to confirm or rule out 1-131 when positive results are obtained by beta-gamma coincidence counting, an energy spectra will be obtained of the gamma signal to identify interfering Ra-226 peaks. Also, reagent blanks will be processed with each batch, instead of quarterly, to obtain statistical information on reagent blank counting data. During previous years of operation,1-131 was not detected in milk samples. During preoperation, positive 1-131 results were 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, respectively. 4-9
-m___.--
I 4.4 -Vegetation The TS call for the gamma isotopic analysis of grass or leafy vegetation collected monthly from two onsite-locations near the site boundary in different meteorological sectors (Stations 7 and 15) and one control location at about 15 or more miles from the plant (Station 37). Grass is collected at each of these locations.- No gardens were found in the land use census where the calculated dose commitment would be 20% greater than that of either of the indicator stations at which vegetation is being sampled. As indicated in Table 3-1, Cs-137 was the only manmade radionuclide i detected. The average level at the control station is seen to be 72 l pCi/kg wet greater than that at the indicator station. This difference is not discernable, however, since it is less than the MDD of 405 pCi/kg ! wet. ! Except for a short-period following the Chernoby1 incident, Cs-137 has been the only manmade radionuclide detected in vegetation samples by gamma isotopic analysis during both the preoperation and operation periods. As a consequence of the Chernoby1 incident: 1-131 was found in nearly all the samples collected over a period of several weeks, some at elevated levels; Cs 137 was also found in nearly all of the samples; and Co-60 was found in one of the samples. The average level of Cs-137 found in vegetation samples in units of pCi/kg wet along with 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 Averaae Fraction Averaae fractions 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-
-Proop 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.
L 4-10 i
4.5 River Water Surface water is composited from the Savannah River at three locations using 1S00 automatic samplers. Small quantities of river water are collected at intervals not exceeding a few hours. River water collected by these machines is picked up monthly; quarterly composites are made up from the monthly collections. The collection points consist 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 90. A tritium analysis was performed on each quarterly composite. As usual, a positive result was obtained from each analysis. As indicated in Table 3-1, the average level of 1142 pCi/l found at the indicator station is 750 pCi/l greater than that at the control station; this difference is not discernable because it is less than the MDD of 766 pCi/1, There was a discernable difference in the tritium level between these two stations in CY 88 and CY 89. At the special station (No. 84), the results ranged from 620 to 1700 pCi/1 with an average of 1081 pCi/1. The LLD is 3000 pCi/1 and the 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 (Li-Lc), the MOD between these two stations and the annual liquid releases of tritium from the plant. All of these values are in units of pCi/l except for the releases which are in units of Ci. Item CY 90 CY 89 CY 88 CY 87 Control Station 392 538 427 524 Indicator Station 1142 1293 843 680 Special Station 1081 1268 1430 1411 Li-Lc 750 755 416 156 MDD 766 518 271 416 Releases 1172 916 390 321 These data show an upward trend for plant releases and some correlation between (Li-Lc) and plant releases. The releases are sufficient to account for the increased levels of tritium at 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 750 pCi/l was conservatively calculated to be 0.078 mrem or 0.78 percent of the TS limit. 4-11 1
On October 12 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 only parties found to be withdrawing river water for drinking purposes were the two downriver water treatment plants (Stations 67 and
- 88) from which samples are collected monthly. As in all previous surveys, no intakes for irrigation use were observed. The survey results were corroborated by contacting the Environmental Protection Divisien of the Georgia Department of Natural Resources and the South Carolina Department of Health and Environmental Control; it was found that no new surface or drinking water withdrawal permits had been issued this year for the Savannah River downstream of the plant.
4-12
4.6 Drinking Water Samples were collected at a control 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 Hill Water Treatment Plant near Port Wentworth, Georgia, which are respectively located about 112 and 122 mt?es downriver. These upriver and downriver distances 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 plants. 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 activity for raw drinking water was 0.02 pCi/l greater for the control station than for the indicator stations. However, this difference was not discernable because it was less than the MDD of 0.72 pCi/1. For finished drinking water, the average gross beta activity was 0.16 pCi/l greater for the indicator stations than for the control station. This difference was not discernable because it was less than the MDD of 0.58 pCi/1. Listed below for each year of operation are the average gross beta levels for 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 (L t - Lc). hriod Indicator [ontrol (Li - Lc) RAW CY 90 2.53 2.55 -0.02 CY 89 2.93 3.05 -0.12 CY 88 2.67 3.04 -0.37 CY 87 2.20 5.50 -3.30 FINISHED CY 90 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 4-13
I 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 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 22.5 percent greater for the raw drinking water than for the finished drinking water; this is expected since the finished water has been filtered. There has not been a discernable difference between the average values at the indicator and control stations during any of the years of operation. As indicated in Table 31, there were no positive results for the radionuclides of interest from 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/l was found in the sample collected for September 1987 at Station 87. 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 pC1/1 collected at the indicator and control stations, the difference between the average levels at these stations (Li - Lc) and the MDD. Period Indicator Cont rol (Li - Lc) 000 RAW CY 90 1320 266 1054 572 CY 89 2508 259 2249 1000 CY 88 2630 240 2390 580 CY 87 2229 316 1913 793 Fil11SHED 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, there was always a detectable difference between the indicator and control stations; as the absolute value of (Li - Lc) exceeded the MDD. The tabulations also show a decided decrease in the tritium levels at the indicator station during CY 90. There was still a detectable difference between the average levels at the indicator and control stations for the raw drinking water, however. During preoperation the results were similar to those for CY 87 through CY 89. As indicated in Table 3-1, there were no positive results from the I-131 analysis of the finished drinking water samples; each result was below its Minimum Detectable Activity (MDA) which ranged from 0.28 to 0.73 pCi/1. Similar results were obtained in previous years of operation. The TS call for a LLD and a RL of I and 2 pCi/1, respectively. 4-14
4.7 Fish The TS call for the 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. Furthermore, 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. B1) 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 27, American shad, an anadromous species, was collected at Station 81. Like CY 88 and CY 89, no posuive 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 24 and October 29, the composition of the catches at the indicator and control stations were as follows. Daig Indicator Control April 24 Channel Catfish Brown Bullhead Redbreast Sunfish largemouth Bass Redear Sunfish October 29 Largemouth Bass Channel Catfish Redbreast Sunfish largemouth Bass Redear Sunfish Redear Sunfish As indicated in Table 3-1, 1-131 and Cs-137 were the only radionuclides of interest found in the semiannual collections of commercially or recreationally important species. Since operation began, no other radionuclides of interest have been detected. In the October catch, I-131 was found at a level of 12 pCi/kg wet in the redear sunfish from the control station and at a level of 13 pCi/kg wet in the redbreast sunfish from the indicator station. These upriver and downriver catches were separated by a distance of approximately 5 river miles. The range for these species is generally less than a mile. The annual thyroid dose that the maximum exposed individual (an adult) would receive from eating fish with an average I-131 concentration of 13 pCi/kg is 0.53 mrem or 5.3 percent of the TS limit. 4-15
Actual releases of 1131 to the river totaled 0.733 mci for the fourth quarter. A conservative estimate of the 1-131 level in fish that might result from these actual releases corresponds to a level of 0.426 pCi/kg wet or about 3 percent of that found in the downriver sample. Since the measured level of I-131 in the downriver fish sample does nnt correlate well with the actual release data, and the 1-131 level in the upriver fish sample is nearly the same as that for the downriver sample (and the upriver fish sample is unlikely to have roamed downstream of the plant discharge structure), it is believed that the presence of I 131 in the fish can be attributed primarily to a source or sources other than plant releases. In October 1989, a positive level of 18 pCi/kg wet was found in one of the samples from the indicator station. The LLD assigned for 1-131 in fish is 15 pCi/kg wet. As seen in Table 3-1, the average level of 103 pCi/kg wet for Cs-137 at the indicator station is 146 pCi/kg wet greater than that at the control station. This difference is not discernable since it is less than the HDD of 405 kCi/kg wet. Since operations began, positive values for Cs-137 have been found in all but one of the 36 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 control stations. Period Indicator Control CY 90 103 249 CY 89 117 125 CY 88 66 116 CY 87 337 119 No trend is recognized in this data, it is noted that a Cs-137 level of 1300 pCi/kg wet was found in one of the fish-samples collected at the indicator station in April. The level was confirmed by a recount. The previous high level found during operation was 446 pCi/kg wet in a sample from the indicator station in CY 87. The RL is 2000 pCi/kg wet. 4-16
l l 4.8 Sediment Sediment was collected along the shoreline of the Savannah River on April 2 and October 2 at Stations 81 and 83. Station 81 is a control station located about 2.5 miles upriver of the plant intake structure at RM 153.3 while Station 83 is an indicator station located about 0.6 miles downriver of the plant discharge structure at RM 150.2. The indicator sample for April was collected at RM 148.6. A gamma isotopic analysis was performed on each sample. Listed below for each year of operation are the average levels of radionuclides of interest in units of pCi/kg dry found in the regular samples collected at the indicator and/or control stations along with the frequency of occurrence and the LLDs. Each of these radionuclides is included in the plant's liquid releases. Period Indicator Frecuency Control Frecuency Be-7, LLD 300 CY 90 465 1.0 545 1.0 CY 89 1300 1.0 415 1.0 CY 88 970 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 C0 58, LLD 25 CY 90 140 0.5 CY 89 135 1.0 CY 88 190 1,0 Co-60, LLD=40 CY 90 46 0.5 CY 89 46 1.0 CY 88 62 0.5 Cs-137, LLD=180 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 4-17
l As in all previous-years of operation, positive readings in CY 90 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 465 pCi/kg dry for the indicator _ station is 80 pC1/kg_ dry less than that for the control station; there is no discernable difference, however, since s this difference is less than the MDD of 2463 pCi/kg dry. .For, Cs 137 the - average reading of 155 pCi/kg dry for the indicator station is 15 pCi/kg dry greater than that for the control station: there is no discernable
-difference since this difference is less than the MDD of 343 pCi/kg dry.
There has also not been a discernable difference between the levels at ' the indicator and control stations for either Be-7 or Cs-137 during any past year of operation. The activation products Co-58 and 0o-60 are seen to be present again this year at .the indicator station at about the same levels as last year but in only half-of the samples, it is also noted that Mn 54 (also an activation product) which appeared in half of the samples from the indicator station the past 2 years was not present this year. Since_the activation products were only found at the indicator stations, their i presence is believed to be due to plant releases. The cobalts were.not found in sediment samples during preoperation. The radiological _ impact due to the presence of Co-58 and Co 60 in the shoreline sediment was assessed by calculating the-whole body dose by direct radiation (from the sediment) to an individual using the methodology and parameters _.of Regulatory guide 1.109, Pavision 1, October 1977 and comparing this dose with that permitted by Section 3.ll.l.2.b of ) the TS (3 mrem per-year). The theoretical dose was conservatively determined to be 2.5 micro rem per year or 0.084 percent of the TS limit. This extremely low dose, although calculable, poses no measurable negative environmental or public health impact. The theoretical doses due to the activation products.in CY 88 and CY 89 were found to be 3.6 and 2.6 micro rem, respectively, i-4 4 L 4-18 g s 1
5.0 INTERLABORATORY COMPARISON PROGRAM Section 3.12.3 of the TS requires that analyses be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program approved by the Nuclear Regulatory Commission (NRC). The Environmental Protection Agency's (EPA's) Environmental Radioactivity i Laboratory Intercomparison Studies (Crosscheck) Program conducted by the Environmental Monitoring and Support Laboratory in Las Vegas, Nevada, provides such a program. Reported herein, as required by Section 4.12.5 of the TS are the results of the EL's participation in the EPA Crosscheck Program. The Crosscheck Program was designed for laboratories involved with REMPs; it includes environmental media and a variety of radionuclides with activities at or near environmental levels. Participation in the >rogram ensures that independent checks on the precision and accuracy of tie measurements of radioactive materials in environmental sample matrices are performed; REMP results can thereby be demonstrated to be reasonably valid. Simulated environmental samples are distributed regularly to the participants who analyze the samples and return the results to the EPA for statistical analysis and comparisons with known values and results obtained from other 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 analyses on the radioactive materials supplied by the program that correspond with those required by Table 2-1. Analyses were performed 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 1-131 analyses of milk samples, and the gross beta, tritium, gamma isotopic, and 1-131 analyses of water samples are summarized in Table 5-1. Delineated in Table 5-1 for each of the environmental media are 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 deviation and normalized range were also provided by the EPA. 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 3 standard deviations was established by the EPA as the control limit. 5-1
TABLE 5-1 (SHEET 1 0F 2) CROSSCHECK PROGRAM RESULTS Date Known . Expected Reported Standard Normalized Normalized Analysis Collected Value Precision Averaae Deviation Deviation Rance Air Filters (pCi/ filter) i Gross Beta 03/30/90 31.0 5.0 '32.33 0.58 0.46 0.12 08/31/90 62.0 5.0 63.33 1.15 .0.46 0.24 Cs-137 03/30/90 10.0 5.0 12.33 2.31 0.81 0.47 , 08/31/90 20.0 5.0 23.67 3.06 1.27 0.71 Milk (pCi/1) I-131 04/27/90 99.0 10.0 114.33 3.21 2.66 0.35 09/28/90 58.0 6.0 52.00 1.00 -1.73 0.20 Cs-137 04/27/90 24.0 5.0 25.00 4.00 0.35 0.95 09/28/90 20.0 5.0 18.67 0.58 -0.46 0.12 Water (pCi/l) Gross Beta 01/26/90 12.0 5.0 9.67 0.58 -0.81 0.12 04/17/90 52.0 5.0 53.00 1.00 0.35 0.24 05/11/90 15.0 5.0 8.33 .58 -2.31 0.12 09/21/90 10.0 5.0 6.67 .58 -1.15 0.12 H-3 02/23/90 4976.0 498.0 5067.00 136.67 0.32 0.31 ' 06/22/90 2933.0 358.0 2963.33 64.29 0.15 0.20 10/19/90 7203.0 720.0 7147.00 130.13 -0.13 0.43 Co-60 02/09/90 15.0 5.0 14.67 2.52 -0.12 0.59 06/08/90 24.0 5.0 20.67 0.58 -1.15 0.12 10/05/90 20.0 5.0 19.33 0.58 -0.23 0.12 Zn-65 02/09/90 139.0 14.0 136.67 6.51 -0.29 0.55 06/08/90 148.0 15.0 142.67 13.20 -0.62 1.04 10/05/90 115.0 12.0 107.00 6.24 -1.15 0.59
t TABLE 5-1 (SHEET 2 0F 2) CROSSCHECK PROGRAM RESULTS Reported Standard Normalized Normalized Date Known Expected Range Value Precision Averaoe Deviation Deviation Analysis Collected 119.67 14.36 -2.39 1.26 Ru-106 02/09/90 139.0 14.0 1.61 21.0 213.67 24.13 0.30 06/08/90 210.0 -1.42 0.51 151.0 15.0 138.67 6.51 10/05/90 37.33 0.58 -0.48 0.10 1-131 08/10/90 39.0 6.0 16.33 3.21 -0.58 0.71 Cs-134 02/09/90 18.0 5.0 0.12 5.0 16.33 0.58 0.46 04/17/90 15.0 0.35 1.12 24.0 5.0 25.00 5.20 06/08/90 11.67 2.08 -0.12 0.47 10/05/90 12.0 5.0 17.67 0.58 -0.12 0.12 Cs-137 02/09/90 18.0 5.0 0.35 5.0 16.67 1.53 0.58 04/17/90 15.0 0.23 0.59 25.0 5.0 25.67 2.52 cn 06/08/90 12.00 1.73 0.00 0.35 52 10/05/90 12.0 5.0 71.33 1.53 -0.66 0.25 Ba-133 02/09/90 74.0 7.0 0.30 10 .' 109.00 2.65 1.73 ' 06/08/90 99.0 -1.05 0 38 110.0 11.0 103.33 3.79 . 10/05/90
i 9 . An absolute vclue of 2 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. 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 for normalized deviation was exceeded for the 1-131 analysis of milk for April 27, for the gross beta analysis of water for May 11, and for the gamma isotopic analysis of Ru-106 in water for February 9. In addition, a trend was recognized in plots of the valua.s of the normalized deviation for the gross beta, C0-60, Ru-106, and Ba-133 in water. Each of the above out of limit and trending cases prompted an investigation. The investigation of the high value of the normalized deviation for the 1-131 in milk revealed nothing unusual about sample preparation or about the calculations. The investigation of the high normalized deviation value and trend for gross beta in water lead to the generation of a new beta efficiency curve and to the counting for gross beta activity before flaming the sample for alpha counting. The investigation of the high normalized deviation value for Ru 106 in water and of the trends of the normalized deviation values for Ru-106 as well as for Co-60 and Ba-133 in water lead to changes in the peak background correction values, a proper preamplifier pole zero adjustment on one of the detector , and the preparation of samples with a more homogeneous mixture. 5-4
6.0 CONCLUSION
S This report has shown the licensee's conformance with Section 3/4.12 of the TS during the year, it has shown that all de'= were carefully examined. A summary and a discussion of the res ..cs of the laboratory analyses for each type sample collected were presented. All results indicate no adverse radiological impact to the environment resulting I from Plant Vogtle operation. ! i 6-1
V GEORGIA POWER COMPANY V0GTLE ELECTRIC GENLRATING PLANT - UNil 1 At4D UNIT 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY OPERATING LICENSE N05. NPF-68 AND NPF-81 ANNUAL ENVIRONMENTAL OPERATING REPORT FOR 1990 (NONRA010 LOGICAL)
V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 AND UNIT 2 ANNUAL ENVIRONMENTAL OPERATING REPORT (NONRAD10 LOGICAL) i (199f) SPECIFICATION In accordance with Section 5.4.1 of the Vogtle Electric Generating Plant Environmental Protection Plan (Nonradiological), Appendix B to Facility Operating License Nos. NPF-68 and NPF 81, this report is submitted describing implementation of the Environmental Protection Plan for the calendar year 1990. REPORTING REQUIREMENTS A. Summaries and Analyses of Results of the Environmental Monitoring Activities for the Report Period
- 1. Aquatic Monitoring - Liquid effluent cionitoring was performed in accordance with National Pollutant Discharge Elimination System (NPDES) Permit No. GA0026786; there was no additional requirement for aquatic monitoring during 1990.
- 2. Terrestrial Monitoring - Not required.
- 3. Maintenance of Transmission Line Corridors
- a. The herbicide Tordon 101 was used for cut surface treatment during the clearing of the Vogtle Scherer 500 KV line on the l section from Plant Scherer to Georgia flighway 15. The product is
! registered by the Environmental Protection Agency for this type of application and approved by State of Georgia authorities. The product was applied in strict compliance with the herbicide label,
- b. There were no c1 caring or maintenance activities within the Ebenezer Creek or Francis Plantation areas during 1990,
- c. Routine maintenance activities within the designated cultural l properties along transmission line corridors were conducted in accordance with the Final Cultural Resource Management Plan.
~
- 4. Noise Monitoring - There were no complaints received by Georgia Power during 1990 regarding noise along the VEGP-related high voltage transmission lines.
B. Comparison of the 1990 Monitoring Activities with Preoperational Studies, Operational Controls, and Previous Monitoring Reports These comparisons were not required because no nonradiological monitoring programs were conducted during the reporting period beyond those performed in accordance with NPDES Permit No. GA0026786 referenced in l Section A above. l 3
l I 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 1990. D. Environmental Protection Plan (EPP) Noncompliances and Corrective Actions There were no EPP noncompliances during 1990. E. Changes in Station Design or Operation, Tests, and Experiments Made in Accordance with EPP Subsection 3.1 Which involved a Potentially Significant Unreviewed Environmental Issue There were no changes in station design or operation, tests, or experiments during 1990 which involved a potentially significant unreviewed environmental question. F. Nonroutine Reports Submitted in Accordance with EPP Subsection 5.4.2 There were no nonroutine reports submitted during 1990. I _}}