ELV-01557, Plant Vogtle Units 1 & 2 1989 Annual Rept Part 2
| ML20042E918 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/31/1989 |
| From: | Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ELV-01557, ELV-1557, NUDOCS 9005040097 | |
| Download: ML20042E918 (184) | |
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owo a rw,, compang 333 Pedmont Avenue Atlanta Gnorg a 30308 Telephone 4D4 LN.3190 Wien;; teess 4D insernon Center PWLwa/
Post Ofice Box 179$
Dantngbrn. / Cat ama 352D1 hHephone 200 N.B $501 May 1, 1990 e,cuwevce w um.
W. O. Hairston, til Seniot Vice Prefedent Nuclear Operahons ELV-01557 0347 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C.
20555 Gentlemen:
V0GTLE ELECTRIC GENERATING PLANT 1989 ANNUAL REPORT - PART 2 In accordance with the applicable regulatory requirements, Georgia Power Company hereby submits Part 2 of the 1989 Annual Report of operating information, it includes the remainder of the 1989 reports not previously submitted.
Sincerely, q)J /4 %
W. G. Hairston, III WGH.lll/JLL/gm
Enclosure:
Annual Report - Part 2 xc: Georoia Power Company Mr. C. K. McCoy Mr. G. Bockhold, Jr.
Mr. R. M. Odom Mr. P. D. Rushton NORMS U. S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. T. A. Reed, Licensing Project Manager, NRR Mr. R. F. Aiello, Senior Resident Inspector, Vogtle mm any R
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GEORGIA POWER
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COMPANY PLANT V0GTLE UNITS 1 & 2 4
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i 1989
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ANNUAL REPORT PART 2 -
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DOCKET NUMBERS 50 - 424/425 LICENSE NUMBERS NPF-68/81 i
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'j GEORGIA POWER COMPANY
- 1 V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET NOS.
50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 1989 ANNUAL REPORT - PART 2 i
TABLE OF CONTENTS-I.
INTRODUCTION l
11.
PLANT MODIFICATIONS AND TESTS OR EXPERIMENTS i
o PLANT MODIFICATIONS o TESTS OR EXPERIMENTS l
111.
EMERGENCY CORE COOLING SYSTEMS OUTAGE DATA REPORT IV. ANNUAL RADIOLOGICAL ENVIROMENTAL SURVEILLANCE REPORT t
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ANNUAL ENVIRONMENTAL OPERATING REPORT l
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r GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 & 2-NRC DOCKET NOS. 50-424 AND 50-425 j
FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 INTRODUCTION i
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, at a site 34 miles southeast of Augusta.
The Unit 1 initial operating license was received on January 16, 1987 and commercial operation started on May 31, 1987.
Unit I completed its second fuel cycle on February 23, 1990. Unit 2 received its initial-operating license on February 9, 1989, and began commercial operation on May 20, 1989. Unit'2 is i
operating in its first fuel cycle.
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GEORGIA POWER COMPANY j
V0GTLE ELECTRIC GENERATING PLANT - UNIT l'AND UNIT ~2 NRC DOCKET NOS. 50-424 AND 50-425 1
FACILITY OPERATING LICENSE NOS. NPF-68 AND NPI-81 PLANT MODIFICATIONS AND TEST OR EXPERIMENTS i
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1989 ANNUAL RDDRT - PART 2 10 CFR50.59(b)
PLANT F0DIFICATI0hS l
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II 1989 N M Rt REPORT - PART 2 10 cm50.59(b) REPORT 87-V1E0019 This Design Change added baffle plates inside the i
1561 (Piping Penetration Filtration System) duct i
work at eight locations to reduce airflows to j
within design tolerance or to within the effective modulation range. of other air balance devices. -
1.
-Addition of the baffle plates inside the duct f
affects no other equipment and does not increase-the probability or consequences of an accident or 1
malfunction with the duct itself, the duct is-i Seismic Category 1, as cited in FSAR section 9.4.3.2.3.c.
adds internal duct baffles for
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The DCP nerely(in addition to the registers and 2.
I flow control danpers already. existing) which do not affect. the seismic qualification of the duct, reference FSAR section 9.4.3.2.3.c and do not create a different malfunction. opportunity not addressed
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in FSAR section 15.
3.
Addition of these baffle plates within the'
.I 1561 duct reduces no margin of safety and has no -
j effect on the bases defined for the applicable Tech Spec, sections 3/4.7.7.-
j 87-VCE0024 Rmove timer fran the start circuit of the l
electric fire punp and replace the handswitch i'
which controls this pmp.. -- The tiner in the start circuit of the electric pwp allows the, fire punps to start out of sequence because of 1
the associated time delay. By removing it and-replacing the handswitch, the overall diesel /
electric fire pmp starting sequence will return to the original design intent.
1.
This change involves renoving a timsr in the start circuit of the electric fire punp to retum the start sequence to the original design intent which does not increase the probability of occurrence or consequences of an accident or any equipment /cmponent malfunction. FSAR sections 9.5.1 and 15.0 were reviewed and require no change.
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1989 AI E AL REPORT - PART 2 10 CFR50.59(b) REPOKI 2.
This change does not increase the probability of any accident or equipment malfunction. No new possibilities or unanalyzed scenarios are created.
This is based on a review of FSAR section 9.5.1 and 15.0, 3.
The safety limits and settings discussed in sections 2.0, 3.0, and 4.0.of the Tech Spec.
do not deal with fire protection.. 'lherefore, there is no decrease in the' Tech Spec. margin of safety.
87-VIE 0032 A.
This DCP adds PABX telephone to Turbine Bldg Chmistry Lab.
B.
Adds PABX telephone to Turbine Bldg Operator's Office.
1.
These phones are be installed in non-safety related areas of the bine Bldg. The PABX system is a non-safety system and modifications to it do not affect any analysis in the FSAR. The probability of occurence or consequences of an accident or eculpment malfunction is not increased.
The detailed cescription of the PAXB systen in Section 9.5.2.2.2 is not altered. The PABX riser diagram will be updated during the annual update-to reflect the new phones.
2.
The PABX modification will provide additional phones. The function of the systen will not change.
The addition of phones does not create the possibility of an accident. It may decrease such possibility by prcviding enhanced ccumunication=
abilities.
3.
The MEX is not described in the basis for Ter
. cal Specifications.
87-V1E0074 Remove Video Tape Recorders from the alann station operaticn sequence of Alarm events.
1.
FSAR sections 3.0 to 12.5 and the Accident-Analysis section 15 were previewed. It was then determined that the removal of the Tape Recorders would not impact these FSAR requirments.
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s 11-1989 ANNUAL KtKitT - PART 2 10 CFR50.59(b) kEPORT' Also, FSAR Section 13.6 details the industrial security requirements, section-13.6 as well as the VDGP Physical Security Plan was, rwlewed.' No decrease in the security effectiveness w uld be realized an a result of impl eenting this change.
2.'
he Accident analysis section 15 of the FSAR was reviewed, it was then determined that the FSAR would not be affected by the r m oval of these-tape recorders.
3.
Sections 2.0, 3.0 and:4.0 were reviewed in the Technical Specification to determine that the remval of the tape l recorders would have no affect on_ che margin of safety as described in the Tech.
Spec.
87-V1E0103 This nod 4.fication involves a piping change co system 2301, project-class 629 and adds a 12"x12"x12" tee and a 12" gate valve to fire water yard' loop piping C-2301-515-12", ; adjacent -
to hydrant 927. The gate valve will be fitted-with a ground level valve box' assembly, and operated with a key wrench. This arrangement basically amounts to an isolatable tee off the main fire water yard loop..
1.
This change involves addition of a isolatable =
tee off tw main fire water yard: loop which does not increase the probability of occurrence or consequences of an accident or any equipment /-
component malfunction.
2.
This change ~does not increase the probability.
of any accident or equipnent malfunction. No new possibilities or unanalyzed scenarios are created.
This is based on a review of FSAR sections 9.5.1 and 15.0.
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The safety limits and settings discussed in L
sections 2.0, 3.0.and 4.0 of the VEGP Tech Spec.-
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there is no decrease in the Tech Spec margin of-safety..
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.II 1989 ANNLEL REPORT - PART 2 10 CFR50.59(b) REIORT 87-V1E0143 Replacanent of servo-valve mounting plates on At2nospheric Relief Valve actuators with an upgraded plate supplied by Paul Monroe. The originally supplied plates do not allow sufficient o-ring ecmpression., The new plates have a shallower o-ring groove allowing for better o-ring ccupression to alleviate leakage.
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This modification inproves the reliability of the Atsnospheric Relief Valves and in no way increases the probability or consequences of en accident.
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. Failure of a relief valve has previously been evaluated in FSAR section 15.1.14 "lnadvertent Opening of a Steam Generator Relief or Safety Valve".
2.
Inadvertent opening'of a Steam Generator Relief Valve has been previously addressed in FSAR section 15.1.4. Also,theprobabilityofinadvertent
' closure or a " stuck ' closed valve is not increased.
by this change. The reliability of the ARV's will be enhanced by reducing hydraulic leakage.
3.
This nodification improves the reliability of the Atanospheric Relief Valves and therefore, does not decrease the margin of safety in Tech Specs..
sections 3/4.3 and 3/4.7.1 were reviewed.
87-VCE0151 Attached a dynamic absorber to.the Fire Protection 4
ptop, C-2301-P4-003, due to vibraticos during operations which exceeded the vendors; reccumendations for long tenn operation. Test performed on 6-12-87 with a tecnorarily installed
' dynamic absorber showed that viarations.were within the acceptable range.
1.
The change involved adding a dynamic absorber-to the Fire Ptop which does not increase:the probability of occurrences or consequences of an accident described in the ESAR. FSAR sections.
9.5.1 and 15.0 were reviewed and require no change. There is no degradation of the " defense-in-death" Fire Protection Program as a result of I
this X:P.
f 2.
This change does not increase the probability -
of any accident or equipment malfunction.. No new possibilities or.unanalyzed scenerios are created.
This is based cn FSAR sections 9.5.1 and 15.0.
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11 1989 ANNl]AL REPORT ~PART 2 10 CFR50.59(b) REFORT 3.
The safety limits and settings discussed in section 2.0, 3.0 and 4.0 of the Tech Specs, do not deal with Fire Protection. Therefore,
.there is no decrease in the Tech Specs, margin of safety.
87-V1E0152 Existing radiation shielding for the Post Accident Sanpling System (PASS) skid does not. allow -
for easy access to the rear of the PASS panel.
This change will modify the shield door to allow -
easier access to the rear of the PASS panel, but:
still keep _ radiation AIARA. Access to the rear-of the panel is necessary for both normal and emergency maintenance. The PASS skid is located on Level A of the Fuel Handling Building.
1.
The change is needed for maintenance purposes and does not increase the probability of occurrence or consequences of an accident described in section 15 of the FSAR.
2.
This modification does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.- Chapter.15 of the FSAR was reviewed for impact.
3.
The margin of safety, as defined in the bases.for Tech. S wes, does not include any specification applicable to this change.. There are no safety limits, limiting safety systan settings, ifmiting conditions for operations, or surveillance requirements involved. The bases for Tech. Specs.
Section 2.0, 3.0 and 4.0 were reviewed for-impact.
l 87-VIE 0162-This change provided battery b'ackup power to Local Zone Indicating Panel A-1813-Q3-F61.
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This change conforms.to criteria stated in FSAR 9.5.1.2.3.1~-B which requires a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery backup systen at the river intake fire detection panel (A-1813-Q3-F61).--No increase in the probability of occurrence or consecuences of '
an accident or malfunction exists due to this change.
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r II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT l
2.
This change ccr: plies with the original design intent of either providing a uninterruptable power
. source for fire panels or a battery backup _ power source. No unanalyzed accident or malfunction can occur.
3.
The fire protection system does not involve l
Technical Specifications.
87-VIE 0165 This change describes core drills on the Unit 1 and Unit 2 boundary and the corresponding penetration seals necessary to restore fire boundary integrity.
1.
This change does not involve any equipment or ccnponent. Penetrations will be made and sealed in such a manner that no degradation in the ability of the facilities's walls and slabs to meet their.
design requirements will' occur. This change will' i-not increase the probability of occurence or
__ consequences-of.an accident as described in FWa section 15 (Accident Analysis). There is no degradation of the " defense-in-depth" Tire -
l Protection Program as a' result of this DCP.
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The change does not create the possibility for any accident or equipment malfunction not previoualy described'and enalyzed in the ESAR.
3.
This design change meets the margin of safety defined by the Bases for the Vogtle Technical Specification. This -is based on a review of the Bases in Sections 2.0, 3.0, and-4.0.
i 87-V1E0169 The Video Anplifiers in the Security system are no L
longer manufactured, therefore require upgrading or replacing.
1.
Sections 3.0 to 12.5 were reviewed and it was determined that this design change _will not increase the possibility.of,_or:the occurrence, or consequences of an accident or malfunction as described in the FSAR. Also,.FSAR Section 13.6 details the industrial security requirenents,-
section 13.6 as well as the V mP Physical-Security Plan was. reviewed., No decrease in the security effectiveness would be realized as a result of inplementing-this change.
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f II 1989 NEAL REPORT - PART 2 10 CFR50.59(b) REPORT 2.
The accident analysis section 15 of the FSAR was. reviewed in determining that the w ssibility of-an accident or taalfunction would not % increased -
i by the installation of this DCP.
3.
The 140 Bases in sections 2.0, 3.0, and 4.0 of.
the Tech Spec, were reviewed. As a result we found the tnargin of safety would not be reduced by the installation of this DCP.
87-V1E0173
-Modify Seven Operator Interface Modules (01M) to replace Linear Scales with thermocouple Type T non-Linear Scales.- Install new dual flex Ineters i
and revise legend Plates. Modify one OIM -
(LIC-4415) as above except replace present-scale with a scale of correct span.- Modifications are associated with the following talperature controls. TI-4130, 4131 indicate tenperature of the outlet Heater f5A and SB. TI-7079 indicates tenperature of Generator Hydrogen Cold gas passage.
TI-5498 indicates temperature of SGFP A. Bearing oil j
coolers. TI-7116 indicates tenp. of Main Turbine
. lube oil. LI-4415 indicates level in Condenser Hotwell "C" and is used-to control valve LV-4415.
1.
These Operator Interface Modules -(OIM) and the associated instruaents are not safety related and have no effect on any equipn.ent or cmponents that i
are safety related and analyzed in the F3AR' Section 15 Accident Analysis.
9-2.
This design change which modifies seven 1
Operator Interface Modules (01M) and one scale with a correct span. Mater scale changes that provide for correct readings will not create an accident or malfunction not evaluated in the FSAR.
3.
These non-safety related Operator Interface Modules (OIM) scale changes have no effect on the I
margin of safety per review of Technical o
Specifications Sections 2.1,2.2,3/4.3 (Instnment) and 3/4.7 (Turbine cycle).
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II 1989 ANNUAL REPORT - PART 2-10 CFR50.59(b) REFORT 87-VIE 0174 h is modification is non-safety related. It adds an in-line thermostat to monitor the line taperature of Heat Tracing lines supplying eye wash and shower units.
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The modification does not affect the accident analysis of FSAR chapter 15, nor affect any syst m required to function or mitigate the effects of any -
postulated accident. The modification only adds an D;
in-line thermostat to monitor the.line tmperature units to ensure the water does, eye wash and shower of Heat Tracing lines supolying t get too hot.
2.
L e proposed change does not create the possibility of an accident or malfunction not analyzed in the FSAR nor is a change to FSAR Q1 apter 15 analysis required.
3.
There is no: change to the margin of safety or basis of the Tech Spec. his. includes a review of the bases for Tech Spec. 3/4.7 87-V1E0182 Add a new 6" vent fr m the Crud Tank (1-1224-T4- -
i 001) to the Waste Holdup Tark (1-1901-T6-002) and '
route a 2" vent line from the new 6" vent line to the Auxiliary Bldg, exhaust. The existing 2" vent.
1 line is abandoned in place. This change will minimize the pressurization of the Crud Tank during backflushing and allow better level'manitoring while maintaining a monitored vent-path to the_
j Auxiliary Building exhaust."
1.
This change does not affect the probability of l
occurrence or consequences of an accident described-nor the malfunction of any equipment assumed to function in FSAR Chapter 15, sections 11.2 and 11.4, and Table 3.2.2-1, sheet 18.'
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his modification does-not impact any accident or.
equipment malfunction not described or implied in' FSAR sections 11.2 and 11.4', and Chapter 15. This change has no system res)anse changes other than to allow more effective'baciflushing.--
3.
This change does not inpact the Tech Spec, bases B 3/4.11 and B 3/4.12 and therefore does not affect the margin' of safety, a
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10 CFR 50.59(b) REPORT 87-VIE 0238 This DCP adds a security Alarm to a door in the PESB, and removes access to the restroom in the PESB Entry Vestibule.
1.
The Accident Analysis section of the FSAR (section 15) was reviewed to determine that the implementation of this DCP would not increase the probability of, or the consequences of an accident as described in the FSAR.
Also, FSAR Section 13.6 details the industrial security requirements.
Section 13.6 as well as the VEGP Physical Security Plan was reviewed.
No decrease in the security effectiveness would be realized as a result of implementing this change.
2.
A review of Section 15 of the Vogtle FSAR was performed in determining that this design change would not create the possibility of an accident or malfunction other than previously evaluated in the FSAR.
3.
A review of the Bases in Sections 2.0, 3.0 and 4.0 of the Vogtle Technical Specification was performed, it was then determined that this design change-would not reduce the margin of safety as defined by the Tech Specs.
87-VIE 0244 This modification added restraint plate:, to support lines 1201-178-1" and 2402-004-1" in the Fuel Handling and Aux Buildings.
The supports are located inside the sleeves of Containment penetrations #62 and #42, on the vendor supplied portion of the flued head assemblies outside the containment pressure boundary.
l 1.
In order to support the process piping portion of the l
penetration assembly in the vertical and lateral l
directions the vendor required a restraint plate be installed in the containment sleeve.
Therefore, the addition of these restraint plates does not increase the probability of occurrence or the consequence of an accident or malfunction of equipment important to safety, but ensures that they meet the design previously evaluated in the FSAR.
FSAR section 15 was reviewed.
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1989 ANIE L REPORT - PART 2 l
10 CFR50.59(b)'REIORT 2.
No new postulated accident is created as a result of adding the restraint plates. FSAR sections 3.6. 3.7, 3.9, 6.2 and 15 reviewed.
3.
The margin of safety described in the Tech Specs bases is not reduced as a result of adding the restraint plates. Sections 3/4.4 and 3/4.6 reviewed.
l 87-V1E0246 Provided design details to seal various penetrations ln the Control Building between' rocrus R-304/R-325, R-307/R-308 and R-305/R-259, these penetration seals are l
for Unit 2 conduit and are addressed on deficiency cards 1-87-1179,~ 1087-1181 and i
1-87-1182. Penetration seals were performed 1
per procedure 00432-C (Penetration Seal-i Control). - Rese seals neet all hazards and fire protection design criteria.
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These penetration seals neet the design l
requirements for the facility where installed.
v he change meets the requirments as described or inplied in FSAR sections 3.4,.3.8.4, 6.4, r
9.4.1and9.5.1withAppendixes9A&9B. Tiere is no degradation of the defense-in-depth" Fire Protection Program as a result of this DCP.
his change does not involve any equipment or component. The hazard analysis is not affected by h e change provides the penetration this change.;d by the plant Fire Protection seals require Program. There is no change to the fire hazard l
analysis of section 9 or Accident Analysis.of r
section 15. No increase in the probability of i
occurrence or the consequences of an accident 3
will occur.
1 2.
This change does not create the possibility for any accident'or equipment malfunetion not previously described and analyzed in the FSAR.
The material used neets the fire protection j
requirements of FSAR section 9.5.1.
3.
This design change meets the margin of safety defined by the Bases for the Vogtle Technical i
Specification.- This is based on a review of the Bases in Sections 2.0, 3.0, and 4.0.
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p II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 87-VIE 0248 A.
This DCP re11 aces the existing 50 Anp Breakers 1NBE71, INB765, 1ABES2, INBR43 & INBS53 with 40 anp Breakers.
B.
Disconnects p' ower supply to EHC cabinet fran the Distribution Panel INYE1 and spare Breaker 11A'E136.
C.
Adds a 480/120V transformer with a 15 amp circuit breaker at the MCC 1NBE64 to provide-power to the EHC Cabinet. This modification
's required to neet Reg. Guide 1.63 i
Electrical Penetration Conductor overcurrent Protection Requirments..
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The proposed changes 1) replaces one class IE bretker with a class 1E breaker of same type, Quality and dimensions and it' does not inpact any safety either directly or indirectly 2) meets' original design intent and specification and does not degrade the reliability of any system component or structure 3) Assuming a malfunction of class-1E equipuent (Such as distribution Panel feeder breakers) the proposed changes would inprove the overall operability of the systen therefore the proposed changes does not cause any malfunction to any equipment or component asstmed to function in accidents analyzed in the FSAR.
(This includes FSAR Chapter 15).
2.
Per Reg Guide'1.63 and FSAR Section 1.9.63 cn electrical penetration conductor shall be protected by a backup device when the primary protective-device fails, #10 AW penetration conductors used for 120V AC power circuit fed fran 120V AC distribution panels are not protected by back-up device.
With the change, regulatory guide and FSAR requirements are net, where backup protective device operates if the primary protective device-fails to operate as shown in revised calculations (X30401) dated 7/14/87. The change does not add any new design changes and does not create any failure nodes which have not been analyzed
.previously.
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II 1989 ANNLRL RDORT - PART 2 10 CFR50.59(b) REPORT 3.
Technical Sweification Section 3/4.8.4 and the
-bases for t w Technical Speci.fications han been reviewed and the changes do not decrease the margin of safety defined by the bases for the Technical Specifications. The changes are made in order to meet the intent of design Technical Specifications and FSAR requirements etc.
87-V1E0256 De-rate two walls located in the Radwaste Transfer--
Tunnel (Area 1-RB-1A), one of which separates the Radwaste Transfer Tunnel from the Radwaste Transfer Bldg and the other which separates the Radwaste Transfer Tunnel fr m the Radwaste Solidification Bldg. Also c abines area-1-RB-1A-A into area l
1-KrB-L1-A.
1.
B is change involves de-rating fire boundaries in areas where no' safe shutdown equipnent or camponents exit, his change will not increase the probability of occurrence or consequences of an accident described in the FSAR Section 9.5.1 and 15.0 were reviewed and. require no change.
There is no degradation of the " defense-in-depth" Fire Protection Progran as a result of this DCP.
2.
h is change does not increase the possibility of any accident or equipment malfunction. No new possibilities or unanalyzed scenarios are created.
h is is based on review of FSAR sections 9.5.1 and 15.0.
3.
The safety limits and settings discussed in sections 2.0, 3.0 and 4.0 of the Tech Specs, do not deal with fire protection.- Therefore, there is no decrease in the Tech Specs, margin
.of safety..
87-VIE 0274 Bis DCP modified the lock configuration on several security system access doors.-
1.
h e Accident Analysis section of the FSAR (section 15) was reviewed to determine that the inplementation of this DCP would not increase the probability or the consequences of an accident as described in the FSAR.- Also, FSAR Section 13.6 12
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II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT details the industrial security requirements, section 13.6 as well as the VEGP Physical Security Plan was reviewed. No decrease.in the security effectiveness would be realized as a result of inplanenting this change.
2.
A review of Section 15 of the Vogtle FSAR was-performed in detemining that. this design change would not create the possibility.of an accident or-malfunction other than previously evaluated in the FSAR.
3.
A review of the Ba'ses in Sections'2.0, 3.0 and-4.0 of the Vogtle Technical Specification was-performed. It was then determined that this design change would not reduce the margin of safety as defined by the Tech Specs.
87-VCE0324
'Ihis change adds a non-Q class 480/208V, 30 KVA transformer, and I conduit (exposed),1with an associated disconnect switch in the hot machine shop. And revises the existing disconnect switch-sizes hi this room to agree with equipment requiranents. 'In addition, the breakers feeding this equipment from MCC ANBH (Aux Bldg.
Level A Eoam 54) are the wrong amperages, and, unst be replaced.
1.
This change will not involve any safety-related components, and will not cause any equipment' assumed to function in an accident to i
malfunction. Reference FSAR Section 15.2.6.
I 2.
This change involves no safety-related conponents or equipment, and is designed / installed per the requirements. Therefore, no accidents or equipment malfunctions will result from this change that is-not described in the FSAR. Reference FSAR section 8.2.
3.
Adding a 208V AC power source to the hot machine shop will not effect the margin of safety as defined in Technical Specifications 3/4.8.1.
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' II 1989 MNJAL REPORT'- PART 2 10 CFR50.59(b) REPORT 87-V1E0329 Provide lube oil sample point on each of the mergency diesel generators in Unit il keep warm skids, he sanple point will be installed -
off the vendor supplied tubing located between the suction and discharge of the lube oil keep wam pmps. This nodification provides chmiistry personnel a safe and precise means of obtaining the required nonthly sanple, t
1.
W e installation of-the lube oil sanple points I
will be done per the aroject class 212 requirments. his caange will not increase the probability of. occurrence or consequences of the malfunction of any equipment or cmponent assumed to function in accidents anclyzed in FEAR section 15.0, 9.5.5 and 9.5.7.
2.'
he installation of the lube oil sanple points will be done per the project class 212 requirements:
therefore, it does not create the possibility of an accident or equipment /cmponent malfunction not described and analyzed in FSAR sections-15.0, 9.5.0
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and 9.5.7.
3.
The proposed change does not decrease the margin of safety defined by the bases of the t
Technical Specification 3/4.8.1.
87-V1N0338 he resistors for indicating lights on the 1-1604-QS-PCP.were replaced with higher wattage resistors. Eis was done to reduce the failurc rate and the operating tenperature of these resistors, i
1.
Le changes were to non-safety related resistors 1
in a non-safety related area of the panel and will not affect any safety related equipment.
~
2.
Le resistors are located in their original.
y locations which is in a metal control' cabinet separated fr a any 1E circuits.. We heat load for this panel remained unciv2nged.
3.
The nodification increases safety by reducing failures and the t aperature of these non-safety y
related resistors.
L l
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11 1989 ANNTAL REPORT - FART 2 10 CFR50.59(b) REPORT l
i 87-V1E0348 Design Change 87-V1E0348 installed, verified and validated (V&V) software in the Plant Safety i
Monitoring System (PSMS), Neutron Flux Monitor System (NFMS) and the Alternate Shutdown Indicat Syst e (ASIS). h e V&V software provides an i
additional level of confidence that the 1
micro-processors will meet their functional
.t recuirements in a highly reliable manner.
Ackitionally, for use with the NEMS as athe modification provid held terminal maintenance tool.-
{
1.
We nodification does not increase the probability of occurrence or the consequences of an accident or.
t malfunction of equipment important to safety previously evaluated in the safety analysis report. The modification installed V&V software, i
in the form of " Burned-In" eproms, which provide an additional level of confidence that the micro-processors will perform their design functions.
Berefore, since the systems' software has under gone a V&V process the probability of malfunction is decreased, the above conclusion:is based on review of FSAR Qiapters 7.5,15, and 7.4.3.3.
he -
hand held teminal is a non-safety related-maintenance tool.
It does not affect qualification margin, safe plant operation or safe plant sluitdown.
2.
Le modification does not create the possibility-for an accident or malfunction of a different type than previously evaluated'in the safety analysis report. - he modification-installed V&V burned-in EPRCNS which provide-an additional level of confidence that the syst es will meet its functional requirements in a highly reliable manner. h is resolution:is based on 4
review of FSAR Chapters 7.4.3.3, 7.5 and 15.
3.-
h e modification to the PSMS,' NFMS, and ASIS does-not reduce the margin of safety as defined in the l
bcses for Technical Specification Sections 2, 3/4.3.3.5,3/4.3.3.6,3/4.3.1.
i 15
II 1989 ANNUAL REPORT - PART 2 t
10 CFR50.59(b) REPORT 87-VIE 0364 Added core drills and penetration seals to Unit 1-Control Building walls between rooms R125/R128 and R125/R131 as required for the installation of Unit 2 electrical conduit inside the Unit 1 protected area.
1.
The' permanent penetration seals were installed.per this DCP and provided a' hazard rating e' qual to or1 greater than required-for the walls penetrated.
There is no degradation of the " defense-in-depth"'
Fire Protection Program as a result of this DCP.
This change ddes not involve any equipment or; component. The building's-hazard anaylsis and structural-design are not affected by this-
. change. Penetrations.will'be made and: sealed'in-such a manner that no degradation in the abilityL I
of the facility?s walls to meet their design requirements will occur. Change will not increase-
-the probability of occurrence or consequences of an accident as described-in FSAR section 15;(Accident
-Analysis). No changes to FSAR Hazard Analysis is
~
required.
2.
The change does not create the possibility for any
-accident or equipment malfunction not jreviously described and analyzed in the FSAR. -Caange does
'not involve any equipment or camponent. The-structural design and hazard analysis are not affected by this change.
3.
E is design change meets the margin of safety-defined by the Bases for the Vogtle Technical Specification. "his is based on a review of the H
Bases in Sections 2.0, 3.0, and 4.0.
87-V1N0367 At "KA8" relay board inside local' Zone Indicating Panel "1NCPFP48" (1-1813-Q3-F48), remove two junpers run between contact "2A" and "3A".
1.
The-change does not directly or indirectly-affect any safety related equipment described-in an accident in the FSAR nor does it impact any equipment required to mitigate an accident.
Therefore, the change does not increase the probability of occurrence or consequences of an accident described in the FSAR. There is no degradation of the " defense-in-depth" Fire Protection Program as a result of this DCP.
16
II 1989 ANNUAL REPORT --PART 2 10 CFR50.59(b) REPORT-he change does not affect die probability of occurrence or consequences of a malfunction of any safety related equipment or cmponent assumed to function in accidents analyzed in the FSAR because the proposed' change reflects the-design described in FSAR Chapters 9A.I.33;-9A~.1.113,
'and Spec. X4AX03.
2.
. W e change does not create the possibility of an accident or equipment /cmponent malfunction not.
described and analyzedtin the FSAR because a ftre in fire zone 141B will not propagate to zone 143A.
Fire detection capability in zone 141B is still available to the. control room operators.
3.
- h is change. sill' maintain the safety margin defined' in the Technical Specification bases for systems and components associated with the affected fire zones because the proposed modification does not change.
the operability and design of these systems and ccaponents'.
l 87-VIN 0379 Selected portions of the Containment purge-
^
supply (1505)'and exhaust (1506) ductworc were y
upgraded to piping for overaressurization.
1 protection. In addition a ly-pass line and orifice were added on minipurge exhaust line to i
allow reduction of pressure below 10" w.g. when-i venting Containment from higher pressures..
-l 1.
Review of FSAR sections 6.2.4 and 9.4.6 reveals this DCP does not involve any aspect of the. Safety Design Bases' Berefore,.this DCP will not increase the probability of occurrence or consequences of an accident described in the FSAR. This review also. included section 15.
2.
his DCP does not affect any valve associated with Containment isolation. Based on a review of FSAR sections 6.2.4, 9.4.6 and 15 this~DCP does not i
create the possibility of an accident or equipment /
component malfunction not already described and analyzed in the F3AR.
l 3.
Review of Tech Spec. section 2.0'and sections j
3.6.1.4 and 3.6.1.7 reveals this DCP does not decrease the margin of safety as defined by the bases of the Tech Spec.
.i 17
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1989 ANNUAL REPORT - PART 2 i
10 CFR50.59(b) REPORT j
.87-V1N0386 h is change adds manual isolation valves upstream and downstream of existing air operated' isolation valves for the Seal.
Injection backflush filters.. The affected lines are-1-1208-22-2",'1-1208-22-3",
1-1208-23-2", and:1-1208-150-2". The valves are project class 212 for pressure boundary-reasons and.are added to allow isolation'of:
existing air operated isolation valves during-maintenance or problans with existing isolation 1
valves.
Reach rods' are used for accessibility.and~ A1 ARA Concerns.
1.
Rese valves are class 212 and are qualified for the pressure boundary conditions. This change impacts no accidents or equipment analyzeo in FSAR section 9.3.4, and chapter 11 and 15.
2.
This change does not create additional malfunctions or accidents that could impact a system required for an analyzed accident in FSAR section'9.3 and-chapter 15.
.i
- 3..
Le addition of these normally open valves-do not impact any equipment in the Tech Spec. including 1
bases B'3/4.4.
87-V1E0387 his design change adds approximately 1.5 inches of ceramic fiber insulation over existing calcium silicate insulation on Steam' Generator. Blowdown inlet piping, condensate piping and the blowdown.
heat excaanger located in Aux,1 Building rooms C-106 and C-108. This insulation is added to line numbers 1-1407-001-3", 1-1407-002-3", 1-1407-003-3",
1-1407-004-3", 1-1304-061-3", 1-1304-062-3",
1-1407-063-3", 1-1304-064-3" and 1-1304-061-6".
The insulation is added to reduce high ambient room.
tencerature in the Aux. Building room C-106 to prevent electrical circuit card failures which are-f located in Steam Generator Blowdown Control. Panel, L
i 18 l
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-II 1989 ANNijAL PEPORT - PART 2 t
10 CFR50.59(b)-REPORT 1.
Additional insulation does not affect any safety related systen. There is no affect on accident i
Analysis of FSAR Chapter 15. Bis design nodification affects Steam Generator Blowdown piping l'
outside the Contaiment which is not required for safe shutdown or assumed to function'during l
accident.
f 2.'
his design nodification to add extra insulation over the existing insulation on the Stean Generator ~
and condensate piping and'the Heat Exchanger.does not create any: accident or does not affe~ct any canponents;which are asstned to function during accidents. h is includes a review of FSAR Sections j
3.6, 10;4.8,.and 9.4.3.-
J a
~
^
3.
Insulation on Stean Generator Blowdown System or condensate. piping at the Heat Exchanger _ rown C-106 i,
located in Aux Building are not addressed'or -
assumed in the margin of safety of' the Technical-o Specification. E is include a review of Sections-
'3/4.7.1 and 3/4.7.13 and 3/4.3 subsectionJ3.3.3.11 of the Technical Specification.-
87-V1N0388 This change _ adds 3/8 inch SS' sample: tubing.
assembly (up to a maxinun length of '24 inches) to-local liquid sample points, waere required ~in the shall not be' connected plant.'his. tubing' assembly / components.
to Seismic Category 1 piping
. ' At -
1 3 resent, most of -the : sample. connections are
- 1orizontal and it is difficultcto take the samples inside the sample bottles without splashing the contaninated liquid.: Adding the sample tubing-assembly will facilitate obtaining grab samples.and decrease the potential for contamination of personnel, equipment and areas.
)
1.
The change reduces contamination of personnel,
.i equipment and areas. Le change does not affect any equipment involved in accidents postulated'in the FSAR'or equipment assumed..to mitigate any 3
accident described in the FSAR. This includes'a review of.all chapter 15 accidents. Berefore, it does not change any accident probabilities or H
J Consequences.
J
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II 1989 AINUAL REIVIC - PAKr 2 10 CFR50.59(b) REPORT 2.
Bis change does not create the possibility of an accident or malfunction not already described in the FSAR because the sanple assembly is Seismic Category.2 and less than 10 pounds in weight. In accordance with paragraph 3.5 of Design Criteria-DC-1005, seismic 2 over 1 analysis is.not required for piping or components weighing less than 10 pounds because failure will not adversely affect
' Seismic Category 1 piping or couponents.
3; he sanple connections -do not affect:the safe operation of the plant.' Therefore, the margin of.
safety defined in the Technical Specification will not be decreased by the proposed change.>
87-V1E0391 W e change requires the drilling of holes in the
~ hrbine Building Unit.1/ Unit 2':Kondary wall and water analysis lab roof inside the Unit 1 Protected.
-Area. These holes were added to provide for the installation.of Unit 2 Electrical Conduit inside the Unit 1 Protected Area. The permanent penetration' seals were installed per the DCP and j
meet or exceed fire barrier rating.
i 1.
Penetrations were made and' sealed in such a'.
manner that no-degradation in the. ability of the facility's wall / slab to meet design requirements i
will occur; This change will not increase the'
. probability of occurrence or consequences of an.
accident as described in FSAR section 15~( Accident
- i Analysis). No changes to FSAR hazard analysis is' required. There.is no degradation of the; j
" defense-in-depth" Fire Protection Program as a result of this DCP.
~
q This change does not. involve any equipment-i or ccuponent.. The building's hazard anaylsis and structural design are not affected by this change.
2.
The change does not create the possibility for
'any.-accident or equipment malfunction not previously described and analyzed in the FSAR.
20
II-1989 ANNUAL REPORT - PART ')
10 CFR50.59(b) REPORT This change does not involve any equiament or component. The structural-design and'aazard analysis are not affected by t b change.
3.
This' design change noets the margin of safety
' defined by the Bases for the Vogtle'Technicel Specification. This is based on a review of the -
Bases in Sections 2.0, 3.0, and 4.0.
-q 87-V1E0394 Replace the existing IIT Barton pressure Rev 1 transmitter. Disconnect the tubing for ipr-405 j
fr a the existing tap located on tae RHR system ~
i recirculation line. Route new instrument tubing to.the RCS hot leg tap associated with:the RVLIS.
instrumentation ILX-1320.
1.
The design change does not increase the probability of occurrence'or consequences of the malfunction of1 any equipment or camponent assumed ~to function in i
accidents; analyzed in FSAR sections 15.0, 5.1, 5.2, 1
7.2, 7.5 and 7.6.
2.
The design change replaces the existing IIT Barton-
~$
pressure transud.tter IPT-405 with a To'aar pressure -
j transmitter. The Tobar pressure transmitter is-1 fully qualified and is better suited to this-l application. The relocation of the instrumentL j
sensing line fr m the RHR system recirculation'line
,1 to the RCS hot leg loop 14 will inprove the. _
1 instrument accuracy and reliability. The instrument i
will no longer be subject to the pressure surges present in the RHR system recirculation line.whichL is' impacting its' readings and reliability. The rerouting of the' instrument sensing line is supported by the rm ised tubing. fabrication iso and the supporting stress calculation. The' design 4
change does not create the possibility of an-1 accident or equipment /cmponent malfunction not described and analyzed in FSAR section 15.0.
I 3.
The design change does not decrease the margin of i
safety defined ay the bases of Technical ~
Specification section 3/4.4.9.
~87-V1N0414 These DCPs involve changes to various pip /or paysical e su) ports 87-V1N0415 (such as deleting / replacing-snubbers and i
87-V1N0416 modifications to pipe support steel) in the 1301 system. There are no changes to the system piping or cmponents.
i 21 i
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i II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 1.
his change does not affect systan function or.
operation and therefore does not affect the accident-analysis, probability or consequences of a Main-C Steam accident' described in Sections 15.1 and 15.2.-
(
i 2.
21s. change does nottaffect i:ny systen, equipment or
. component a function'or operation and therefore does--
3 not affect:the accident analysis, probability or
~
consequences described in Section 15.1 and 15.2.
3.
his change;does not affect:the Main Stean systan/ '
1 equipment function or operation and therefore does not affect the safety margin defined by Tech Sepc;
- 3/4.7.8.. Bere is no change -in the bases of Tech Spec 3/4.7.8. L e list of individual snubbers referred to the bases of Tech Spec. 3/4/7.8 will' change, but the margin of safety is not affected, a
87-VIN 0436 Revise the high alarm and low alarm / pump trip j
setpoints for.the following tanks in the Liquid-1 Waste Processing System (system 1901)r The Floar Drain Tank, (equi'xnent. tag no. 1-1901-T6-008), - the -
Waste Holdup Tanc (1-1901;T6-002), the Laundry and-Hot Shower Tank (A-1901-T6-007), the Waste: '
Evaporator Condensate Tank (1-190-T6-003), the<
' Waste Monitor Tanks (1-2901-T6-009 and!
1-1901-T6-010) and the Chemical Drain: Tank (A-1901-T6-005)._ Eese components are non-safety related and their failure will not compromise a-safety-related system or affect the safe shutdown of the plant.:. All' tanks are project class 417--
and 427.
r 1.
This Liquid Waste Processing system is a non-safety related systan. Failure of the c mponents of this-
- i
' system will not affect the ability of the plant to.
accmplish a safe shutdown, nor will it empromise a safety-related systen.. This is based on.a review of FSAR sections-15.0 and 11.2. h ere is no affect
-on the consequences,of a radioactive liquid waste' systen leak.as described in FSAR section 15.7.2.
- 2.
Ee. changes are to c mponents of the' Liquid Waste Processing. system (systan 1901)'. This. system is-i non-safety related~and not required to function in an accident as described in FSAR chapter 15.0.
\\
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II~
1989 ANNUAL REPORT - PART 2' 10 CFR50.59(b). REPORT
. Inplementation'of these changes' does not increase the. possibility of an accident or malfunction.
This includes a review of FSAR sections 11.2 and 15.7.2.
. 3. -
Based on.a review of the Tech' Specs basis,-.
including section B 3/4.11, this modification will
- not decrease 1the margin of safety.
87-V1N0440 Nitrogen supply header line'l-1224-012 will be cut-
- and a blind flange /s al 31ece will be added toL separate one header aranci from the other to prevent contamination from high pressure process:
a leakage into the Nitrogen header and other process.
filter sections. A tee with a pressure indicator--
(PI-41310) will.be added'(along with the PI isolation valve): downstream of valve 1-1224-U4-015
- to monitor the Nitrogen header for process -
i backleakage. A nameplate will be added to the 1
~
Backflush Filter Control Panel identifying PDI-41304 as crud tank pressure.
1.
Eis change' does notlinpact any eqdipment/ accident important to_ safety.' h is' review included FSAR sections 9.3,.11'.2, 11.4 and chapter 15.-
i 2.
The equipment associated with thik change is j
non safety-related and does not inpact any-safetyL related equianent. There is no new malfunction created by tais' change. 'Ihis was determined by -
]
review of.FSAR sections 9.3, 11.'2,:11.4 and chapter 15.
l 3.
There are no applicable Tech Spec bases associated with this change. This included review of bases B 3/4.4., and B 3/4'.7.
87-V1N0454
'Ihis modification added personnel access control turnstiles to the PESB1 Bldg.
1.
'Ihe Accident Analysis.section of the FSAR was i
reviewed. It was then determined that the installation of the security turnstiles to the PESB.does not effect the areas of the FSAR
~
discussed in section 15. Also, FSAR Section 13.6 details the industrial security requirements,.
t section 13.6'as well as the VEGP Physical Security Plan was reviewed. No decrease in the. security effectiveness would'be realized as a result-of implementing this change.
'23 i
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- 1989 NNUAL REPOKf - PART 2 t
10 CFR50.59(b)' REPORT 4
4 2.
The installation of Security turnstiles in the PESB does not increase the possibility of an' accident as described in the FSAR,_specifically section 15 " Accident Analysis".
3.
The LOO Bases were reviewed in sections, 2.0, 3.0 and 4.0 of the Tech. Spec. It was then detennined that ' the. addition of-these' Security '
+
turnstiles will not reduce the margin of-safety as-describsd in the Tech.--Spec.-
j l Core Drill Requests :2330'and 2331 required pen 87-V1N0455 t
seals to Unit -1 Control Bldg.-walls betwen Rooms R180/R185, R131/RA63, and-RA63/RB38.for-i installation of Unit 2 electrical conduit inside the Unit 1 protected area.
1, Penetration's will be made and~ sealed in such a j.
manner that no degradation in the ability of the
-facility's floors / walls to meet their design.
-i requirements will occur. This change will not-increase the probability of occurrence or- ~
consequences of an accident as described in FSAR section 15. No changes to FSAR hazard Analysis is required.- This change does not' involve-t any equipment or canponent.. The building's hazard -
analysis and structural design are not affec_ted by this change. There is no degradation of the
" defense'in-depth" Fire Protection Program as a result of this DCP.-
2.
The change does not create the possibility for j
any accident or equipment. malfunction not previously.
L described'and analyzed in the FSAR. Change.does y
not involve any equipment or camponent...The structuralidesign and hazard analysis are not ~
,j affected by this change.
3.
This design change meets the margin _of safety.
defined by-the Bases for the Vogtle Technical' Specification.: This is based.on a-review of the bases in Sections 2.0, 3.0,;and 4.0.
4 E
7 24 1
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11 1989 R@UAL REPORT - PART 2 4
10 CFR50.59(b) REPORT 87-V1N0458 he design change adds a time relay.in parallel with 3
the starter interposing relays on Service Air compressors 1-2401-1E-501, 502, 503 and 504, his -
addition will allow the start circuit to be maintained until the residual bus transfer
-is complete and the selected compressor / compressors have auto restarted. his eliminates the need for operators to manually restart cmpressors following a transfer to restore instrument air-system header pressure.
1.
The design change allows for auto start capability of air compressors following a residual bus i
transfer. This allows for system restoration without manual' operator actions. ' he system is required for plant startup and normal operations:-
however pneumatically operated valves essential for.
safe shutdown and accident mitigation are-designed.to assume fail-safe position upon loss-of air pressure. Therefore, the compressed air system is not required for safe shutdown or following a design bases' event. Therefore, this modification does not increase the probability of i
occurrence or the consequences of an accident or
.i malfunction previously evaluated. Bis response included a review of FSAR Section 9.3.1.1.1 and 9.3.1.3 and Chapter 15, 2.
All valves that require instrument air for.
operaticn that are ecsential for safe shutdown and.
accident mitigation are designed to-assume a fail L
safe position following a loss of air pressure..
The auto restart capability does not affect normal system operation.
It will-allow system restoration to be accomplished without operator action. This addition does not create any new malfunctions not previously evaluated. This response follows a review of FSAR. sections 9.3.1.1.1 and 9.3.1.3.
3.
Plant design is such that no plant equipment relies upon the canpressed air system to perform its safety functions thus there is no safety design basis for the system per FSAR section 9.3.1'.1.1.
The canpressed air system is not addressed in the basis for any Technical S3ecification. 'Ihis included a review of the basis for Tech Spec.
section 3/4.7.
(
25 Y
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F II 1989 M NJAL REPORT PART 2 10CFR50.59(b) REPORT 87-VCN0460' h is modification made changes in Normal Chiller Circuitry (1591) such that the chiller will autmatically reset after a loss of control power.
1..
.The normal chillers-(1591) are not considered in any.section 15 FSAR. accident and are not important to safety.
'2.
We normal chilled water system described in FSAR 9.2.9.2 has.no safety design bases.
3.
The normal chilled water system is not a_part r
of any Tech Specs.'. bases.
l 87-VIN 0465 This change involves the assignment of penetration seal nmbers and selection of appropriate seal j
details for existing unsealed conduit penetrations.
1.
'Ihe proposed change does not' increase the probability of occurrence or consequences of,the.
malfunction of any equipment or component assumed:
to function in accidents analyzed in the FSAR. :The change involves the assignment of penetration seal numbers.. There is no degradation of the " defense-in-depth" Fire Protection Program'as=a result of this DCP.
2.
This change involves the assignment of penetration
~
seal numbers. Eis modification does not create the possibility of any accident-or malfunction-of a different type than previously evaluated in the FSAR.
3.
This design. change meets the margin of safety.
defined by the Bases for the Vogtle. Technical Specification. This is based on a review of the Bases in Sections 2.0,'3.0, and 4.0 -
87-VIN 0466 The change required the drilling of holes in the Unit 1 Aux. Bldg. floor between roms UC-C07/R-C49 and Control Bldg Walls between roms R-128/R-132, R-132/ Unit 2 R-117-E & R-307/R 308. These holes were_added to provide.for the' installation of Unit 2 electrical conduit inside the'. Unit I l
protected area. The permanent penetration seals were installed per this DCP and provided a hazard-rating equal to or greater than required for the-l walls penetrated.
j t
26 J
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,3 l
-11 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b): REPORT f
1.
Penetrations will be made'& sealed in such a manner
.that no degradation in the. ability of.the
. facility's floor / walls to meet their design'-
requirements will occur. B is change will not increase the probability of. occurrence or consequences of an accident as described'in FSAR section 15. No changes to FSAR hazard analysis.is-required..This change does'not involve any equipment or components. h e building's hazard:
analysis and structural design are not'affected by tk.a change. : h ere'is no degradation of the
" defense-in-depth" Fire Protection Program as:a-
' result of this DCP.
2.
B is change _does not create the possibility for any-accident or equipment malfunction not previously_
-described'and analyzed in the FSAR..- Change does
~
not involve any equipment or components. he structural design and hazard analysis:are not-affected,by this change.
3.
This design change meets the margin of safety.-
defined by the. Bases for-the Vogtle Technical;.
Specification.- This is based on a review of the Bases in Sections 2.0, 3.0, and 4'.0.
87-V1N0468 The change required theLdrilling ofl holes in the
~
Unit 1 Control Building walls between' Rooms R-126/R-122, R-199/R-160 & R-164/R-128.- These I
holes were added to provide-for the installation of Unit 2' electrical conduit:inside the Unit.1 protected area. ~ h e-permanent penetration seals were installed per this DCP and provided..
a hazard rating equal to or greater than required for the walls penetrated.
T 1.
Penetrations were made and sealed ~in~ such a manner
~i that no degradation in the ability of the.
facility's walls to meet their design requirements 3
will occur. This change will not increase the probability of occurence or consequences of an accident as described in FSAR section 15'(Accident Analysis). No changes to FSAR hazard analysis ~is
]
required. This change does not involve any equipment or cculponents. The building's hazard i
analysis and structural design are not affected~
by this change. Bere is no degradation of the
" defense-in-depth" Fire Protection Program as a result of this DCP.
V 27
1 II" 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT-i 2.
This change does not create the possibility for any accident-or. equipment malfunction not' previously described and analyzed in the FSAR.
1 Change does not involve any. equipment or c mponents.
3.
Tliis design change meets the margin of safety defined by-the Bases for_the Vogtle Technical Specification.. This is based on a review of the Bases in Sections 2.0,- 3.0, and 4.0,.
'87-V1E0469
- Added breaches and penetrations seals to Control'
' Building, plaster walls between rooms R-325/R-325,.
R-117/R-117 and R-120/R-119 Aux. Bldg.= Plaster:
~
walls between rooms: UC-D06/UC-D06 and Turbine
- Bldg concrete block wall"between' Units 1 &' 2.
These' breaches were recuired for installation of Unit-2 conduit insice the Unit _1 protected area.
- 1..
Penetrations were made and. sealed-in such a manner thatnoTdepadationintheabilityofthe facilities walls to' meet their design requirements will' occur. This change will not increase the -
probability of occurence or consequences of an accident as described in FSAR section'15. No changes to FSAR hazard analysis'is required. This change (does not involve any equipment or.-
components. The building's hazard analysis and.
r structural design are not affected by thisLchange.
There is.no degradation of the " defense-in-depth" Fire Protection Program as a' result of this DCP.
2.
This' change does not' create the possibility for any accident or equipment malfunction not greviously described and analyzed in the FSAR. Cange does not involve any equipment or c mponents.- The structural design and hazard analysis are~not affected by this-change.
3.
This: design change neets the margin of safety-defined by the Bases for the Vogtle Technical Specification. This is based an a review of the -
Bases in Sections 2.0, 3.0, and 4.0..
88-V1N0002 This modification replaces the existing ATI power supply located in the ESF sequencer panels (1-1821-U3-001,002) with a more reliable 28-
s III 1989 ANNUAL REPORT - PART 2-10 CFR50.59(b)_ REPORT-
- power supply by the recamended vendor (Eaton:
Consolidated Controls), as~a result of previous-failures. Tne sequencer panels are safety-related-class >lE and are located in the Control Building.
in a mild envirorment. The change affects internal:
wiring only. The replacment power supply is qualified las class-1E.
1.-
As shown in FSAR Table 15.0.8-1 the emergency' power syste is assmed to function in accidents in
<FSAR 15.6 Also per FSAR 15.0.8 the Diesel Generator is assmed to start in seconds upon-an L.O.P..The accidents described in FSAR 15.6:'
(based upon the asstoptions of Table. 15.0.12-1) state that no single active failure will prevent the reactor protection syst e fr a functioning
-properly. Therefore, this changefdoes not increase-
)
the probability.of malfunction of required-equipment, i
2.-
Accidents' described in FSAR-15 do not discuss-.
the malfunction of the ESF sequencer or standby:
power system. Worst case assumptionsf(Table 15.0.12-1)- consider the failure of entire protection trains. A single failure of the sequencer would not adversely affect thel consequences,of an accident.due to redundancy.- This change does'not affect the.
function of the' sequencer and therefore does not increase the probability of an accident.
1 3.
Tech Spec, bases-3/4.3.2,: 3/4.8.1 and 3/4.8.3 inply the requirements of the sequencer in relation to AC sources and ESF c mponents actuated by ESFAS; This-change does not adversely affect the. sequencer function and therefore does not decrease the margin of safety per the Tech Spec.
J 88-V1N0003 Backdraft dampers 1-1593-D7-101 thnt 108 which are located in the Auxiliary Feedwater Pumphouse have q
the blades. The blades were removed fr m the' backdraft depers in response to deficiency-1-87-3440 which identified the dampers as not in I
29
t II 1989 AMRIAL REPORT - PART 2, 10 CFR50.59(b) REPORT:
the open position as required by note 5 on P&ID' 1X4DB227. The openings in which these backdraft-dampers are located provide for a natural circulation airflow path for the accident mode of ventilation of the turbine driven Auxiliary Feedwater pump rom and.the normal /mergency mode of the motor driven Auxiliary Feedwater ptmp-roms. The backdraft dampers are-intended to-prevent excessive inflow of air and are'not required to prevent nonnal infiltration -since :
the original design required the-damper blades to -
remain normally open for natural circulation.- This DCP will make the changes provided by Temporary-Modification 1-87-481 permanent.
1.
The removal of blades from the subject backdraft danpers will not affect-the safety function of the l
Auxiliary Feedwater Punphouse WAC system or-
)
increase the probability of a malfunction of'any i
equipment assumed to function in accidents
^
analyzed in the FSAR. The design temperature will not be affected by reoval of the damper'--
blades and thus, will not prevent the Auxiliary l
Feedwater Ptmphouse WAC-system fra performing.
its intended safety function. This includes a review of FSAR Sections 2,-3, 9.4.8 and 15.
2.
Removal of the blades will. assist the natural:
circulation within the Auxiliary.Feedwater P mp House. This will further assure the maximum rom temp is not exceeded. All current WAC design parameters will be maintained with a
the backdraft damper blades removed.
3.
This change will not decrease the margin of safety as defined in the Tech Spec. bases.for the Auxiliary Feedwater system (TS 3/4.7.1.2) and the Auxiliary Feedwater Pumphouse ESF WAC system j
(TS 3/4.4.7.13). The Tech Spec. bases for the i
Auxiliary Feedwater Pumphouse ESF WAC system 1
requires that the system maintain an ambient air
~
temperature which does not exceed the allowable i
temperature for continuous duty rating for i
equixnent served by the WAC-system. Removal i
of t2e backdraft damper blades will not affect-the ability of the ESF WAC system to perform its intended safety function.
I s
30
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' II-1989 ANNUAL REPORT - PART 2 '
10 CFR50.59(b) REPORT, 88-V1N0010 Relocation of the line supervision cmponents-in the security system.
.1.
'Ihe Accident Analysis section of the FSAR'
-(Section 15) was reviewed to determine that thet implementation of this DCP would not increase the probability of, or the consequences of an accident as described in the FSAR. Also, FSAR Section 13.6 details the~ industrial security requirements,-section 13.6 as well-as the VEGP Physical Security Plan was reviewed. No decrease u
in the security effectiveness would be realized'as i
a result of inplementing this change.
- 2. -
A review of Section 15 of' the Vogtle FSAR was performed in determining that this design change would not create the possibility of an accident.or malfunction.other than previously 1
evaluated in the FSAR-
~
3.
A review of the Bases in Sections 2.0, 3.0 and=
1 4.0 of the Vogtle: Technical Specification was performed. It was then determined that this; design change would not reduce the margin of safety as defined by the Technical Specification.
88-VIN 0013 Ibis modification eliminates the automatic. runback i
of turbine power to 50 percent on-a trip ~of.one circulating water pump..
1.
The change does not involve safety related equixnent and only effects safety related equipment j
to tae extent that the possibility of.a reactor.
trip due to a secondary plant transient is changed.
This change is being performed to minimize the-magnitude of the transient which would result =from a cire water pmp trip, since under most.
atmospheric conditions operation could continue.-
')
at much more than.50 percent power. Accordingly, the probability of a~ reactor trip is decreased, i
2.
Any sequence.of-events remains bounded by a:
turbine trip, which is analyzed in Chapter 15.
If turbine-load is not manually. reduced by a 3
' sufficient amount, a turbine trip'will occur _on high condenser pressure'.
3.
Turbine load control is not discussed in the
~
1 1-
, Tech Specs. or basis. No turbine or reactor protective trips.are being changed.-
]
1 l
31
9 t
)
t II 1989 ANNUAL REPORT - PART 2:
10 CFR50.59(b). REPORT 88-VIN 0019 This DCP added access control and alarm annunciation caaabilities to Diesel Fuel Oil Storage Tark Door D-101.
l.
The Accident Analysis-section of the FSAR
[
was reviewed to detennine that the t
implementation of this DCP wuld not increase t'ae probability of, or the consequences of an accident as described in the FSAR Also, FSAR Section 13.6-1 details the industrial security requirements,.
section 13.6 as well.as the VEGP Physical Security Plan was reviewed. No decrease in the security :
4 effectiveness w uld-be' realized as a result of inplmenting.this change.
2.
A'heviewofsection!15oftheFSARwas-performed in determining that this design-l change would not create-the possibility of an 4
accident or tralfunction other than previously evaluated in the FSAR.
3.
'A review of the Bases in section 2.0, 3.0_and 4;0 of the Tech Saecs. was performed.. It was then determined that tais design change muld not -reduce -
.the margin of safety asLdefined:by the Tech
, 4
. Specs.
88-V1E0027 Added Core drills and penetration seals to Control Building walls between: rooms R-C06/
R-C07,'R0164/ Exterior, R-185/ Exterior anC R-A44/R-A23 and Auxiliary Building floor l
between rooms UC-D07/R-D53, these. penetrations:
are required for thetinstallation of Unit:2~
electrical conduit inside.the Unit l' protected area. 'The permanent ">enetration seals were -
installed per this DC? and provided.a hazard rating equal to or greater than required'for-the walls' penetrated.
4 1.
This change does not involve any equip 2ent or.
components. Penetrations were unde arxl sealed in such a manner that no degradation in the-ability of the facilities' walls / slab to neet
(
their design requirements will occur. This j
change will not increase the probability'of' occurence or consequences of an accident as described in FSAR section 3.5 and 15 -(accident analysis). The building's hazard analysis 'and structural design are not affected by this change.
1 This included review of FSAR sections 3 and15.
32
A ye i.
II 1989 N MIAL REPORT ~- PART 2 10 CFR50.59(b) REPORT h ere is no degradation of'the " defense-in-depth"-
Fire Protection Program as a result of this DCP.
2.
E is change does not involve any equipment or-cmponents. The building's hazard analysis and structural design are not affected by this change.
- h is included review of FSAR sections 3 and 15 3.
21s design change neetsithe margin of safety i
-defined by the Bases for the Vogtle Technical-
= Specification. 2 1s is based on a review of the' Bases in Sections 2.0, 3.0,_and 4.0.
t 88-VCE0029 Four (4) of the Unit:2 ERF_ Computer _CRT Displays ~'
(2t41) were deleted and-four (4) of the Unit 1-ERF Cmputer 11 MIS were made ccxmni to both units.
This was accomplished by.dnstalling a manual A-B Coax switch A-HS-6277,.in the Technical Supoort Center,(TSC) Computer Rocan. The Coax switc'a allows. the four (4) conman Rt4IS to be connected -
to either the Unit l'or. Unit 2 ERF Computer via
- the ccnnunication ports 1(Linkports) LP-2A:and '
LP-2B, 1.
This nodification does not increase the probability of, occurrence or the consequences of an: accident or.
malfunction of equipment inportant -to safety -
previously evaluated in the FSAR. Bis modifi -
i cation does not. affect.any conponent/ equipment - -
that mitigates the effect of any. accident. described i
in the FSAR'Section 15. nis nodification does not increase the probability of occurence or conse-quences of the malfunction of_any component /-
equipment assumed to function in accidents analyzed in Section'15 of the FSAR. Though-'the ERF Computer System is assumed to be operational in all nodes of operations, including' accident and-post accident conditions' (Emergency Plan Section H
^
and FSAR Section 7.5) ANU is designed to be highly reliable and qualitatively camparable (with regards:
to accuracy) with class 1E systems it is non-Q, 62J, and performs no safety-related functions.
2.
This nodification to the ERF Couputer : System does not create the possibility of an accident or..
malfunction of a different type than previously L
evaluated in the FSAR Section 15.:
- c IIL 1989 ANNUAL REPORT - PART 2 10 CER50.59(b) REPORT 3.
h is m dification to the ERF C m puter System does J
not reduce the margin of safety as defined in the basis'for any Technical Specification, he ERF Ccruputer System is not addressed;in the Technical Specifications, h is included a review of Tech-Spec. 3/4.3-
{
88-VCN0032-his change'added early warning fire detection (one ionization smoke detector) to Water Treatment Building storage room.105.
I
.1.
his change ~ involves. adding an additional' smoke.
- detector. in the Water Treatment Building, which -
does not increase'the probability of occurrence or consequences of an accident' described in the
.FSAR. FSAR section 15.0 was reviewed and requires no change. This change involves adding an; additional smole detector in the Water Treatment 1
Building, which does not increase the probability his included;quipment/
of occurrence or consequences of y
review of component malfunction.
FSAR chapter 9 and 15.
2.
Thisl change does not: increase the probability'of any' accident or' equipment malfunction.--No new possibilities or unanalyzed scenarios are created.
-t his included review of FSAR_ chapter 9 and 15, E
3.
The safety limits and settings discussed in section 2.0, 3.0, and 4.0'of the Tech Specs. do not deal with-Fire Protection. Therefore, there is no decrease'in the Tech Specs. margin-of safety.
88-VIE 0038 Reroute ftre protection line :2-2301-L4-211-2 1/2" which supplies fire hose station 2-2301-R4-159.
to allow sufficient clearance away from main-steam line 2-1301-L4-005-26", Aux R159 south MSIV room.
1.
B is change involves rerouting a fire protection line that supplies a. Unit;2 fire hose station, which'does not increase the probability of1 occurrence or consequences of an accident described in!the FSAR.' FSAR sections 9.5.1'and 15.0'were reviewed and require no change.
2.
This change:does not increase the probability of-any accident or equipment malfunction. No new I
34
a II j
1989 ANNUAL REPORT - PART-2::
10 CFR50.59(b) RE NRT possibilities or unanalyzed scenarios are created.
- his is based on a review of FSAR sections 9.5.1 and 15.0, a
3.
The safety limits and settings discussed in section 2.0,- 3.0, and-4.0 of the Tech Spec. do not deal with Fire Protection. Therefore, there is no' decrease in the Tech Spec. margin.of safety, c
88-V1E0039 h is Design Change added tie-ins to the Unit 1-j scoped service air headers for' Unit 2 users in the 1
Unit 2 Control Building. The change involved a-tie-in from a service air. header in the Unit l-j Control Building to a service air header in the Unit 2 Control-Building. The loads that are being t
- supplied by this tie-in are fortsupervision of fire protection pre-actio,n valves in sprinkler systems in the Unit 2 Control Building ;
1.-
he service air systen supplies compressed,-
' filtered, dry and oil free air to outlets t
throughout the plant for-the operation of..
~
pneumatic; tools and-other service' air requirements.
t The service air' system is not required for the safe shutdown of the plant. The probability of-occurrence or consequences-of an accident or malfunction is not changed. :This portion of the-service air system is not assuned to function in accidents analyzed in the FSAR. This-included a-review of-FSAR section 9.3.1.1.1 and Chapter 15.-
2.
The plant is-designed such that no plant _ equipment' relies upon the ' service air? system to perform its -
safety function. This tie-in effects only super-a vision of fire protection pre-action valves in; sprinkler systems. The design charige does not-create -the possibility of any. accident or.
i malfunction of a-different type previously-evaluated. This is based on a review of FSAR section 9.3.1 and Chapter 15.
3.
This modification does not involve any. change'to the' Technical Specifications.- The service air system does not have al safety design basis as it is not expected to operate during accident conditions.
This is based on a review of FSAR section 9.3.1.1.1 and a review of the bases for Tech Spec. 3/4.3.
4 35 s
^
.. c; II, 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REIORT 88-VIN 0047 Modify fire suppression sprinkler system 058 and 102'to remedy spray pattern obstructions. System 098 will also'be modified to add an inspector's test connection.1 These fire suppression systems are non-safety related Project Class 629.
1.
- 1his change involves the modifications of sprinkler i
systems _due to obstructions which presently block l
spray patterns and inpact coverage. ~ This change will not' increase the probability of occurrence or consequences of an accident described in=the FSAR.- Sections 9.5.1'and 15.0 were reviewed and-require no change. There.is no degradation of-the-f
" defense-in-depth" Fire Protection Program as a-result of this DCP.
l 2.
This change does not increase the possibilities or unanalyzed scenarios.= This is. based on a review of FSAR sections 9.5.1 and 15.0..
3.
The safety limits and settings discussed in section-2.0, 3.0, and.4.0 of the.VEGP Tech Specs, do not deal with' fire protection. Therefore,.there is no-decrease.in the Tech Specs. margin-of safety.
88-VlN0050 ThisDCPprovideddetailsforinstallinh penetration seals in the north east wall of t
the Turbine Building Level A area that leads-into the Service Bldg. ccmuunications cunnel.
This provided a 3 hr fire separation barrier between the Service and Ibrbine Eldgs as L
required by Nuclear Mutual Limited Insurance.
1.
The change does not increase the probability of occurrence or consequences of an accident described in the FSAR.' This included a review of sections 15 and 9.
The change does not increase the probability of occ.urrence, or consequences of the malfunction of any equipment a
or component assumed to function in accidents analyzed-in the FSAR. This included a review of sections 15 and.9.
2.
The change does not create the possibility of an accident or equipment / component malfunction not described and analyzed in the FSAR. This is based on review of sections 9 and 15.
i 1
l 36 3
II' 1989 M M ML REPORT PART 2-10 CFR50.59(b) REPORT 3.
W e change does not decrease the margin of safety:
' defined by the bases of the Technical-Specifications. E is is based on review of the-1;
, basis of Tech Specs. 3/4.
88-V1N0051 h is modification is part of the human factors i
evaluation to eliminate-selected annunciator windows which appear on other auxiliary panels
-and display the same basic'information found on-
'f other Control-Rocm: sources or have been determined
,to be a nuisance to the operating staff.~
Additionally varioussetpoint changes were madec to decrease the rate of alarm incidence.-
1.
The modification 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.-1This conclusion is based on review of. chapter 15 of the FSAR~and-other sections.
2.
W e modification does not create the possibility-
[
of an accident or malfunctico 'of a different t,fpe; than previously. evaluated in the FSAR.-1This conclusion is based on review of chapter 15 of the i
FSAR and other sections.
l 3.
We~ modification does not reduce the margin of-safety as defined in the bases for Technical Specifications sections 2.0,: 3/4.2, 3/4.3 and 3/4.7.
88-V1N0059 The modification was inplemented to prevent the I
possibility of tripping the Steam Generator' Feed Punp Turbines when bulbs were replaced on the tri) solenoid valve status indicator for the turbines. Previously, if an incandescent.
light bulb was inadvertently used instead of the u
l neon bulbs, full voltage would be applied across the trip. solenoid and a turbine' trip would occur. h e change adds a time delay relay and resistor so that incandescent bulbs will be used in the future. This will eliminate the trip possibility 1-)
associated with changing a light bulb.
J 37 e
n
J 1
.II 1989 ANNUAL REPORT - PART 2' L
10 CFR50.59(b): REPORT l
l:
l 1.
- The modification changes the type of light bulb used for the SGFPT trip solenoid status lights.-
l The addition of the resistor and time delay does not inpact the purpose of the status light. The a
SGFP is assmed to fail-in ~section 15.2.7 of the FSAR.. This. change does not change the probability,
~ f any equipment malfunction considered in FSAR
- i o
chapter-15 or 10.4;7..
2.
The addition of this modification does not. affect-
.i nomal system operation. The function of the -
status light circuit does not change. Based on a-s review of FSAR chapter 15 and 10.4.7, this change, does not create-the possibility of an accident or malfunction not already described in the FSAR.
3.
Tech Spec. bases'for 3/4.4 and 3/4.7 does not.
I' address the type of lighting required and as the -
oystem function or: operation has not changed, the margin of safety as defined in the bases for Tech Specs._is not affected. V1E0060 Lower Fire Protection isolation valve 2-2301-U4-191 from elevation 208'-0 to elevation 203'-0".- lIn-addition', correct: associated line and valve nmbers l
1 which were erroneously designated. LThis change allows practical installation and accessibility of isolation valve 2-2301-U4-191 and sprinkler system-4 068 preaction valve. This change also corrects-inconsistencies in line/ valve numbers, t
1.
This change involves _' changing the elevation of a fire protection' isolation valve which does not increase the probability of occurrence of.
consequences of an accident or any equipment /
canponent malfunction. There is no degradation of the " defense-in-depth" Fire Protection Program as a-result of this DCP.'
2.
Thischange.doesnotincreasetheprobability_of_
any< accident or equipment malfunction. "No new possibilities or unanalyzed scenarios are created. This is based on a review of FSAR sections 9.5.1 and 15.0.
1 l
38
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m II' 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT P
3.
- Le safety limits and settirgs discussed in sections'2.0, 3.0 and 4.0 of the VEGP Tech Frec.-do not' deal with fire protection. Werefore,.
o dene is no decrease in the Tech Spec. margin:
of safety.
88-VCE0061
- h is modification provides a second shielded-
.r Rev 1 vault for.a:second set of demineralizer vessels,. relocation of process radiation shields, 1"
and removal-of an existing hydraulic pedestal crane from the Alternate Radwaste Building (ARB).-
1.
Eis change does not change the operability of the liquid radwaste process system and will not increase j
the probability _of-occurrence or consequence of an accident described in the FSAR. Addition of the-second demineralizer vault.does not effect the heavy.
load anal,61a of Chapter 9.
There is no change to the other accidents postulated in the FSAR' Chapter
-15.
2.
A review of Chapter 15 shows that this change does not effect the possibility of an accident or equipment / component ma) function not presently.
described and analyzed in the.FSAR.
3 3.-
This change does not decrease the margin of safety defined in the bases of. Technical-Specification section 3/4.11.-
88-V1N0063 his change involves modifications >to the:
Rev 1 Circulating Water Chemical Injection-system (1410),
Utility Water system (2419), NSCW ChemLcal Injection system (1413)'and Wrbine~ Plant Sampling system (1311). Specifically,.
A.
Add a Chemical injection' station to Circulating Water system.
B.
Add biocide,; dispersant and anti-corrosive injection capability to Nuclear Service Coolire Water syaram.
C.
Replace Turbine Plant Sampling system corrosion coupon rack.
.l D.
Addition of safety shower and eyewash to Circulating Water chemical injection skid.
39
II<
.1989 M E AL REPORT - PART 2-10 CFR50.59(b) REPORT 1.
'Ihis change does not affect the results of the toxic gas evaluation of FSAR Section 2.2.
This
-change has no effect on the design performance of requirements of the NSCW system. For these reasons,.
this change has no effect on the consequences or probability of any accident described in FSAR section 2.2.-'or 15.-
-2.
The change does not create a new component malfunction that effects any system required to mitigate the effects of-any. accident analyzed in FSAR Chapter 15.
3.
There is no change to the margin of safety as" defined in the basis of Technical Specification 3/4.7.
88-VCE0067 Three Catalytic Hydrogen Rec mbiners are provided; for Units 1 and 2.
One recombiner per unit is used-
.i in each main process loop to remove hydrogen fr a j
the hydrogen-nitrogen fission gas mLxtures by-oxidation to water vapor which is removed by condensation.- he third reembiner is available on a standby basis. The units are self-contained and are designed for continuous operation.c The _
existing circuitry currently provides a permissive for 1HV-0115 and-2PV-0115 if the third recombiner i
is in use. This condition could lead to an-explosion because the unit recombiner, if it trips, l
will continuo to receive hydrogen. ' Installing a l
" Unit 1/off/ Unit 2" selector switch on Recombiner No. 3 control. panel (A-1902-P5-CHC) will'give a.
j permissive to HV-0115 for the particular unit a
which the rec mbiner unit (A-1902-D6-002) is-servicing.
y 2
1.
The proposed change does not-increase the probability of occurrence or consequences of any accident as analysed in the FSAR Chapter 15.
I 2
'Ihe change does not create the possibility of any malfunction not already analyzed in the FSAR. 'This design change will give a permissive to HV-0115 for l
the particular unit which the Catalytic Hydrogen Recmbiner unit (A-1902-D6-002)'is servicing.-
This is based on review of FSAR section 11.3 and 15, i
40 i
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..II 1989 ANNUAL REPORT 'PART 2--
10 CFR50.59(b) REPORT 0
3.-
There is no change in the margin of safety as.
defined by/4.3.3 and Tablesthe bases of the Tech Specs, sectio 3.3.10, 3 3.3-10, 4.3-6.
l 88-VCE0071 This change revises' air flws to certain areas /=
Unit I roms in the 1539 system (Startup designator GK-05) to meet the heat loads per the Project -
-1 Calculations X4C2111V01 Rev.:6 and X4C2111V02 a
Rev. 3.
The work scope'will involve adjusting:
the fan speed and_ balancing dampers: removing temporary duct caps and opening of fire dampers'.
J i
l 1.
This change:in the' air flows'does not increase i
the probability of occurrence or conse ences of an accident or equipment /cmponent mal ction as-analyzed in FSAR section 9.4.1.
The changes 1are:
very minor in nature and'are being made to meet--
- the requirements of, dual-unit operation per the flow diagrams.
-2.
This design change in flow rates does not create the possiaility of an~ accident or equipment-i malfunction'not already malyzed in section l
9.4.1 of the FSAR. System performance is not'
.inpacted.
3.
This change does not_ decrease the margin of safety as. defined by the bases of Technical Specification section'3/4.7.
88-VCE0072 This change involves revision of'the air flow-t diagram to redistribute conditioned air to serve both Unit 1 and Unit 2 Control Rooms for~ normal i
and mergency modes of operation. This will involve adjustments to fans and volume dampers to achieve design air flow.. This change is-in accordance with the original design concept ~of L
a dual unit control rom operation. The physical boundary wall erected to separate Unit 1 operation is removed to reconfigure the area as a dual unit embined control room during the norral-and emergency modes of operation.
41 i.
.o
d II 1989 ANNUAL REPORT PART 2 10 CFR50.59(b) REPORT
- 1. -
his-change to redistribute conditioned air for a.
dual unit combined control roam does not increase-the probability'of occurrence or conse ences of' i
an accident or equipment / component mal
- etion, h is is based on a review of control room-ventilation isolation' system described in'section-7.3.6 of the FSAR and the HVAC FD4A of. FSAR table -
i 6.4.4-1.-
4 i
2.-
This design change does not create the possibility of an accident or equipment malfunction not already r
analyzed in the FSAR. This is based on. review of-FSAR sections 6.4,-7.3, 9.4 and'15, 3-here is-no change to. the margin of. safety. defined '
'in the basis of Technical Specification 3/4.7.6.
a
' he Technical Specification change to section i
~3/4'7.6 has been accepted by.the NRC in amendment' 9 to the operating license.
88-VCE0073 21s change revises air flows :(Flow diagrams AX4DB254-1 and AX4DB256-1) to certain areas /
Unit I rooms in the 1533, 1535, and;1537.'
systems to meet the heat loads per the project
' calculations X4C2111V01 Rev. 6 and'X4C2111V02 Rev'3.
to providc. consistency between the documents i.e.
the project calculations and the air flow diagrams.
4
- /
El.:
he change in the-air flows 1does not increase the probability;of occurrence or consequences of an accident or equipment / component malfunction-analyzed in-section 9.4.1', 6.4-and 15 of the' FSAR. The: changes are very minort in nature t
and are being made to meet the. requirements of i
dual Unit operation 2.-
he change in the flow rates does not create-
- the possibility.of an accident or equipment-j 1 malfunction not described or analyzed inl
.section 9.4.1 of the FSAR..Overall~ system.
performance is not impacted.-
I 3.
Eis change does not decrease the margin of safety defined by the bases of Technical 1 3
Specification lsection 3/4.7.
42-s Y
y
-11 1989 ANNUAL REPORT -'PART 2 10 CIE50.59(b): REPORT 88-V1N0074 Remve' support VI-1214-028-H001 frcan Containment and Auxi Building Drains ~ - Radioactive System;- The j
support is in the Aux. Bldg, area 3G of level A.
1 Due to a stress reanalysis the support-is no longer required.
1~.
This. change does not affect system' function or-operation and-therefore does not increase the'
- probability _of. occurrence'or the consequences of j
any accident described in' sections 9.3 or 15.0.- ~
j Stress analysis indicates that pipe'. stresses are within code allowables.--
i 2.
This change does not affeet:any system, equipment
-j or component'sifunction or operation and-based on i
a review of FSAR sections 9.3 and-15.0, would not v
create the possibility of an unanalyzed 'or undescribed accident or equignent/ component -
malfunction.
33 3.
This change does not affect any syst!em/dquipment.
-(
function or operation and'therefore does not affect'
'the safety margin defined in Tech Specs, sectioni 3/4.11.
e a
88-VIN 0082 This is'a non-safety related change to relocate settlement markers to areas of_ easier access. -The new markers will-be installed using the standard details shown of: drawing AX2D94V001.. 'Ihe original l
markers will remain in place and(will be available
-(
if any' future correlations are needed. FSAR section 2.5.4.13.2 details the monitoring requirements for' settlement of the power blocit; structures.-. Many of:
the markers are located within vital areas and radiation control zones. Ihis change relocates markers outside of these areas,;thereby reducing i
the need for assistance'from' Security and Health Physics, and' reducing surveyor man-hour requirements for each monitoring period.
1.
There are no changes to any equipment or ccruponents assuned to function in any accident analyzed in the FSAR. This=is based on a review of FSAR section 15.
0 5
1 43 q
m II 1989 NE E L REPORT - PART 2 10 CFR50.59(b) REPOKI 2.
The design change does not create the possibility of any canpanent malfunction not already analyzed in the FSAR. This change does not alter any equipment or systan empenents. This is based on review of FSAR section 15. There is no effect on the settlanent progran discusted in FSAR section-2.5 and SER section 2.5.4.4.3.
3.
This change has no effect on the Tech Spec.
There is no decrease in the margin of safety as defined in basis of Tech Spec. 3/4.6. The
'gL settlanent markers will not alter any equipaent included in the Tech Spec.
88-V1N0084 A.
This DCP relocates two ladders in the Auxiliary Building to provide better access fran roan R108 to RA06.
1he u to be installed as ramveble. pper ladder is 31 B.
. Raoves part of the nonorail, and two k
supporting beans in RA06 of the Auxiliary Building. During normal operations these beams and nonorail will be stored on the floor of rocm RA06 and will be reinstalled whenever the hoist for the Letdown Heat lbtchanger needs to be operated.
7 Relocating the ladders will provide inproved access to RA06 fran R108. Part of the monorati and two supporting beams have to be renoved to relocate the ladders.
1.
This change will not increase the probability.of occurrence or consequences of an accident as described in FSAR section 15 and section 6.
All materials uwet fire protection and Seismic Category I requiremente. There is no change to the heavy loads analysis of 9.1.5.
2.
This change does not involve any equipment or camponents required to function after an accident.
The hazards analysis is not affected by this change. The proposed change will inprove access to roan RA06. There is no change required to the accident or hazard analysis of FSAR sections 9.1.5 end 15, 44
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x n II 1989 Alt G L REPORT - FART 2 10 CFR50.59(b) REPORT 3.
his change has no effect on the basis for the Technical Specification. Le ladders and nonorail have no effect on equi <:nent ration.-' his is
' based on review of Tec:vtical ecification 3/4.7.
88-V1E0099 h e Hot Machine Shop and surrounding area was served with ventilation by the Fuel Handling-Building WAC system with taqwrary duct during -
b.
Unit 2 construction. his DCP pemanently ~
connected the Hot Machir.a Shop to the' Unit 2-Atmiliary Bldg. WAC syste. Sme of the design flows were also increased to reduce the =4' man normal taip from 1004 to 94 F.
1.
Since this change reconfigures the WAC system for the referenced areas to be in accordance with the.
original WAC system desip, this proposed change-does not increase the probability of occurrence or consequences of an accident described'in the FSAR,
-sectiona 9.4.2 and chapter 15.0. he proposed change will only rewori non seismic WAC ductwork in a non seismic area. Rus,'the aroposed-change does not increase the probaaility of occurence or consequences of the malfunction of any equipment or caponent asstned to function in' accidents analyzed in the FSAR, section 9.4.2 or chapter 15.0.
2.
h e proposed change reconfigures the referenced areas to be in accerdance with the original WAC system design and thus this change does not create-the possibility of an accident or. equipment /
ccnxnent malfunction described and analyzed!in the FSAl section 9.4.2 or chapter _15.0.
-3.
. We aroposed change restores. the WAC system for the lot Machine Shop and its adjacent areas to be in accordance with the original WAC system desigt, h us, this change does not decrease the margin of safety defined by the bases of the Technical Specification 3/4.7 and 3/4.9.12.
)
45 1
Il 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT l
l 88-VCE0104 This DCP modified several of the Unit 2 CCTV camera
- towers, l.
The accident analysis section of the FSAR (section 15) was reviewed to determine that the implementation of this DCP would not increase the probability of or the consequences of an accident as described in the FSAR.
Also, FSAR Section 13.6 details the industrial security requirements.
Section 13.6 as well as the VEGP Physical Security Plan was reviewed.
No decrease in the security effectiveness w'f.id L realized as a i
result of implementing this ch.nge.
2.
A review of Section 15 of the bit!
FSAR was performed in determining that th4 design change would not create the possibility of an accident or malfunction other than previously evaluated in the FSAR.
3.
A review of the Bases in Sections 2.0, 3.0 and 4.0 of the Vogtle Technical Specification was performed.
It was then determined that this design change would not reduce the margin of safety as defined by the Technical Specification.
88-VIN 0lll This DCP installed a security barrier on Unit 1.
1.
The accident analysis section of the FSAR (section 15) was reviewud to determine that the implementation of this DCP would not increase the probability of or the consequences of an accident as described in the FSAR.
Also FSAR Section 13.6 as well as the VEGP Physical Security' Plan was reviewed.
No decrease in the i
security effectiveness would be realized as a result of implementing this change.
2.
A review of section 15 of the Vogtle FSAR was performed in determining that this design change l.
would not create the possibility of an accident or malfunction other than previously evaluated in the FSAR.
l 46 o
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II 1989 ANNUAL REPORT - PART 2 10 CfK50.59(b) REKET 3.
A review of the Bases in Sections 2.0, 3.0 and 4.0 of the Vogtle Technical Specification was perforned. It was then determined that this design change would not reduce the margin of a a safety as defined by the Technical Specification;-
88-VCE0112 his DCP renoved tenporary ductwrk and caps, installed duct sections betwen the units, and capped tenporary openings, -as required to sever the tanporary connections between the Unit I and Unit 2 Auxiliary Building HVAC systems (1551 Supply and 1553 Exhaust) and conplete the Unit 2 systen for operation in its permanent configuration, his work was performed in conjunction with the renoval of the' Unit 1/ Unit 2 temporary security barriers per DCP 88-VCE0100.
1.
Eis change brings the duct systen to the permanent configuration asstmed in section 9.4.3 and in the section 15. accident analyses of the FSAR.
2.
his DCP pern:its coupletionldf the systan in its nonnal permanent configuration for which'tle 1
evaluation of FSAR sections 9.4'.3 and.15 were perforned.
3.
h is DCP has no effect'on the' margin of safety defined in the' Tech Spec. section 3/4.7. bases.
88-VIE 0113 Fire protection line 2-2301-L4-205-2 1/2" will be-routed to allow novament of main feedwater line 2-1305-L4-057-16"- without interference.. h e fire protection line is non-safety related project class 629 and supplies water. to Fire Hose Station (MIS) 2-2301-R4-123. his nodification will not inpact the operation'or response of either the fire protection or main feedwater system, and does not 1
require a Tech Spec or FSAR change.
1.
his change involves rerouting a fire protection.
line that supplies a Unit 2 MIS which does not increase the probability of occurrence or-consequences of an accident or equipment /
com)onent malfunction described in the FSAR.
FSAA sections 9.5.1 and 15.0 were reviewed and require no change. '1here is no degradation of the " defense-in-depth" Fire Protection Progran as a result of this DCP.
47
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II 1989 ANNUAL REPORT - PART 2 10 CIR50.59(b) RU GKi 2.
This nodification creates no new possibilities or unanalyzed scenarios. This is based on a review of FSAR sections 9.5.1 and 15.0.
3.
- he safety limits and sattings discussed in section 2.0, 3.0 and 4.0 of the VEGP Tech S not deal with fire protection. -Therefore, pecs. do there is no decrease.in the _ Tech Specs, nargin of safety..
88-VIE 0116 h is change'provided desi m details-to seal various-aenetrations in the Fuel hndling Building Rocxn 1-C05, Auxiliary k ilding Room UC-A06 and Control-Building Rooms R-121 and R-128. W ese penetrations' were addressed m Deficiency. Cards 1-88-3074, 1-88-3129, 1-88-3512 and 1-88-3880. These seals-meet all hazards and Fire Protection design criteria.
1.
This change will not increase the probability of occurence or consequences'of an accident as described in FSAR section 1.5 and section 6.
' All-materials used neet the Fire Protection-Requirements._. h is change does not involve any equipment or canponent. The hazard analysis is not affected by this change; h e change provides-the penetratim seals required by the plant-Fire Protection program. There is no change to the Fire Hazard analysis of chapter 9 or AccidentL Analysis of chapter 15.
2.
This change does not create the possibility for
- any accident or equipment malfunction not previously described and analyzed in'the FSAR.
The material used neets the Fire Protection requirenents of FSAR section 9.5.1.
3.
This design change neets the margin of safety
. defined by the Bases for the Vogtic' Technical-Specification. - his is based on a review of the -
Bases in Sections 2.0, 3.0,_and 4.0.
89-V1N0001'
'No lateral restraints added to line 1-1201-068-3/4" to prevent excessive vibration. ' This line is the drain line for the pressure relief valve loop seal'...
1 48 i
II 1989 ANIE L REPORT - PART 2 10 CFR50.59(b) REPORT 1.
he design change does not. affect the probability or consequences of any accident described in the FSAR.
B ere is no change to the pipe break analysis of.
section 3.6 since the pressurizer loop seal stress analysis (hence break analysis) does not change and the 3/4" drain line is less than the mininun size (1") considered for the HELBA. Le addition of
-the two pipe supports does not affect the accident analysis of chapter 15.
2.
here'is no change to the HELBA analysis of chapter 3.6.
Since addition of the pipe supports does not affect operation'of the pressurizer, PORV or pressure instrumentation, there is no new cmponent malfunction created that is not already analyzed in' the FSAR.
3.
Bere is no change to the' margin of safety as defined in basis of Tech Specs. 3/4.4.3, 3.4.9.2, 3/4.4.6 and 3/4.7.8.
89-V2E0004 A.
Three (3) welds on ISO 2J4-1201-068-01 are to be built up and ground to confonn to ASME Sec. III NC-3673.2 (b) - 3 sketch d, to reduce SIF @ weld from 2.1. to'1.3.
B.
Modify spring support at D.P. 20, V2-1201-068-H609 to r id vertical restraint, to make systen more ri id.
C.
Modify lateral support @ D.P. 30 to' include a new vertical support @ 30A.
1.
E is change does not affect the occurrence or consquences of any failure or equipment asstaned
'to function in the FSAR accident analysis.
h ere is no change to the pipe break analysis of Section 3.6.
h e grinding of 3 welds or modification of supports does not effect operation of the pressurizer (chap 5),
power operated relief valve, or pressurizer instnznentation (chap 7)=.
Therefore, there is no effect on the accident analysis of Chapter 15.
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II 1989 N NUAL REPORT - PART 2 10 CIR50.59(b) REIORT 2.
There is no change to the HELIA analysis of chapter 3.6.
Since grinding welds or nodification of supports does not effect operation of the pressurizer, PORV or pressurizer instr mentation, there is no new component malfunction created that is not already analyzed in the FSAR.
3.
W ere is no change to the margin of safety as defined in basis of Tech Specs. 3/4.4.3,3.4.9.2, 3/4.4.6 and 3.4.7.8.
89-V2E0005 Changes were made to_several mating surfaces in the Unit 2 heater drain ptmps to reduce the potential-
.for steam Icaks frcan the shell of the ptmp to at2nosphere.
1.
We operation of the ptmps is not affected.,
There is no impact on any mapter 15 accident.
analysis.
2.
The-change only effects non-safety related equipment and does not change its operation or.
transient response. Accordingly, no new accidents or malfunctions are created.
3.
The heater drain 2mps are not included in the basis for any Tecmical Specification.
89-V2E0006 Replacment of original steam dtmp drain pot instrunent trees with a conductivity level detectico systm.
1.
This change does not affect any safety related.
equipment nor does it increase the probability-or consequences of an accident.. The steam dtmp.
syst e is not safety related and'as'such is not-relied upon in any accident analysis. 'FSAR-sections 15 and-10,4.4-reviewed.-
2.
This modification inproves reliability of the-steam dump system. It does not increase the possibility of an accident other than previously addressed in FSAR.. New materials installed meet specifications stated in section 10.3.6 " Steam and Feedwater System Materials"; 1here is also no increase chance of flooding per review of section 3.4.
50
l II-1989 ANNUAL REPORP - PARr 2 10 CFR50.59(b)LREPORT 3.
This modification will not-reduce the margin of safety as defined in Tech Spec.' sections _2.0.
(bases), 3/4.3.4 (Turbine Overspeed Protection) and 3/4.7.1 (Turbine Cycle) reviewed.
89-V2E0008
. ("9000cfm) point of 2FSL-12045 from 510 mV Change set to 590 mv ('.4800cfm). Eis will allow the flowswitch to. actuate during the alloted 30. seconds following an autostart of the Control Room Emergency HVAC s st m.
2FSL-12045 is located in Control B1
. R-B18, 1.
This change affects the setpoint of 2FSL-12045 only and does not affect its other design requirenents for postulated accidents and therefore does not' increase the' probability of occurrence or consequences of an accident or equipment /ccoponent malfunction described in the FSAR. FSAR sections 6.4, 7.3 and 15 were reviewed and require no change.
2.
his change'does not create the possibility of an accident or equipment malfunction. No new possibilities or unanalyzed scenarios are created.
.7his is based on' review of FSAR sections 6.4, 7.3 and 15.
3.
Based on review of.the basis for Cmbined Technical Specifications,~section 3/4.7'.6.there is no change to the margin of safety.
89-V2E0009 This DCP removed the Harmonic Vibration Dampeners from the NSCW Transfer Punp Motors 2-1202-P4-007 &.
008 and reinstalled the shorter motor head bolts.
For each transfer ptmp additional supports were installed for the ptmp driver stand. 'These supports consisted of (2) horizontal members from the ptmp house walls to a ring around the.ptsp driver stand.
1.
The NSCW transfer ptmps are assumed to function in accidents as described in FSAR sections 9.2.1 and 9.2.5.
This design change is made to enhance the reliability of the ptmps and has no effect on any equipnent or-ccmpanent other than the transfer atop. The modifications were vendor approved and aave been analyzed and found not to adversely affect the seismic qualification of the pimps.
51
II 1989 NEAL REPORT - PART 2 10 CFR50.59(b) REPORT 2.
Le design change affects only the operating vibration characteristics of the NSCW Transfer unp/notor assenblies. hechangeisenvejoped
- ry the existing descriptions and analysis an FSAR section 9.2.1, 9.2.5, and 15, and no new accident possibilities or malfunctions are created. The nodifications are designed so there are not 2/1 possibilities.
3.
Based on review of bases section 3/4.7.5, this change does not decrease'the margin of safety as defined by the Tech. Specs.
89-V2E0010 Reroute non-safety related nitrogen supply lines 2-1224-109-2" to 2-1224-110-2" to seal injection backflushable filter system to facilitate the removal of bonnets on valves 2W-41232B and 2HV-41326B respectively. The backflushable filter system is not given credit for operation in the FSAR accident analysis. The reroute of the piping did not affect the seismic integrity of these class 212 lines.
1.
By providing adequate maintenance access, this change enhances equipment reliability and decreases the probability of equipment /couponent.
malfunction and occurrence or consequences of an accident analyzed in FSAR chapter 15 and sections 11.2, 11.4 and 9.3.
2.
h is change only reroutes existing piping. No new malfunction or accident can be created by this change. Pipe supports are (nutffected. Lese lines are class 212 for seismic and pressure boundary reasons. This included review of FSAR sections 9.3, 11.2, ada 11.4 and chapter 15..
3.
No credit is given to the backflushable filter systun for any accident analyzed in the FSAR. The Tech Spec bases of B 3/4.4 and B 3/4.7 are not impacted for this change.
89-V2E0012 Changing the setpoint valve fran 120 F to 1357 on the BQP Rack Switches -
(Plant Tag Nos - Description) 2TSH-15212G - Isolates 2HV-15212C SGB/D 21SH-15212L - Annunciator Blowdown 21SH-15216L - Annunciator Blowdown 52
II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 2TSH-15214L - Annunciator CVCS 2TSH-15215L - Annunciator CVCS Switch setpoints were changed due to temperatuias experienced in the rocin being close to 120'T in normal plant operations, This proximity to.
setpoint of switch to roca environment temperatures had led to unnecessary isolation and nuisance alarms.- Raminder of associated switches had been.
changed under Field Change YFCRB-7473 and 7474 during construction /startup activities.
1.
Raising the steam generator blowdwn and CVCS letdown isolation temperature switch setpoints do not effect the consequences of any accident analyzed in FSAR Section 3.6,
- p. 3F and Chapter 15 as listed in the safety anal sis report.
2.
man the setpoint of the tanperature switch wit ts design range does not create any potential for a camponent malfunction not previously_
reviewed in FSAR Section 3.6, App 3f or Section 15 as stated in the safety analysis report.
3.
Since there has not been any change in the-safety channels _or equipment quali.ficiation,.there is no decrease in the margin of safety for Technical Specification 3.3.3.11.
89-V2E0020 The two existing porous snubbers are on the main high and low legs of f1 w instruments in the Stator Water Cooling Panel 2-1326-P5-HSC. The porous snubbers are to be added between the branch line of each instrument and the 3-way manifold valve.
This is to facilitate calibration of instrtunants without affecting the pressure drop to.others.
1.
This change does not affect any safety related component nor does it increase the probability or consecuence of failure of such cmponents as described in FSAR section 15.0~and 3.5.1.3.
2.
This change does not create new accident or equipment / component malfunction possibilitics.
This change does not affect any turbine trip-functions of pressure switches or turbine over-speed protective functions. FSAR sections reveiwed 53
[
11 1989 At W A1. REPO E - PA R 2 10 CFR50.59(b) REPORT 3.5.1.3, 10.2 and 15, 3.
h is change does not affect Tech Spec, bases 3/4.3.4 or 3/4.7.1.
89-V2E0025 h is DCP changes the control circuitt,/ for 13.8 KV Rev 1 breaker 2NAA09 (2-1825-S3-0AA feeder to Field Support Building transfonner kNAA09X) in the following manners (1) the breaker auxiliary contacts for RCPs #1,13 are removed from the breaker trip circuit (2) the time delays for reclosure pennissives are changed to delay reclosing (3) the actuation circuit _for TD relay 262-1 is nodified to prevent inadvertant reclosure.
1.
W e accidents analyzed in FSAR 15.2.6 and 15.3 consider the loss of the non-1E 13.8 W systs.
his DCP has no detrimental inpact upon the 13.8 W or related systems and actually inproves reliability.- herefore, this change does not increase the probability of occurrence or consequences of the malfunction of any equipment-assuned to function in FSAl!. accidents.
2.
his DCP inproves the reliability of the 13.8 W-system and the automatic bus transfer schane and therefore does not create the possibility of non-analyzed accident or equipnent malfunction.
3.
The non-1E 13.8 W system is not discussed in the Tech Spec, therefore, this change does not decrease any margin of safety defined therein.
89-V2E0026 Route and support 1-1/2" drain lines frcin the Turbine Plant Sanple Coolers (equipment tag nos.
2-1311-P5-CRA and 2-1311-P5-CRB) collection trays directly to the Turbine Building drain system.
1.
Routing and supporting non-safety related drain lines to the floor drain in the Turbine Building:
does not increase; the probability or consecuences-of an accident described in the FSAR, based on a-review of section 15.0. he ability of the plant to accmplish a safe shutdown will not be affected by this modification.
2.
It does not appear that this nodification in any '
way will increase the possibility 'of an accident or malfunction occurring that will lead to an unsafe-54
{
i
II 1989 AMulAL REPORT - PART 2 10 CFR50.59(b) REPORT condition. h is is based on a review of FSAR section 15.0, 3.
' Based on a review of the Tech Spec basis, including section B 3/4.7,-there is no applicable margin of safety for these non-safety related piping changes.
89-VCE0033 his change involves an addition of a " free field" strong mH.on accelerograph near the River Intake Structure. We instr eent will be located approximately 2000 ft, northeast of the Unit 1 Contaiment. A small concrete retaining wall and pad is required. An enclosure box and a heater /thermatat are provided to protect the instrument and a battery charger and power source is provided for power. This design change allows for direct free field m asurem nts of a seismic occurrence for use in emparing observed -
seismic response of particular structures to the design response.
1.
his change has no inpact on accident analyses or equipment imaortant to safety, his includes a review of FSA3 sections 3.7, 8.3, and chapter,15.
2.
his change has no impact on any accident analyzed or not analyzed, he added instrment has no control functions and is a " stand alone" non-safety related unit. This review included FSAR sections 3.7, 8.3 and~ chapter 15.
3.
This modification increases the confidence level for the systen data and has no safety function role. herefore, the change does not inpact the unr in of safety defined in the Tech Spec bases B 3 4.3.3.3.
89-VCN0037 This m dification involved the removal of the Rev 1 secondary locking device-(mechanical canb lock) fnxn the Burnable Poison Rod Assanbly (BPRA)
Handling Tool for both Units.. W e. secondary locking devices were removed in accordance with Westinghouse camb lock removal procedure FCN GAEO-40514 and GBEO-40502.
-)
55 l
l' II ON.
R Le old design configuration had two locking features used to hold the Burnable Poison Rods in proper aligment. The primary locking feature is acccuplished by the pull down forces _ associated with existing magnets. The secondary locking device feature was acccmplished by use of a mechanical ccxnb lock.
1.
W e probability of occurrence or consequences of accidents previously evaluated in the FSAR will not increase since no mitigating credit is taken in the accident analyses for the BPRA tool. he BPRA handling tool.is not relied upon to perform a safety function nor does it effect any equipment relied upon to perform a safety function..
This included a review of FSAR section 9.1 and all Chapter 15 accidents.
2.
Since the tool will be used for the same intended design function and its performance will not be degraded, there is no possibility of creating a new accident other than previously evaluated in the MiAR.
3.
There is no Technical Specification which has its bases for the margin of safety. determinate on the BPRA handling: tool. This included a review of Technical Specification 3/4.9.6 and 3/4.9.7.
89-V2N0055 Provide reinforcing sleeves on 1-1/2" discharge-pipes of the Dic pumps 2-1615-S4-501-P1 & -P2.
Add horizontal restraints for DiC tubing 2-1615-537-1-1/2" and 2-1615-541-1-1/2'.-
1.
These proposed changes do unt nodify the design-or operaticn of any equipment assumed to function in accidents analyzed in the FSAR, Chapter 15.
Therefore, these changes do not increase the probability or consequences of a malfunction of.
any equipment assumed to function in accidents analyzed in the FSAR. This was based on review of FSAR section 3.5.1 56
t if;
_p H '_
, y _,-'*'
j i
p l
4 1
i2 s
i II:
x 1989 ANNUAL REPORT - PART_2 10CFR50.59(bbREPORT 21 These changes do not add any new equipnent or change the operatim of'any equipment:that would y
a create'the possibility of:a new type of accident or j
an pasnt malfunction not described and ' analyzed in section 10.2.2 and chapter 15.
= 3.i 1The margin a safety-described in Technica16 Specification bases 3/4.7 and 3/4.3.4 are not 3-
- affected by these changes.
. 89-V2E0056f Feedwater HeatersT4A &-4B high level switches, Rev 1 Plant Tag Nos. 21EH-4341 and 21RI-4342 were! raised
' to actuate at 12' inches:below centerline of heater l-yyJ instoad"of orginal design point of = 26-inches below-heater centerline. rihis modification was done' because. Unit 1: normal level at 100%' xwer,in these' htrs isLat 22"'below centerline.4 Hadng the.
- hi level 10" above. the." normal level-will-el te nuisance alarms;and heater isolation-
~
while providing proper heater protection.
1.-
The'feedwater heater setpoint' change;ddes not1 inpact the heater protecticn provide and does not;
'inpact'any; accident analysis performed in FSAR section 15, 2.
The feedwater heater setpoint change does:not
'inpact the. switches' function to-isolate Lthe-heaters'sinlet valves,-nor does it create thel possibilityand 15 b-of an accident not analyzed:in:the FSAR section'10-a failure'of-.these
.non-safety related.y-couponents.a 3.
Tech Specs margin'are not. effected as described in Section 3/4.3 and 3/4.7. Heaters are not required.
This chan can not effect.any-item or component required or safe shutdown of the plant 89-V2E0065-
' Modification of pipe support V2-1301-176-H001 to prevent binding and allow free axial; pipe movement <
cl.
The piping stresses due to the proposed change.
are _still within code allowables. VEGP Design Criteria DC-1017 Rev. 5 specified "g". values = ;'
57 I
4
II.
1989 N EUAL REPORT - PART 2 10 CFR50.59(b) REPORT for valves are not exceeded. 'Therefore, the nodification of this pipe support does not increase the probability of occurrence or consequence of an accident described the FSAR Section 15.-
2.
the proposed change does-not nodify the design or operation of any equipnent asstmed to function in accidents analyzed in the FSAR and there is no impact on the failure node and effects analysis described in FSAR Section 10.3. Therefore, this change does not increase the probability or consequences of a malfunction of any equipment asstuned to 5.niction in accidents analyzed in the FSAR Section 15.0.
3.
Technical Specifications 3/4.7.1.5 and 3/4.7.8 do not define any applicable margin of safety for this pipe support change.
89-V2E0068 Add rigid supports to line nos. 2-1303-007-2" and 2-1303-008-2" to alleviate excessive pipe vibration d en valves 2HV-4302B & 2HV-4303B are open.
Pipe support tag ntubersi 2J1-1303-007-03-N03 2J1-1303-007-02-N07-2J1-1303-008-03-N03-2J1-1303-008-02-N08' 1.
The piping stresses due to the proposed change are still within code allowables. The pro3osed change is to non-safety related equipment. Taerefore, addition of pipe support does not increase the probability of occurrence or consequence of.an accident described in FSAR Chapter 15;-
2.
The proposed change does not nodify the design or operation of any equipment assumed to function in accidents analyzed in the FSAR. Therefore, this change does not increase the probability or-consequences of a malfunction of any equipment-asstmed to function in accidents analyzed in FSAR Chapter 15.
1 58
-l l
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~
11 1989 ANNUAL Kd u a - PART.2 10 CFR50.59(b) REPORT 3.
Technical Specification 3/4.7 does not define any applicable margin of safety for this pipe support nodification.
89-V2E0073 Modification of pipe support V2-1305-081-H013 to prevent binding and allow fire pipe novement by increasing the gap between the pipe and support
. steel.~
~
1.
W e piping stresses due to the proposed change are still within code allowables, he proposed change is to non-safety.related equipment.
Therefore, nodification of pipe support does not increase the probability of occurrence or consequence of an accident described in FSAR Chapter,15 and Section 3.6.
2.
Le proposed change does not nodify-the design or operation of any equipment assmed to function in accidents analyzed in the FSAR and there is no inpact on the failure node and effects analysis described in FSAR Section 10.4.7. herefore. this change does not' increase the probability or consequences of a malfunctim of any equipment assmed to function in accidents analyzed in a
FSAR Chapter 15 and Section 3.6.
3.
Tech Spec. 3/4.7 doec not define any applicable margin of safety for this pipe support nodification.
89-V2E0076 Pressure test point valves 21301X4874 and 21301X4941 upstream of the main turbirse steam-valves were deleted and the connection ont the steam line plugged. Vibration of.these-valves had caused a steam leak.
1.
his change has no effect whatsoever on
_ plant operation or transient response. Ttm test connectim is not discussed in FSAR section 10.2.
2.
This change does not involve any safety related equipment. It does not present any new accident possibilities based on a review of -
section 15 of the FSAR.
4 59-
.II 1989 AMEAL REPORT - PAIE 2 -
10 TR50.59(b) REPORT 3.
This pressure test point is not discussed in any Tech Spec. or in any basis,
. including-3/4.7.
89-V2E0078 This change involves installation of a pemanent-sheet metal structure outside the Unit 2 Contaiment equipment hatch. ' This includes provisions for pemanent normal lighting, security lighting, and power receptacles. Design methods are in accordance with Project Category 2 requirements.for a non-safety related installation.
This structure will provide a protected work space for equipment staging during outages.=
1.
It is highly.unlikely-that a cmponent of the sheet metal structure could became a tornado missle as defined in the FSAR sect',on 3.5 and appendix 3C. '
In the event that a missle is generated the seismic Category 1 structures are designed for postulated missle loading which enccupasses this occurrence.
Therefore, the change does not increase the:
probability of occurrence or consequences of.an-accident as described in-the FSAR section 15, 2.
The sheet netal structure is a permanent -
non-safety related installation which is not-associated with any safety related system. The-only Category 1 system in-the vicinity-of the structure are the containment shell and the missle shield door, which are designed to withstand loading in excess of any loads-generated by the collapse of the' sheet metal structure. Therefore, the proposed change does not increase'the likelihood or consequences of the malfunctico of any couponent or. equipment assmed to function in accidents analyzed in the FSAR chapter 15, 3.
There is no decrease in the safety marg'in of any plant syst es per review of the basis of the Technical Specification.
60 u _ __
o
l II 1989 AI N AL REPORT - PART 2 10 CFR50.59(b) REPORT 89-V2E0096 The design ch e replaced the existing Veritrax Pressure Tr tters with Rosecont nodel 1151GP transmitters for instrutents 2PT-507 and 2PI-508 (Feed Ptrup Discharge / Main Steam pressure). This change' eliminated oscillation in the Feedwater control systems by providing an instrument with danpening provisions which eliminate output oscillationduetosourceinputoscillations.
1.
The change replaces the type of transmitter being used to supply main steam / feed pump discharge pressure to the feed punp speed control system.
These transmitters are class 62J serving no safety-related function. The change does not affect systan operatice or the.way the system operates or responds. These pressure transmitters have no effect on the operation of any ecxnponent/
equipment assumed to function in the Accident analysis and therefore the change will not increase the probability of occurrence or consequences of an accident or malfunction previously evaluated. Response based on a review of FSAR section 15.2.7, 15.2.8 and FSAR section 7.0, 10.3 and 10.4.7.
2.
The transmitter changeout has no effect on safety related equipment, adds no new equipment nor requires any safety related equipnent to function differently.than previously analyzed. The change will not create the possibility of an accident or malfunction not previously. described in the FSAR.
System operation remain-the sane. Response was.
based on a review of FSAR section 10 and 15.
^
3.
The design ch e installs a different nodel pressure tran tter for better input to system control. These transmitters (2PT-507, 2PI-508) are not involved in a safety related function and do not affect the Tech Spec, bases 3/4.7,.
3/4.3 or LCO sections 3/4.7 and'3/4.3.
89-V1N0100 Modify"A" bus ducts, so that the penetration seal and renove the current xnetration frcxn Train between Diesel Generator rocxn R-103 and south wall of Ibnnel 1T4A are configured to allow 1" seismic novment. Likewise, provide for the similar configuration for Diesel Generator room R-101 to south wall of Ibanel IT4B penetration for Train "B" bus ducts.
61
II 1989 AtEUA1. REPORT - PART 2 10 CFR50.59(b) REPORT 1.
Le proposed change does not' increase the probability of occurrence or consequences of an accident or eouipment/couponent malfunction described in the FSAR. FSAR sections.9.5.1 and 15.0 were reviewed and require no change.
he change only relocated penetration seals from one side of a wall to the o h ere is no degradation of:the~"pposite side, defense-in-depth" Fire Protection Program as'a result of this DCP.
2.'
Ihis mdification. creates.no new possibilities or unanalyzed scenarios. ' Wis is based on a.
review of FSAR sections 9.5.1 and.15.0.
3.
The safety limits and settings discussed in section 2.0, 3.0,'and 4.0 of the VEGP Tech-S xc. do not deal with fire protection.
Taerefore, there is no decrease in the Tech Spec margin of safety.
89-V2N0103
.1he main turbine control cabinet was modified to-
~
renove the closing bias signal from'the control
- valves, three. intercept valves, and one stop valve. The change reduces'the probability
- of-valve closure due to minor voltage transients in the ERC power supply, a
1.
The change effects only the main turbine' steam control valves. Le change reduces the probability of an undesirable valve closure event. Based on a review of FSAR sections 10.2, 15.1,: and 15.2, accident probability is. not effected.
2.
The change only reduces.the chanses'for the turbine steam valves to unintentionally close.
Valve failures.in both-the opeofand closed position are bounded by secondary heat recoval increase and decrease events evaluated in.
chapter 15.
- 3..
The margin of safety is not reduced based on a review of the Tech Specs, bases for section 2 and 3/4.3.4.
62:
f' II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REIORT MI!OR DEPARTURES FRQ4 DESIGN (MDD's) 88-V1M008 Upgrade the Security system data input boards.
1.
The accident analysis section of the FSAR (section 15) was reviewed to determine that the inplementation of this DCP would not increase the probability of or the consequences of an accident as described in the FSAR. Also, FSAR Section 13.6 details the industria1' security requirements, section 13.6 as well as the VEGP Physical Security Plan was reviewed. No decrease in the security effectiveness would be realir,ed as a result of inplementing this change.
2.
A review of section 15 of the Vogtle FSAR was performed in determining that thin design change would not create-the possibility of an accident-or malfunction other than previously evaluated.in the FSAR.
3.
A review of the bases in Section 2.0, 3.0 and 4.0 of the Vogtle Technical Specification was performed. It was then detennined that this design change would not reduce the avgin of safety as defined by the Technical Specification.
88-V04011 Replace the carbon steel manway cover of the' demin, storage-tank with a stainless steel cover fitted with piping and-valves to allow for-sanpling. The stainless manway cover will have a 6" opening _at center fitted with a baffle on the liner side to prevent liner from being sucked into-opening. The cover will also have a 2" opening fitted with a sanple valve.
Also, the 4" underground distribution piping is to be reflected on design drawings.
1.
'1his change does not impact any accident analysis evaluated in cha?ter 15 of the FSAR and.does not increase the pro mbility of occurrence or the consequences of an accident or malfunction of equipm.ent important to safety previously evaluated!
in the FSAR.
63
I II 1989 ANNIRL REPORT - PART 2 10 CFR50.59(b) REK RT-2.
his nodification does not create the possibility of an accident or malfunction of a _different type than previously evaluated in the safety analysis report. Chapter 15 of the FSAR was reviewed.
3.
his nodification involves no safety limits, limiting safety syst e settings, limiting conditions for operation or surveillance requirements. It does not reduce the nargin of safety.as defined in the bases for the Tech Spec.
bases for section 2.0, 3.0 and 4.0 were reviewed.
88-V1M013 he nodification was to review the vital DC breaker.
settings to provide proper coordination and to ccxoply with the requirements.of Branch Technical position 04E3 9.5-1.
1.
Changing the setpoints'of the DC circuit breakers to provide proper coordination will not affect nonnal operation of the DC systm..his nodification only affects the response of the DC system to short circuit conditions,. in that selective sequential triap will now occur to isolate faults from the systm.z This change:is consistent with the failure analyses contained in FSAR sect. 8.3.2 and FSAR chapter 15:
therefore, this nodification will not' increase the probability of occurrence-or the' consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.
2.
Changing the setpoints of the DC circuit breakers to provide proper coordination will not affect normal operation of the DC syst m.. h is nodification only affects the response of.the DC systs to short circuit conditions in that, selective sequential' tripping.will occur to isolate a fault. This nodification is consistent with the failure analyses contained in FSAR sec 8.3.2 and FSAR chapter 15r therefore this nodification does not create the possibility of an accident or malfunction of a different type-than previously.
evaluated in the Safety Analysis Report.
3.
Changing the setpoints of the DC circuit breakers to provide proper coordination will not affect normal operation of the DC system.- This nodification only affects the response of the DC system to short circuit conditions in that, lj 64 i
i
.m_.:.....-,
o II 1989 A RIAL REPORT - PART 2 10 CFR50.59(b) REPORT selective sequential tripping will occur to isolate a' fault. his nodification is consistent with the bases for Tech Specs. 3/4.8.2: therefore the nodification does not reduce the nargin of-safety as defined in the basis for any Technical.
S cification in particular basis for Tech Specs.
3 4.8.2.
88-V1M015 Proteus Conputer Systen - Replace existing 2 L megaward drums and controller cards with new controller cards and 4 negaword solid state amory units.. Also install a tape backup device on one of the two solid state umory units, his change inproves system response time and reliability of the syst s. (he additional 2 megawords of nuory cannot be accessed until a resysgen of'the operating system is done).
1.
W e Proteus C m puter is not safety related and is not assmed to serve a safety related function in the plant's design bases analysis. _ herefore, this change does not increase the probability of occurrence or causequences of an accident or malfunction of equipment inportant to safety -
previously evaluated in the FSAR. Reference FSAR section 7.5.
2.
The replacement of the mass nuory devices ~ on the-plant cmputer does not affect the software or system operation, and therefore does not create the possibility of en accident or malfunction of a different type tluci previously evaluated in the TSAR. Reference FSAR section 7.5 and 15.0, 3.
his change is to the Proteus nuory hardware only and does not affect any software controlled functions. B erefore this change does not reduce the nargin of. safety defined by the basis for the-Technical Specifications. Reference Tech Specs, 3/4.1.3 and 3/4.2.1 -
88-V1M018 Disable the contact 3/4 of 1HS-0276A by lifting; wires C5 and-C6 of cable 1ABD47SA at terndnation cabinet 1ACPIO7. -(Tenninal Points TB4.64- & TB4.65)
Disable the contact 3/4 of 1HS-0277A, by lifting wires C5 & C6 of cable 1BBD47SA at tenninal points TB3.69 & TB3.70 of Tennination Cabinet 1BCPIl0.-
his change will prevent the Systen Status
.y Monitoring Panel (SSMP) frcxn illuminating during j
1 65
)
~
l i
II 1989 ANNVAL REPORT - PART 2 10 CFR50.59(b) REPORT normal operation of the Boric Acid Transfer System.
However, when the control of the Boric Acid Transfer motor is transferred to the shutdown panel by placing 4
transfer switch TRS LR to local position, the window on SSMP will be illuminated.
1.
Disabling the SSMP alarm function associated with the Boric acid transfer pumps has no impact on the function of any safety related equipment or components as evaluated in the FSAR accident analysis.
Therefore, the-probability of or consequences of an equipment malfunction have not been increased.
2.
Removal of the Boric Acid Transfer pumps from the SSMP alarm logic brings the system into compliance with section 7.5.5 of the FSAR.
The SSMP is only intended to alarm for automatically actuated ESF functions.
3.
This MDD has no impact on Boration System Technical Specification (3/4.1.2) or any other Technical Specification.
89 V2M001 This modification will enable the Unit 2 fault recorder j
to save information recorded during a generator undervoltage condition.
This wiring addition is internal to the fault recorder panel and connects the existing generator P.T. input to an undervoltage sensor.
1.
The' fault recorder is not safety related and is used only as an information gathering device.
A review of FSAR chapter 8.0 shows no mention of the plant fault recorder.
2.
This modification has no effect on plantLoperation or safety related equipment.
Reference FSAR chapter 8.0.
3.
The fault recorder is not mentioned in Technical Specifications (3/4.8).
It does not affect any equipment required for operation or safe shutdown of the i
plant.
89 VCM002 Three doors presently have windows measuring 96 square inches each.
This MDD will install an improved barrier.
L 1
I I
66 t
t e
1989 ANNUAL REPORT - PART 2 10 CFR:i0.59(b) REPORT-1.
A reviev of sections 3.0 to 12.5 of the FSAR was used in determining that the addition of the im) roved barrier in MDD 89 VCM002 will in no way increase tie probability of occurrence or the consequences of an accident or malfunction of equipment important to the safety of the plant, as previously evaluated in the FSAR.. Also, FSAR Section 13.6 details the' industrial' security o'
requirements, section 13.6 as well as the VEGP Physical Security Plan was reviewed.
No decrease in the security effectiveness would be realized as a result of -
implementing this change.
2.
The addition of the improved barrier on the doors will in no way create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR. A further review of FSAR sections 3.0 to 12.5 was performed to aid in this determination.
3.
The modification performed here was reviewed against Technical Specifications sections 2.0, 3.0 and 4.0 which are the bases for LCO conditions.
This review was used to determine that the addition of the improved barrier-does not reduce the margin of safety as defined in the Technical Specification, 89-VlM003 Changed wiring on IMSH 13205, steam packing exhaust filter high moisture alarm to provide annunciation on relative humidity rising above setpoint.
1.
The SPE filter as described in FSAR 9.4.4 is not safety related.
2.
FSAR 9.4.4 states that SPE system is not considered important to safety. The change cannot create an accident as the alarm circuit only is affected.
3.
The alarm circuit only is affected. The ability of the unit to function as ventilation exhaust treatment system as defined by 3/4.11.2.4 is not affected.
67
II 1989 A!MIAL REPORT - PART 2 10 CFR50.59(b) REPORT 89-V04004 To revise nany of the secondary plant annunciator setpoints to a nore useful value. Presently see are too conservative and sme are too liberal. Existing setpoints are based on
=
chemis control antici ated when the setpoints were ori inally develo 1.
The FSAR gives the ch mistry. limits in Tables 10.4.6-4 and 1.4.6-2 and 10.3.5-1. All existing and new setpoints are within the limits
=
prescribed in the FSAR,. thus they have been
=
previously evaluated.
=
2.
We acknowledgment or actions taken in resaanse to these annunciators is not associated wita any accident evaluated in the FSAR. Since the new setpoints are bounded by the see criteria as the
=
old setpoints, no new accidents or malfunctions z_
are created.
=
3.
Secondary plant chmistry specs are not a part of any reviewed Tech Spec. bases. However, basis-3/4.4.7 requires steady state steam generator chmistry -to be within the " steady state limits (as provided by the vendor)". h is annunciation is within EPRI, Westinghouse,.and the FSAR limits.
89-V1M005 On Relay 160 (P.T. failure relay for Emergency'.
Diesel Generator), switch wires at points.(5 & 7)-
and (15 & 17). W is change is required to allow operation of the 160 relay per its intended..
design, h e relay came fr m the vender wired per normal A-B-C rotation. his change will correct to the Georgia Power C-B-A phase rotation 4 1.
his change will allow operation of the 160 relay in the Diesel Generator relay protection schme per the intended design. It will inhibit unwanted tripping of the Diesel Generator due to PT failure.
Wis change, therefore, decreases the probability of an accident.
68
II 1989 MNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 2.
W e 160 relay will now be configured as required by the original design. It orovides a blocking fmetion for the 140 relay ("oss of Field) and 151 relays sovercurrent) in case of a blown FT fvae. he generator protective functions are discussed in section 8.3.1.1.3 of the 1% R and are not affected by this Design Change.
3.
All protective devices will function as specified by original design and are not affected by this change, he Diesel Generator System will provide its intended safety function and therefore, the margin of safety as discussed in Technical Specification Bases section 3/4.8 (Electrical Power Systans) will not be reduced.
89-V2M006 This nodification changes the original parts specification for the fabrication of the seismic su3 port brackets for penetrations 2-1818-H3-;P71 (ports 1, 5,.12,18) and 2-1818-H3-P19 (ports 1, 5, 11, 12) to support PERMS and NIS cable connections, he support bracket changes frcin.25 to.312 and the spacing between these holes changes from 1,56 to 2.00.
1.
This design change involves a minor nodification.
to the PERMS feedthrough seismic su3 port (Conax) which includes U-Bolt and bracket caanges. :Bese changes have no impact on the functions and characteristics of the PERMS systans as described in section 11.5 of the FSAR. This description' does not address PERMS in so far as the level of detail that this nodification involves. his nodification does not increase the probability of occurrence'or the consequences of an accident or malfunction of ecuipnent inportant to safety previously evaluatec in the FSAR..
2.
his nodification makes minor design changes to the.
physical dimensions to the U-Bolt and bracket for the PERMS feedthrough seismic. support. These design changes adhere to applicable IEEE design-codes and installation specifications, f
69 6~
II 1989 AIM AL REPORT - PARr 2 10 G R50.59(b) REPORT 3.
A review of Tech Spec, section 3/4.3.3 (monitoring instr oentation) was done. These nodifications to-the U-Bolt and bracket emponents of the PERMS support does not reduce the margin of safety'as defined in the basis for any Tech Spec.
89-V2M007 This nodification involved replacing the Unit 2 Sigma Refueling Machine cable reel with a similar reel which has a distributor. We distributor assures that the cable will be evenly distributed along the reel as it is retracted. h is nodification resolved the problem of the cable
' jaming as it was retracted.
1.
The Sigma' Refueling Machine is a Non-Safety related Seismic Category 2 systen as described in DC-1010, Rev. 5.
The nodification improves the Refueling Machine without affecting its load capacity.
This nodification of the Refueling Machine will not degrade its perfonnance and, therefore, will not
-inpact the Fbel Handling Accident Analysis or operation of Safety Related Equipment as described in the FSAR.
2.
B e modification' involved replacing ths cable reel with a similar reel whch has a distributor to inprove maintainability and reliability.- The modification did not degrade: the perfomance of.
the Machine. Also,'it does not inpact the Fuel Handling Accident Analyses or the operation of the Safety Related Equipment,-
Based on the above, this nodification does not create the possibility of an Accident or Malfunction of a different type.
3.
The nodification has no effect on the margin of -
safety defined by the bases of Tech Spec. Section 3/4.9.6.
70-t1 A
II 1989 REAL REPORT - PART 2 10 CFR50.59(b) REPORT 89-Vm008 Revise Waste Gas System Hydrogen reccubiners oxygen and hydrogen circuitry and tubing to event analyzer micro-fuel cells'from being ed by overpressurization or burn-out, his wi prevent the analyzers from being damaged when a waste gas canpressor trips.
1.
FSAR section 15.7.1 deals with loss'of waste gas due to waste gas decay tank failure or failure of associated piping, his change modifies the -
sanple tubing without canpromising the accident analysis of the FSAR. h e modification does not increase the probability of occurrence or the consequences of an accident or malfunction of.
equipment important to safety as previously evaluated in the FSAR.
2.
his nodification does not create the possibility of an accident or malfunction of a different type previously evaluated in the FSAR.
FSAR accident analyses of chapter 15,were reviewed with attention to section 15.7.1.
3.
This nodification does not reduce the margin of.'
safety as defined in the Bues for Tech Spec.
section 3/4.6.4. B is change,_in no way, restricts the capability of maintaining operable the hydrogen reccxnbiners with regards to detection and control as defined in the Technical Basis section 3/4.6.4.
89-V2M009 Change type of light fixture in.. room RB-88 from a mark W F to a mark M I and change the mounting detail to "B".
W e circuit feeding the light will be changed to 2NLP37-1 from 2NLP38-1. In room RA-74, change the light fixture from mark WF to mark WI, change the mounting-detail to "8", and change the feeder from 2NL?33-3 to 2NLP37-3.
1.
We arobability of an accident will not be increased-by the change of fixture type, nounting detail, or feeder. he FSAR does not address specifics of these lights in it's analysis. Reference FSAR s
section 9.5.3 and chapter 15.
i 2.
Le modification does not create the possibility of an accident, he change in feeder and nounting type of these lights does not' affect function of these lights.- Reference FSAR section 9.5.3 and chapter 15.
71 e
r 11 1989 ANNUAL REPORT - PART 2 10 CFR 50.59(b) REPORT 3.
The lighting in these rooms is not described in any Technical y.
Specification or bases for any Technical Specification.
Therefore, this char.ge in no way reduces the margin of safety as defined in the bases for any Technical Specification.
89-VCM010 This modification changes breaker setpoints to prevent spurious tripping of lighting breakers.
I 1.
The modification 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. The modification changes only breaker setpoints tc prevent spurious tripping and to provide proper breaker coordination (Ref. section 8.0 and chapter 15 of FSAR).
2.
The modification does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
The modification changes only breaker setpoints to prevent spurious tripping and to provide proper breaker coordination (Ref. section 8.3 and chapter 15 of I
FSAR).
3.
The modification does not involve a change to Technical Specifications.
The modification changed only breaker setpoints to provide coordination and prevent spurious tripping.
Breaker setpoints are not covered in Technical Specifications.
Reference Technical Specification section i
3/4.8.
89-V2M0ll This modification installed a lanyard support on hanger V2-1201-053-H006 in order to accommodate the addition of one I
lanyard at Node Pt. 50 on line 2-1201-053-16" for thermal growth and vibration testing.
This support will house the lanyard for monitoring snubber movement on hanger V2-1201-053-H006.
I 72 1
i
II 1989 ANNUAL REPORT - PAKr 2 10 CFR50.59(b) REPORT 1.
This nodification was installed on V2-1201-053-et006 to nonitor pressurizer surge line urnenent as aart of power ascension testing requirements of FSA1
]
section 14.2.8.2.48 (thernni expansion test).
l This lanyard and support were designed and installed i
to teet the requiranents of FSAR section 3.9 B.3.4 (canpanent supports) and comply with the applicable ASME Sect III codes. This modification did not increase the probability of occurrence or the I
consequences of an accident or um1 function of i
equipment in1portant to safety previously evahated j
in the FSAR.
2.
The lanycrd support us installed by 89-V2M011 was s
used to monitor pressurizer surge line novement as l
part of power ascension testing of Unit 2.
This 1
support was seismically designed and installed to I
meet ASME requirements, and therefore, creates no new accidents or unlfunctions not addressed in FSAR section 15.
3.
Tech Spec. bases 3/4.7.8 gives the LOO and surveillance requirements for snubbers. The installation of this lanyard support per 89-V2M011 does not reduce the unrgin of safety as defined in the Tech Spec.
89-V1M012 The addition of the " Start Delay" and "Mcnitor
^l Delay" setpoints for 1XI22564A and 1XI22564B to the Instrument Setpoint Sheets.
1.
The modification to add vibration monitor setpoints does not affect operation of the ESF chillers at all. The vibration nonitor provides Control Room annunciation only and will not cuase an ESF chiller to trip off or fail to start. Therefore, no increase in the probability of occurrence or consequences of an accident or equipment malfunction previously evaluated in the FSAR will result.
2.
As stated previously, the subject vibration provides Control Roan annunciation only and can l
not impact ESF chiller operation.
73
i 1
-l II 1989 ANNUAL REPORT - PART 2 4
10 CFR50.59(b) REPORT
'1 3.
Revi w of Technical Specification 3/4.7,11 indicates that valves, rom coolers,. chiller and pmp-are checked per surveillance requirements for position-I and actuation on' SI signal.' 'Ihe setpoints for the '
- vibration monitors does not change Tech Spec.
requirements. Vibration monitors provide alarm functions only. h erefore, the margin of safety.
l is not reduced.
89-V2M013 Changed model n mber of Pyco RTD from.-
122-3046-12-6 tc 122-3027-24". 21s.is a 24" long probe with a protective sheath.-- h is change will place the sensing tip of 2TE-2618:
y into the active airflw region _ of 2-1505-A7-001'-
e plenm to more accurately control the :tmperature t
of the air passing through the ductwork.'
1.
'Ihe proposed change does not degrade the safety of the system and will not increase :the probability of occurence or consequences of an accident 1 described in the FSAR.. The longer probe'will allow contaiment purge supply: system to supply' design _
air. temperature to Containment. DThe Containment:
Purge Supply is not important.to' safety as^
described by 9.4.6 and is not;a part of section 15 Accident Analysis..
. 2.
The proposed change involves a non-safety related.
system and will not increase the probability of:
occurence or consequences of;a malfunction of safety-related equipment.or c m ponent previously 1
evaluated in the FSAR.-<By supplying proper
- temperature air into Contaiment a more desirable evnironment will be ' established'for equipment
inside.
~ '
3.
The Containment Normal Purge Supply._ unit and associated temperature control: loop is=not a part of any Technical Specification design bases.
89-V1M014 Revise wiring in the BETA annunciator panel-(i.e, i
Condensate Demin Panel-1-1414-P5-FDP) to allw
~
Control Room alarm AKB17D03 to clear once the t
local alarm at!PS-FDP is acknowledged. This will 4
allow the. Control Rom to be aware of subsequent alarms at the Condensate Denin. Panel..
i 74
y h
II
'1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT:
1.
Revis of the FSAR accident analyses of chapter 15 shows that the modification does not' increase the probability of occurrence or the consequences of an' accident or malfunction of equipment inportant to safety previously evaluated.
2.
Review of this modification condensate system description of FSAR section 10.4.6 and the accident analyses of FSAR chapter 15 reveals that this modi-fication.does not create,the possibility of an accident 'or malfunction of a-different type-then previously evaluated.in the FSAR.~
3.:
h is modification does not involve any safety:
-related equipment nor does-it involve any limiting-conditions for operations.or surveillanceL requirements'as described by Tech Specs..h us, this mod does not reduce the margin of. safety as-
-defined in the Bases of.the Tech:
Specs.' Tech Specs. iwr section 2.0, 3.0 and-4.0 were reviewed.--
-]
e 89-V2M015
. Revise the wiring in the Beta Annunciator P'anel at Panel 2-1414-P5-FDP to allow Control Room i
annunciator ALB17D03 to clear once the' alarm at' i
the local panel is acknowledged by an Operator..
1.
L e Condensate Demineralicer System.is a non-safety.
related system'. 5 Failure of the ccanponents of. this -
systen will not affect the ability of the plant.to 4
accomplish a safe shutdown; This is based on a review of ESAR sections 15.0 and'10.4.6.-
o!
1, 2.
This modification will not in'any way increase the possibility of an accident or malfunction occurring that will lead to an unsafe conditions i
E is is :>ased on a review of FSAR sections 15.0 and 1
10.4.6.
3.
Based on a review of the Tech Spec. basis,.
)
including section B 3/4.7.1, this modification will not decrease the margin of safety at the plant.
1 75 q
4 II:
1989 ANNLIAL REPORT - PAIG' 2 10.CFR50.59(b) REPORT'-
89-V1M016 This mdification remved the filter element from the Waste Evaporator Feed backflushable filter and.
the Floor Drain Tank backflushable filter. These' filters are not needed since filtration for. the Liquid Waste Processing System is performed at thet
' Alternate Radwaste Building. The liquid waste evaporator is not utilized therefore there'is no-need to separate solids / crud from the liquid.
1 All wastes go,to the Alternate Radwaste Building for processing.
l 1.-
Increase in the probability of occurrence or thei consequences of an accident or malfunction of 1
equipment important to safety is not caused by this' change. This change-does not inpact the pressure-boundary of the system and included a review of FSAR chapter 15.
p 2.
Removal of the' filter el ments will not create
.the possibility of an accident or malfunction different from FSAR evaluations of chapter 15 since no credit is taken for the elements.and the rem val will not affect the pressure boundaries.
3.
. Based on a review of thd Tech Spec. bases, including section-B-3/4.11 Radioactive Effluents',
i this modification will not reduce the margin of Iafety at the plant.
89-V2M017 Reroute the conduit 2NE525RS00llto acemmodate
-relocated flow switch when the Auxiliary CNNI.
Cooler fan 2-1515-A7-002 was replaced..
~
1.
This change affects a Non-Q conduit attached to i
a Non-Q cooling unit. The conduit' attachment.is sufficiently evaluated to ensure no impact to other Q equipment (EFCRB 8210F) and therefore does not increase the probability or the consequences of 1
an accident or malfunction of. equipment.
2.
Conduit affects only Non-Q equipment. However, change is consistent with original design criteria, Therefore, this m dification does not create the 1
possibility of an accident or malfunction of -
l equipment.
1 i
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II 1989 ANNI]AL REPORT - PART 2L 10 CFR50.59(b). REPORT
-3.
ReroutingL the subject conduit does not affect the omration of the-Auxiliary CtMr cooler fans and :
taerefore the margin of safety does not_ decrease..
89-V2M018-Ratove the filter _ elements fran the Waste Evaporator Feed backflushable filter and the:
-Floor Drain Tank backflushable filter. The:
filtering capability is not required since filtration;is'perfonned at the Alternate-Radwaste Building.-
=!
1.
The Licuid' Waste Processing System is'a non-safety 3
- relatec system. Failure of the ccmponents of this-systen will not affect the ability of the plant to i
acconplish a safe shutdown This'is based on a review of FSAR sections 15 and 11.2.
4 i
2.
This modification will not in any way will inbrease-i the possibility of-an accident or malfunction
{
occurring that would lead to an unsafe condition..
This is based on a review of FSAR1 sections 15.0 and 1
11.2.
3.
Based on a review of the Tech Spec basis, j
including section B 3/4.11 Radioactive Effluents,
{
this nodification will not reduce the margin of; j
safety at the plant.
1 89-V2M019 Modify the Nuclear Sanpling tubing supports to -
l match field ~ requirements.; One support will be 1
attached-to'an existing hanger., Another,supaort's base plate is being made smaller to fit in.tae available space.
1.
The support nodifications will' not increase the probability =of occurence or consequences of an i
accident or malfunction of equipment required for
-j safe-shutdown.' The Nuclear Sanpling' system l
is not required for safe-shutdown of the plant.
The subject tubing was and still is~ seismically.
qualified.
2.
The Nuclear Sampling Panel is not involved in 1
or used to mitigate any presently evaluated chapter 15 accident. Since this change does not alter the function or qualifications of the Nuclear R
Sanpling Panel, no new accident possibilities are-4 created.
77 j
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II 1989 ANIRIAL REPORT - PART 2 10 CFR50.59(b) REPORT
- 3.-
Implementing this m dification does not affect the
- function or reliability of the Nuclear Sanpling Panel, nor does it' affect our ability /4.4.7.
to meet Tech Specs. surveillance 4.4.7 or Basis 3 t
89-V1M020
'This e dification changes breaker ICD 1M01,-02,-03,.
04; and 05 frmt 15A to 30A thermal magnetic breakers-per table 10 of DC-1823-There should be a-30A.
thermal m gnetic breaker installed.
1.
This modification only involves changing-
' breakers ICD 1M01, 02, 03, 04 and 05.fr a 15A to 30A thermal magnetic breakers to conform to.
DC-1823. It will not cause any e pment assumed l
~
to function in an accident to mal etion.
Reference FSAR' sections 15.1=and-15,2, 2.
This m dification involves.only changing thermal magnetic breakers to conform with DC-1823.'
'It was-originally.a design error. No accidents or
. equipment malfunction will result from this change-that is not. described in the'FSAR.- Reference FSAR sections 15.1, 15.2 and=8.2.
1 3.
his mdification does not involve _a change _.
to Tech Specs (Ref. 3.7~.1.2).
It:only. involves
. changing thermal magnetic breakers to conform to 1
DC-1823. Eis change will not affeet the margin of 1
safety as definedtin Tech Spec. 3.7.1.2 and 3.8.1.
89-V2M021 h is modification changes breaker 2CD1M01', 02, 03,.
04, and 05 from 15A to 30A thermal magnetic breakers.
per table 10 of DC-1823.. B ere should.be a 30A:
thermal magnetic breaker installed.-
1 1.
E is modification involves only changing; j
4 breakers.2CD1M01, 02, 03, 04 and'05 from 15A to:
30A thermal magnetic breakers to conform to -
DC-1823. It will not causeLany e
'prmnt assuned to function in an accident to mal ction.
Reference FSAR sections 15.1 and 15.2.
.)
1 1
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-II 1989' ANNUAL REPORT ~- PART 2 10 CFR50.59(b) REPORT 2.
The nodification involves only changing thermal-magnetic breakers to confonn with DC-1823. It-was originally a; design error. No accidents or:
equipment malfunction will' result from this change that is not described in the FSAR.. Reference FSAR sections 15.1, 15.2 and 8.2.
3.-
This nodification does not involve a change _ to l
Tech S3ecs section Ref. 3.7.1.2.
It involves only. clanging thennal magnetic breakers to conform to.DC-1823. This change will not effect the margin of safety as defined in Tech Specs.;3.7.1.2-and 3.8.1. V010022 Add a note to the typical detail drawings to allow ladders and handrails to be added, modified, or deleted. 'An approved safety a luation will-.be-required and ladder and handrail' typical: details j
shall'be-followed.- The typical-details are shown on drawing AX2D94V004. These details were used during plant construction for ladder and handrail-1 fabrication and installation, 1.
This change will allow future: ladder and handrail char.ges... All changes will be performed 1
in accordance with approved typical details. These
~!
details were used during the original construction of the plant for ladder and handrail' installation.
These' details meet all seismic design criteria.
The-safety evaluation required for each change will; a
insure no increase'in the probability of occurrence or'the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report can occur.
2.
This change cannot create an accident or j
malfunction of a different type than previously j
evaluated in the-safety analysis report. This
)
change allows changes to ladders.and handrails i
providing typical details are'followed.and an approved safety evaluation is completed. These steps-insure that no safety related equipment can 1
be affected by this change.
l j
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-II 1989 #EUAL REPORT - PART 2 10 CFR50.59(b) REPORT 3.
Ladders and handrails are not included in the basis for the Technical Specification.:-This change-will-have no effect on.the basis.for the Technical
~ Specification', 'Ihe required safety evaluation will.
ensure no future change-can affect the basis-for the Technical Sepcification.
8S-V2M023 This change allows a variation from the mounting.
l detail-drawing for mounting of 2PDI-1222.(Steam Generator Blowdown Cartridge filter differential pressure-indicator)._ It is necessary to allow welding on one side of'the nounting plate and anchoring to the wall for the opposite nounting j
plate ~ side. This is a nonsafety, nonseismic
-j class 62J instrument.
3 1.
Making an exception to the nounting detail; drawing for this nonsafety, nonseismic class instrument has no impact on an accident or i
-malfunction of' equipment imaortant to safety
.and evaluated in the FSAR caapter115,
- 2. -
'1he pressure differential indicator has no-.
control or operational function'that could' i
affect any equipment or accident of any type including one not previously analyzed-in the FSAR chapter 15.
~
1 3,
2PDI-1222 is a new instrument, nonsafety-related..
nonseismic, and is'not included in the Tech Specs.
-t including Bases.B 3/4.7..It is an indicating instrument only with no impact on safety.-
~
89-VCM024 To add Potable Water Building pump' casing drains-l in the Potable Water Puup House.1.'Ihis will keep packing leak-off off the floor.--
1.
Neither the Potable Water Building, any equipment in it or the Potable Water-System interfaces with any equipment that' is important'to safety, i
Implementing this change-does not alter that i
fact.
1 2.
The Potable Water System is not involved:in any Chapter 15taccident. Since this nodification does not link the Potable Water system with any' other system, no new malfunction possibility is created.
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1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 3.
-No Tech Spec basis assumes anything about the' function or malfunction of any equipment in 5
or related to the Potable Water Building.
89-V04025
. Modified control loop's AT-12470 and AT-12471 to provide proper operation of Fuel Handling Building normal-supply reheat coils'. Current to resistance converter made reverse acting.
1.
he IHB Normal Supply Air System is not important to safety as outlined in FSAR section q
9.4.2.
2.
The RIB Normal Supply System is non-safety _
related-and the control loop change cannot cause any. type of accident.-
j 3.
he -affected equipment is not. part.of any :
1 Technical-Specification basis and_specifica11y' not a-part of 3.9.12; 89-VQ4027 This modification adds four' receptacles to the u
north side of the Water Treatment Building to provide power to the NRC Mobile Lab. These-120V AC Receptacles will be powered from normal?
lighting panel ANLP79,- breakers 30,_32,:34:and 36.-
1.
Eis change will not involve any safety-related-c aponents, and will not cause any equipment:
assumed to function in an accident-to malfunction..
j Reference FSAR section 15.2.6..
2.
This change involves no safety-related'~
components or equipment ~it only. adds 120V AC j
receptacles to a normal lighting panel in the Water Treatment Bldg.. No accidents or equipment malfunction ~will result fran this-change that is not described ~in the FSAR. Reference FSAR sections i
15.2.6 and 9.5.
3.
The addition of four 120V~AC receptacles to a norma 1' lighting panel in the Water Treatment Building will not effect the margin of safety; as defined in Tech. Spec 3/4.8.1.
89-VC4028 This modification involves changing the setpoints on the level switches for the 81 5
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II 1989 ANNGL IEPORT - PART 2 '
10 CTr# 59(b)~ REPORT g
Imbe Oil Storage Area Sunp Per A-1420-P4-508.,
he pump start setpoint is to m lowered for b
instrment ALSH7680. The High-High alarm is also to be lowered for AISH7680; -In. addition, the' spare switch setting is to be revised on-AISL7680.
1.
The Waste Water Effluent' System, system-1420, _
J is a non-safety related system. Failure of the components'of thistsytem will not affect thet
.j ability.of the plant to accomplish a safe shutdown. This is based on a review of section 15,-
Accident Analyses'.
2.
Based on a' review of section 15.0 of the FSAR,
' Accident Analyses','it.does not appear that this modification in any_way will increase'the E
possibility of an a cident or malfunction-. occurring that would lead:to an unsafe condition.
3.
Based on a review of the Tech Spec. basis, this -
' modification will not_ reduce the margin of safety-i at the Plant.
89-V1M029 Removed existing #2 torque switch limiter plate and replace with a 13 limiter plate set at 3' i
1.
This change does not exceediany manufacturer i
design criteria and-in fact establishe.s proper, valve operation.,Therefore, there is no increase' l
in the probability of: occurrence or consequences-of a malfunction of safety related equipment or
- ]
components.. Section 9.2.1 of+the FSAR reviewed.
2.
The installation of the f3 limiter plate-o provides correct-operation of NSCW Tower Return a
Valve 1HV-1668A and does'not change any. previous-system or component operation'. _ This approved 4
vendor replacement limiter plate does not create the possibility'of an accident or. equipment / component _
l malfunction other-than that evaluated in the FSAR.
i Section 9.2.1 of the FSAR reveiwed.
1 i
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'II-1989 ANNUAL REPORT - PART 2
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10 CFR50;5's(b)-REPORT.
1
- 3. -
hebasesfor-operation loftheNSCWsystem.
is maintained by this change which effectively E
provides correct operation of valve 1HV-1668A. _
The margin of safety. defined by the bases of the Tech Specs.lis unchanged by the installation of the new limiter plate Tech Spec section 3/4.7.41 reviewed.
89-V2M031 Change; time dial setting for RCP Time Overcurrent Relays'fram'960 cycles at 500%.of Tap to 1080, y
-cycles-at 500% of Tap. ; Change High Dropout unit' q
T.J
-setpoint for.RCP Time Overcurrent. Relays fram.
j N3 Tap.6' (480) to Tap 7 (560 anps)'.. ; This change will y
allow all'RCP's to start reliably without tripping j
when used.as the:last punp to start.
y y
4
.1.
The< change adjusts the:RCP; Time;-Overcunent' 1
Relay'setpoints to start reliably without nuisance.
j tripping, while~'still gravidingLadequate locked-Rotor protection' for tae RCP notors and overcurrent i
protection for the containment 7enetrations. - 17he new setpoints are still below tae design limits J
of the penetration Ref. Fig. 8.3.1-7 sheet-.12 of
,19.
1 2.
The new setpoints;for.theRCP Time Overcurrent
' Relays sinply allow alllthe RCP's to start; reliably j
without nuisance tripping while still aroviding 1
adequate locked. rotor protection for the RCP's and-adequate protection for the containment--
penetrations, r
3.
Thechangedoesnotdecreasethhmarginof' safety defined by the basesffor the Tech Specs.'(see_ sect i
3/4.8.4) the new setpoints are still below the design limits of the Contaiment penetrations.
Ref FSAR Fig 8.3.1-7' sheet 12 of 19.
89-V2M032 Underfrequency relay model #222A1175~has became i
obsolete and replaced by model #422B1275 which is the same form, and' function as the original model #222All75. Drawings 2X3D-AA-C03A &
2X3D-BD-B01X need to be' updated to' reflect the change in nodel d's and rewiring due to a change in connection points on the new relays.
83
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=1989 ANNUAL REPORT - PART 2-
=
10 CFR50.59(b) REPORTj L1, the new relays (model #422B1275) represent the same form and function as'the old relays and will' be calibrated to the same setpointias the old-relays and, therefore, will not' increase the-probability of' occurrence or consequences of an-accident or malfunction of equipment'inportant to safety previously evaluated in the Safety.
a-Analysis Report. This includes a review of FSAR Chapter-15.
2.c ~ The new Relays-(nodel #422B1275) represent the.
same form and function.as the old relays'and will be' calibrated;to the same setpoint as the old-relaysi:-therefore, will not create'the possibility:
3 of an accident or malfunction of a different type
.L than previously evaluatbd'in the Safety Analysis.
. Report.- This includes'a review of chapter 15.
3.
The new Relays (model'f422B1275) represent the same form and function as the old relays and will-be calibrated t'o the: seme setpoint as the' old; relays:-therefore, this nodification will not:
reduce the margin of safety as defined in the Basis for any Technical Specification.-)The new
?]
relays will be calibratedeto the:same setpoint.'
as the old relays which will'be-consistent'with Tech Spec '2,2.1 basis.of U/F Reactor Trip signal'~
reaching the Reactor Trip Breakers in 0.3 secs.
89-Vm033 The Waste Gas System Hydrogen Rec mbiner is subject i
to transients which cause pressure fluctuations,at:
the Rupture Disc. This. leads-to fatiguing of the disc and, eventually, cracking. In order to alleviate the conditions-leading to fatigue, a vactnn support has been installed within the i
rupture disc. It will brace' the disc against j
back pressures as they occur.
]
1.
The Waste Gas Systen is a ncn-safety related system. Failure of the c mponents of this system will not affect the ability of the plant to acceplish a safe shutdown. This is based on a i
review of FSAR sections:15.0 and 11.3.
y e
S4 1
J
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- II 1989 ANNUAL REPORT - PART 2'
-10 CFR 50.59-(b) REPORT 2.-
This modification does not increase the possibility of an-ll accident or malfunction occurring that would~ lead to an unsafe condition. This is based on;a reviewfof FSAR. sections 15.0-and-11.3.-
- 3'1
-Based on a-review of the Technical Specification basis, 1) including SectionLB 3/4.ll,:this. modification will not reduce.
the margin:of safety.
~
q
' 89-VCM035 The. instrument air inlet and outlet tubing to, current to ti
- pneumatic converter AFY-12777 will be disconnected. - This tubing will, be rerouted toirun'from the-air regulator-
. presently' upstream of AFY-12777.directly to the positioner -
for the -inlet-vanes AFV-12777.
1.
This change does not; involve.any safety-related equipment.'
1 It will.only change the method by-which_ the1 inlet vanes are g.
positioned?to obtain design air-flow from thel fume hood' exhaust fan. This change willLnot increase 1the probability; of an accident.or consequences ofiany' accident previously q
evaluated'in the FSAR. '
1 2.
This change does not affect any safety-related eouipment, nor -
will it create-the possibility of any accident'not evaluated 4
in the FSAR'
~1 i
3.
The-fume hood exhaust fan is not addressed specifically in the Section 3/4.11.2.4 of the Tech Spec. which-addresses the " Ventilation Exhaust Treatment, System" of which the fume hood exhaust is a part, will.not be'affected by this l
change and therefore the margin of safety will'not be-reduced.
89-VCM036 Upgrade the Security Card Reader.-
1.
The Accident Analysis ~ section~ of the FSAR (section 15) was reviewed to determine-that the implementationLof.this DCP-would not' increase the probability of-or -the consequences' of an accident:as.iscribed-in the FSAR.
.I
.i-
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85-a
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!q
-II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT Also, FSAR Section 13.6 details the industrial-security requirements, section 13.6 as well as the E
VEGP Physical Security Plan was reviewed. No-decrease in the-' security effectiveness would be
. realized as a result of inplementing,this change..
2.
A review of section 15 of the FSAR was performed in determining that this design change would not create the possibility of an accident or
-malfunction other than the previously-evaluated in B
'the FSAR..
3.~ -
A review of sections 2.0. 3.0 and 4.0 of the Vogtle Technical Specifications wse performed.- It was then determined that this' design change would not' reduce'the margin of. safety as defiried by'the:
Technical Specifications.-
B 89-V2M038 his modification adds a CS lug to pipe just.
l above the strap on support V2-2406-004-11601 -
1 i
'to provide dead load support of relief valve.
1 2-PSV-9697-and connects the relief valve'to existing tailpipe using'1/2" diameter stainless
,~
steel-tubing. The lug. addition-to the support puts-the weight'of the tailpipe on the' support anchor
-- i bolts.instead of.on the Relief Valve threadede connection.
1.
W e purposes of the Aux. Hydrogen Gas System-includes assuring a continuous supply.of gas to the Reactor Coolant drain tank as described in i
FSAR section 9.3.5.1 W e. design changes'in this modification ~ ensure that 2-PSV-9697 can i
perform its.intendedLdesign function. Le support lug addition was designed ~and installed H
per applicable codes:of ANSI'B31.1 and AISC.
]
2.
The Design Changes to the Aux. Hydrogen Gas l
System per 89-V2M038 ensure that 2-PSV-9697 can j
relieve pressure from the hydrogen header as.
i designed. Neither the design nor installation-of this nodification creates the~ possibility.
of an accident *or malfunction of a'different type than previously evaluated in the FSAR.
j 3.
Review of Tech Spec. 3/4.4 shows that tSe Aux. Hydrogen Gas System is not addressed in the level of detail that this nodification involves.
86
11 1989 M E AL REPORT - PART 2 10 CW50.59(b) REPORT 89-V2M039 This change deletes the Boric Acid transfer pumps handswitch position interlock fr m the System Status Monitor Panel logic. This is accmplished by lifting and sparing the appropriate conductors at the Main Control Board termination cabinets.
1.
Disabling the Boric Acid transfer pump frm the SSMP logic has no inpact on the function of any safety-related emponent evaluated in chapter 15 of the FSAR. Therefore, the probability of occurrence of any accident is not increased.
2.
This MDD brings the SSMP into cmpliance with section 7.5.5 of the FSAR by removing the manually actuated ESF functions fr m the SSMP which is intended to alarm for autmatically actuated ESF functions.
3.
This change has no inpact on the boration system Technical Specification (3/4.1.2) or any other Tech Spec.
89-V2M041 This MDD substituted spare contact 11 for contact 1 on transfer switch 2HS-5106C. This was accomplished by moving the field side wires of cable 2CD1M05SF presently connected to TBF-43 and 42 to TBF-65 and 66.
1.
This modification changed the field wiring from a failed contact 1 to the spare contact 11 of transfer switch 2HS-5106C. This enabled the circuit to operate as designed and thus reduced the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. FSAR sections 7.3.7 and 10.4.9 reviewed.
2.
As this wiring modification made no functional or operational changes in the system it does not create the possibility of an accident or malfunction other than those which were previously evaluated in the FSAR. FSAR section 7.3.7 and 10.4.9 reviewed.
l l
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f II-1989 ANNUAL REPORT - PART 2 -
10 CFR50.59(b) REPORT 3.
h is change does not decrease the safety-margin as defined by the Tech Spec bases.: he -
modification leaves the system operationally and q
functionally identical:to the originalidesign i
condition. Tech Spec' section 3/4.1.2 -reviced..
89-V2M042
' A..
Bis MDD-installs a jumper from CR42 point 207
.to the coil'on TDR 28 in condensate demin panel-2-1414-P5-FDP. This allows a -delay start -
j
.of the spent resin transfer puup-
[
2-1414-P4-502.
B.
Changes the backwash high flow setting fran 500 GPM to 375 GPM by adjusting regulator 2-HICV-03298B in condensate demin panel 2-1414-P5-FDP.
j C.
. Removes pins 6,7 and 8 at' tenor drun switch -
l 44 -(TDS-44) to eliminate spent resin auto I
start.
1.
The modificationsrinvolved do not increase the-probability of occurrence orfthe. consequences ofs an accident or malfunction of equipment important>
l to safety previously evaluated in the FSAR. h e accident analysis, section 15.0 was reviewed with.
1 aarticular attention to section 15.1 Increase In -
leat Removal.By The Secondary Systan" ~andL15.2l
" Decrease In Heat Removal By h e Secondary System".
2.
The modifications enhance system operability _with mgards to Item A and Item B and provides manual i
control of spent resin transfer with regards to l
Itan C.
Section 15.0 of the FSAR was. reviewed, l
he modifications do not create the possibility of an accident or malfunction of a different type than previously evaluated in the safety analysis report.
3.
The modificaitons do not reduce the margin of l
safety as defined in the Basis for any Technical Specification. The bases for Tech Specs.: sections l
2.0, 3.0 and 4.0 were reviewed.
i 88
II' 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b): REPORT l
1 89-V1M048=
Revise the backwash high flow setpoint on the
- condensate demins from 500 GPM to 357 GPM.
h is change prevents the backwash recovery tank from overflowing.
1.
, he Condensate Demin system is not safety-relat'd..
1 e
Since the lower backwash flow rate _ yields an
_ equally effective backwash with less stress on-the system, not only is any safety-related.
equipment unaffected by.the change, but the-.
non-safety related equipment associated with this system is less likely to malfunction; j
~
2.
Re-Condensate D ein system is not involved in any:
. chapter 15 accident.--Since implementing-this change only enhances the operation of-the j
Candensate Demin system,;no_new accident i
possibility is created.
t 3.
Implementing this change does not alter our. ability i
to maintain proper, secondary plant chemistry as y
assmed in Tech Spec Basis 3/4.4.7.
~
89-Vm049 To provide system overpressure protection for the.
Normal Chilled Water System by lowering the relief-valve. set pressure on the pump suction. Change the set pressure of relief valve APSV-22302 fr m 135 psig to 45 psig. To-accomplish this=a new spring nust be: installed in the valve.
i 1.
The Nomal Chilled Water-Systiem is not included y
in: accident-scenario in chapter 15'of FSAR.
2.
Le Nomal Chilled Water Systs has no' safety 1
design basis per paragraph 9.2.9.2.3 of the FSAR.
Nevertheless, the. change in relief valve set pressure was conservative in relation to possible system failure.
3 3.-
The Normal Chilled Water System is not considered in any Tech Spec basis. Technical Specifications 4"
3/4.7.-11 addresses the ESF Chilled Water System =
only.
- i 89-Vm050 Changed the fire water tank level setpoints and~the instruments used to monitor these setpoints. he.
1 level setpoint was raised to increase the amount of-water available in the fire water tanks. The instrument type was changed to one with a smaller deadband.
'l 89 i
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L II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT-1.
The setpoint level changes assured that mihinun. _
amounts of firewater are maintained:as required by -
FSAR.: The new switches-give more accurate control-of level. There is no degradation of.the
" defense-in-depth" Fire Protection Progran as a result of this MDD, 2.-
The new switches'are similar to those.already installed but with a smaller deadband and will.
therefore ore accurately control: tank level within FSAR requirments.
~
3.
The fire water tanksLare not included in the
' Technical Specification base'.;
j s
1 89-V1M051 This MDD corrected Automatic Rod Control wiring for
{
autmatic insertion of control rods. The' incorrect j
wiring would make. the. rods nove out then stop.for l
-an auto insertion signal. 'The. rewiring of a J control card will allow the proper operation of the Autmatic Rod-Control'Systen..
i 1
- 1.
- Automatics Rod Control is not a safety system but-.
can be used-during some events to prevent the plant frm exceeding Reactor. Trip limits.-
j 2.
The nodi.fication changed the plant -design to match -
y/
proper operation of Automatic Rod Control'. TheJ d
FSAR assumes correct Automatic Rod Control per section 7.7.7.1.
1 i'I 3.
Auto Rod Control is not required for-safety of the1
. plant, therefore correct operation of' Auto Rod Control does not reduce the safety margin,of the plant. Per review of Tech Specs, 3.4.8.
i 1
89-V1M052
. Addition of a hanger between the RHR suction line-and the RHR suction line vent,.1-HV-10466.. This i'
-will restrict movement relative between.the RHR
. Suction line and 1-HV-10466, the suction line vent.
O 1.
The addition of the hanger will prevent an y
overstressing of the associatedEpiping during a-SSE or an OBE. The removal of the possibility of overstressing will prevent the failure of the-1 piping under a' seismic event as required by FSAR 3.7.B.
90 l
V; II 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 2.1 This addition of a hanger.will not create the possibility of an accident different.from those described in section 15.0'of the FSAR. This hanger was required _to ensure that the piping to-the RHR suction line vent could. survive a seisatic~ event..
The: design of this hanger agrees with the original.
. installation specifications for hangers. The ccupliance of: the hanger design with the original' installation requiraw/ats ensures that the hanger will function as intended and will not conpranise=
the systan which it supports.:
3.
The margin of safety established for the -
associated system in:the Tech Specs is-assured by Engineering analyses and installation i
requirements aa described in the FSAR 3.7.B.
This' hanger addition ensures that the piping' involved can survive a seismic event as required by section 3.7.B of the FSAR.
89-VQ4059 Add one fuse to'each neutral leg for 120V AC d
power feeding Loop-2, Zone-1, Leg-A_(Unit 1),'and
" cop-4, Zone 2, Leg-A -(Unit 2) paging-systan in 3
Containment. The. type of fuse is Bussman MDA-15, i
This nodification adds the additional ~ fuse-i protection so that testing of breakers CB6 and CB12
- i in the Gai-Tronics control cabinet-(A-1701-U3-TDC) is eliminated.
-1.
The additional fuse will not affect the-menetration protection since the' fuse curve is-l mlow the penetration capability curve.. (Ref FSAR Figure 8.3.1-7, Sheet 16'of 19). Therefore, 1
the fuse should open prior to penetration damage i
occurring and thus maintain the integrity of the containment pressure boundary (Ref. FSAR
,l 8.3.1.1.12).
1 Also,-the plant is designed to shutdown without H
relying'on connunications equipment.- (Ref. FSAR 9.5.2.2.6).
2.
This modification does-not affect operation of the plant page and provides_ additional protection of the containment penetrations on this leg, s
i 91
II!
-1989 ANNUAL REPORT - PART.2 10 CFR50.59(b). REPORT 3.
Le addition of the. fuses will provide the redundant penetration protection required (Ref~
Tech Spec. 3/4.8.4).-
89-V1M062 We quick disconnects of the Nuclear Sanpling Panel (Unit 1) were changed to eliminate the need-for an adapter for the gas stripper, to be
' consistent with the Unit'2-~ disconnects, and-f increase the pressure rating. This change will
. allow interchangeability of sanple banbs for.
both units and eliminate-the threat of:
contamination should the pressure boundary fail with the previous disconnects.-- The affected-sanple' bombs are 1-1212-NPS-501, 502, and 503, he Swagelok disconnects are nodel QT4.-
1.
Eischange'doesnotaffectconsequences/-
-occurrences'of any accident or malfunction of equipment important to safety. - L e Nuclear-Saupling Panel is' not safety-related and.is' not taken credit for:in FSAR chapter 15...
q 2.
The Nuclear Sanpling-Panel is not -safet'-related i
j and is not required to mitigate any: accident of a i
type similar or different-from the analyisLof
- l FSAR chapter.15, 3.
Changing; quick-disconnects on the Nuclear Sampling j
Panel does not affect the Tech Spec.= Bases B 3/4.4.7 or-B 3/4.4.8.
1 89-V1M06S Le changes.acconplished-by this MDD are threefold.
l The intent of these changes are to decrease the' i
likelihood =of contaminated water leavitig the RWST
-j valve room.. The mods are-A.
Add threaded caps to the~ lines used'to take chemistry samples of the RWST.'
B.
Revise the design of the floor joint' a
seal to ensure that any contaminated-
'1 water in the room cannot leak into the-s soil'below, i
.p u
~
l l
92 a
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II 1989 NMIAL REPORT - PART 2 L 10 CFR50.59(b) REPOKr C.
Design and install a curb at the door toi the rom such that any water on the floor of the room cannot leak under the door' ara-onto the ground outside the' RWST valve rom.
'I.
%ese three changes will in no way affect the,
functional operation of the plant as described-in FSAR. These changes will serve only.to ensure:
the leaktightness of the RWST smpleilinesiand the' RWST valve roan itself Since these changes are
" nonfunctional",in nature, they will not result -
.in the increase of-either the probability or-consequences.of;any. accident or malfunction.
~
' described in the FSAR-(Ref 15.0 of the FSAR).,
q 2.
- R ese nonfunctional changes to the plant will-serve only to inprove the leaktightness: of-the RWST q
valve roan and: the RWSr sattple lines. i Due' to' theD natureofthesemodifications,Jtheywillnotresult.
)1 in the possibility of t an; accident or malfunction of:
j a different. type than previously: evaluated-in the
.- {
FSAR.
a 3.
hese changes will not result in a change to the" ll level of safety established by the Bases for the RWST Tech Spec. 3/4.9..
jl
.i 89-V2M066 he changes ademplished by this.MDD are threefold. -
l Re intent of these changes are to' decrease the l
likelihood of contaminated water leaving the RWSTL
.l valve roan.' The mods are:-
i a
A.
Add threaded caps to the: lines used j
to take ch mistry samples of the RWST.
1 B.
Revise the design of the floor joint -
l seal to ensure that any contaminated' c'
water in the roan cannot leak into-the soil below.
.a C.
Design and install a curb at the door to-1
'the roan such that any water on the floor
..of. the roan cannot leak under the door:andL onto the ground outside the RWST; valve
_j roQn.
l.
93
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1 11 1989 ANNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 1.
%ese three changes will in nolway affect: the
~~y functional operation of the plant as described in-in the Ff"R.
These changes will serve only to ensure the leaktightness of the RWST samale lines and the RWST valve room itself.' Since taese changes' are;" nonfunctional'.' in nature,. they will not result in the increase of.either the probability or-consequences of.any accident or malfunction described in;the FSAR( Ref. 15.0 of the E AR).
2.
These nonfunctional changes to the plant will serve.-
a only to inprove.the!leaktightness of the RWST valve room and the RWST sample lines..-Due to the nature of these modifications,1they will not result
-in the possibility of an accident or malfunction
.of a'different type'than previously' evaluated in the; FSAR.
3.
W ese changes will not result in a change to the.
level of safety established by the Bases for the RWST Tech Spec. 3/4.9.:
89-V2M067 Tank pressureiindication; 2-PI-0932,pr y' Additive Change the span of the Containment S cm 0 to'60-psig to 0_to 15 psig. This was done:to achieve-a clearer indication of the. cover pressure on the SAT.
1.
This change results in'an enhanced reading ofLthe pressure in the SAT. Since the design pressure of the' SAT is 10 psig,.this' change will have no
)j detrimental operating effect.-.. The probability of' occurrence' and/or severity of consecuences of 'an-accident or malfunction as describec, in the FSAR i
is.not affected by this change.
.2.
This change is to the ' scale of the instrumnt only.
The-internals are not significantly changed frcm the original design. The'only effect this-change will' have is to enhance the readability of'the gauge by the operations staff when adjusting the ' cover -
pressure.of the SAT. For these reasons, this. change-will not create the possibility of an' accident or malfunction of a different. type than previously' evaluated in the.FSAR.
3; 94
L II 1989 At E AL REPORT:- PART 2_'
s 10CIR50.59(p) REPORT.
3.
Based on the foregoing answers,' this instrument scale change will not reduce the margin of safety-established by the: Bases of the Contaiment Spray. Tech Spec.:
89-V1M068' This modification 'added en individual volume control' for Control Rom page snaker iS46.to stop feedback probles associated with this speaker.
1.
This modification affects'only the volme ofE Control-Roam speaker #S46. The installation of:the
- volume control will enhance Control' Room conmunications by eliminating feedback aroblems.
The volume control is located between tae anplifier and the speaker and therefore, reduces -
the. volume offthe emergency tone signals at-that-speaker (Ref, FSAR 9.5.2.2.1).. However; these:
emergency tone signals' are still heard with j
clarity and the origin of the signals is'fr a i
the Control Roam.- Furthermore,:the Control Room i
and Shutdown Panels are designed and-instr mented-i to bring the plant to a safe shutdown condition,;
_j assuming a single' failure of safety-related-1 equipnent, without -relying-on cmnunications equipment:(Ref. FSAR 9.5.2;2.6).-
l 2.
This modification affects only the volume of Control Rom speaker fS46. The plant page is in j
routine use which willLensure the availability of the speaker at a volume-level appropriate for this location. (Ref FSAR 9.5.2.3) 3.
The only canunications required by Technical
~;
Specifications is between the Control Rom and -
personnel at the Refueling. station during core alterations.(Ref. Tech Spec 3/4.9.5). This conuunication is provided by the sound power system (Ref. FSAR 9.5.2.2.3) which is not affected by this modification.
1 i
89-V2M069 1he main-transfomers have two sets of contacts
~
for-annunciationofhighoiltempe5^tur*-
ne
.segmakesat65Candother;at90 C.
The 65 C contacts are to be disconnected.
1.
This change does not involve'a safety related
.)
equipment / system. It does not increase probability of accident or malfunction not 3
analyzed in section 8 of the FSAR.
i i
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II 1989 AWiKL REPORT - PART 2:
10 CFR50.59(b) REPORT 2.
The change does not create possibility of an accident or malfunction not previously evaluated.
In FSAR sections 8.12, 8.1.4.2, 8.3, 1.2.8.-
3.
Themarginofsafetyisnotgffectedperreview'of annunciation only. point _(65 C) is used for
- sec 3/4.8. We set 89-V1M072.
Add an "in-service" permissive in the low
~ differential pressure alam annunciator for the.
Unit 1 Stean Packing Exhaust and Steam Jet Air-
- Ejector Filter Units...To prevent the low differential pressure annunciators fra alarming if the filter units are not in service.-
1.
Le SPE and SJAE systems as describe in FSAR l
' 9.4,4' are not inportant to; safety.
2.
FSAR 9.4.4 states that SPE and SJAE exhaust j
filters have no safety design bases.;
~
3.
he change -is'in alarm circuitry only and would i
not-affect the-ability;of the units to. function as a ventilation exhause treatment _ system as I
defined by 3/4.11.2.4.
89-V2M073 Add an "in-service permissive in theLlow:
' differential pressure-alarm annunciator for; 8
the Unit'2 Lteam Packing Exhaust and Steam Jet Air Ejector Filter Units. 1To prevent the low differential pressure annunciators from alarming j
if the filter units are not in service.
l 1.
Le SPE and SJAE systems as' described <in FSAR 9.4.4 are not importantito-safety.
q i
2.
FSAR.9.4.4 states that SPE and SJAE exhaust i
filters'have no: safety design bases.
3.
h e change is in alarm circuitry only and would not affect the ability of the units to function as a ventilation exhaust treatment system as defined by'3/4.11.'2.4..
I 4
1 96
11 1989 NNJAL REPORT - PART 2 10 CFR50.59(b) REIORT 89-V1M074 Install a check valve in the hydrogen supply line to the main generator (Unit 1).
1.
The change does not require change to FSAR sectica 9.3.5 which describes the Aux Gas system.
The change dos not increase the probability or cons ( pences of accident.
2.
Addition of a check valve in supply line improves the reliability of machine it will prevent generator depressurization due to failure of hydrogen skid. This does not create any accident not analyzed in section 15 of the FSAR.
3.
A review of section 3/4.7 does not indicate that safety margin will be affected.
89-V04079 This nodification involved adding knee braces to all 50 Boraflex Coupon holders on the four Boraflex Caupon Trees in accordance with Holtec International Engineering order no. 70810-1 dated 7/14/89.
1-The probability of an accident will not be increased by the weight addition to the Boraflex Coupon Trees. The weight of the trees is bounded by the Spent Fuel Pool rack analysis for full and epty racks at each location. Also, this change will not increase the probability of malfunction of a safety related ccxuponent. This modification has significantly reduced the likelihood of Boraflex Coupon Holder failure.
2.
This nodification has decreased the probability of Boraflex Coupon holder failure, a malfunction not evaluated in the FSAR.
3.
The Boraflex Coupon Trees are not described in any Technical Specification or bases for any Tech Spec.
Therefore, this change in no way reduces the margin of safety as defined in the bases for any Tech.
Specs.
89-V1M080 The cabling from the Rod Control System to the Proteus computer was changed such that Proteus would track all rod movement correctly. Before this MDD was installed, if any rods in shutdown banks C, D, or E were stepped out, all three of these banks would be indicated as stepping out.
97 l
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1 11 1989 ANNUAL REIORT'- PART 2 10 CFR50.59(b) REPORP 1.
Rod Control and Proteus are not safety related systems. - The nodification was rolling six(6)-
teminations in one non-safety related panel.-
^
2.
his change to the plant design reduced the-possibility of an accident by having proper-indication of control rod position displayed and tracked by Proteus.
3.
He Protous ccuputer can be used as a non-safety related verification of the Rod Control demand counters and digital rod displaye(DRPI). This
'"back-up" will aid in trouble shooting problems with any of these syst es.;
89-V2M083 his nodification corrects a wiring error, resulting q
fran a wiring diagram error, in the Reserve Auxiliary Transfomer (RAT)' 2NXRA differential -
relay.(487RA) scheme <
1.
- his modification corrects a wiring error so I
that the-differential relay scheme works as originally intended. Asipreviously wired,?a fault or load surge on switchgear 2NA05 would have erroneously triaped RAT 2NXRA instead of the switchgear breamr. This nodification conforms to FSAR 8.3.1.1.2 (K.3) in'that'the differential relay now only reacts to faults within the' relay zone.
l 4
2.
his nodification corrects a wiring.erro::.so i
that the differential relay scheme works;as originally intended. Reference FSAR chapter 8.0.
q
.3.
This nodification corrects a design error :
J such that the relay scheme' functions as originally 1
intended. This correction actually improves-the l
availability of power frm transfomer 2NXRA.--
}
Therefore, no change to the original design assunptions have been made.
A' review of the bases
)l to Technical Specifications section'3/4.8-show:that this change does reducelthe margin of safety.
i I
)
1 98
II
-1989 ANNUAL REPORT - PART 2-10 CFR50.59(b) REPORT 89-V m0086
--Add a note to drawing AX4DD000 Rev-14 stating that unions may be added without design change ~ if the following criteria has been neti-a) Seismic category 2 piping systen, b) ANSI pressure and tanperature rating is not greater than'150 lbs.
c) Non radioactive piping system-However, an approved Request for. Engineering Review followed by an As-Built Notice is' required._
1.
Installation'of unions per this MDD wil1~be done on systems-described above as allowed by piping'and-material classification manual-AX4DR001 Rev 9..
Installation of unions will be reviewed on an' individual basis before approved for installation.
Individual safety evaluations will-be perfonned as required by procedures, his design change will-not be used on any piping system with nuclear class 1&2 and seismic class 1 systans and will not increase the probability'of occurrence or the consequence'of an accident or malfunction of-equipment important to safety listed _in FSAR section 3, Table 3.2.2-1.
2.
'Ihis modification is only a drawing change design change, which allows installation of unions in seismic category - 2 piping 'systan which do -
not carry radioactive fluid and have ANSI pressure
.j' and temperature rating less 150f, _ Failure of:the system is,within the bounds of accident' analysis described in chapter 15 of the FSAR.
a 3.
This nodification will not involve-any safety.
1 systan and will not affect margin'of safety as j
defined basis of Tech Spec. described in bases 1
section of Tech Spec.
Ia 89-V2M087 Renove door 2-2111-L1-A90 from between rooms-N R-A22 and R-A81 on level A of the Control Bldg.
l to allow transfer of air between rooms so that i
temperature switch 2-TSH-12819 can adequately-a monitor the tenperature :of the Auxiliary Relay -
Panel Room, R-A22.
.l 1
99
e 11-1989 /&'UAL REPORT - PART 2 10 CFR50.59(b) REFORT 1.
The renoval of door 2-2111-L1-A90 does not increase the probability of occurrence or the consequences of an accident or equipment /
cmponent malfunction because this door is not a fire barrier, flood barrier, tornado barrier,
~
security barrier, or missle shield / barrier mer review of drawing AX1D94A20, Rev 6, FSAR c1a 15 has been reviewed and requires no change. pter 2.
Removal of door 2-2111-L1-A90 will have no impact on the operation and proper function of either the Auxiliary Relay Panel or the Auxiliary Relay Panel Room Cooler and therefore, will not create the possibility of an accident or safety-related equipmnt malfuncticn not described in the FSAR.
3.
Removal of the subject door does not affect the operation of the Auxiliary Relay Panel Room Cooler and therefore does not decrease the margin of safety defined by the bases for Technical Specification.
89-V2M088 Swap indicator lights at stator water control aanel 2-1326-P5-HSC to follow plant convention, led on right and Green on left. Existing bulbs to be replaced with LT-16 type lights to arevent breaker tripping due to short circuit witain the
{
bulb holder.
l 1.
This is not a safety related system and does not affect any other safety related systans as evaluated in chapter 15.9 of FSAR.
2.
'Ihis change does not create possibility of an accident or malfunction other than previously evaluated in chapter 15.
3.
Swapping indicator lights will not reduce the
- j margin of safety as defined by the bases for -
Tech Spec. section 3/4.7.
L 89-V1M089 Swap indicator lights at stator water control aanel 1-1326-P5-Hsc to follow plant convention, 3ed on right and Green on left. Existing bulbs to be replaced with ET-16 type lights to prevent breaker tripping due to a short circuit within the bulb holder.
100 I
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II 1989 ANNUAL PZPORT - PART 2 10 CFR50.59(b) REPORT 1.
h is is not safety related systs and does not affect any other safety related systems as evaluated in chapter 15.0 of FSAR.
2.
h is change does not create possibility of an accident or malfunction other than previously evaluated in chapter 15.
3.
Swapping indicator lights will'not reduce the margin of safety /4.7.as defined by the bases for Tech Spec section 3 89-V2M090 The Burnable Poison Rod Assembly (BPRA) Inserts fabricated for use in Unit 1 Westinghouse spent fuel racks were modiffed for use in Unit 2 Holtec International spent fuel racks. '1he modification involved decreasing the external dimensions of BPRA inserts frcxn 13.0 x 13.0 inches to 10.30 X 10.30 inches.
1.
The BPRA inserts that are modified for use in the high density racks (Holtec International) will decrease the probability of a fuel handling accident due to the proper fit without any overlapping. The BPRA inserts are not safety related and do not affect any other safety related equipment.
2.
Le center to center saacing of the Unit 2 high density spent fuel rac'
previously evaluated in the FSAR Reference FSAR-1 chapters 8.0 and 15.0.
j 2.
We wiring change will only provide proar
.j indicaticri of Bus #2 voltage at the QEA3.. No 2
accidents or equipment malfunction will resnit:
l from this change that is not described in the FSAR, reference chapter 8.0 and 15.0.
3.
Le proposed transformer wiring change does not involve Tech Specs or decrease the margin of safety defined by the Bases for any Technical 1
Specification.
89-V1M101 Bere is a blanked off opening / hole on the north; side of Unit 1 generator. A transition spool piece and 3/4" ball valve is to be installed over the.
1 opening.. The valve is to be installed with Garlock gaskets and ' capped closed with a blank flange.
I 1.
his change does not affect safety related.
equipnent.
It does not create increased probability of occurrence or the consequence of an accident or malfunction not evaluated in' FSAR section 10.
i 2.
This change does not create the possibility of an i
accident or malfunction not evaluated in section 10.1 & 10.2 of FSAR.
l l
3.
Be change does not affect safety related equipment and is not addressed in Tech. Specifications.
89-VIM 103 Lift cable NGV660XA frcxn RTD TE-6800.and'1and on RTD TE-16173, which is a spare tanperature element:
measuring the same system data as TE-6800,- This a
will supply controller TV-6800 a'nore accurate l'
signal for control-of' stator cooling. water tanperature.
1..
Le change does not affect any safety related systems as described in FSAR nor is stator
~
L cooling water required for safe shutdown of.
lant. Reference system definition section 0.2 of FSAR.
1022 l
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11-1989 AIR 1AL REPORT - PART 2 10 CER50.59(b) REPORT j
2.
Le change cannot cause an accident or malfunction
)
of different type not evaluated in section 10.0, 10.2 and 15.1..
3.
L e change does not effect the margin of safety as
. defined in Techaical Specifications. Ref, section 3/4.7..
89-V2M104 Life cable NGV660XA from RTD TE-6800 and land on l
RTD TE-16173, s ich is a spare temperature element measuring the same system data as TE-6800. his j
will supply controller 'IV-6800 a nore accurate,
I signal for control of stator cooling water j
temperature.
1.
The change does not effect any safety related system as described in the FSAR. Stator cooling water is not required for safe-shutdown of the plant. Reference FSAR section 10.2.
2.
21s. change does not, and can noti cause 'an accident or malfuncticn of a cmponent-that is different from those described in FSAR-sectica 15.-
-i 3.
his change does not effect the margin of safety as defined in Tech Spec, section 3/4.7.
89-V1M107 Mr the maghine gas high temperature alarm from:
134'T to 1207. Set ITIS-6846 located in hydrogen and stator gooling water cabinet 1-1326-P5-HSC at 120'T.
1.
he change does not increase the probability of-occurrence or consequences of malfunction of safety related system / equipment. FSAR sections reviewed 10.2, 15.1 and 15.2.
+
2.
Change does not create the possibility of an.
j accident or equipment malfunction not described and analyzed in FEAR, sections reviewed 15.0.8, 15.1 and 15.2.
L 3.
Per review of 2.2.1 and 3/4.3.4 the change.does'not decrease the margin of safety, defined by the bases of the Technical Specifications.
I 103 j
r E
II 1989 N MIAL REPORT - PART 2 10 CFR50.59(b) REPORT 89-V1M104 his nodification lowers the taperature and alarm setpoints for 4 heat tracing circuits. Re existing setpoints are based on the need to inaintain the associated piping at 1707 for 12% boric acid-concentration, iowever, the system operates at 4% boric acid concentration. W e new setpoints are the required netPoints to maintain a mininun tenperature of 65T for 4% boric acid concentration.
1.
h is change decreases the operating tenperature of piping associated with 4 heat trace circuits to maintain the required mininun of 657. Eis does not affect the operation of safety related equipment. Reference FSAR section 8.3.1.1.9.
2.
h is change will not create the possibility for an accident or equipment malfunction. his' change will allow the under tenperature alarms to clear and the systen to function correctly. Reference 8.3.1.1.9 and chapter 15, 3.
h is change will not decrease the margin of safety defined by the bases for the Tech Specs, h is is a setpoint change that reduces the pipe taperature from that recuired for 12% boric acid solution to that requirec, for 4% boric acid solution. Reference 3.3.3.10 of the Tech Specs.
89-V2M108 mange time overcurrent relay tap setting fr:xn TAP 3 to TAP 4 and cycles to operate at 500% of TAP frcan 65 to 55 cycles for breaker 2NAA09 due to addition of a 500 KVA transformer under DCP 89-VCN0038; 1.
his nodification does not increase the probability of occurrence or the ccmsequences of an accident or malfunction of equipment inportant to safety-previously evaluated in the Safety Analysis Report.
We change aniy adjusts the time-overcurrent relays setpoints for breaker 2NAA09 to reflect the addition of a 500 KVA transformer being added under DCP i 89-VCN0038. he new setpoints are coordinated with upstrean crarcurrent relays. his includes a review of FSAR section 8.3, 15.0.8 and 15.2.6.
104
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II 1989 ANNUAL REPORT - PART 2' 10 CFR50.59(b) REPORT 2.
Le modification does not create the possibility of an accident or malfunction of a different type
~
than previously evaluated in the Safety Analysis Report. _ The change only adjusts the time l
overcurrent relays setpoints for breaker 2NAA09 l
to reflect the addition of a 500 KVA transformer
. being added under DCP l 89-VCN0038. his breaker will nw feed the Field Support Building transformer and the Alternate PESB transformer. h e new setpoints are coordinated.
with upstream overcurrent relays, his includes a review of FSAR sections 8.3, 15.0.8 and 15.2.6.
3.
Le change does not decrease the margin of safety defined by the bases for the Tech Specs (See Section 3/4.8.4)._ 1he new setpoints for breaker 2NAA09 are coordinated with upstream overcurrent relays',
i 89-V2M109 Igr the machine gas high taperature alarm fra 1347 to 120 T.
Set 2r1S-6846 located in the hydrogen and stat - cooling water cabinet
?
2-1326-P5-HSC at 120 i
1.
We proposed change does not increase probability of occurence or consequences of malfunction of safety related syst e/ equipment sections reviewed 3
10.2, 15.1 and 15.2.
t 2.
Le proposed change does iat create the_ possibility of an accident or equignent malfunction not a
described and analyzed in FSAR section reviewed 15.0.8, 15.1 and 15.2.
l-3.
Per review of 2.2.1 and 3/4.3.4 the proposed change does not decrease the margin of safety defined by l-the bases of the Technical Specifications.
89-VQ40112 Imer the setpoint of the Electric Fire Pum)
(C-2301-P4-002) safety relief valve, CPSV-13076 fr m 175 psig to 138 psig, his will allw the puun to be o>erated safely with little or no flw L
witaout overaeating or dam ging the punp. A i
bypass flw of 13.75 gpm at dead head pressure fra the pwp discharge is available.
L I
105
~
L II-1989 AtNUAL REPORT - PART 2 10 CFR50.59(b) REPORT 1.
This change involves lowering the setpoint of the safety pressure relief valve for the electric fire ptmp which does not increase the probability of occurrence or consequences of an accident or
. equipcent/ component malfunction described in the FSAR. FSAR sections 9.5.1 and 15.0 were reviewed and require no change. There is no degradation of the " defense-in-depth" Fire Protection Program as a-result of this MDD.
2.
This nodification creates no new possibilities or unanalyzed-scenarios. This is based on a review of FSAR sections 9.5.1 and'15.0.
3.
The safety limits and settings discussed in section 2.0, 3.0, and 4.0 of the VEGP Tech Spec do not deal with fire protection. Therefore, there is no decrease in the Tech Specs margin of safety.
1 106 j
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1989 ANNUAL REPORT - PART 2 i
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10 GR50.59(b) r e
TEST OR EXPERIMENIS L
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I i
II 1989 A!MIAL REPORT - PART 2 l
10 CFR50.59(b). REPORT T-DU-87-10 his arocedure tested the door alams to the PESB JPS area. Alam and tanper supervision -
1 were tested, j
- 1. -
h e FSAR did not analyze the Security system for1 accident probabilities. h e addition of these alams will increase the effectiveness of the security syst e and therefore decrease the probability of an accident. Reference FSAR section i
13.6. Also,-FSAR Section.13.6 details the-
-i industrial security requirements, section 13.6-as well as the VEGP Physical Security Plan was reviewed >No decrease in the security
.l effectiveness would be realized as a result of inplementing this change.
l 2.
h e FSAR did no accident analysis of the Security systm. We addition of alarms will not create the possibility of an accident other than that analyzed in the FSAR.-
3.
Se alarms cm the Security UPS will enhance phvaical security and increase the margin of safety. Tech Specs. did not address security in the basis.
=!
T-ENG-87-14 This procedure tested the installation of ACAT controlled tumstiles in the PESB. Tested were 4
alam capability and turnstile operation.
' Reference FSAR section 13.6.
1.
The addition of turnstiles at the PESB has no j
direct effect on Accident Analysis. We turnstiles i
will decrease the probability of sabotage due to increased security of the main entry point. -
Also, FSAR Section 13.6 details the industrial security requirements, section 13.6 as well as the VEGP Physical Security Plan was reviewed.,No decrease in the security effectiveness would be realized as a result of inplementing.this change.-
2.
L e FSAR did not analyze the Security syst e since it has no direct effect on plant safe shutdown capability. L e addition of turnstiles does not create the possibility of a different type of accident.
3.
The test increases the margin of safety by increasing security at the PESB main entry point.-
. Tech Specs. did not address security in the basis.-
107:
in e
11 1989 At M AL REPORT - PART 2 10 CFR50.59(b). REPORT T-ENG-88-10 his test deacustrates the operability of the Condensate Filter Danin Systan and Transfer Systan, his is a temporary procedure to test the j
system operation Which udmicks a preoperational test that would have been done during plant l
start-up.
1.
Testing this systan does not inpact any equirent '
.1 important to safety or accident analysis. Tais J
system is non-safety related; This included a review of FSAR section 10.4.6;and chapter 15.
3 2.
his nodification involves a non-safety related systan with no-impact an any safety functions or -
equipment. This review included FSAR section 10.4.6 and chapter 15.
i i
3.
his change does not imaact the margin of ' safety as descri:ed in the Teca Spec section B 3/4.7L i
T-ENG-88-13 To provide test nethod for performing a-functional t
test on.the BTRS cannon chiller upon inplementation of DCP 87-VCE0230. Test will verify proper. limit indicaticn on handswitches-1-HS-390 and 2-HS _0390 i
as to the unit operating-the chiller.
- j 1.
During performance of.this test the BTRS chiller, l
A-1208-E6-008, is not actually started because
- 3ower leads.to the ccupressor motor are lifted.
1 3ecause the Bhs systan is not required for -
reactivity. control, his test will not inpact or-affect plant operation, nor will it. increase the a
probability of occurance or consequences of an accident. This is based on a review of FSAR section 9.3.4.
2.
We BTRS systen will not actually be placed'in service during this test and it is not required to mitigate any accident.. h e possibility of an accident or malfun: tion of a different type than i
that evaluated in the FSAR is not created.
3.
Le BTRS systan does not have any associated Technical Specifications based on review of 3/4.1 for. reactivity control in Technical Specification and therefore does not reduce ~the l
margin of safety.
l l
N
II 1989 ANNUAL REPORT - P/RT 2 10 CFR50.59(b) REPORT T-HC-89-02 To test the Control Rocxn logic of the Control Rocxn unergency filtration units to insure that no nore than one(1) of the two(2) fans will start on a Safety Injection / Control Rocxn Isolation signal.
1.
his test verifies that the systen will operate within the design parameters after installation of nodification DCP 83-V1N0079. This test sinulates events described in FSAR section 6.4 to assure ccinpliance, herefore, this test does not increase the probability of occurrence or the consequence of an accident or malfunction of equipment.
2.
his test sinulates events described in section 6.4 of the FSAR to assure ecuiinent response is as evaluated in the FSAR anc. twrefore, does not create the possibility of an accident or malfunction of equipment.
3.
This test verifies that the correct requirenents of Technical Specifications 3/4.7.6 are met by design change package 88-VIN 0079. W erefore, tim unrgin of safety is not required.
T-DU-89-06 he purpose of this. experiment is to determine if there is a tanperature prob 1mi in the Unit 1 and Unit 2 north and south MSIV areas.
1.
histestinvolvesthetemporarfinstallationof thermocouples and a recorder in each MSIV roan.
It will not' affect the operation of permanent plant equipment.
2.
he testing eculpment installed in the MSIV rooms will be placec, to avoid any possible impact on f
pennanent plant equipcznt.
3.
his test will not affect any pennanent plant equipment nor will this margin of safety for the Tech Spec. be affected.
109 1
L II 1989 AIM RL REPORT - PART 2 10 CIR50.59(b) REPORT T-ENG-89-08 he test establishes and maintains unit coerating conditions required for performance of a baseline '
secondary alant performance test, he test requires t wt the plant remain stable at approximately 100% pcwer with significant leakage paths into or out of the Turbine cycle isolated.
1.
If the cycle isolation wal'akun identifies the'.
main steam dmps as excessively leaking valves,
-the isolation of these valves would increase the probability of lifting a main steam safety valve upon a reactor trip. Ikuever, since administravely only a few steam dmps would be allowed to be
'solated at one time, the increase in probability of lifting a main steam safety is very very small.
No credit is taken for the main steam dumps in the safety analysis.
In addition, a reactor trip we' tid have to occur during the short period that these valves would be isolated.
2.
This test does not manipulate or modify any safety grade equipment and thus does not create the possibility of an accident or malfunction of a different type than previously evaluated in the-safety analysis report.
3.
We test does not manipulate any safety grade equipment and therefore does not reduce the margin of safety as defined in the Tech Specs.
T-ENG-89-09 h e test procedure valves-in and measures the zero shift of temporary high precision test feedwater flow differential pressure transmitters. Rese test transmitters are used in a baseline secondary plant performance test.
1.
We applicable feedwater flow channel-is reaved fram service prior to valving 3n and measuring the zero shift of-the test transmitter. Therefore,' the test does not increase the probability of occurrence or the consequences of an accident.or malfunction of any plant equiptwnt.
110
3,.
~
- v
\\
II..
n-1989 ANNUAL REPORT - PART 2.
10 CFR50.59(b) REPORT 2.
Since the' applicable feedwater flow channel is out-J of service Wwn-this-test is performed, it does not create the possibility of an accident or malfunction of a different type than previously evaluated in the-safety analysis report.
.l 3.
Since the test is performed on equipment which is.
-out of service, there is no reduction in the margin!
q of safety as defined in the Tech Spec.
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III 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 8
V
III i
V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 & 2 1989 ANNUAL REPORT PART 2-I EMERGENCY CORE COOLING SYSTEM OUTAGE DATA REPORT j
This report contains:
a outage dates and duration of outages b
ECCS systems or components involved in the outage i
e cause of the outage, and d
corrective actions taken j
- UNIT 1 -
Unit 1 Emergency Core Cooling System components were out of service a totai of 620 hours0.00718 days <br />0.172 hours <br />0.00103 weeks <br />2.3591e-4 months <br /> and 11 minutes in 1989.
1.
a 1 7-89 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> and 53 minutes b &c)
RHR isolation Valve found with wrong size interlock fuse, d
Fuse changed and valve returned to service.
2.
a) 1 20 89 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 6 minutes b) & c)
Safety injection Pumps removed from service to comply with Technical Specifications, d)
Pumps returned to service.
3.
a 1-21 89 115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> and 18 minutes b &c)
Accumulator Isolation Valve torqued closed, d
Valve manually opened off of closed seat, and returned to service.
4.
a 2 21-89 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and 24 minutes b &c)
Train A Safety. Injection Pump maintenance outage.
d Maintenance completed and pump returned to service.
l l
l' i
5.
a) 2 28 89 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> and 2 minutes b)&c)
Train B Safety Injection System preventive maintenance service, d)
Servicing completed and system restored to service.
6.
a) 3 8 89 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> and 50 minutes b)&c)
Train B Component Cooling Water Pump has various small leaks, d)
Leaks repaired and pump returned to service.
7.
a) 3 10-89 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 40 minutes l
L b) & c)
RHR Heat Exchanger Outlet Valve requires calibration.
I d)
Calibration complete and valve restored to service.
l l
8.
a) 3-11-89 39 minutes b) & c)
RHR Heat Exchanger Outlet Valve removed from service for I&C inspection, j
d)
Inspection completed and valve returned to service.
9.
a) 3 20-89 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> and 20 minutes b)&c)
Boron Injection Tank Discharge Isolation Valve limit switch out ofadjustment, d)
Limit switch adjusted and valve restored to service.
10.
a 4 13 89 47 minutes b &c)
RHR Hot Leg Isolation Valve removed from service for testing.
d Testing completed, valve restored to service.
11.
a 4 13 89 7 minutes b &c)
Train A RHR Pump placed in pull-to lock for valve testing, d
Testing completed, pump returned to service.
12.
a 4-13 89 39 minutes b &c)
Train B RHR removed from service for testing.
d Testing completed, Train B returned to service.
13.
a 4 13-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 38 minutes b &c)
RHR Hot leg Isolation Valve removed from service for testing.
d Testing completed, valve restored to service, i
14.
a) 5 16 89 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> and 56 minutes b) & c)
Train A Centrifugal Charging Pump removed from service for maintenance outage.
d)
Outage completed, pump restored to service.
l 15.
a) 5 22 89 21 minutes b) & c)
Charging Pump Miniflow to Refueling Water Storage Tank Isolation Valve breaker racked out to replace thermal overload, d)
Thermal overload replaced, valve restored to service.
i i
l 16.
a) 5-23-89 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> and 6 minutes l
b) & c)
Train B RHR removed from service for maintenance outage, d)
Outage completed, Train B returned to service.
17.
a) 5-26-89 20 minutes l
b) & c)
Refueling Water Storage Tank to Charging Pump Valve Breaker turned off to replace thermal overload.
d)
Thermal overload replaced, valve restored to service.
18.
a) 5-26-89 14 minutes b) & c)
Chaging Pump Miniflow Isolation Valve Breaker racked out to replace thermai overload.-
d)
Thermal overload replaced, valve restored to service.
19.
b))&c) a 6 5-89 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 15 cinutes Train B Safety Injection Pumn Suction Header Valve and Charging Pump Header to Safety Inject ion System Valve thermal overload bypasses require replacement.
d)
Thermal overload bypasses replaced, valves returned to service.
20.
a) 6-7 89 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 21 minutes b)&c)
Train A Safety Injection Miniflow Isolation Valve removed from service'for thermal overload jumper replacement.
d)
Jumper replaced, valve restored to service.
21.
a) 6 7-89 41 minutes b) & c)
Train A RHR Pump Inlet Valve removed from service for thermal overload jumper replacement.
d)
Jumper replaced, valve returned to service.
22.
a) 6 7-89 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 25 minutes b)&c)
Train A Safety Injection Discharge Isolation Valve removed from service for thermal overload jumper replacement, d)
Work stopped, valve returned to service.
23.
a) 6-12-89 26 minutes b)&c)
Train A Safety injection Discharge Isolation Valve removed from service for thermal overload jumper replacement, d)
Jumper replaced, valve returned to service.
24.
a) 6 13 89 14 minutes b) & c)
Charging Pump Miniflow to Refueling Water Storage Tank Isolation Valve removed from service for thermal overload-jumper replacement, d)
Jumper replaced, valve returned to service.
25.
a) 61389 16 minutes b) & c)
Volume Control Tank Outlet Isolation Valve removed from service for thermal overload jumper replacement, d) jumper replaced, valve returned to service.
26.
a) 6-14-89 12 minutes b) & c)
Train B RHR Containment Sump Check Velve removed from service for testing.
d)
Testing completed, valve restored to service.
27.
a) 6 14-89 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 2 minutes t
b) & c)
Train A RHR check valve removed from service to inspect for possible loose pin.
d)
Valve inspected OK and returned to service.
28.
a) 6 14-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 11 minutes 4
b)&c)
Train A RHR check valve removed from service for_ testing, d)
Testing completed and valve returned to service.
I L
- -_ __ _ -_N
i 29.
a) 6 14 89 22 minutes l
b) & c)
Charging Pump to Reactor Coolant System Isolation Valve removed
{
from service for thermal overload jumper replacement.
d)
Jumper replaced, valve returned to service.
]
i 30.
a) 6-14 89 20 minutes b)&c)
Refueling Water Storage Tank to Charging Pump Valve removed from service for thermal overload jumper replacement.
d)
Jumper replaced, valve restored to service, e
i 31.
a) 6 25 89 38 minutes b) & c)
Train B RHR Heat Exchanger to Safety injection Pump Valve removed from service for thermal overload jumper replacement.
1 d)
Jumper replaced, valve restored to service.
32, a).
6-25 89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 30 minutes b) & c)
Train B Containment Sump Isolation Valve removed from service for thermal overload jumper replacement.'
t d)
Jumper replaced, valve restored to service.
33, a) 6-25 89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 59 minutes b) & c)
Train B RHR Pump Inlet Valve, Train B RHR Hot Leg Isolation Valve, and RHR. Pump Miniflow Valve : removed from service for thermal overload jumper replacement.
d)
Jumpers replaced, valves restored to service, i
34, a) 6 30-89 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 58 minutes b)&c)
Train B Centrifugal Charging Pump Discharge Valve closed for i
- testing, d)
Testing completed, valve returned to service.
(
35.
a) -
7 9 89 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 41 minutes h
b) & c)
Train A RHR Heat Exchanger Outlet Valve removed from servirs for testing, d)
Testing completed, valve returned to service.
36, a) 7-17 89 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and 44 minutes b) & c)
Train B Centrifugal Charging Pump removed from service due to inboard motor bearing oil leak.
d)
Leak repaired, valve restored to service.
-37.
a 7-20 89 13 minutes-i b & c)
Train B RHR removed from service for valve testing, d
. Testing completed Train B returned to service.
38.
a) 8 24 89 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 17 minutes-b) & c)
Train B RHR Heat Exchanger Outlet Valve. removed from service for preventive maintenance.
d)
Maintenance completed, valve returned to service.
39.
a) 9-12-89 22 minutes b) & c)
Train B RHR Containment' Sump Check Valve removed from service for testing.
d)
Testing-completed, valve returned to service.
40.
a) 10-2 89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 12 minutes b) & c)
Train A RHR System removed from service to fill and vent letdown system.
d)
Fill and vent complete, RHR System returned to service.
41, a) 10-3-89 less than one minute.
b) & c)
Train A RHR electrical breaker removed from service to search for ground.
~
d)
Breaker closed and restored to service.
42, a
10 13-89 4 minutes b & c)
. Train B RHR Pump put in Pull to-Lock for testing.
d Testing completed, pump restored to service.
43, a
10-14-89 162 hours0.00188 days <br />0.045 hours <br />2.678571e-4 weeks <br />6.1641e-5 months <br /> and 14 minutes-b & c)
Train B Centrifugal Charging Pump Alternate Miniflow Relief Valve leaking.
d)
Valve replaced, new valve entered service.
44.
a 10-19 89-42 hours and 14 minutes b & c)
Train B Centrifugal Charging Pump removed from service to repair the Centrifugal Charging Pump Miniflow to Refueling Water Storage Tank Isolation Valve.
d)
Valve repaired, pump returned to service.
45, a) 10-24-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 31 minutes b) & c)
Train A Centrifugal Charging Pump removed from service for
- testing, j
d)
Testing completed, pu p returned to service.
m 46, a) 10-31-89 32 minutes b)&c)
Train A Centrifugal Charging Pump Discharge Valve removed from service for testing.
d)
Testing completed', valve restored to service.
47, a
12-7-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 18 minutes b & c)
RHR Hot leg Isolation Valve removed from service for testing.
d Testing completed, valve restored to service.
48.
a) 12-18 89 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> and 38 minutes b) & c) 1 rain A RHR Pump, Heat Exchanger Outlet Valve, and Heat Exchanger Outlet Valve removed from service for testing.
d)
Testing completed, valves returned to service.
I i
- UNIT 2 -
l t'
Emergency Core Cooling System components'were out of service a total of 283 l
hours and 15 minutes in 1989.
k 1.
a 2 19-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 10 minutes b &c)
Train E RHR Pump taken to Pull to Lock for valve testing, d
Testing completed, pump restored to service.
l l
2.
a 2-20-89 134 hours0.00155 days <br />0.0372 hours <br />2.215608e-4 weeks <br />5.0987e-5 months <br /> and 40 minutes b &c)
RHR Loop 4 Isolation Valve failed to calibrate.
i d
Valve repaired and returned to service.
3.
a) 2-20 89 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 51 minutes b)'&c)
Train B RHR Pump taken to Pull-to-lock for pressure transmitter testing.
3 d)
Testing completed, pump returned to service, j
4.
a 2 21-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 54 minutes b & c)
Train B RHR Pump taken to Pull-to Lock for testing.
d Testing completed, pump returned to service.
5.
a) 2-25-89 27 minutes b) & c)
Train B RHR Pump taken to Pull-to-lock for valve testing, i
d)
Testing completed, pump restored to service.
6.
a) 2-27-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 17 minutes I
b) & c)
RHR Suction Relief Valve removed from service.for response time testing.
d)
Testing completed, valve restored to service.
L 7.
a 2-28-89 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 16 minutes I
b & c)
Train A RHR Loop Suction Valves shut for response time testing.
l d
Testing completed, valves restored to service.
i l
8.
a) 3 9 89 5 minutes
[
L b) & c)
Train A Safety Injection Miniflow Valve closed for testing.
d)
Testing completed, valve re opened.
9.
a) 3 9 89 3 minutes b) & c)
Train B Safety Injection Miniflow Valve closed for testing.
d)
Testing completed, valve re-opened.
1 10.
a) 3-9-89 13 minutes b) & c)
Both Trains of RHR declared inoperable when isolation valves found open.
d)
Valves closed, system restored to service. -
11.
a) 3-12-89 18 minutes b)&c)
Train A RHR Pump stopped and removed from service when discharge pressure dropped.
d)
Bad instrument reading verified, pump returned to service.
10.
a 3 18-89 35 minutes b &c)
SI Pumps inoperable upon Mode 3 entry.
d Breakers racked in, pumps returned to service.
13.
a) 3-19-89 33 minutes b) & c)
Automatic Safety Injection signal blocked while troubleshooting.
d)
Problem corrected, block removed, automatic SI signal capability restored.
14.
a) 3 21-89 6 minutes b) & c)
Train A RHR to Cold leg -Safety Injection Isolation Valve closed for. testing, d)
Testing completed, valve reopened and returned to service.
15.
a) 3 21-89 10 minutes b) & c)
Train B RHR to Cold leg Safety Injection Isolation Valve closed for testing.
d)
Testing completed, valve reopened and returned to service.
16.
a 3 30-89 10 minutes b &c)
Train B RHR Pump taken to Pull-to-lock for valve testing, d
Testing complete, pump restored to service.
17.
a 4-7-89 33 minutes b &c)
Accumulator #2 removed from service-due to low pressure, d
Pressure restored, accumulator returned to service.
18.
a 4 14-89 41 minutes l
b &c)
Accumulator #2 removed from service due to low pressure, d
Pressure restored, accumulator returned to service.
19.
a 5-2-89 13 minutes b &c)
Accumulator #4 removed from service due to low pressure, d
Pressure restored, accumulator returned to service.
20.
a 6 9-89 14 minutes b &c)
Accumulator #3 sample valve opened for sample, d
Sample valve closed, accumulator restored to service.
21.
a) 6-13-89 13 minutes j
b) & c)
Train A RHR Pump removed from service for testing.
i d)
Testing completed, pump returned to service.
1 m
22.
a 6-14 89 7 tinutes b &c)
Train B RHR Pump removed from service for testing.
d Testing completed, pump restored to service.
23.
a) 6 23-89 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and 25 minutes b) & c)
Centrifugal Charging Pump Miniflow to Refueling Water Storage Tank Isolation Valve removed from service for maintenance.
d)
Maintenance completed, valve restored to. service.
24.
a) 7 25 89 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 12 minutes b) & c)
Train A Safety Injection Pump Miniflow Isolation Valve removed from service fer repair.
d)
Repair complete, valve restored to service.
25.
a 7-29 89 50 minutes b &c)
Accumulator #4 removed from service due to low pressure, d
Pressure restored, accumulator returned to' service.
26.
a) 8-30 89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 59 minutes.
b)&c)
Train A RHR removed from service for miniflow control calibration.
d)
Calibration completed, Train A returned to service.
27.
a) 9-25-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 1 minute b) & c)
Train A RHR Transfer Switch removed from service for repair.
d)
Repair completed, switch restored to service.
28.
a) 10-1-89 91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> and 34 minutes-b)&c)
Refueling Water Storage Tank Sludge Mixing Isolation Valve removed from service when indicator light falls.
d)
Replaced bulb sockets, valve returned to service.
i 29.
a) 10-10 89 18 minutes l
b) & c)
Train A RHR to Safety Injection Cold leg Isolation Valve i
removed from service for testing.
l d)
Testing completed, valve restored to service.
30.
a) 10-14-89 7 minutes b) & c)
Train B RHR Pump taken to Pull-to-Lock for testing, d)
Testing completed, pump returned to service, i
31.
a) 10-14-89 17 minutes b) & c)
Train B RHR valves removed from service for ISI_ testing.
d)
Testing completed, valves returned to service.
-r
l
- 32..
a) 10 14 89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 18 minutes b) & c)
Train B Chemical and Volume Control System Valves removed from service for ISI testing.
d)
Testing completed, valves returned to service.
33, a
10-15 89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 14 minutes b &c)
Si valves removed from service for stroke testing, d
Testing completed, valves returned to service.
34.
a) 10-19-89 19 minutes b) & c)
Train B Centrifugal Charging Pump Discharge Valve removed from service for testing.
d)
Testing completed, valve restored to service.
35.
a) 10-19 89-5 minutes b) & c)
Train A Centrifugal Charging Pump Discharge Valve removed from service for testing, d)
Testing completed, valve restored to service.
36, a) 10-23-89 38 minutes b) & c)
Train A Centrifugal Charging Pump Discharge Valve removed from service for testing, d)
Testing completed, valve restored to service.
37, a) 10 31-89 32 minutes b)'&c)
Train A Centrifugal Charging Pump Discharge Valve removed from service for testing, d)
Testing completed, valve restored to service.
38, a) 11 9 89 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 59 minutes b)&c)
Train B RHR Heat Exchanger Discharge Valve removed from service t.
for preventive maintenance.
l d)
Testing completed, valve restored to service.
39, a
11-26-89 5 minutes l
b &c)
Train A RHR Pump taken to Pull-to-Lock for valve testing.
d Testing completed, pump returned to service.
40, a) 12-12-89 46 minutes b) & c)
Train A Centrifugal Charging Pump Discharge Valve removed from service for testing.
d)
Testing completed, valve restored to service.
41, a) 12-14-89 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 54 minutes b) & c)
Train B Centrifugal Charging Pump Discharge Valve closed for testing.
1 d)
Testing completed, valve reopened and restored to service.
42.
a) 12-15 89 30 minutes i
b) & c)
Train B Centrifugal Charging Pump Discharge Valve closed for testing.
d)
Testing completed, valve reopened and restored to service.
43.
a) 12-15 89 14 minutes b) & c)
Train A Centrifugal Charging Pump Discharge Valve closed for testing.
l d)
Testing completed, valve reopened and restored to service.
l 44.
a) 12-29 89 9 minutes b) & c)
Chemical and Volume Control System Valves removed from service for ISI testing.
d)
Testing completed, valves restored to service.
I i
i i
i
IV 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 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT CALENDAR YEAR 1989 i
e i
7-V0GTLE ELECTRIC GENERATING PLANT RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT TABLE OF CONTENTS SECTION TITLE EAGE
1.0 INTRODUCTION
1 1 2.0
SUMMARY
DESCRIPTION.........,........
2-1 3.0 RESULTS
SUMMARY
3-1 4.0 DISCUSSION OF RESULTS 4-1 4.1 Airborne 4-3 4.2 Direct Radiation 4-5 4.3 Milk........................
47 4.4 Vegetation.....................
48 4.5 River Water...................
4-9 4.6 Orinking Water....................
4-11 4.7 Fish........................
4-13 4.8 Sediment 4-15 4.9 Aquatic Vegetation 4-16 5.0 INTERLABORATORY COMPARIS0N PROGRAM............ 5 1
6.0 CONCLUSION
S 6-1 l
l i
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- .i..............
' j.
^
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LIST OF TABLES TABLE TITLE PAGI 2-1
SUMMARY
DESCRIPTION OF RADIOLOGICAL l
ENVIRONMENTAL MONITORING PROGRAM 2-2 2-2 RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS 2-7.
31 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
3-2 4-1 LAND USE CENSUS RESULTS 4-2 4-2 RESULTS SUMMA 2Y FOR AQUATIC VEGETATION 4-17 S-I CROSSCHECK PROGRAM HESULTS 52 4
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LIST OF FIGURES i
FIGURE TITLE g
2*I TERRESTRIAL STATIONS NEAR SITE BOUNDARY 2 10 2-2 TERRESTRIAL STATIONS BEYOND SITE BOUNDARY OUT TO APPROXIMATELY SIX MILES AND RIVER STATIONS.
2-))
23 TERRESTRIAL STATIONS BEYOND SIX MILES 2 24 DRINKING WATER STATIONS p.13 I
l l
l ili l
I
. ACRONYMS CSRA Central. Savannah River Area
'. j
.i CY Calendar Year EL Environmental Laboratory i
EPA'-
Environmental Protection Agency GPC Georgia Power Company l
'LLD-Lower Limit'of Detection l
MDD Minimum Detectable Difference
-MDA
~ Minimum Detectable' Activity NA Not Applicable
'i
_NDM
' No: Detectable Measurement (s)
NRC Nuclear Regulatory Commission l
- 0ffsite Dose Calculation-Manual i
1 REMP Radiological Environmental Monitoring' Program RL Reporting Level
]
RM River-Mile j
SRS-Savannah River Site TLD-
-Thermoluminescent Dosimeter j
TS Technical Specifications for Unit l'and Unit 2 VEGP
' Alvin W. Vogtle Electric Generating _ Plant i
i a
iv
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(;,7
-V0GTLE ELECTRIC GENERATING PLANT RADIOLOGICAL ENVIRONMENTAL-SURVEILLANCE' REPORT
1.0 INTRODUCTION
This is the third annual Radiological Environmental Surveillance Report for the Alvin W. Vogtle Electric Gen'. rating Plant (VEGP).
It covers activities of the Radiological Environmental Monitoring Program-(REMP) duringlcalendar year (CY) 1989. Hence all dates in this report are for; the yearL1989 unless otherwise-indicated. The specifications for the
- REMP are provided by. Section 3/4.12 of the? Technical: Specifications _ for -
-.1 1
Unit I and Unit 2 (TS).
The objectives'of the REMP are to ascertain,the levels of radiation.and-the concentrations of radioactivity in the environs of VEGP'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 i
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 j
this assessment, j
~
o The preoperational stage of the REMP started in' August of 1981 when the j
initial collections of the radiological environmental _ samples were made; there-was a phase -in period of_ a few years-before the preoperational-l program was fully implemented. The transition.from the preoperational 1
stage to the operational. stage hinged-about. initial criticality for Unit--
1 1 which occurred on March 9,:1987 ; A low power operating ~ license for i
Unit 2 was obtained on' February 9,1989; initial criticality for Unit 2 1
occurred on March 28.
~
L 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 i
main samples utilized for environmental-monitoring;is presented in Section 3.
A discussion of-the results including: assessments of any-radiological impacts upon the environment is provided in Section 4.
The results of the Interlaboratory Comparison Program are presented in Section 5.
The chief conclusions are stated in Section 6.
- i 1-1
2.0
SUMMARY
DESCRIPTION a
'A summary description ~of the REMP:is-provided in Table 2-1..This-table portrays the program in the manner by which it11s being regularly carried out; =1t is essentially a copy of Table 3.12-1 of the TS which delineates-the program's-requirements.: Sampling-locations specified by
.l Table 2-l'are described in Table ~2 and are shown on maps in Figures
' 1 through 2-4.
This description of the sample' locations closely followsithat found in the table and figures of Section 3.0'of the ;
Offsite Dose Calculation Manual -(00CM).
It is stated in Footnote (1) of Table 3.12-1 of the TS that deviations-are' permitted from the required sampling-schedule (which is delineated.
in Table:2-1 herein), 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~. -
During 1989, all: the-laboratory analyses except for the reading of the thermoluminescent dosimeters :(TLDs) were performed by Georgia' Power Company's (GPC's) Environmental Laboratory-(EL). in' Smyrna, Georgia. The -
EL was previously called the Central Laboratory. TheLreading of the TLDs was-provided by~Teledyne Isotopes Midwest Laboratory'in Northbrook, Illinois.
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TABLE 2-1 (SHEET I 0F 5)
SUttiARY DESCRIPTION OF RADIOLOGICAL-ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE' SAMPLES ~
SAMPLING AND TYPE AND FREQUENCY.
AND/OR: SAMPLE AND SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS
- 1. Direct Radiation (l)
Thirty-seven routine monitoring Quarterly Ga m h dose quarterly.
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; and
'?
The balance of the stations to be placed in -special interest areas
~
such as population centers, nearby
. residences, schools, and in one or two-areas to serve as control stations.
_ _ _ _ _ _ _ = _.
- -..
= = -
1 2
MEEME _
TABLE 2-1-(SHEET 2.0F 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES SAMPLING ~AND-TYPE AND FREQUENCY' AND/0R SAMPLE
- AND SAMPLE LOCATIONS
' COLLECTION FREQUENCY-0F ANALYSIS
- 2. Airborne Radioiodine and-Samples from seven locations Continuous-sampler oper-Radioiodine Cannister:
Particulates ation withl sample collec--
I-131 analysis weekly.
tion weekly, or more:
Five samples from close to the frequently if required by five site boundary locations, dust loading' Particulate SamDier; in different sectors; Gross beta' analysis (2):
following filter change;-
and gamma isotopic' analysis (3) of composite y
(by location) quarterly.
One sample from the vicinity of w
a community'having~the highest calculated : annual average ground-level-D/Q;_and ~
One sample from a control location,
'as for example, a population center-
~
at_a distance of 10 to 20 miles..
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TABLE 2-1 (SHEET 3 0F 5)
SUP9lARY DESCRIPTION 0F RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.
EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND.
TYPE AND FREQUENCY ~
AND/0R SAMPLE AND SAMPLE LOCATIONS COLLECTION FREOUENCY OF' ANALYSIS
- 3. Waterborne
- a. Surface (4)-
-One sample upriver Composite sample over Gamma isotopic' analysis (3) 1-month period (5)-
monthly. : Composite for.
Two samples downriver
' tritium ~ analysis quarterly.
- b. Drinking Two samples.at each of the two Composite sample of'
.I-131 analysis on each nearest water treatment-plants river water near the
' sample when the dose that could be affected by plant intake'at each water
. calculated for the discharges.
treatment plant over consumption of the water.is 2-week period (5) when.
" greater than 1-arem par
- ro
' year (6). Composite for-1.
Two samples at a control.
I-131 >tnalysis is!
location.
required to be performed;
. gross beta and. gamma-isotpic on each~ sample, monthly
' analyses:(3) on raw water-
-composite otherwise; and monthly. Gross beta, gamma
-grab sample of finished isotopic and I-131 analyses:
water at'each water on grab sample of finished treatment plant every 2 water monthly. Composite-weeks or monthly, as-
- for. tritium analysis on raw-appropriate
.and finished water quarterly.
- c. Sediment from One sample from downriver area Semiannually.
Gamma isotopic analysis (3)-
Shoreline with existing or potential semiannually.
recreational-.value.
One sample from upriver area with existing or potential recreational value.
_ _ _ _ _ _ _ ~. _
TABLE 2-1 (SHEET 4 0F 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPREGENTATIVE SAMPLES SAMPLING AND
' TYPE AND FREQUENCY AND/0R SAMPLE AND SAMPLE LOCATIONS COLLECTION FRE00ENCY OF ANALYSIS
- 4. Ingestion
- a. Milk Two. samples from milking animals (7) 6~ weekly-Gamma isotop)ic-analysis (3,8 biweekly.
at control locationslat a distance of about 10 miles or more and preferably in a wind direction of
. lower prevalence.
~b. Fish At least one: sample of any commer-Semiannually Gamma isotopic analysis (3) cially or recreationally:important.
on edible' portions species in' vicinity of plant semiannually
.?
discharge area.
. u, At least one sample of any species in areas.not influenced by plant discharge.
At least one sample.of.any
.During spring spawning anadromous species in vicinity of season.
. plant discharge.
. Gamma isotop)ic
- c. Grass or Leafy.
One' sample _ from two onsite locations Monthly during growing analysis (3,8. monthly Vegetation.
nearLthe site boundary:in different season.
. sectors.
One sample from a control location at about 15 or moreLmiles distance.
~:--
= = - - =. -
- = - -
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1,3 TABLE 2-1:(SHEET.;5 0F 5).
I
SUMMARY
. DESCRIPTION 0F-RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM j
' TABLE NOTATIONS.
4
.(1)
One or more instruments, such as a pressurized ion chamberi.for continuously. measuring and recording acquired' dose may be used in place of, or in addition to, integrating dosimeters.. For.the purposes of this
. table, a TLDJs considered to be one phosphor; two.'or more phosphors in-4
- a. packet _are-considered as two or more dosimeters.
Film badges-shall not be used as dosimeters for measuring. direct radiation.:
(2)
Airborne particulate sample filters shall be analyzed for gross beta i
radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or. more after sampling. to. allow for radon and -
thoron daughter decay., If gross beta = activity in. air particulate; samples is greater than 10' times the' yearly mean of control? samples, gamma isotopic analysis: shall be performed on the individual samples.
]
-(3).
Gamma isotopic analysis means the identification:and quantificiation of.
i gamma-emitting radionuclides that may be attributable to the effluents-from.the facility.
(4)
- The upriver samples are taken at. distances beyond significant; influence -
of thesdischarge. The downriver.' samples are.be taken in: areas; beyond:
and near the mixing zone..
j F
(5)'
Composite sample aliquots.shall be collected at time: intervals that are -
very:short (e.g., hourly) relative to the compositing period-(e.g.,
1
~
monthly) in order to assure obtaining a-~ representative sample.;
(6)
The dose shall be calculated for the maximum organ and-' age: group, using the methodology and parameters in the.ODCM.~
1 (7)
A milking animal is a cow or goat producing milk for human consumption.
(8)
If gamma isotopic analysis is not sensitive enough to meet the' Lower-r E
Limit of Detection = (LLD)-for I-131, & separate-analysis'for I-1311will be performed.
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I TABLE 2-2: (SHEET'l 0F 3)
RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station:
Station Descriptive-Direction (2) Distance-(2)'. Sample Number
'Tvoe (1) location (miles)-
~Tvoe (3) 1.
I Hancock Landing Road N
1.1 D
i 2
I-River Bank' NNE 0.8 0
Discharge Area-NE 0.6' A.
it
-I s:
.3-I.
River Bank
.NE 0.7
-0 4'
I River Bank ENE 0.8-
'D 5
-I River Bank E
1.0 D
6 I
Plant Wilson ESE 1.1 D:
7
~I Simulator Building SE 1.7-0,V,A-8.
I River Road.
-1.1 D-9
-I River Road S
1.1 0-f 10 I
Met-Tower SSW 0.8-A 10 I-
. River Road SSW
- 1. I' D
.11 I
River Road:
SW-
.l.2 D
12 I
River Road WSW l.2
.D.A 13 I
River Ro'ad W
l.3 D.
1 14' I
River Road
.WNW 11.8-D-
15.
I Hancock Landing Road NW 1.5 D,V
-l 16 I
Hancock Landing Road NNW-1.4.
D,A 1
17 0
' Savannah River-Site (SRS) River Road N
5.4 D
18-0 SRS D Area.
NNE 5.0 D
'19 0
.SRS Road ~A.13 NE.
4.6~
D 20 0
SRS Road A.13.1-ENE 4.8 D.
21~
0 SRS Road ~A.17 E~
' 5.3
'D
- l 22-0 River Bank-Downstream-1 of Buxton Landing-ESE' 5.2 D
H 23 0
' River Road SE:
'4.7 D
24-0 Chance Road SSE 4.9 D
25 0
Chance Road and Highway 23 S 5.2 D
l~
26 0
Highway 23, Mile-15.5 SSW 4.6-
'D 27 0-Highway 23, Mile 17
_SW-4.8 0
28 0
Claybon Road WSW 5.0 0
29 0
Claxton-Lively Road W
5.0 0
L 30 0
Nathaniel Howard Road WNW 5.0 D
31 0
River Road at Allen's-Church Fork NW-b.s D
32 0
River Bank NNW 4.6 0
33 0~
Nearby-Permanent Residence.SE.
~3.3-D-
s E
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TABLE 2-2 (SHEET 2 0F 3).
RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station' Station
-Descriptive-Direction (2) Distance (2)
Sample Number,_
Tvoe-(1) location-
~
(miles)
Tvoe-(3)
+35 0
Girard SSE 6.6 D,A 36 C:
Waynesboro' WSW 14.9 D,A-37 C
Substation-(Waynesboro)
'WSW 17.5 D,V 43-0 Employees Recreation Area' SW-2.2' D
80 C
Augusta Water Treatment j
P1 ant-
- ". NNW-27.5
'W(4)'
81-C Savannah River N
2,2-
_F(5),S(6)
-82 C-Savannah.. River-(RM 151.2)
NNE-0.8 R
83 I-Savannah River'(RM 150.6)-
ENE __
0.8 R,S(6)
'84 0
Savannah River (RM 149.5)-
ESE_ '
l6-R-
i 85 I
. Savannah River
~
ESE-5.0-
-F(5) 87 I
Beaufort-Jasper County-Water Treatment Plant; Beaufort, SC-
.SE.
76' W(7)-
-88 I-Cherokee Hill Water Treatment Plant; i
Port Wentworth, GA SSE 72 W(8)-
98-C W. C. Dixon Dairy SE-9.8~
M i
99 C-Boyceland Dairy
.W-24.5 M
j 4
TABLE NOTATION:
-i (1)
Station Types
'C - Control' I - Indicator Other 1
(2)
Direction and distance are reckoned from a point midway between the two reactors (3)
Sample Types A - Airborne Radioactivity
-l 0 - Direct Radiation j
F - Fish i
M - Milk
. j R - River Water 1
S - River Shoreline Sediment 1
W - Drinking l Water-1 V - Vegetation
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TABLE 2-2'(SHEET 3-0F 3)
RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Y
TABLE NOTATIONS (Continued) l
. (4)'
The intake for the Augusta Water-Trestme'nt'P1 ant-is located on the\\
Augusta Canal...The entrance to this canal is at' River Mile. (RM) 207 son:
the Savannah River. The canal-effectively parallels the river.
The.
intake to the pumpingostation is 3.6 miles down the canal and only_ a -
%j tenth of a mile across a narrow neck of land to the river.
I (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 M
149.7 and;150.7 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 153.2:for upriver collections.
1 (7)-
The intake for the. Beaufort-Jasper County Water Treatment Plant ~is -
located at the end of a canal which begins;at _RM 39.2 on the Savannah
~~
River. This intake is about 16~ miles by line.ot sight down..the canal from its-beginning on the Savannah River.
(8)
The intake for the Cherokee Hill Water Treatment Plant is located on I
Abercorn Creek which is about one and a quarter creek miles from its mouth on the Savannah River at.RM 29.
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- u MAMP1'ON O.
e'
'=
i
=== aam
,[
- . (
(
f 8
\\
8***
TERRESTRIAL STATIONS voens g h g g a e n e sses asw BEYOND 6 MILES FIGURE 2-3 sas.,
2-12 l
1
-1 AUSWTa es& Ten Tasaftesoff PLANT SOUTH CAROLINA savawNaN neven P6aw
(
VEGP 4 t
i 3
Consi es aussey nieur :
asaw = eemusen 1
I namenenw user Tsunament pennt -
l N
s i
l
\\
GEORGI A -
p
'#' \\
mm O
'"'"*""'" e v+g 0
10 20 t
waren vasafesswT e
i I
I e6am
[D4 mi1es spony weemsontN) t 0'
i-l-
k l
'887'k DRINKING WATER STATIONS stactne esmanavnse part h
I UMt? I aAID UNffI
(
l FIGURE 2-4
~
an -
2-13
9
.+
1 1
3.0 RESVLTS
SUMMARY
In accordance with Section 6.8'1.3.o'f 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 3-1 in the format of the table in-the Radiological Assessment Branch Technical Position, Revision.
1, November;71979; Results ~ forf samples collected at locations other than indicator or control stations' or in addition to those stipulated-by-Table 2-1 are included in Section 4, the discussion of results section,-
for the. type sample.
Naturally occurring radionuclides which' are not -included. in.the= plant's effluent releases are not required to be reported.: Naturally occurring.
Be-711s produced in the reactors; miniscule quantities are found-in the
' liquid releasesa No other naturally 1 occurring radionuclides are known:
to be included-in -the plant's effluent releases. - Hence,-the 1
radionuclides 1of-interest for the radiological environmental samples
-l monitoring liquid releases (river water, drinking water, fish,: sediment l '
and: aquatic vegetation) are manmade radionuclides plus Be-7, while only-manmade radionuclides'are'of interest for the other radiological enviror. mental, samples.
a j
l.
3-1
~'
TABLE 3-1-(SHEET 1 0F 10).
RADIOLOGICAL-ENVIRONMENTAL MONITORING PROGRAM ANNUAL"SUP9iARY Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1989 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
-(Fraction)
Direction (Fraction)
Airborne Gross Beta
~10 19.1 No. 12 20.0:
18.2
'O
~
Particulates 306 5-40 River Road 5-36 7-33 (fCi/m3)
(25d/254);
1.1 mils:L (49/49)
. (52/52).
WSW Gamma Isotopic 24 Cs-134 50 NOM-(c)
NOM NOM 0
Airborne 131 70 NDM.
. NDM.-'
NOM -
'O
.Radioiodine-308
^(fCi/m3)
Direct Gamma Dose NA (d)
.17.9 No.-15 21.1-18.4 0-Radiation 71 10-25 Han tan Rd 19-24 16-21~
(63/.63) 1.5 miles (4/4)~
(8/8) 7(mR/91 days) 2 m.
W
--w-s.uma--man,-
mene p p m w-rw.-m
-eww.r-._
-ns.ari.e-..ew-
.mur-ame,q-ie C=8 a dwwyn,'oo
--^**mmanmem su J. - - -
.ah:
a f_.
^
'..---,i%.-.-
e,-
e',,
e
e-se;
~~.
. TABLE 3-1;(SHEET'2 0Fl10)
RADIOLOGICAL ENVIRONMENTAL' MONITORING PROGRAM AN.JAL St# MARY
?
Vogtle Electric Generating-P1 ant, Docket Nos.:50-424'& 50-425-Burke. County, Georgia, Calendar Year'1989' Medium or._
Type and Lower Limit All Indicator Location with Highest-
- Control Locations ~ Number of-
~ Pathway' Sampled 16tal Number of-Locations Annual Mean ".
Mean (b)-'
Reported ~
Nonroutine
.(Unit of.
of Analyses Detection (a)
Mean (b).
Name-Mean"(b)
.. Range
.. Measurements.
Measurement)
Performed (LLD)
Range Distance &
Range
- (Fraction)'
(Fraction)
Direction (Fraction)
' Mil k Gamma Isotopic
-(pCi/1) 54 Cs-134 15 NA NDM --.
NDM 0
Cs-137 18 NA No. 98 7.0 7.0 0
m Dixon' Dairy 5.8-7.7
- 5. 8 -7.' 7 --
9.8 miles SE-
. (3/27)'
-(3/54)
Y Ba-140
'60-NA.
NDM.
NDM 0
La-140
'15 NA NDM NDM 0
I-131 1
NA NDM.
NDM 0-53 Grass
' Gamma Isotopic (pCi/kg wet)
-36 I.-131
- 50-NDM NDM NDM 0
Cs-134 60 NDM NDM :
'NDM
.0 E:
-L.
. _ _ _ _., ~ _., _...
..m
~.
-TABLE'3-1 (SHEET 3-0F 10)
L RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUPfiARY a
Vogtle Electric Generating-Plant, Docket Nos. 50-424.& 50-425 -
-Burke County, Georgia, Calendar Year 1989 i 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 r (Unit of of Analyses
~ Detection (a)
Mean (b)
Name.
.Mean (b)
Range Reported.
Measurement)
Performed (LLD)
Range-Distance &
Range
' (Fraction).
Measurements-(Fraction)
Direction (Fraction)
Cs-137 80 9.7 -
No. 7 9.7.
NDM-0-
9.7-9.7 Simulator 9.7-9.7
~
(1/24) 1.7 miles
-(1/12)
SE River Water Gamma Isotopic (pCi/l) 36 Be-7
' 80 (e)
NDM_
NDM NDM 0
V' Mn-54 15
- NDM
-NDM NDM 0
Fe-59
- 15 NDM--
NDM.
NDM 0
Zn-65
.30 NDM-NDM NOM 0~
Zr-95 30 NDM
'NDM NDM 0
Nb-95 NDM-NDM NDM 0
M maw r==*.
++
+mwem n
.m.
- -w ew nrn-=4-r_mmw 2-
--d-W.ms rs w-W-eww - -
-w-
.mw+=ww---vC
-mawxrm--w'
-a rMh
--e
--s m
--pr-h wz s
-s
~ TABLE'3-1-(SHEET.4.0F 1G)-
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Vogtle Electric ' Generating -Plant, Docket Nos.l50-424 &' 50-425 Burke County, Georgia,_. Calendar,_. Year 1989 Medium or Type and Lower Limit--
All Indicator
' Location.with Highest Control Locations-Number of-Pathway Sampled ' Total Number of Locations Annual Mean Range Reported-Mean (b)'
,NonroutineL
.(Unit of of. Analyses Detection (a)
Mean (b)
Name Mean'(b)-
Measurement)
Performed (LLD)
Range Distance &
Range..
'(Fraction)
(Fraction)
Direction' 4 Fraction)
. Measurements Cs-134 15 NDM MDM NDM 0-
1 Tritium 3000 1293
- No. 83 I 1293 538 0
i y
'8 1010-1590 Downriver 1010-1590 281-770 im
_(4/4)
.0.3 miles (4/4).
(4/4) i Water Near Gross Beta 4-2.93 No. 88 3.19 3.05 0
JIntakes to Water 36-1.5-5.1
~ Port.Went 2.3-5.1
~ 2.2-5.2 Treatment P1 ants
-(24/24)
Downriver.
-- (12/12)
(11/12)
(pCi/1) 122 miles H
Gamma Isotopic 36' Be-7 80 (e)
-15
.NDM NDM NDM-
- 0 1
Fe-59 30
- NDM NDH NDM O
t
- m 1-%
C'"
'*T M
"WI
- 4*C
"'I
Y7
'"7
"'"""t"F WS
'MT
-*a m
u.
r t--.
a-.v.
efP
~ TABLE 3-1.(SHEET.5'0F 10)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUPMARY Vogtle Electric Generating P1 ant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar. Year'1989 Medium or Type and:
Lower Limit All: Indicator Location with Highest
- Control locations-Number of5 Pathway' Sampled Total Number of Locations Annual Mean.
Mean (b)
Nonroutine (Unit of
-of Analyses '
Detection (a)
Mean (b)
Name-Mean (b)
Range Reported Measurement)
. Performed (LLD)
Range Distance &.
Range (Fraction).
Measurements (Fraction)~
Direction (Fraction)
C0-58.
15 NDM NDM -
NDM 0
=NDM x0 i
Zn-65 30 NDM NDM-NDM 0'
-Nb-95 15 NDM NDM
'NDM 0
- 0 Cs-137 18
La-140 15-NDM -
NDM.
NDM 0
Tritium 3000
'2508-No. 88
- 2752 259-0 12 1490-3976 Port Went~
1650-3970-182-390 (8/8).
Downriver (4/4)
(4/4) 127. miles 1sr
._wy_
q
==
w-m, ma,-
an m+
W,y e
7.pp---d e
n,,.ahw-3
- -"'~f'Fhv;
'M W9 e
y.p-[=>-
--"p--
D*"*
e fy 4 g
- y+*q,-
utow..Wa*'g-q.
5,gy+4 g_r.2p-+-
s
-TABLE 3-1 (SHEET.6 0F 10)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUfflARY Vogtle Electric Generating Plant, Docket;Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1989 Medium or-Type.and Lower Limit All Indicator Location with Highest.
Control Locations-Number of Pathway' Sampled Total Number of Mean-(b)
Name Mean-(b)
. Range-Reported--
Locations Annual Mean -
Mean (b).
Nonroutine (Unit of of ' Analyses Detection.(a) i Measurement)
Performed (LLD)
.. Range Distance &'
Range (Fraction)
Measurements (Fraction)
Direction (Fraction)
Finished Water at Gross Beta 4
2.36 No. 87 2.51 2.38
'O.
Water' Treatment 36 1.4-3.6 Beaufort 1.4-3.6 1,4-3.6 P1 ants (22/24)
Downriver
- (12/12)
(11/12)
(pCi/1) 112 miles-.
Gamma Isotopic 36 Be-7 80-(e)
NOM NOM NOM
-0
?
Mn-54 15 NOM
- NOM NOM
.0 Fe-59 30 NOM
- NOM NOM 01 i
Co-58 15 NOM.
' NOM NOM 0;
1 Co-60 15-
- NOM-NOM
. NOM
-01
- l Zn-65 30 NOM NOM NOM'
' O..
Zr-95~
30
- ' NOM NOM
.: NOM 0
Nb-95 15
' NOM NOM -
NOM.
.'O-i Cs-134 15 NOM.
NOM -
NOM 0
.M-
.. [,
y, -,,
_L,u._.__
,,,,,,,y_,__
,.p_
6
,E%y,
-+,,-..-.%-m e
f-
..r
_m
.-c+
,,,q
,..sp.
-f
TABLE 3-l'(SHEET 7 0F 10)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL St#91ARY--.
Vogtle~ Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1989 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-(Fraction)
Direction (Fraction)
'O Ba-140 60 NDM NDM
~ NDM 0
La-140 15 NDM NDM NDM-0 3 -
I-131 l-36 1
NDM NDM-NDM 0
w Tritium-2000 2236 No. 87 2325 259 0 '-
i 12~
1320-2960 Beaufort
.1810-2960 179-390 l
(8/8)
Downriver (4/4)-
(4/4) 112 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
' O Co 130
~ NDM NDM NA
Zn-65 260 NDM NDM' NA 0
4
.., ~
~
.,~%
o:g 4
=7-i p.
gw, wr-v a-m m
...r.2 m
--u-u c
~-
TABLE 3-1 (SHEET 8 0F 10)
RADIOLOGICAL ENVIRONMENTAL MONITORING ' PROGRAM ANNUAL SUP9lARY Vogtle Electric Generating Plant, Docket Nos. 50-424 & 50-425 Burke County, Georgia, Calendar Year 1989 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)
Measurementsi (Fraction)
Direction (Fraction)
Cs-134 130 NOM -
NOM NA 0
Cs-137 150 NOM NOM-NA 0.
Fish Gamma Isotopic (pCi/kg wet) 9 Be-7 100 (e)
NOM NOM NOM 0-
-[
Mn-54 130 NOM-NON NOM 0
Fe-59 260 NOM NOM NOM 0'
Co-58 130 NOM
- NDM -
NOM 0
Co-60 130' NOM NOM'
. NOM 0-Zn-65 260 NOM NOM NOM-0 I-131 15-(e) 18 No. 85 18 -
NOM 0
18-18 Downriver 18-18 (1/4)
.3 miles (1/4)
Cs-134 130 NOM-NOM NOM 0
Cs-137 150-
.117.3 No. 81:
124.6 124.6 0-32-280
. Upriver 47-310 47-310 (4/4) 2.2 miles (5/5)
(5/5)
= - -. -
TABLE 3-1 (SHEET 9 0F 10)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL SUPMARY Vogtle Electric 1 Generating Plant, Docket Nos. 50-424 & 50-425' Burke County, Georgia, Calendar Year.1989 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' Meanl(b)
Range-Reported Measurement)
Performed
- (LLD)
' Range Distance.&
Range '
~(Fraction).
Measurements.:
(Fraction)
Direction (Fraction)
Sediment Gamma Isotopic (pCi/kg dry) 4 Be-7 300 (e) 1300 No. 83 1300 415 0
1200-1400
.Downriver 1200-1400
- 270-560 (2/2) 0.7 miles-(2/2)
(2/2)
Mn-54 50 (e) 18 No. 83_
18-
'NDM 0
18-18 Downriver 18-18 (1/2) 0.7 miles (1/2) m Co-58 25 (e) 135 No. 83 135 NDM.
0 130-140 Downriver 130-140 (2/2).
0.7 miles-(2/2)
Co-60 40 (e) 46 No. 83 46
'NDM 10 22-70 Downriver 22-70
~"
-(2/2) 0.7 miles (2/2)
Cs-134-150
.NDM NDM
--NDM
-0 Cs-137
'180
~230-No.'83 230 125 0
210-250 Downriver.
210-250
.120-130
-(2/2).
0.7 miles
-(2/2)
(2/2) w _____:.-
-:w L--.
- -.,-x
=
=
!~
L l
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 1989-TABLE NOTATIONS
~
a.
.The LLO is defined in' table Notation' 3 of5 Table ~ 4.12-1 of the' TS.
j Except as noted.otherwise, the values listed in.the column are t. hose.
found in thatLtable.. :In' practice,s the' LLDs attained are generally. much-lower than the' values listed, b..
.Mean and. range;are~ based upon detectable measurements only, c Fraction of'
- detectable measurements at'specified locations is indicated'in
-parenthesis, c.-
NoDetectableMeasurement(s).z 1
i d.
Not Applicable, j
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 I-131 analysis is not required:to be perfarmed'on these samples when the ' dose calculated from the consumption of water is less_ than-1 mrem per year.
i
.j
'f l
3 1
l F
3-11 l
4 i
.]
'4.0 DISCUSSION-0F RESULTS.
L 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.< 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 preoperations. - The results' 5
were examined to perceive ~any trends. To provide perspective, a result ~
_might also be compared with its LLD or. Reporting Level (RL). Attempts-
. were made to. explain any RLs or other high radiological levels _ foundLin; the samples.t There were.no failures in the laboratory _ analyses of each of the samples in attaining the LLDs required by Table 4.12-1; of the TS for this report per.iod.
Unless otherwise. indicated, 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-preoperations these c
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 of11987_'after operations
-commenced, these included abnormal ~releasesifrom SRS..'There were'no obvious extraneous short term impacts during=CY 88 and CY-89 ' 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 was conducted on April.10:and:ll. -The:
~,
' locations.of the. nearest; milk animal', residence and garden:of greater L
than.500 square feet producing broad leaf vegetation in each of the 16-meteorological sectors within a distance of 5: miles are tabulated in' i
Table 4-1.
Land within SRS.was excluded from the census. ' Any L
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 l.
purposes are presented in Section 4.5.
l i
I I
p 3
L 4-1 i
- j i
h'.
. TABLE 4-l' p.
LAND USE CENSUS RESULTS:
1 Distance in Miles to Nearest Lo' cations'in-Each Sector-
- i SECTOR-MILK:
RESIDENCE.
. LEAFY ANIMAL GARDEN N
j NNE NE ENE E
ESE
~ "
2
I s_
4.5 SSW 4.6-SW l '. 3
. 4.' 4 -
- ]
WSW 1.2
' 3' 1 -
1 i
w
-1.5-
.3 WNW 1.8
~i.:
NW 1.8-l NNW 1.6 i
- None within 5 miles and outside of SRS.
~l
- l a
k li a
e
.- j t
4-2 4
L 4
9 4.1 Airborne As indicated by Tables 2-1 and 2 2, airborne particulates and airborne p
radioiodine 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 radioiodine, respectively. The filters and canisters are collected weekly.
Each of the air particulate filters is counted for gross beta activity. A gamma isotopic analysis is performed quarterly on a composite of the air particulate filters for each station.. Each charcoal canister is analyzed for I-131 by gamma spectroscopy.
On four occasions, both the airborne particulate and radioiodine samples were deemed to be unacceptable due to a very low volume of air drawn through the filter and canister. When collecting the samples at Station 10.on June 6 and 13, the pump was found to be off; there did not appear to be a problem with the pump as it would restart easily; the pump was replaced on June 13. The power was found to be off at Station 12 on July 10 and 17; the fuse had blown on July 17; the problem was attributed to thunderstorms.
In addition to the above failures, the particulate samples collected on February 20 at Station 10 and on August 28 at Station 12 were also j
deemed unacceptable as in each case the filters had inadvertently been mounted off-center.
Consequently, little dust had collected on the filters. Those who install the filters were reinstructed in the steps to be followed for proper sta11ation.
1 As seen in Table 31, the average weekly gross beta activity during.the 1
year for the indicator stations was 0.9 fCi/m3 greater than that for the l
control station.
However, this difference n 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 89 are compared i
below with those attained during previous years of operation, with the j
entire.preoperational period (which began in September 1981 for the air i
I monitoring stations) and with the range'of annual averages during the calendar years-of preoperations.
Preop Preop
.G.t01m
[1.32 -
(LBA CY 87
-Overall Ranaes
'I Indicator 19.1 24.7 23.0 22.9 18.1-28,1-Community 18.8 22.8 22.3 21.9 19.3-25.5 i
control 18.2 23.7 23.5 22.1 18.3-26.5 i
i 4-3 t
L..
i' The average weekly readings for CY 89 are seen to be roughly 80% 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 preoperations. No trends were recognized in these data.
Like CY 88, no positive results for manmade radionuclides were found during CY 89 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 was During preoperations 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
.l.0 (Ci/m3, respectively. The required LLD is 60 fC1/m3 Also, during preoperation Cs-134 was foutc in about 8% of the indicator composites; the average level was 1.2 fC1/m3 1131 was not detected in any of the charcoal canisters during the year.
There were no positive results during the previous. years of operation.
During preoperations, 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 fC1/m3; however, the LLD usually: attained was about 30% of this value. The Rt.is 900 fCi/m3, A
i 4-4 i -
i i i ai ii
4.2 Direct Radiation Direct (external) radiation is measured by TLDs.
A TLD badge placed at each station; each badge contains 4 calcium sulfate TLD car u.
Hence, each of the TLD badges consists of 4 tiosimeters.
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. The 2 control:
stations-(Nos. 36 and 37) are well over 10 miles from the plant.
S)ecial interest areas consist of a nearby permanent residence-(No. 33),
t.1e Town of Girard (No. 35), and the GPC employees' recreational _ area (No.43).
Station 34, a special interest area station at 6,3 miles in the SSE sector, was discontinued at the end of CY 88 as the Girard Elementary School (Hjacent to Station 34) closed at the end of the 87 88 school year. To enhance the-statistical base for the control stations, consideration is being given to adding additional stations.
During the third and fourth quarters, TLD badges were olaced on a trial basis at a location 10.4 miles from the plant in the SE sector adjacent to the Oak Grove Church, and at a location 10.3 miles from the plant in the NW sector adjacent to the McBean Cemetery; these were designated as Stations 47-and 48, respectively.
Frequently, TLDs are lost due to theft and damaged due to vandalism.
A total of 5 badges was found to be missing during the year A sixth badge was lost in shipment to the contract laboratory.
As may be sean from Table 3-1, the average quarterly dose of 17.9 nR acquired R the indicator stations over the, year was 0.5 mR less than that arquired at the control stations; this difference was not discerncble, however, since it was less than the MDD of 2.5 mR.
The average tuarterly dose at the trial control stations No. 47 and No. 48 were 17.0 and 16.5 mR, respectively.
The quarterly doses acquired at the outer ring stations ranged from 9.9 to 26.4 mR with an average of 17.2 mR for the year which is 0.7 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 1.3 mR.
The quarterly doses in. units of mR acquired at the special interest areas were as follows.
Station No.
Averaae Minimum Maximum 33 21.2' 19.9 22.6 35 18.7 16.8 20.4 43 17.4 13.9 19.8 4-5
1 mi The doses acquired at the special interest stations are seen to be somewhat typical and within the range of those acquired at the other stations.
Listed below for the indicator, control and outer ring stations, as well as for the special interest areas., are the average ' levels in units of mR/91 days obtained during each year of operations and the entire period of preoperations along with the ranges of annual averages obtained Aring the calendar years of preoperations, i
Preop Preop
.Qrma CY $3 (J,_6A (LS2 Qrenll Ranges Indicator 17.9 16.8 17.6 15.3 15.1 - 16.9 Control 18.4 16.1 17.9 16. '.
14,1 - 18.2 Outer Ring 17.2 16.0 16.7 14.7 12.5 - 16.2 No. 33 21.2 19.7 21.3 16.6 13.6 - 19.9 No. 34 18.4 20.1 15.1 12.5 - 18.1 No. 35 18.7 18.1 18.5 15.1 12.6 - 17.6 No. 43 17.4 14.8 15.2 15.3 13.9 - 25.0 Overbl?, the doses for CY 89 were roughly 4% greater than those found during previous years of operation and nearly 17% greater than those found during preeperations.
No trend is recognized in these data, however.
4-6
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 ir the land use census. The availability of milk within 5 miles of a plant was meager throughout preoperations. A milk animal is a cow or goat producing milk for human consumption.
The only manmade radionuclide found during CY 89 from the gamma' isotopic analysis of the milk samples was Cs-137.
Listed below are the average, minimum and maximum levels in units of pCi/1 for the control stations along with.the fraction of detectable measurements during preoperations and each year of operations.
Period Averaae Minimum Maximum Fraction Preoperatiuns 18.0 90 27.0 7/194 CY 87 10.4 9.)
10.8 2/39 CY 88 6.9 4.9 8.1 3/52 CY 89 7.0 5.8 7.7 3/54 Although the fraction of detectable measurements during operations is more than 50% greater than that during preoperations, the average level has become less than 40% of that during preoperations.
The LLD and RL as required by the TS are 18 and 70 pCi/1, respectively.. All but two of the 15 positive results were obtained from samples collected at Dixon Dairy; chese two were collected at Boyceland Dairy, one during 1
preoperations and the other during CY 87.
A positive I 131 level of 0.81 pCi/1 with an uncertainty of 0.26 pCi/1 at the 95% confidence level and with a maximum detectable activity-(MDA) of 0.61 pCi/l was found in the sample collected on May 23 at Boyceland Dairy. An investigation of these analyses, and other related analyses for the same time period, indicated the strong likelihood that the samples had been cross contaminated in the laboratory from glassware used to prepare I-131 Stanaards.
For this reason these results are not considered valid and are not shown.in Table 3-1.
To diminish the probability of a recurrence, laboratory personnel were reminded not to i
use the same glassware to process standards and samples.
The glassware for standards is to be labeled "For Standard Use Only".
1-131 was not detected in' any other milk sample during the year, nor has 1-131 been detected otherwise in milk samples during operations. During i
preoperations, I-131 was detected only during the Chernobyl incident.
The LLD-and RL required by the-TS are 1 and 2 pCi/1, respectively.
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't 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
-l stations at which vegetation is being sampled.
As indicated in Table 3-1, Cs-137 was the only manmade radionuclide detected; it was detected only in one-sample which had been collected at an indicator station.
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 the period of preoperations and each year of operations.
Indicator Stations Control Stations Period Averaae Fraction Averaae Fractions Preoperations 54.6 0.573 4.37 0.193 r
CY 87 24.4 0.318 61.5 0.250 CY 88 38.7 0.280 0.0.
0.000 CY 89 9.7 0.042 0.0 0.000 These data show an overall downward trend in both the average level and the fraction of detectable measurements.
The LLD and RL are respectively 60 and 2000 pCi/kg wet.
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4.5 River Water
-Surface water is composited from the Savannah River at three locations using 1500 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 0.3 miles upriver of the plant intake structure, an indicator. station (No. 83) which is located about 0.3 miles downriver of the plant discharge structure and a special station (No. 84) which is located about 1.4 miles downriver.
A gamma. isotopic analysis was made on each monthly collection.
Like CY 87-and CY 88,' there were no radionuclides of interest detected in CY 89.
A tritium analysis was performed on each quarterly composite. A positive result was obtained from each analysis. As indicated in Table 3-1, the average level of 1293 )Ci/1 found at the indicator station is 755'pCi/1 l
greater than that at t1e control station; this difference is discernable i
because it.is greater than the,MDD.of 518 pCi/1.
There was also a discernable difference in the tritium level between these two stations in CY 88. At the special' station (No. 84), the results ranged from 905 to 1780 pCi/l with an average of 1269 pCi/1.
The LLD is 3000 pCi/1 and the RL is 10 times greater.
Listed below for each year of operations 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 MDD between these two stations and the annual liquid releases of 4
tritium from the plant. All of these values are in units of pCi/1 except for the releases which are in units of Ci.
llam LLB2 CY 88 GL61 Control Station 524 427 538 Indicator Station 680 843 1293 Special Station 1411 1430 1268 Li Le 156 416 755 MDD 416 271 518 Releases 321 390 916 l
These data show an upward trend for the levels at the indicator station and some correlation between (Li-Lc).and plant releases. The releases are sufficient to account for the increased concentration of tritium at the indicator station. The CY 89 level at the indicator station'is modest in comparison to those which have generally been found further downstream on the river during the past three decades or so, it is shown in Table 3-1 that the average tritium levels at the intakes Rr the indicator water treatment plants which are more than a hundred miles, 4-9
v_-.
t downriver are nearly twice the level for the indicator station shown above.
The annual organ dose that the maximum exposed individual (a child) would receive from drinking water with an average tritium concentration of 755 pC1/1 was conservatively calculated to be 0.078 mrem or 0.78% of the TS limit.
On September 26 the annual survey of the Savannah River was conducted downriver of the plant for_approximately 130 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 downrive* water traatment. plants (Stations 87 and
.88) from which samples are collected monthly. As in all previous surveys, no intakes for irrigation use were observed.
On September 22, the survey results were corroborated by contacting the Environmental Protection Division of.the Georgia Department.of Natural Resourcet and
' the South Carolina Department of Health and Environmental Control; it was fcund that no new surface or drinking water withdrawal permits had been issued for the Savannah River during the previous 12 months..
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4.6 Drinking Water i
Samples were collected at a control station (No. 80), the Augusta Watdr 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 miles 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 riverweter which was composited near the plant's intake (raw drinking water)ites are made up from the monthly collutions.and of grab samples of finished compos 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 4.12 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 0.12 pC1/1 greater for the control station-than for the indicator stations. However, this difference was not discernable because it was less than the MDD of 0.85 pCi/1..For finished drinking water, the average gross beta activity was 0.02 pC1/1 greater for the control stations than for the indicator station.
This difference was not discernable because it was less than the MDD of 0.57 pCi/1.
There were no positive results for the radionuclides of interest from the gamma isoto)ic analyses of the monthly collections. Only one positive result has )een found since operations began; Be-7 at a level of,68.2 pCi/l was found in the sample collected for September 1997 at Beaufort.
L 1
Positive results were obtained from the tritium analysis of each of the quarterly composites. Furthermore, there was a discernable difference between the average tritium values for the two type stations for both the raw and finished. drinking water since these differences were each greater than their MDDs.
As indicated by Table 3-1, the average values of the tritium levels for the indicator stations were 2249 and 1977 pCi/l greater than those for the control station for raw and finished drinking water, respectively; the MDDs were correspondingly 1000 and-627 pCi/1.
1 Similar results were obtained during all previous years of operation and i
during preoperations.
]
i 4-11
Each result for the I-131 analysis of the finished drinking water samples was below its MDA which ranged from 0.21 to 0.75 pCi/1.
Similar results were obtained in CY 88.
The TS call for a LLD and a RL of 1 and 2 pCi/1, respectively.
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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 enalysis on the edible portions of each sample collected.
About a five mile stretch of the river is generally needed to obtain adeouate fish samples.
For the semiannual collections, the control station (Ilo. 81) extends from approximately 2 to 7 miles upriver of the plant intake structure and the indicator station (No. 85) extends from about 1.5 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, Americen shad,-an anadromous species, was. collected at Station 85. : Like CY 88 no positive results for the radionuclides of interest were obtained from the gamma isotopic analysis.
In CY 87, Cs-137 was found in one of the three shad collected at a barely detectable level of 10 pC1/kg wet. The LLD and RL for Cs-137 in fish as specified by the TS are 150 and 2000 pCi/kg wet, respectively.
On April 24 and October 23, the composition of the catches were as follovs:
D.ittti Station 82 Station 85 April 24 Large Mouth Bass Large Mouth Bass Red Ear Sunfish-Red Ear Sunfish October 23 Chain Pickerel Chain Pickerel large Mouth Bass Large Mouth Bass Red Ear Sunfish Red Ear Sunfish As indicated in Table 3-1, I-131 and Cs-137 were the only radionuclides of interest found in the semiannual collections of commercially or recreationally im>ortant species; since operations began, positive results had only seen found for Cs-137.
A positive I-131_ level of 18 pCi/kg wet with an uncertainty of 11 pCi/kg wet at the 95% confidence level was found in one of the two samples collected at the indicator station in October.
The LLD assigned for I-131 in fish is 15 pCi/kg wet. The annual thyroid dose that the maximum expos 9d indisidual (an adult)f 18 pCi/kg is 0.737 mrem or 7.37% of =the TS-would receive from eating fish with an average I-131 concentration o limit.
4-13 C-a
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- . t l' Actual releases of I-131 to the river tota 19d 0.631 mci for the year.
Conservative calculations of radioactivity in fish that might result from 4
i
.' i these actual releases correspond to a level of,0.718 pCi/kg or about 4%
of the one sample. -In addition, aquatic vegetation samples, discussed in 1
sectioni4.9 of this report, indicate other sources of radioiodines-exist.
1 Since the measured levels of I-131 do not corre11 ate with actual release
~
data, and there are indications that other sources of I-131 besides Plant Vogtle exist,4 we do not believe these fish results are the result of a
c plant operations.
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It can be seen' from Table 3-1, that the average level for Cs-137 at the l
indicator-station of 117.3 pCi/kg wet is 7.3 pCi/l:g less1than that at the j
control station. This difference is not discernable since'it is
.j less than the MDD of 219 kCi/kg wet.
Since operations began, positive i
values for Cs-137 have been found in all but one (an indicator sample in CY 87) of the 25 samples colletted. The. levels found in CY 89 are '
1 typical of those found in the previous years.
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L 4.8 Sediment Sec" rent was collected along the shoreline of the Savannah River on April 4 and October 23 at Stations 81 and 83.
Station 81 is a centrol station located about 2.3 miles upriver of the plant intake structure at RM 153.2 while Station 63 is an indicator station located about 0.7 miles downriver of the plant discharge structure at RM 150.2. The indicator sample for October was collected at RM 149.5. A gamma isotopic analysis.
' was. performed on each sample, As in all previous years of operation, positive readings for Be-7 and Cs-137 were found in each sample and the readings were on the same order as found in those. years.
For Be 7, the average reading of-1300 pCi/kg-dry.for the indicator station is 885 pCi/kg-dry greater than that for the
. control station; there is no discernable difference, however, since this difference is less than'the,MD0 of 1227 pCi/kg dry.
For Cs-137 the t
average reading of 230 pC1/kg dry for the indicator station is 105 )Ci/kg dry greater than that for the$ control station; there is no discernaale difference since this difference is less than the'MDD. of 144 pCi/kg dry.
Also indicated in Table 3-1 is the presence of the activation products, Mn-54, Co-58 and Co-60 at the downriver station.
Each of these radionuclides were found at slightly lower levels than in CY 88.
The radiological impact due to the readings of Mn-54, 00-58 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 l'.109, Revision 1, October 1977 and comparing this dose with that permitted by Section 3.11.1.2.b of the TS (3 mrem per year).
The theoretical dose was determined to be
.0026 mrem per year or 0.087". of the TS limit. This dose is nearly 30%
lower than that calculated for last year. This extremely low dose, although calculable, poses no measurable negative environmental or publit health impact.
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4.9 Aquatic Vegetation On six occasions dur'ing the year, a sample of aquatic vegetation (eaeria im u )h a downriver indicator station and an upriver control station to
' hut commonly known as water weed, was collected on a trial basis at bot t
determine the suitability of its use as an environmental sample to monitor'eny radiological impact due to liquid release. This vine like densely foliaged plant grows underwater at depths of 3 meters or less and acts somewhat like a filter.
Gamma isotopic analyses were performed on each samplei.
For the first collection, the indicator station was located at RM 149.7 4
and the control stations at RM 151.
Subsequently, the indicator station was at the GPC landing which is located at approximately RM 149.5 and the control station at Hancock Landing which is located at approximately RM 151.7.
The results in units of pCI/kg wet are summarired in Table 4 2.
To be noted is the presence of positive results for Mn 54,'Co 58 and Co-60 at the indicator station and the absence of these radionuclides at the control station;-this tends to support-the suggestion that tha presence of t./ - > radionuclides in sediment samples at the indicator station is due co plant releases. Also, ithe presence of I-131 at the control nation and its absence at the indicator station supports the contentinn that the presence of I-131 in a fish sample collected at the indicator station is not due to plant releases.
It may also be seen from Table 4 2 that the average reading of 219.4 pCi/kg wet for Be-7 at the control station is 56.2 pCi/kg wet greater than that for the indicator station; this difference is not discernable' l
since it is less than its MDD of 178.1 pC1/kg wet.
Similarly, the average reading of 19.0 )Ci/kg wet for Cs-137 at the control station is 3.3 pti/kg wet greater t1an that for the indicator station; the 1
difference is not discernable since it is less than its MD0 of 40.5 l
pCi/kg wet, it appears that water weed would be a suitable radiological environmental monitoring sample.
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TABLE 4 2 RESULTS
SUMMARY
FOR AQUATIC VEGETATION Indicator Station Control Station Radionuclide LLQ(a) hann BAngg Fraction lit.gn Rangg Fraction Be-7 150 163.3 130 218 4/6 219.4 85-391 5/6 Mn 54 15 58.1 29-145 4/6 NDM 0/6
=
Co 58 15 39.8 24-68 5/6 tlDM 0/6 C0 60 15' 43.1 43 43 1/6 NDM 0/6 I-131 30 NDM 0/6 15.5 14-17 2/6 Cs-137 20 15.7 9-23 2/6 19.0 13-30 3/6 Table Notation l-a.
The El has determined that these values may be routinely attained.
=
b 1
l 5.0 Interlaboratory Comparison Program Section 3.12.3 of the TS requires that aWvses shall be nerformed on radioactive materials supplied as part (. -
'nterlaboratory Comparison Program that has been approved by the Nu Regulatory Commission z
(NRC).
The Environmental Protection Agency?s (EPA's) Environmental Radioactivity Laboratory Intercomparison Studies (Crosscheck) Program conducted by the Environmental Monitoring and Support Laboratory in Las Vegas, Nevada provides such a program and the El participates in the program.
Reported herein, are only those results where the type analysis and sample in the EPA Crosscheck Program are the same as that delineated in Table 2-1.
The crosscheck program was designed for laboratories involved with REMPs; the program involves environmental mettia and a variety of radionuclides with activities at or near environmental levels.
Participation in the program ensures that independent checks on the precision and t. curacy of the measurements of radioactive materials in environmental sayle matrices are performed as part of a quality assurance program tc demonstrate that the results are reasonably valid.
Simulated environmental samples are distribut ed regularly to the participants who analyre the samples and return the results to the EPA for statistical analysis and comparisons with known values and with results obtained from other participating laboratories. The crosscheck program thus provides each participant with documentation on the p.ecision and accuracy of its performance; the program helps in indicating instrument or procedural problems; the program clso provides each participant with a comparison of its performance to that of other laboratories.
The El performed the analyses called for by the program on each sample provided by the EPA.
Analyses were performed in a normal manner.
Each sample was analyzed in triplicate as required by the program.
Table 5-1 provides a summary of the relevant results of the EL's participation in the program.
The results listed in Table 5-1 were obtained:
from the gross beta and gamma isotopic analyses of air filters; from the gamma isotopic analysis of a milk sample; and from the gross beta, tritium, gamma isotopic and 1-131 analyses of water samples.
Not shown in Table 5-1 are the results from the gross beta analysis on the air filters collected on August 25 and the 1-131 analysis on the milk sample collected on April 28.
The EPA invalidated the gross beta results on the air filters for the August 25 collection for all participants due to problems they had with I-131 in these air filters.
The EPA also invalidated the results of the 1-131 analysis of the milk samples collected on April 28 for all participants as the activity placed in the sample by the EPA was much less than the activity routinely measured by most of the participants.
5-1
=
TABLE 5-1 (SHEET I 0F 2)
CROSSCHECK PROGRAM RESULTS
~;- ~~
~
Date Known Expected Reported Standard Normalized Normalized Analysis Collected Value Precision Averece Deviation Deviation Rance Air Filters (pCi/ filter)
Gross Beta.
3/31/89 62.0 5.0 63.33 0.58 0.46 0.12-Cs-137 3/31/89 20.0 5.0 25.67 0.58 1.96 0.12 8/25/89 10.0 5.0 9.33 0.58
-0.23 0.12 Milk (pCi/1)
Cs-137 4/28/89 50.0 5.0 49.00 2.00
-0.35 0.47
'3" Water (pCi/1) so Gross Beta 1/20/89 4.0 5.0 3.00 0.00
-0.35 0.00 4/18/89 57.0 5.0 52.67 1.15
-1.50 0.24-5/12/89 50.0 5.0 46.00 2.00
-1.39 0.47 9/22/89 6.0 5.0 6.00 0.00 0.00~
0.00 10/31/89 32.0
'5.0 35.67 1.53 1.27 0.36 H-3 2/24/89 2754.0 356.0 2696.67 90.74
-0.28 0.28 6/23/89 2754.0' 356.0 2696.67 90.74
-0.28 0.28 10/20/89 3496.0 364.0 3240.00 120.00
-1.22 0.39 Cr-51 2/10/89 235.0 24.0-217.67 13.01
-1.25 0.64 Co-60 2/10/89
' 10.0.
5.0 9.67 58
-0.12 0.12 6/09/89 31.0 5.0 28.33 1.53
-0.92 0.36 10/06/89 30.0
- 5.0 31.33
.l.53 0.46 0.36 I
'M lW M
TABLE 5-1 (SHEET 2 0F 2)
CROSSCHECK PROGRAM RESULTS Date Known Expected Reported Standard Normalized Normalized Analysis Collected Value Precision Averaae Deviation Deviation Rance Ru-106 2/10/89 178.0 18.0 171.00 17.69
-0.67 1.29 6/09/89 128.0 13.0 112.00 13.89
-2.13 1.26 10/06/89 161.0 16.0 140.67 35.23
-2.20 3.88 Cs-134 2/10/89 10.0 5.0 9.67 2.08
-0.12 0.47 4/18/89 20.0 5.0 20.00 1.73 0.00 0.36 6/09/89 39.0 5.0 38.00 2.00
-0.35 0.47 10/06/89 29.0 5.0 29.33 5.03 0.12 1.35 10/31/89 5.0 5.0 4.67 0.58
-0.12 0.12 Cs-137 2/10/89 10.0 5.0 10.67 2.89 0.23 0.59 4/18/89 20.0 5.0 18.67 4.16
-0.46 0.95 T
6/09/89 20.0 5.0 20.67 1.53 0.23 0.36 10/06/89 59.0 5.0 62.67 3.21 1.27 0.71 10/31/89 5.0 5.0 5.33 0.53 0.12 0.I2 Ba-133 6/09/89 49.0 5.0 45.33 2.31
-1.27 0.47 10/06/89 59.0 6.0 51.33 4.51
-2.21 0.89 l-131 8/04/89 83.0 8.0 80.67 3.79
-0.51 0.52 l
e
The acceptance criteria used by the El are warning limits and control limits defined as the 95% and 99% confidence levels, respectively, for both the normalized deviation and the normalized range.
The normalized deviation is a measure of the accuracy of the data. The normalized range is a measure of the precision of the data.
Results are evaluated for trends and out of control limit conditions.
It is noted from Table 5-1 that the normalized range for Ru-106 in the water sample collected on October 6 exceeded the control limit.
It was also noted thet the Ru-106 and Ba-133 results in water samples exhibited evidence of negative bias.
Evaluation of these analyses demonstrate that bias and precision are not due to sample preparation, instrument quality control or instrument calibration.
The decay schemes for Ru-106 and Ba-133 suggest possible negative bias due to summing losses from analytical peaks.
Corrections for losses due to summing are being evaluated, in past years, the NRC's " Criteria for Comparino Antlytical Measurements" was used in this report to determine agreement witn known values.
It was decided to adopt the more restrictive criteria, described above, that was already being employed by the EL.
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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 data were carefully examined. A summary and a discussion of the results of the laboratory analyses for each type sample collected.were presented.
No measurable radiological impact upon the environment as a consequence-of plant discharges to the atmosphere was established.. Although low levels of tritium in river water samples and of Mn 54, 00-58 and-00 60 in sediment samples were found downriver of the plant, and their presence might (at-least. )artially) be due to liquid effluents from the plant, evaluations show t1ey pose no measurable negative impact upon the environment or public health.
An aquatic vegetation plant collected during the year on a, trial basis shows promise as a radiological environmental sample for monitoring liquid releases.
The results of the EL's participation in an Interlaboratory Comparison Program were presented. One result exceeded a control limit, an-investigation was made, corrective actions are being' evaluated.
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V GEORGIA PCWER COMPANY-V0GTLE ELECTRIC GENERATING PLANT - UNIT I AND UNIT 2 NRC DOCKET NOS, 50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 ANNUAL ENVIRONMENTAL OPERATING REPORT FOR 1989 (NONRADIOLOGICAL)
1 V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 AND UNIT 2 ANNUAL ENVIRONMENTAL OPERATING REPORT (NONRADIOLOGICAL) 1989 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 1989.
REPORTING REQUIREMENTS A.
Summaries and Analyses J ic Results of the Environmental Monitoring Activities for the Report Period 1.
Aquatic Monitoring. - Liquid effluent monitoring was performed in accordance with National Pollutant Discharge Elimination System (NPDES) Permit No. GA0026786; there was no additional requirement for aquatic monitoring during 1989.
2.
Terrestrial Monitoring - Not required.
3.
Maintenance of Transmission Line Corridors a.
There was no herbicide use within the VEGP transmission line corridors during 1989.
b.
There were no clearing or maintenance-related activities within the Ebenezer Creek or Francis Plantation areas during 1989, c.
Routine maintenance activities within the designated cultural 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 Company during 1989 regarding noise along the VEGP-related high voltage transmission lines.
B.
Comparison of the 1989 Monitoring Activities with Preoperational Studies, Operational Controls, and Previous Monitoring Reports These comparisons were not required because no nonradiological environmental monitoring programs were conducted during the reporting period beyond those performed in accordance with the NPDES Permit No.
GA0026786 referenced in-Section A above.
C.
An Assessment of the Observed Impacts of Plant Operation on the Environment There was no significant adverse environmental impact associated with plant operation during 1989.
s
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D.
Environmental. Protection Plan (EPP) Noncompliances and Corrective Actions There were no EPP noncompliances during 1989.
E.
Changes in Station Design or Operation, Tests, and Experiments Made in Accordance with EPP Section 3.1 which Presented Significant Environmental Impact or Involved a Potentially Significant Unreviewed Environmental Question There were no changes in station design or operation, tests, or experiments during 1989 which presented signficant environmental impact or 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 1989.
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