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From Renee M.
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Date:
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Subjects-Vogtle Meeting j
Yo. Bruno.
1 A-e.vou aware of the. meeting being held tommorrow at 9:00 to oiscuss Vogtle?
More impotantly, are you clanning on listening?
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i socke with' Dave Matthews today & he didn't know if yot were planning on listenino...I was a little surprized!
Based on what
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believe will be "an oral OI report." I think that it very beneficial for you to listen in.
T-e-evere. let me know if you are available & I'll make sure that tne room nas a speaker-ohone in it & will tie you in..You may
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want to fanc a room for yourself with a speaker ohone because.it j
will probably be a-long meeting.
.If OI nands out any oriefing
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naterial. I'll make sure that you get it faxed.
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.et. me tnow.where I can.rea:h you tommorrow if you're going to be
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September 30,1993 LCV 0163 i
Docket No.
50 424 l
TAC No.
M87782 M47783 F
l U. S Nuclear Regulatory Commission ATrN: Document Control Desk Washington, D. C. 20555 Gentlemen-a VOGTLE ELECTRIC GENERATING PLANT l
10 CFR 50 APPENDIX J EXEMPTION REQUEST AND l
R" JUEST TO REVISE TECilN! CAL SPECIFICATIONS REGARDING AUX 1LIARY COMPONENT COOLING WATER I
l SUPPLY AND RETURN CONTAINMENT 1 SOLATION VALVES in accordance with the prosisions of to CFR 50.12, Georgia Power Compacy (OPC) requests a f
one time exemption ikom the requirements of 10 CFR 50, Appendix 2, Section Ill.D.3 relate to the Unit I auxiliary component cooling water (ACCW) supply and return contain l
isolation valves. Section Ill.D.3 of Appendix 3 requires that Type C tests be performed each reactor shutdown for refueling but in no case at intervals greater than 2 years I
exemption would allow the required test interval for valves IIV 1974 (and asso l
1 1217.U4 113),1975,1978, and 1979 to be extended ikom 24 months to prior to entry i
l Mode 4 following the next scheduled reibeling outage (or the next forced outage requir l
into Mode 5), but no latct than November I,1994, la addition, in accordance with th of 10 CFR $0.91, GPC proposes to amend the Vogtle Electric Generating Plant (V i
Technical Specifications (TS). Appendix A to Operating License NPF48 as they subject valves. The proposed antendment would affect TS 4.6.1.2d by adding l
would extend the Type C test interval for the subject valves consistent with the exempt l
request. As with the exemption reqbest, the proposed amendawnt would be a l
extension of the Type C test interval for these valves.
4 Georgia Power Company submits that the proposed exemption and license exigent circumstances and respectibily requests that the NRC process the p
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under the provisions of the regulations applicable to an salgent request. These circ as follows in February 1992, a LM= lag Document Change Request (LDCR) tha i
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, U. S Nuclear Regulatory Commlulon LCV 0163 l'
Page 2 6 2 41 of the VEGP Final Safety Analysis Report (FS AR) was processed under the prov 1
l 10 CFR $0 59. This 1.DCR (FS 92 007) revised the subject table to show the post accident position of the ACCW supply and retum containment isolation valves to be open rat closed, and the leakage testing requirements were revised from Type C to Type A as def CFR $0 Appendix 3. As a result, the subject valves were not Type C tested during th refueling outage in the Spring of 1993. Recently, during a followup review to revise ano l
documentation to correspond to LDCR FS 92 007, it was determined that the Type C test 1
requirement was inappropriately revised to Type A. The required 24 month interval s l
10 CFR $0, Appendix J and TS 4.6.1.2d will expire on October 28,1993, for the subject va and the unit must be in Mode 5 for the testing to be performed. Therefore, unless the proposed esemption request and license amendment can be processed on an exigent basis, Unit l
forced into Mode S in sulticient time prior to October 28,1993, so that the required testing can be per fornwd Unit 2 is paciently in a refueling outage, and the corresponding valves ha l
i tested during this outage Therefore, the proposed exemption and license amendment are not applicable to Unit 2 i
'Ibe proposed exemption and ita basis is provided in enclosure 1. The proposed license l
amendment and its basis is provided in enclosure 2 including a detailed discussion of the circumstances discuued above that have necessitated this request. An evaluation pursuant to i
l CFR $0.92 showing that the proposed changes do not involve'signincant hazards conside prosided as enclosure 3, an environmental assessment is provided as enclosure 4 i
up TS page is provided as enclosure S. In accordance with to CFR 50.91, the designa i
l ollicial will be sent a copy of this letter and all enclosures l
The Plant Review Board has resiewed and recommended approval of this request, in additio GPC han determined that the proposed exemption and license amendment will not have a 1
signi6 cant effcet on the emironment Mr. C K McCoy states that he is a Vice President of Georgia Power Company and is auth to execute this oath on behalf of Georgia Power Company and that, to the best of his know i
and belief, the facts set forth in this letter and enclosures are true.
GEORGIA POWER COMPANY BY
/
~ C. K McCoy,/
l Sworn to and subscribed before me this h, ay of imllAA1993
&n Ad j
Notary Public I
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U. S. Nuclear Regulatory Ccmmission LC%0163 Page 3 I
CKM/NJS f
Enclosures:
4
- 2. Proposed TS Amendment 3.10 CFR 50.92 Evaluation
- 4. Environmental Assessment
- 5. Marked Up Page
. xc Qcornia Power Company Mr. J. D. Deasley Jr.
Mr M. Sheibani NORMS U. S. Nuclear Reuulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. 5. Ilood, Licensing Project Manager, NRR Mr. D. R Donser Senior Resident inspector, Vogtle i
Slattaf_Grpizia Mr J. D Tanner, Comminioner, Department of Natural Resources
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ENCLOSURE 1 VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES f
BASIS FOR PROPOSED EXEMPTION
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Section Ill.D.3 of Anpendix J to 10 CFR 50 requires that Type C tests be performed at each reactor shutdown for refueling but in no case at intervals greater than 2 years. Georgia Power Company (GPC)is requesting a one time extension of this intesval as it applies to the auxiliary component cooling water (ACCW) supply and return containment isolation valves liv 1974 (and associated check valve 1-1217-U4113),1975,1978, and 1979. The current 2 year intervalis scheduled to expire on October 28,1993, and GPC is requesting that the interval be extended to prior to entry into Mode 4 following the next scheduled refueling outage (or the next icreed outage requiring entry into Mode 5), but no later than November 1,1994.
e 14dstound in February 1992, Licensing Document Change Request (LDCR) FS92-007, prepared under the provisions of 10 CFR 50.59, was reviewed and recommended for approval by the Plant Revi Iloard in accordance with section 6.4.1.6 of the Vogtle Electric Generating Plant (VEGP) TS.
On February 21,1992 LDCR FS 92 007 was approved by plant management. The subject 1,DCR revised table 6.2.4-1 of the VEGP Final Safety Analysis Report (FSAR),in part, with respect to the ACCW vupply and return containment isolation valves. Prior to the change, ta 6.2.4 1 stated that the subject valves were subject to 10 CFR 50, Appendix J Type C leakage testing requirements, and that they were normally open during operation but closed under post-accident conditions. Ilowever, as noted in footnote "g" to table 6.2.4 1, it is highly desirable that ACCW flow be maintained to the reactor coolant pumps (RCPs)if possible. Under most accident scenarios, ACCW flow would be maintained to support operatiori of the RCPs. Therefore, the I.DCR revised the leakage testing requirements to Type A and stated that the post accident position was open in addition, the associated penetrations were added to FSAR table 6.2 penetrations that are not vented or drained during Type A testing. As a result, these valves not Type C tested during the Spring 1993 refueling outage. They have, however, been teste during previous outages on both units.
The basis for the LDCR was that the subject salves do not receive a containment isolation signal (they are remote manually operated), and the associated penetrations are considered es to the desirability of maintaining cooling water to the RCP under post-accident conditions. In addition,it was thought that the ACCW was a closed system since it does not communicate directly with the containment atmosphere or primary coolant. Therefore, it was thought that A testing was sullicient for these penetrations.
Ilowever, the safety evaluation failed to consider that, while the ACCW system is scismic category I and the hard piping is fabricated of ASME Section 111, Class 3 materials, the El-1
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ENCLOSURE I (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES B ASIS FOR PROPOSED EXEMPTION i
installation was in accordance ANSI B31.1 and an N stamp was not aflixed. In addition, there are some components such as motor coolers and flexible piping that are not of Class 3 materia i
Therefore, the ACCW system does not meet the ANSI standard criteria for a closed system. In i
consequence, the supply and return isolation valves must be considered to perform an isolatio function and should be subject to Type C testing. This was discovered recently during a subsequent review for the purpose of making corresponding changes to associated documentation.
Technical Basis J
The subject valves have been Type C tested during all previous refueling outages with the exception of the Unit i Spririg 1993 outage. A review of the maintenance work order (MWO)
^
history was performed on the ACCW containment isolation valves. This review showed that there were MWOs for seat leakage, packing leaks, flange leaks, preventive mr.intenance, and several inspections, but there were no "as found" Type C local leak rate test (LLRT) failures after the initial entry into Mode 4 on both units, An MWO was written against 2ilV-1979 for high seat leakage during the period between the preoperational LLRT and initial Mode 4 entry, however, this condition has not reoccurred.
In addition, a review of the valves' LLRT history aner initia! Mode 4 entry demonstrates the reliability and low leakage trends of these valves. Listed below are the maximum values for b l
the "as found" and "as IcR" LLRTs performed after initial Mode 4 entry. This data was taken from six refueling outages between the two units. Note that penetration 28 is the ACCW supply j
line and penetration 29 is the ACCW return line.
PENETRATION 29 PENETRATION 28 i
MAXIMUM LEAKAGES MAXIMUM LEAKAGES lilV-1974 = 152 sccm' tilV-1978 = 20.5 seem tilV-1975 - 62.0 scem l
lilV 1979 = 40.4 seem 2iiV-1974 = 99.6 sccm' 211V 1978 = 49.2 scem 21{V-1975 = 136.3 seem 211V-1979 = 90.6 sccm l
Includes leakage through associated 113 check valve
'I The Inservice Inspection Program currently specifics a maximum allowable leakage of 1000 s
l for each butterfly valve and 1500 seem for the check valve. (The Icakage limit for the combination of valve llV 1974 and check valve 113 would be 2500 secm.) T based on Appendix J, but were established based on the low leakage history of these 1
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ENCLOSURE 1 (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT j
10 CFR 50 APPENDIX J EXEMPTION REQUEST l
REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES B ASIS FOR PROPOSED EXEMPTION serve the purpose of defining the point at which repair would be required. The Appendix J
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leakage limit for all penetrations subject to Type B and C testing (0.6La) at VEGP is 228,2 scem. The current total for Type B and C test leakage at VEGP as of September 10,1993, is i
14398.8 secm. As of the last Type C LLRT, the leakage for each of these four valves was as follows: HV-1974 - 152 seem (this includes leakage past check valve 1-1217-U4-113 in parall scem. In with HV-1974); liv-1975 - 11.6 seem; HV-1978 - 9.3 secm; and HV-1979 - 11.4 addition, it should be noted that the test pressure, Pa, was 45 psig at the time these numbers w obtained. The test pressure has since been revised to 37 psig, so it would be reasonable to assumJ l
that the leakage would be less at the lower pressure.
During the last outage for Unit 1, maintenance was performed on llV-1979 that could have afTected its leakage, and no LLRT was performed since it was not required by the FS AR at t time. The motor and gearbox were removed and the limit switch settings were altered, but no i
work was done that would have afTected the valve seat The standard work practice for setting limit switches on this type of son-seated butterfly valve following this type of maintenance is as follows. First, the valve is manually closed using the hand wheel until 00 (closed) is teached, a the limit switch is set. Then the limit switch is tested by manually operating the valve again.
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Finally, the valve is stroked using the motor until the limit switch actuates. At this po l
wheel is used to ensure that the valve is seated properly after the limit switch actuates. As a reference point, in the Spring of 1992 this type of work was performed on Unit 2 valve HV and premaintenance and post maintenance LLRTs were performed. The premaintena l
and the post-maintenance leakage was well within the leakage limits for this valve. The GPC is confident that Unit I valve llV-1979 remains Icak tight.
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Furthermore, the probability of an event that leads to core damage and a failure of the AC piping inside containment with a failure to isolate containment is not considered t The probability of containment isolation failure following a core damage accident 4
l the VEGP individual plant examination (IPE). In order to model a more conservative containment isolation failure than was considered in the base case VEGP IPE, it was a the occurrence of any core damage scenario would cause a break in the ACCW flow p the operator would be required to isolate the ACCW system for successful conta Based on a Type C test interval of 2 years, the frequency of core damage with containme l
isolation failure was found to be on the order of 10 7 er reactor year. Extending the requi p
Type C test interval for these valves as proposed has a negligible impact on tha j
Finally, the ACCW system is seismic category 1, and the hard piping is fabiicated o 1
l Section 111, Class 3 materials. (There are some components such as motor coolers and l
pipmg that are not fabricated of Class 3 materials.) Therefore, even though the
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ENCLOSURE I (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST REG ARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES 4
D ASIS FOR PROPOSED EXEMPTION meet the ANSI standard criteria for a closed system,it can be considered to be highly reliable and there is reasonable assurance that for most events its integrity would be maintained.
4 1
i Justification 10 CFR 50.12 states that the Commission may grant exemptions from the requirements of the regulations contained in 10 CFR 50 provided that: (1) that exemption is authorized by law, (2) the exemption will not present an undue risk to the public health and safety, (3) the exemption is consistent with the common defense and security, and (4) special circumstances as defined in 10 j
CFR 50.12(a)(2) are present.
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- 1. The requested exemption is authorized by law.
No law is known to exist that would preclude the activities covered by this exemption request.
Therefore, the Commission is authorized to grant thi: exemption.
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- 2. The requested exemption does not present an undue risk to the public health and safety.
The subject valves are considered to be leak-tight, and based on the maintanance history of the valves, the proposed extension of the Type C test interval will not impair valve operability or significantly degrade leak tightness. The probability of an event that leads to core damage 6
with a failure to isolate containment accompanied by a failure of the ACCW piping inside containment is not considered to be credible. Extending the Type C test interval for these valves as proposed has a negligible impact on that probability. Therefore, there is no undue risk to the health and safety of the public.
The requested exemption will not endanger the common defense and security.
3 The common defense and security are not an istue in this exemption request.
Special circumstances are present which necessitaM the request for a one-time exemptio 4.
i the regulations of 10 CFR 50, Appendix J. Sectior. III.D.3.
Compliance with the regulation would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted. The requirement for a shutdown solely for the purpose of testing the subject valves, especially when the expected leakage is considerably less than requirements, would result in excessive i
costs in the form of tast revenues.
El-4
ENCLOSURE I (CONTINUED)
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VOGTLE ELECTRIC GENERATING PLANT 9
10 CFR 50 APPENDIX J EXEMPTION REQUEST
]
REGARr " " WX1LIARY COMPONENT COOLING WATER EfURN CONTAINMENTISOLATION VALVES i
SUPPL' h
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DAS!S FOR PROPOSED E)GMPTION l!
The exemption would provide only temporary relief from the applicable regulation, an good faith effort has been made to comply with the regulations. All other pene to Type C testing have been faithfully tested. When the discovery was made that l;
inappropriately removed the Type C test requirements from the subject v,alves, GP l'
l prompt action to investigate and correct this condition.
i Safety Imoact Georgia Power Company has reviewed the proposed exemption and has made the fo determination:
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- 1. The proposed changes do not involve a significant increase in the probability or conseque j
of an accident previously evaluated. The proposed change is a one-time only extensio Type C leakage test interval for the Unit 1 ACCW supply and return containmen l
valves. As such, it has no effect on the probability of any accident previously evaluated.
Furthermore, based on the past leakage test history of these valves, there is reasonable l
assurance that extending the test interval to no later than November 1,1994, (or the next forced outage that requires entry into Mode 5) will not adversely affect the ability of th valves to perform their isolation function. Therefore, the proposed change will not in 1
significant increase in the consequences of any accident previously evaluated.
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- 2. The proposed change will not create the possibility of a new or different kind any accident previously evaluated. The proposed change does not change t l
method of operation of any plant equipment, and no new failure modes have be j
any plant system or component. Furthennore, no new limiting failure has b l
result of the proposed change.
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1 The proposed change does not involve a significant reduction in a margin of saf l
continues to be reasonable assurance that the subject valves will remain capable their isolation function. In addition, the proposed change avoids a plant shutdown l
t the purpose of performing Type C testing of these valves.
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1 El-5 1
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ENCLOSURE 2 VOGTLE ELECTRIC GENERATING PLANT l
REQUEST TO REVISE TECilNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES B ASIS FOR PROPOSED Cil ANGE P_tspssd Changs j
The proposed amendment would revise the existing Technical Specification (TS) survei!!anc requirement 4.6.1.2d for Unit I by adding a footnote that would extend the surveillance in for the next required Type C leakage test of the auxiliary component cooling water (ACCW) 1-1217 supply and return contsnment isolation valves llV-1974 (and associated check valve 113),1975,1978, and 1979 to prior to entry into Mode 4 from the next scheduled refueling outage (or the next forced outage requiring entry into Mode 5), but no later than November 1 i
1994. The proposed amendment would be a one-time only extension of the surveillanec inter for the subject valves. The current surveillance interval expires October 28,1993, t
GrmnstancCjLSurrounding the Proposed Chanug j
i In February 1992. Licensing Document Change Request (LDCR) FS92-007, prepared under l
provisions of 10 CFR 50.59, was reviewed and recommended for approval by the Plant R Board in accordance with section 6.4.1.6 of the Vogtle Electric Generating Plant (VEGP) TS.
l On February 21,1902, LDCR FS92-007 was approved by plant management. The subject LDCR revised table 6.2.4-1 of the VEGP Final Safety Analysis Repon (FSAR), in part, with respect to the ACCW supply and return containment isolation valves. Prior to the change,
'6.2.4-1 stated that the subject valves were subject to 10 CFR 50, Appendix J Type C IcakaBe l
testing requirements, and that they were normally open during operation but closed under pos accident conditions. Ilowever, as noted in footnote "g" to tabic 6.2.4 1, it is highly desirable that ACCW flow be maintained to the reactor coolant pumps (RCPs)if possible. Under most accide scenarios, ACCW flow would be maintained to support operation ofthe RCPs. Ther fore, the LDCR revised the leakage testing requirements to Type A and stated that the post accident position was open. In addition, the associated penetrations were added to FSAR tab penetrations that are not vented or drained during Type A testing. As a result, these va not Type C tested during the Spring 1993 refueling outage. They have, however, been 4
l during previous outages on both units.
The basis for the LDCR was that the subject valves do not receive a containment isolation s l
(they are remote manually operated), and the associated penetrations are considered l
to the desirability of maintaining cooling water to the RCPs under post accident conditions. In J
addition, it was thought that the ACCW was a closed system since it does not communicate directly with the containment atmosphere or primar coolant. Therefore,it was thought th A testing was suiTicient for these penetrations.
Ilowever, the safety evaluation failed to consider that, while the ACCW system is seismic category I and the hard piping is fabricated of ASME Section !!!, Class 3 materials, the a
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ENCLOSUlW 2 (CONTINUliD)
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECilNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER l
SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES I
i B ASIS FOR PEQPOSED Cll ANGE mstallation was in accordance ANSI B31.1 and an N-stamp was st aflixed. In addition, there i
l are some components such as motor coolers and flexible piping that are c ofClass 3 materials.
i Therefore, the ACCW system does not meet the ANSI standard criteria for a clowd system. In consequence, the supply and return isolation valves must be considered to perform an.solaiie function and should be subject to Type C testing. This was discovered recently during a j
subsequent review for the purpose of making corresponding changes to associated d
l documentation.
Georgia Power Company (GPC) requests that the proposed change be processed on an exige basis The required 24-month surveillance interval expires on Unit I on October 28,1993. In the absence of the proposed relief, Unit I would have to be placed in Mode 5 suniciently prior to October 28,1993, so that the required testing could be performed.
1 Dash 3
As stated above, the subject valves have been Type C tested during all previous refueling outages with the exception of the Unit 1 Spring 1993 outage. A review of the maintenance work order l
(MWO) history was performed on the ACCW containment isolation valves. This review show that there were MWOs for seat leakage, packing leaks, flange leaks, preventive maintenance, and l
several inspections, bui there were no "as found" Type C local !cak rate test (LLRT) failures ane the initial entry into Mode 4 on both units. An MWO was written ag inst 211V-1979 for high seat leakage during the period between the preoperational LLRT and initial Mode 4 entry, howeve this condition has not reoccurred.
in addition, a review of the vakes' LLRT history aner initial Mode 4 entry demonstrates the reliability and low leakage trends of these valves. Listed below are the maximum values the "as found" and "as len" LLRTs performed aner initial Mode 4 entry. This data was taken frorn six refueling outages between the two units. Note that penetration 28 is the ACCW s j
line and penetration 29 is the ACCW return line.
j PENETitATION 29 PENETitATION 28 MAXIMUM LEAKAGES MAXIMUM LEAKAGES tilV-1974 = 152 scem*
liiV-1978 = 20.5 scem tilV-1975 = 62.0 sccm lilV-1979 = 40.4 scem 211V-1974 = 99.6 seem' 211V-1978 = 49 2 seem 2ilV-1975 = 136.3 seem
?llV-1979 = 90.6 scem Includes leakage through associated i13 check valve E2-2 o.
Ets C L u d L K u 2 n. u m.. m m i
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS l,
REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES BASIS FOR PROPOSED CH ANGE The Inservice Inspection Program currently specifies a maximum allowable leakage of 1000 seem for each butterfly valve and 1500 seem for the check valve. (The leakage limit for the combination of valve HV-1974 and check valve 113 would be 2500 scem.) These limit based on Appendix J, but were established based on the low leakage histo y of these valves an serve the purpose of defining the point at which repair would be required. The Appendix J leakage limit for all penetrations subject to Type B and C testing (0.6L ) at VEGP is 228,273 i
a l
secm. The current total for Type B and C test leakage at VEGP as of September 10,1993,is 14398.8 secm. As of the last LLRT, the leakage for each of these four valves was as follows:
l IIV-1974 - 152 scem (this includes leakage past check valve 1-1217-U4-113 in parallel with HV-1974);llV 1975 - 11.6 secm;ilV-1978 - 9.3 secm; and ilV-1979 - 11.4 scem. In addition,it l
should be noted that the ic:t pressure, Pa, was 45 psig at the time these numbers were obtained.
l The test pressure has since been revind to 37 psig, so it would be reasonable to assume that the leakage would be less at the lower pressure.
h During the last outage for Unit 1, maintenance was performed on liv 1979 that could have FSAR at the alTected its leakage, and no LLRT was performed since it was not required by th:
p time. The motor and gearbox were removed and the limit switch settings were altered, but r:e l
work was done that would have afTected the valve seat. The standard work practice for setting limit switches on this type of soft-seated buttedly valve following this type of maintenance is as l
follows. First, the valve is manually closed using the hand wheel until 00 (closed) is reached, and i
the limit switch is set. Then the limit switch is tested by manually operating the valve again.
Finally, the valve is stroked using the motor until the limit switch actuates. At this point, the whcciis used to ensure that the valve is seated properly after the limit switch actuates. As a j
reference point, in the Spring of 1992 this type of work was performed on Unit 2 valve HV-19 and premaintenance and post-maintenance LLRTs were performed. The premaintenance
)
and the post maintenance leakage was well within the leakage limits for this valve. Therefor GPC is confident that Unit i valve HV-1979 remains leak-tight.
Furthermore, the probability of an event that leads to core damage and a failure of the ACCW f'
piping inside containment with a failure to isolate containment is not considered to be c The probability of containment, isolation failure following a core damage accident is mod the VEGP individual plant examination (IPE). In order to model a more conservative scenario of 4
containment isolation failure than was considered in the base case VEGP IPE, it was assumed tha
,I the occurrence of any core damage scenario would cause a break in the ACCW flow path and t
)
the operator would be required to isolate the ACCW system for successful containment i Based on a Type C test interval of 2 years, the frequency of core damage with containment isolation failure was found to be on the order of 10-7 per reactor year. Extending the required Type C test interval for these valves as proposed has a negligible impact on that pro E2 3
j ENCLOSURE 2 (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECIINICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES B ASIS FOR PROPOSED CH ANGE
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l Finally, the ACCW system is seismic category 1, and the hard piping is fabricated of AS Section 111, Class 3 materials. (There are some components such as motor coolers and flex piping that are not fabricated of Class 3 materials.) Therefore, even though the ACCW d l
i meet the ANSI standard criteria for a closed system,it can be considered to be highly reliable a there is reasonable assurance that for most events its integrity would be maintained.
J In conclusion, GPC submits that from the standpoint of safety, waiting until the Fall of 1994 (or until the next forced outage that requires entry into Mode 5) to perform !eakage testing on these valves is preferable to the alternative of shutting Unit I down solely for the purpose of 1
l Type C testing of these valves.
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ENCLOSURE 3
)
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAI RPECIFICATIONS REGARDING Al>*XILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES 1
10 CFR 50.92 EVALUATION i
Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to I
l proposed changes and has made the following determination:
- 1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change is a one-time only extension of the Type C leakage test interval for ae Unit 1 ACCW supply and return containment isolation l
valves. As such, it has no effect on the probability of any accident previously evaluated.
l Furthermore, based on the past leakage test history of these valves, there is reasonable l
assurance that extending the test interval to no later than November 1,1994, (or the next forced outage that requires entry into Mode 5) will not adversely affect the ability of these valves to perform their isolation function. Therefore, the proposed change will not involve a l
significant increase in the consequences of any accident previously evaluated.
- 2. The proposed change will not create the possibility of a new or different kind of accident any accident previously evaluated. The proposed change does not change the configur method of operation of any plant equipment, and no new failure modes have been defined fo l
any plant system or component. Furthermore, no new limiting failure has been identifie result of the proposed change.
- 3. The proposed change does not involve a significant reduction in a margin of safety. Ther i
continues to be reasonable assurance that the subject valves will remain capable of perfonning their isolation function. In addition, the proposed change avoids a plant shutdown solely for I
the purpose of performing Type C testing of these valves.
Conclusion j
Ilased on the preceding analysis, Georgia Power Company has determined that the propo change will not significantly increase the probability or consequences of any acciden evaluated, create the possibility of a new or different kind of accident from ay accident I
previoc;Jy evaluated, or involve a significant reduction in a margin of safety. There proposed changes ined the requirements of 10 CFR 50.92(c) and do not involve a s i
hazards consideration.
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ENCLOSURE 4
[
VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST AND REQUEST TO REVISETECIINICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMEN'T ISOLATION VALVES ENVIRONi{ ENTAL ASSESShENT 4
Identification of Proposed Action i
l The requested exemption and license amendment would grant temporary relief from the 2 year schedular requirement associated with Type C periodic localleak rate tests (LLRTs). The j
~
proposed action would allow the LLRTs to be performed for valves HV-1974 (and associated check valve 1-1217-U4-113),1975,1978, and 1979 prior to entry into Mode 4 following the next l
scheduled refueling outage (or next forced outage requiring entry into Mode 5), but no later than November 1,1994. The current surveillance interval expires October 28,1993.
Need for Procosed Action 4
One of the conditions of the Vogtle Electric Generating Plant (VEGP) Operating License, as specified in 10 CFR 50.54(o),is that primary reactor containments shall meet the containment L
]
Icakage test requirements set forth in 10 CFR Part 50, Appendix J. Appendix J to 10 CFR Part 50, Section Ill.D.3, requires, in part, that Type C LLRTs shall be performed in no case at intervals greater than 2 years.
Compliance with Appendix J to 10 CFR Part 50, Section III.D.3 would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was i
adopted. The requirement for a shutdown solely for the purpose of testing the subject valves, especially when the expected leakage is consid:rably less than requirements, would result in i
excessive costs in the form oflost revenues. Additiona!!y, shutdown solely for the purpose of testing the subject valves would result in an increase in occupational radiation exposure and a additional transient on the plant.
1 Environmental Imoact of Proposed Action The proposed action will not increase potential radiological environmental effects due to containment leakage beyond those already permitted by the regulations. Testing of Type C i
components under Appendix J to 10 CFR Part 50is intended to demonstrate that containme leakage from these components is within defined acceptable limits. These limits provide information used to calculate the maximum radiological consequences of a design basis accident.
i Appendix J limits the combined leak rate for all penetrations and valves subject to Type B test to less than 0.6 times the maximum allowable containment leakage rate with the containment i
l pressurized to its design limit (commonly termed "0.6 La")-
The subject valsc:, usociated with the auxiliary component cooling water (ACCW) system, been Type C tested during all previous refneling outages with the exception of the Unit 1 E4-1 4
ENCLOSURE 4 (CONTINUED) 1 VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST AND REQUEST TO REVISE TECIINICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES 1
i ENVIRONMENTALASSESSMENT i
1993 outage. A review of the maintenance work order (MWO) history was performed on the ACCW containment isolation valves. This review determined that there were MWOs for s leakage, packing leak, flange leaks, preventive maintenance, and severalinspections, but their were no "as found" Type C LLRT failures after the initial entry into Mode 4 on both Units. An j
MWO was written against 211V-1979 for high seat leakage during the period between the l
preoperational LLRT and initial Mode 4 entry; however, this condition has not reoccurred. A l
review of the LLRT history for the subject valves after initial Mode 4 entry demonstrates the l
reliability and low leakage trends of these valves.
l Furthermore, the probability of an event that leads to core damage and a failure of the ACCW piping inside containment with a failure to isolate containment is not considered to be credible.
The probability of containment isolation failure following a core damage accident is modeled in the VEGP individual plant examination (IPE). In order to model a more conservative scenario of containment isolation failure than was considered in the base case VEGP IPE,it was assumed that the occurrence of any core damage scenario would cause a break in the ACCW flow path and that the operator would be required to isolate the ACCW system for successful containment isolation.
11ased on a Type C test interval of 2 years, the frequency of core damage with containment isolation failure was found to be on the order of 10-7 er reactor year. Extending the required p
Type C test interval for these valves as proposed has a negligible impact on that probability.
Therefore, radiological releases will not differ from those determined previously, and the proposed exemption does not otherwise afTect facility radiological effluent or occupational exposures. With regard to potential nonradiologicalimpacts, the proposed action does not affe plant nonradiological efiluents and las no other nonradiological environmentalimpact. Therefo Georgia Power Company (GPC) concludes there are no measurable radiological or nonradiological environmentalimpacts associated with the requested exemption.
Alternatives to Proposed Actiort Since GPC has concluded there is no measurable environmental impact associated with the requested action, any alternative with equal or greater environmentalimpact need not be evaluated. The principal alternative would be to deny the proposed action. Such action would not enhance the protection of the environment.
Alternative Use of Resources This action does not involve the use of resources not considered previously in the Final Environmental Statement for Vogtle Electric Generating Plant, Units I and 2.
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I ENCLOduRt. 4 (wn nauwj j:,
I VOGTLE ELECTRIC GENERATING PLANT ll, 10 CFR 50 APPENDIX J EXEMPTION REQUEST AND l=
REQUEST TO REVISE TECHNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER l
SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES ENVIRONMENTAL ASSESSMENT l
Conclusion l
Based on the foregoing environmental assessment, GPC concludes that the proposed action will l
not have a significant efTect on the quality of the human environment.
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E4-3
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ENCLOSURE 5 i
l VOGTLE ELECTRIC GENERATING PLANT i
REQUEST TO REVISE TECHNICAL SPECIFICATIONS l
REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES PROPOSED TECHNICAL SPECIFICATION CHANGE t
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9310040341 930930 4
PDR ADOCK 05000424
Y CONTAINMENT SYSTEMS i
SURVEILLANCE REOUIREMENTS (Continued)
I If any periodic Type A test fails to meet 0.75 L the test schedule i
for subseyent Type A tests shall be reviewed an8 approved by the l
b.
j Comission.
If two consecutive Type A tests fail to meet 0.75 L, a l
Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed; The accuracy of each Type A test shall be verified by a supplemental l
c.
test which:
i Confirms the accuracy of the test by verifying that the absolute i
3 j
1) value of the supplemental test result, L,, minus the sum of the Type A and thE superimposed leak, L, is equal to or less than i
0.25 L,;
l Has a duration sufficient to establish accurately the change in 2) leakage rate between the Type A test and the supplemental test; and i
i Requires that the rate at which gas is injected into the 3) containment or bled from the containment during the supplemental j
l i
test is between 0.75 L, and 1.25 L,.
j Type 8 and C tests shall be conducted with gas at a pressure not less than P, 37 psig, at intervals no greater than 24 months *except for
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tests involving:
l 1)
Air locks and l
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Purgs supply and exhaust isolation valves with resilient 1
il 2) material seals.
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i Air locks shall be tested and demonstrated OPERABLE by the require-i e.
ments of Specification 4.6.1.3; I
Purge supply and exhaust isolation valves with resilient material l
l seals shall be tested and demonstrated OPERABLE by the requirements
)
f.
I of Specification 4.6.1.7.2; l
The provisions of Specification 4.0.2 are not applicable.
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CUl4jC ! Of f [
l~c 01Llif} bHl no kfW fd6/t5l Amendment No. 63 (Unit 1) i V0GTLE UNITS - 1 & 2
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3/4 6-3 Amendment No. 42 (Unit 2) i
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i UNITE 5 $?Afe8 i
NUCLEAR REGULATORY COMMISSION wasenetom.s.c.smsem l
October 26. 1993 1
i s...e I
Decket No.30-414 I
i l
Mr. C. K. Nctey Vies President - Nuclear i
VogtleProject Soorgia peper Ceepany P. 4. Som 1896 j
Straingham, Alabama 38f01 Dear Mr. McCoy
$UBJECT: !$$UANCE OF EXENPf!0N TO 10 CFR PART 50 APPtNDt1 J.
TYPt C TESTS - V087LE ILICTile 80tERATIN8 PLANT, SECTION!!!.D.3.M7783)
UNIT 1(TACNO.
,i ten from the se$edule regelresents of Ig CPR Par (GPC),
80,1993 Georgia Power Company et al.
j ty letter dated t se, requested an en III.D.3 for the aust11ary c cet cooline water steply Appendiz J. Secti and return containment isola,tten valves 1W-1974 associateR shock valve i
and 1 W-1975 the Vogtle Electrie 1-1117-U4-113).15-1973. INV 1978, The regulation reqeiros that Type B and C
. Beneratine Plant (Vogtle). Unit 1.
local leak rate tests be conducted at intervals no greater than 24 months.
l GPC requested the exemption to avoid a Vogtle Unit I outage solely for the purpose of performing these tests.
The Nuclear Regulatory Consission has erasted the requested schedule' exemption until prior to entry into Nede 4 fram the next scheduled refueling outage (or the nart forced outsee requirina entry inte Nede 5), but no later than Merester 1 1994. The Nec staff find's that the increased eenfidenne in 1
containmend inteerity fo11 swing testing is not' sufficient to offset increased l
personnel radiation exposure and other risks associated with perfernine these tests at power, er the undue burden of a forced outage to perfore the testing while shut doen. The staff believes there is a high degree of confidence that the components affected by this amenption will not degrade to an acceptable i
extent during their extended operatine interral between tests.
The igIC staff finds that anting the exemption from the requirements of is autherised by law, will not 10 CFR Part 50 Ascendia Section 111.0.3, d safety and is censistent with present an sedu,e risk to tthe commen defense and security. The sta lads that special sublic health an d
that applicatten of the circumstances, justify the exemptions namelyIs not necessary tc achieve the I
resulation in the particular circumstances underlying purpose of the Rule.
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20*d BPS 6 tsS 904 3115 311000 Dd450 le6i:00 S66i-0E-10
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Mr. C. K. McCoy 4 i
A copy of the esemption and the supporting Safets Evaluation Iw the staff is i
enciesed. The exempties has been forwarded to tw office of the Federal j
Register for publicatica.
3 Sincerely.
i h,
' ht te - - -
Robert A. Hermann Acting Pro,tect Director ProjectDirectorahII-8 DivisionofReacterProjects-1/II 1
Office of Nuclear Reactor Regulaties i
l i
Enclosures:
j 1.
Emesption to 10 CFR Part 50 Appendix J 2.
Safety Evaluation l
ccw/encleeurest i
See next page i
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I seiss resi-es * *
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russ's CO'd OPS 6 PSS E 311S 3Y000 DdNSO W961:00 S661-0E-TO 4
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l Mr. C'. K.
J Georgia Power any Vogtle (tectric Genenting Plant i
i cc Director Mr. J.-A. antley Harold Rehets, Natural Resources l
Managee - Licensing arteent of i
la Power Company 1
tutler Street, St. Suite 1388 i
P.
Sea 1895 Atlanta, Georgia 30334 i
Birmingham, Alabama 15:01 Attorney General i
l Mr. J. 8. Seasley Las artment I
General Manager. Vogtle Electric 18t J leial h ildine Generati Plant Atlanta, Georgia 303:4 i
P. c. tem 1 Wayaashem, Georgia $0034 Nr. Alan R. Herdt Pro oct Branch M i
Reglenal Administrater, Region !!
U.
. Nuclear Regulatory Cesatssion U. 8. Nuclear Regula C4 emission Marietta Street lei. Suite 1900 101 Marietta Street
., Suite 1900
~
anta, Georgta 3 as Atlanta, toergia 3 23 Mr. Dan H. Smith, Vice Pmsident Office of Planalng and Budget Peuer Supply Operatiana Asen 8188 Ogle Power CorpoMtion SW.
ILOG Enchange Place 270 Washington Street i
Atlanta, toergia 303$4 Tuchar, hergia 80085-1849 E
im 1
01Tice of the County Commissioner Charles A. Patriaia.fs j
Burke County Cassiesten Paul,FleerNastings,Jano
& Walker Waynesbere, Seergia 30830 lith 1080 Connecticut Avenue 10036, 101.
Mr. J. D. Veedard Washtegten,DC senior Vice President -
Itaclear Operatione Arthur H. Deshr. Esquire i
ta Power Company Troutman Staders i
P 0. Sex 1895 NationsBank Plua Blmingham, Alabama 35201 600 Peachtree Street, Nf.
Suite 5805 i
i Attenta, toergia 30306-1216 Resident Inspector i
U. S. Nuclear Regulatory Cassission i
P. O. Sex 571 Waynesbere Georgia 30830 4
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WITf8 ETAfft of AMl!ca i
l locLEAR RtalLATORY cGel1111M 4
4 l
In the Matter of S
h 450MIA POWEA COMPAllY
?
l 08LETWORpt POWER CORPORAT!0ll 1
Decket No. 50-424 IUl!CIPAL ELECTRIC Alml0RITY OF GEORBIA h
CITY OF DALT0R, GEORI!A h
(Unitfle.1)Vestle Electric Generating Plant,
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,1 EIDIPTIM
- s. -
The leergia Power Company, et al. (CPC, the licensee), is the holder of 4
l Operating License No. NpF-44, which authorizes operation of the Vogtle l
Electric Generating Plant, Unit 1, at steady state reacter core power levels i
not in exeess of H11 segawatts thema1. The license provides, assag other l
things, that the licensee is subject to the rules, regulations and orders of the Ceemission new or hereafter in effect.
i 1
The plant is a pressurited water reacter located at the licensee's site i
i in turke County, Georgia.
l f
Section 50.54(e) of 10 CFR part 50 requires that primary reacter I
containmentsforwatercooledpowerreactersbesubjecttotherequirementsof 1
Appendix J to'IO EFR Part 50. Appendix J eentains the leakege test i
requirements, schedules, and acceptance criteria for tests of the leak tight
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integrity of the primary reacter containment and systems and compements which
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l penetrate the containment.
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t Section !!!.D.3 of Appendix J te le CFA Part 50 requires that Type C t u k rate tests be performed daring each reactor shutdown for refueling but in 1
i no case at intervals greater than 2 years. Type C tests are intended to asasure oestainment isolatten valve lankage rates for certain containannt isolation valves.
i i
4 l
Sy letter dated September 9,1993, the licensee requested a ese-time exemption from the requirements of Appendix J. Section !!!.O.1. free 14 months l
l to prior to entry into Mode 4 following'the next scheduled refueling outage f
l (orthenextforcedoutagerequiringentryintoMode4),butnelaterthen November 1, 1994 for the Unit I auxiliary camponent coeltag water (ACQO i
supply and return containment isolation velves 15-1974(andessectatedeheen valve 11817-U4-113),1W-1975,1HV-1978.and1W-1979. In their request, the licenses provided the date when the leak tests had last been performed and the i
date when the current teak test will septre.
The leak tests for which the licenses has requested schedular esempties I
were last conducted during the fall 1991 refueling outage, based on the I
i infomatten provided in the licensee's application. The liceasse has stated L
that, in the absence of the proposed relief. Unit i usu1d have to be placed in i
f Mede 5 aufficiently prior to October 28,1998, se that the required testing I
could be performed.
a IV.
2 i
The licenses presented information in support of their request for an l
extension of the Type C test intervals. The Appendix J 1eakage limit for all j
peastrations sebject to Type I and C testing (0.6L.) at vogtle is 228,873 j
seca. The current total for Type 8 and C test leakage at Vogtle as of i
l) i suiss smes-ee-et cyse.,
90*d 8PS6 PSS E 3115 311000 DWSn WOO 2:00 S66!-0E-10
4 3
l September it. 1993, is 14,398.8 seca. As of the last Type C local leak rate
{
test (LLRT), the leakage for each of these four valves was as follows:
1 1W-1974 - ilt occa (this includes leakage past check valW 1-!!!7-04-113 in parallel with 1W-1974); 1W-1978 - 11.8 seen 1 W-1978 - s.3 seen; and i
I 1W-1979 - 11.4 seca.
The licenses stated that, based on the past leakege test htstory of i
these valves, Wre is reasonable assurance that utending the test interval
- t. no ia.r a. Wo.nr 1,1,u (.r m next for.ed outa.e na re, sires entry into Mode 6), will not adversely affect the ability of these valves to i
perform their isolatten function.
V.
Resed en the above, the staff finds there is reasonable assurance that l
the containment leakege-limiting function will be maintaind ted that a forced l
Therefore, the staff finds l
sutage to perfem Type C tests is not necessary.
l the requested temporary exemption, to allou the Type C test interval for the ACCW supply and retum containment isolttten valus to be extended to prior to l
entry into Mode 4 following the next scheduled refueling outage (or W next l
forced outage requiring entry inte Mode 5), but ns later than Nevenber 1, l
1994, to be acceptable. The exemption requst has been evaluated in a safety l
evaluation dated October 26,1993.
Accordingly, the Commission has detemined that, pursuest to 10 CFA f
50.lt(a), thd'fWluested exemption is authorised by law, will not present an f
undue risk to public health and safety, and is consistent with she cessen l
f defeare and security. The Commission finds that the special circumstances as As specified in le CFR f
requiredby10CFR$0.lt(a)(2)arepresent.
50,1t(s)(1)(11), special circumstances are present whenever the a 4
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l the reculation in the particular circumstance would not serve the underlying purpose af the rule er is not necessary to achieve the underlying purpose of l
the rule. The underlying purpose of the rule is to ensure that the components coeprising the primary containment boundary are maintained and leak tested at 1
periodic and appropriate intenals. The 24-eenth maximum interval was l
eriginally expected to bound the typical operating cycle, including a limited amount of mid-eycle estage time. Strict adhe nnes to the to-month maximus interval is not necessary to meet the underlying purpose of the rule in that, i
j taking into consideretten the requested extension, the components that i
comprise the primary containment boundary will still be tested at a frequency.
i l
that is appropriate to these coepenents and their application.
t j
Therefore, the staff finds the requested temporary exemption, to allow f
the Type C test interval for the ACCW supply and nturn containment isolath valves, as described in the licensee's September 30,1998, to be extended to prior to entry into Mode 4 following the next scheduled refueling outage (or the next forced outage requiring entry into Mode I), but ne later than November 1, 1994, to be acceptable.
i t.n exemption is hereby granted from the requirements of section !!I.D.3
}
of Appendix d to le CFR part le, which requires that Type C tests be perfomed during each reactor shutdown for refueling but in no case at intervals greater l
l than 1 years, te prior to entry into Mode 4 following the next scheduled refueling outige (or the next foned estage requiring entry into Mode l), but i
i no later than November 1,1994. for the subject valves.
4 i
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98:ST East-eE41
-.a 80*d 8?S6 PSS 906 3M S n 006 O WSn WOIE:00 566T-0C-T0
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E l.
Pursuant te le CFR 61.31, the Commission has determined that the granting of this Exception will have no significant tapact on the quality of l
l thehumanenvireneont($$FA84406datedOctober18.t993)
J i
This exemption is effective upon jssuance.
FOR THE RUCLEAR REGilLATORY Cs WIssION
- bit, i
ivision of Reacter tjects - !/11 Office of Nuclear Reactor Regulation v
Dated at Rockville, Maryland l
this 5th d4y Of October 1993 i
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60*d BPS 6 PSS 906 311S ETHOOh OdNSn 1J22:00 S66T-0E-TO 1
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NUCLEAR REGULATORY COMMIS810N j
s waamneten.e.c.musaset l
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s&FETY EVALUAT10N BY TMt ePFfet 0F NUCLEAR REACTot RFaul ATION i
TIMp0RARY EttWf!M ftoM ApptNDft J fMTttVAL l
FOR LOCAL LEAM A&It TESTIM OF CONTA15tMT pfMETRATIONS V08TLE ELEcfttt RENERAffM PLANT. LETT 1 DGCKET IEL 60-4s4 i
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1.0 jgIgBgilig l
1 Ry letter dated September 30,1993 Seergia Power et al. (the licensee), requested a license amendment to t
le Electric ificati surveillance s a feetnote(TS)t extends the 8enerating Plant Unit 1 (Vogtle). Technical mquirement 4.6.I.2d. The requested change-tha i
surveillance interval for the next trod c leakage test of the auxiliary component coeli water (AC s
y and return containment ivalve11117-U4-113; i
isolation valves 1W-1974 and associa 1W-1978 lW-1978 and 1 -1979 to arter to entry into Mede 4 fol kng the d
next sche,duled refu,elies outage er the next forced outage requiring entry into Nede 8), but no later then ovember 1, 1994. The amendment provides a i
ens-time only extension of the surveillance interval for the sub est valves.
As presently written, TS 4.6.1.2d reentres that 10 CFR Part 80, pendiu J, j
section !!1.0.3 Type 8 and C tests for the subiect valves be ce acted at j
intervals ne greater than 24 months.
In February 1991 the licenses prepared and iM FS99-007 under the are siens o lamented Licensi 59 and in Change Request (Vogt's TS 6.4.1.4.
The LOCR revised Table 5.t.41 of the i
accordance with i
Vogtle Final Safety Analysis Report CFSAR)lves.part with respect to the ACCW in supply and rotun containannt isolatten va Prlertothechange, Table 8.t.4-1 stated that these valves were subject to le CPR part 80 hey wonAppendix Section !!!.0.3, type C leakage testing requirements, and that t l
i normally open during operation but closed under post-accident conditions.
However, as noted is festnote 'g' to Table 4.1.4-1. ACCW flew should be mainteird to the roaster coolant pump (ACps)underseatpost-accident i
conditions, if possible. Therefore, t LDCR changed the leaksee testing requirteents free Ty ed the post-accident position of thevalvesto" opes.pCtoTyeeAandIn addhien, the associated sonstrations were F8AR Table 6.t.6-1 as trationsthatarenotvonfederdrainedduringType A testing. As a resul e this LOCR, these valves were not Tybe C tested i
during the Vogtle Unit 1 spring 1993 refueling outage, althousi they had been j
tested during previous eutages sa both units.
t i
The licensee's basis for the LOCR was that the sub,iect valves de not receive a and the containment 1selstion signal (they are remote manually operated),he RCps.
j The associated penetrations are needed to maintsin cooling water to t 1
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?ggatom Wp estST ES$1-e2-91 l
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11eennes thought that the ACCW was a closed system because it does not i
l communicate 41mstly with the containment atmosphem er primary toelant.
Thus, when appmving the LOCR, the licensee had concluded that Type A testing was suffielent for these penetratiees.
However, during a recent document review, the licensee discovered that the safety evaluat en for the LDCR was flowed. The evaluatten failed to son 1 der while the ACCW system is seisets catetery 1 and the hard piping is thatlcated of A$NE Sulien !!! Class 3 mterials, the systen had been l
fabr tastalled in esserdance with AASI 831.1 and no R-stamp was affixed. In addition, f Class 3 materials, heefers, the ACCW systes does not meet theso i
composed o l
ANSI standard criteria for a closed systeu. Consqquently the supply and retura isolaties valves est be considered to perfem an Isolation function I
and should be subject to Type C testing.
i l
2.0 SERijg a
h subject valves have been free C tested durise all previous refooling estages with the exception of the Unit 1 Spring 1998 outase. The Itcensee i
reviewed the maintenance work order (1910) mistory of the ACCW containment i
isolatten valves. This review found llf0s for seat leakage, packing leaks, preventive saintenance, and several inspections, but food no flanga lagkg,pe C local leak rate test (LLRT) failures after the initial setry i
'as found y
1 into Mode d on either Vogtle mit.
l
)
m Ilconsee aise reviewed the LLRT history of the valves after initial Nede 4 entry and found thts history to desenstrate the reliability and low leakage l
trends of these va' vos. Listed below are the maximum values, taken free six l
refueling outases between the two unitsIry.r both the 'as found* and 'as lef fo The below values indicate the LLAfs perfereed after initial Nede 4 m
'werst case' lukage. penetratten is is the ACCW supply lies and penetration 29 is the ACCW retum line.
ptRtfRAf!0N 38 PflETRATION tl marmuuta= %
maciane i rmee l
1M-153 = E0.5 som 15-1No e 158 secua 1HV-1979 = 40.4 sces 1 W -1975 = 82.0 seca l
IW -1976 = 49.8 sces tW-1974 s 99.8 scas*
l SW -1979 = 90.8 se e tW-1975 = 134.3 sce i
- Includes leakage through associated check valve 1-1t17-U4-113 le latervice Inseettien program currently specifies a maximus The V le leaksee of 1000 sece for each butterfly valve and 1500 seen for the all Tlie leakase limit for the combinatlen of valve 1W-1974 and i
j check valve.
would be 8500 eten. These limits were not based en check valve 1 1817 04-113but were established based on the low leakage history Appendix J requirementsIne the seint at which tir would be required. The of these valves and def Appendix 41dakage limit for all penetrations ect to Type 8 and C testing 2
1 for Type I and C test i
1eakat)e at Vagtle, as of,273 scos. The current 10,1993,is14,398.8 (0.5L at Vogtle is 238 i
september jl I
sEIST sest-et-e1
}
ET/1s'sl NE 3MS 31.!XM MtGO Wt1E2:00 S66T-02-TO
I l:
- i 1W-1974 - 1st LLRT,{the leakepe for eash of these four valves was as fellows:
l seem this includes leakees past check valve 1-1217-04-113 in pan 11e1 with IW-1TF4); 1M 1975 - 11.5 seca 1W-1978 - 9.3 scca; and 1 p -1979 -
was el pais at the time these numbers were 11.4 sces. The test pressure $a.. ince been reduced to 37 esi with previous license Asendments 43 (Unit 1) and 42 (Unit 1),l in acc shtained. h test pressure s
and the leakage l
would be less at this lower pressure.
f Ouries W 1ast ea for Unit 1 the lieeuee perfereed maintenance en IW-1975 that seuld e affected,its leaka f
d LLAT since it Was net required by the FSAR at the time. pe, but per eree nehe saintenanc 4
the meter and esattes and altering the limit switek settings, but M work was The standard work practice fbr done that woeld have affected the valve seat.
setting limit switches en this tsee of soft-seated betterfly valve vellowim first, the valve is manual'y closer j
this type of maintenance is as f"elIows:
using the hand wheel until g' (fully closed) is macher', and the limit switch i
again. Finally,he limit switch is tested by manually operating the is set. Then t actuates. At this point, the haul wheel is used to ensure'that the valve is in the seated preserly after the limit switch actuates. As a reference point 1
Spring of 1992 this type of work was erfereed on Unit t valve IW-1976 and.
pre eatatenance and post-maintename.LRTs were serforced. The pre and post-maintenasse leakage was well within the leakage 11eits for this valve.
l The rebabil of contalment isolatten failure fellent! tion (!p he Ipt a core led in the Voltle individual plant asent i
accident is 1933. In order to sedel.a more was schmitted by letter dated December 23,tica failure than was considered in conservatin scenario of containment isola the base case Vestle Ipt, the licensee assened that the occurrence of any core damage scenarie would cause a break in the ACCW flev path and that the operater would be required to iselste the ACCW system for successful the centstament isolatten. Based on a Type C test interval of 1 years, d by the I
frequency of core desage with cost'pineont isolation failure was founper licenses to be en the order of 18 that extending the required Type C test interval for these valves beyond the 4er perfed has a neglisible impact en that probability. Thus, i
l Appendix J of en event that ends to care damage end a fallere of the I
t w probabili de containment with a failure to toelate containment is not l
ACCW piping i the licenses h staff concurs that the l
considered to be credible between empiration of the current leak tests to l'
additional operatten peri prior to entry inte Nede 4 11evinethenextscheduledrefuelingoutage(or
, but no later then l
the next fereed outage requiring entry inte Mode 5) decrease the margia between i
is not expected to significantly i
Nevesiber 1,19g4, leak rate and L,.
l expected as-feend and the hard piping is fabricated of N ACCW aystem is ' seismic catenery 1,Sese componentslarials.
such as ester coolers ASRE lection !!! Class 3 materLals.
The licessee and flexible piplag, are not fabricated of Class 3 ma cone)uded that, even theegh the ACCW does not meet the ANS! standard criteria for a alesed system, it can be considered to be highly reliable and that there natatnined. h staff concurs with this conclusion. grity would be is reasonable assurance that for sost ever.ts its inte i
ss 51 spot-et-et N*E 31!S 3 YD00 OdNSn Wtf2:00 S66i-0E O
- -. - - ~. _ _. _
- ~ -. _ -. _. _ _. _ _ _ _ _ _,,
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I The NRC staff alas finds that the t-hear interval requirement for Type 8 and C components is sufficient for timely estectlen of sigstficant deterioration l
while, at the seas time permitting the teste to be performed during piant Leak rate teettag of the penetrations during shutdowns is preferable Sees penetrations outages.
because of the leser radiation esposere te plant personnel.For these penetrations con not be tested at power.
during power operaties er for which testing at power is inadvisable, the increase in eenfidents of sentainment leattieht integrity following a l
successful test is slight and does set justify a plant shutdows specifically to perform the tests w thin the t-year ties period, considering the factors i
discussed obeve.
i 3.0 M f
mation, the NRC etsff finds the requested one-time eel Based on the above eval11ance requirement 4.8.1.2d, is acceptable. AsIrevided i
l trod i fuglen to TS survei festnote,tne = swi and r.erval for the n.extafttien,me, the surveillance int en so. ui.e nt e
!=g=(and associated ch'ch, valve 3-1317-M.1135,131975,1W-1978' and t of l
~
W s
}Wi-i ine.= sen.or Vestl6 nit 1 ts
- prior %,e nog i,nte M, ode.4 1979, is extended f U
ett l
si 1ed ro set er t or,e e requiringentryinteNedeI).butne9aterpanNevenber1,itM.a l
k Principal Centributer C. E. Carpenter. Jr.
l Date:
October 26,1993 i
O@ 9 9
W 48tET EseT-95-e1 crect.,4 6 MS E 31gg 311000 DM&#
Wt'2:00 S66T-EE-10
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f UNITED STATES I
f.). i j
NUCLEAR REGULATORY COMMISSION l
- , kb ff WASHINGTON. D.C. 20086-0001
'+
October 27. 1993 Docket Nos.
50-424 l
and 50-425 Mr. C. K. McCoy l
Vice President - Nuclear Vogtle Project Georgia Power Company i-P. O. 80x 1295 l
Birmingham, Alabama 35201
Dear Mr. McCoy:
SUBJECT:
ISSUANCE OF EXIGENT AMENOMENT - V0GTLE ELECTRIC GENERATING PLANT, j
UNIT 1 (TAC NO. M87782) 4 i
I The Nuclear Regulatory Commission has issued the enclosed Amendment l
No. 70 to Facility Operating License NPF-68 and Amendment No. 49 to j
Facility Operating License NPF-81 for the Vogtle Electric Generating Plant, i
Units I and 2.
The amendment consists of changes to the Technical Specifications (TS) in response to your application dated September 30, 1995, and is applicable to Unit 1 only.
i The amendment is a one-time only revision to the TS surveillance requirement 4.6.1.2d for Unit I that adds a footnote extending the surveillance interval i
for the next required Type C leakage test of the auxiliary component cooling water supply and return containment isolation valves 1HV-1974 (and associated l
check valve 1-1217-U4-113), 1HV-1975, 1HV-1978, and 1HV-1979 to be extended from 24 months to prior to entry into Mode 4 following the next scheduled j
refueling outage (or the next forced outage requiring antry into Mode 5), but no later than November 1, 1994.
i A copy of the related Safety Evaluation is also enclosed. A Notice of i
Issuance will be included in the Commission's biweekly Federal Resister notice.
i Sinc ely, C. E. Carpenter Jr.,
ng Project Manager Project Direc orate II-3 Division of Reactor Projects - I/II 1
Office of Nuclear Reactor Regulation b[
Enclosures:
1.
Amendment No. 70 to NPF-68 2
(,( b 2.
Amendment No. 49 to NPF-81 3.
Safety Evaluation
(
cc w/ enclosures:
See next page
]
,pCom, n
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l Mr. C. K. McCoy Georgia Power Company Vogtle Electric Generating Plant i
cc:
Mr. J. A. Bailey Harold Reheis, Director Manager - Licensing Department of Natural Resources Georgia Power Company 205 Butler Street, SE. Suite 1252 P. O. Box 1295 Atlanta, Georgia 30334 Birmingham, Alabama 35201 4
Attorney General i
Mr. J. B. Beasley Law Department l
General Manager, Vogtle Electric 132 Judicial Building Generating Plant Atlanta, Georgia 30334 j
P. O. Box 1600 Waynesbore, Georgia 30830 Mr. M aa R. Herdt Project Branch #3 Regional Administrator, Region II U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission 101 Marietta Street, NW. Suite.2900 101-Marietta Street, NW., Suite 2900 Atlanta, Georgia 30323 Atlanta, Georgia 30323 Mr. Dan H. Smith, Vice President Office of Planning and Budget Power Supply Operations i
Room 615B Oglethorpe Power Corporation 270 Washington Street, SW.
2100 East Exchange Place Atlanta, Georgia 30334 Tucker, Georgia 30085-1349
)
l Office of the County Commissioner Charles A. Patrizia, Esquire Burke County Commission Paul, Hastings, Janofsky & Walker Waynesboro, Georgia 30830 12th Floor 1050 Connecticut Avenue, NW.
Mr. J. D. Woodard Washington, DC 20036 4
Senicr Vice President -
4 l
Nuclear Operations Arthur H. Domby, Esquire l
Georgia Power Company Troutman Sanders i
P. O. Box 1295 NationsBank Plsza Birmingham, Alabama 35201 600 Peachtree Street, NE.
Suite 5200 Atlanta, Georgia 30308-2216 4
l Resident Inspector U. S. Nuclear Regulatury Commission P. O. Box 572
(
Waynesboro, Georgia 30830 k
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UNITED STATES
)
j-f j
NUCLEAR REGULATORY COMMISSION g
j WASHINGTON D.C. 20085-0001 1
/
j GEORGIA POWER COMPANY
}
OGLETHORPE POWER CORPORATION i
MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA
{
CITY OF DALTON. GEORGIA V0GTLE ELECTRIC GENERATING PLANT. UNIT 1 j
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. NPF-68 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
~
i i
A.
The application for amendment to the Vogtle Electric Generating.
1 Plant, Unit 1 (the facility) Facility Operating License No. NPF-48 1
filed by the Georgia Power Company, acting for itself, Oglethorpe
{
Power Corporation, Municipal Electric Authority of Georgia, and l
City of Dalton, Georgia (the licensees), dated September 30, 1993, i
complies with the standards and requirements of the Atomic Energy i
Act of 1954, as amended (the Act), and the Commission's rules and i
regulations as set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the i
provisions of the Act, und the rules and regulations of the 1
Commission; i
8 C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set i
forth in 10 CFR Chapter I; i
D.
The issuance of this amendment will not be inimical to the common i
defense and security or to the health and safety of the public;
}
l E.
hissuanceofthisamendmentisinaccordancewith10CFRPart i
51 of the Commission's regulations and all applicable requirements have been satisfied.
de-
l.
l i
! 1 l'
2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
)
NPF-68 is hereby amended to read as follows:.
i Technical Soecifications and Environmental Protection Plan i
j The Technical Specifications contained in Appendix A, as revised through
)
Amendment No. 70, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated i
into this license. GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
4 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMISSION 4
k-Robert A. Hermann, Acting Director Project Directorate 11-3 Division of Reactor Projects - I/II j
Office of Nuclear Reactor Regulation
Attachment:
l Technical Specification j
Changes i
j Date of Issuance: October 27, 1993
)
a i
e
f
- t UNITED STATES i
j NUCLEAR REGULATORY COMMISSION
- a ff WASHINGTON. D.C. 20E05-0001
%; c/x; GEORGIA Polder C(BtPANY OGLETHORPE Polder CORPORATION MUNICIPAL ELECTRIC AllTHORITY OF GEORGIA CITY OF DALTON. GEORGIA V0GTLE ELECTRIC GENERATING PLANT. E I
AMEfEMENT TO FACILITY OPERATING LICENSE l
Amendment No. 49 License No. NPF-81 1.
The Nuclear Regulatory Commission (the Commission) has found thatI A.
The application for amendment to the Vogtle Electric Generating l
Plant, Unit 2 (the facility) Facility Operating License No. NPF41 l
filed by the Georgia Power Company, acting for itself, Oclothorpe Power Corporation, Nunicipal Electric Authority of Georg< a, and' city of Dalton, Georgia (the licensees), dated September 30, 1993,*
complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurarr.e (1) that the activities authorized by this amendment can be conducted withcut endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E.
W ' issuance of this amendment is in accordance with 10 CFR Part l
51 of the Commission's regulations and all applicable requirements have been satisfied.
i i
e
=
l
~
l
l a
2-4 i
2.
Accordingly, the license is hereby amended by page changes to the i
Technical Specifications as indicated in the attachment to this license j
amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby amended to read as follows:
4 l
Technical Specifications and Environmental Protection Plan i
The Technical Specifications contained in Appendix A, as revised through Amendment No. 49, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated j
into this license. GPC shall operate the facility in accordance with j
the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its ' ate of issuance and shall d
i be implemented within 30 days of issuance.
i FOR THE NUCLEAR REGULATORY COMISSION l
l I h.
w-j obert A. Hermann, Acting Director l
Project Directorate II-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation i
l
Attachment:
i Technical Specification j
Changes Date of Issuance:
October 27, 1993 i
i l
4 1
m...e
w d
ATTACHMENT TO LICENSE AMENDMENT NO.70 FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND TO LICENSE AMENDMENT NO. 49 FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 4
i 4
i j
Replace the following page of the Appendix "A" Technical Specifications with the enclosed page.
The revised page is identified by Amendment number and contains vertical lines indicating the areas of change.
Remove Pace Insert Pane 4
l 3/4 6-3 3/4 6-3 i
i t
1 l
I
~
j
~
I 4
4 4
e
I l
i CONTAINMENT SYSTEMS I
SURVEILLANCE-REOUIREMENTS b.
If any periodic Type A test fails to meet 0.75 L the test schedule for subsequent Type A tests shall be reviewed an8 approved by the
)
Commission.
If two consecutive Type A tests fail to meet 0.75 L Type A test shall be performed at least every 18 months until tw, a o
consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed; j
c.
The accuracy of each Type A test shall be verified by a supplemental test which:
1)
Confirms the accuracy of the test by verifying that the absolute value of the supplemental test result, L, minus the sum of the Type A and the superimposed leak, L,, is, equal to or less than 0.25 L,;
2)
Has a duration sufficient'to establish accurately the change in leakage rate between the Type A test and the supplemental test; and 3)
Requires that the rate at which gas is injected into the containment or bled from the containment during the supplemental test is between 0.75 L, and 1.25 L,.
d.
Type B and C tests shall be conducted with gas at a pressure not less I
than P, 37 psig, at intervals no greater than 24 months
- except for testsinvolving:
1)
Air locks and 2)
Purge supply and exhaust isolation valves with resilient material seals.
e.
Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3; f.
Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.2; g.
The provisions of Specification 4.0.2 are not applicable.
i
- The Type C test interval for Unit I valves HV-1974 (and associated check valve 1-1217-04-113), HV-1975, HV-1978, and HV-1979 may be extended to prior to entry into Mode 4 following the next scheduled refueling outage (or the next forced outage requiring entry into Mode 5), but no later than November 1,1994.
V0GTLE UNITS - 1 & 2 3/4 6-3 Amendment No. 70 (Unit 1)
Amendment No. 49 (Unit 2)
.p" %g\\
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UNITED STATES
[#
i NUCLEAR REGULATORY COMMISSION ei WASHINGTON. O.C. 20686-0001 s
N.... f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO 70 TO FACILITY OPERATING LICENSE NPF-68 j
AND AMENDMENT NO.49 TO FACILITY OPERATING LICENSE NPF-81 j
q 5
V0GTLE ELECTRIC GENERATING PLANT. UNITS I AND 2 DOCKET NOS. 50-424 AND 50-425 4
1.0 INTRODUCTION
i l
i By letter dated September 30, 1993, Georgia Power Company, et al. (the licensee), requested a license amendment to change the Vogtle Electric l
Generating Plant, Unit 1 (Vogtle), Technical Specification (TS) surveillance requi*ement 4.6.1.2d.
The requested change adds a footnote that extends the
- surve'llance interval for the next required Type C leakage test of the auxiliary component cooling water (ACCW) supply and return containment isolation valves'IHV-1974 (and associated check valve 1-1217-04-113),
1HV-1975,1HV-1978, and 1HV-1979, to prior to entry into Mode 4 from the next i
scheduled refueling outage (or the ne' t forced outage requiring entry into x
Mode 5), but no later than November 1,1994.
The mendment provides a i
one-time only extension of the surveillance interval for the shbject valves.
As presently written, TS 4.6.1.2d requires that 10 CFR Part 50, Apperidix J.
4 Section III.D.3, Type B and C tests for the subject valves be conducted at intervals no greater than 24 months.
Also in the September 30, 1993, letter, the licensee requested an exemption from the schedule requirements of Section III.D.3 to Appendix J of 10 CFR Part 50 for the auxiliary component cooling water supply and return containment isolation valves. _The regulation requires that Type B and C local leak rate tests be conducted at intervals no greater than 24 months.
By letter dated October 18, 1993, the Commission issued an environmental assessment which determined that the proposed change does not alter any initial conditions assumed for the design basis accidents previously evaluated nor change operation of-safety systems utilized to mitigate the design basis accidents, and that there are no significant environmental effects that would result from the proposed actions.
By letter dated October 26, 1993, the Commission granted the requested schedule exemption until prior to entry into Mode 4 following the next scheduled refueling outage (or the next forced outage requiring entry into Mode 5), but no later than November 1, 1994.
i-l In February 1992, the licensee prepared and implemented Licensing Document i
Change Request (LDCR) FS92-007 under the provisions of 10 CFR 50.59 and in i
accordance with Vogtle TS 6.4.1.6.
The LDCR revised Table 6.2.4-1 of the l
Vogtle Final Safety Analysis Report (FSAR), in part, with respect to the ACCW l
supply and return containment isolation valves.
Prior to the change, Table 6.2.4-1 stated that these valves were subject to 10 CFR Part 50, Appendix J, i
Section III.D,3, Type C leakage testing requirements, and that.they were l
normally open during operation but closed under post-accident conditions.
i However, as noted in footnote "g" to Table 6.2.4-1, ACCW flow should be l
maintained to the reactor coolant pumps (RCPs) under most post-accident i
conditions, if possible. Therefore, the LDCR changed the leakage testing requirements from Type C to Type A and changed the post-accident position of j
the valves to "open."
In addition, the associated penetrations were added to i
FSAR Table 6.2.6-1 as penetrations that are not vented or drained during Type A testing. As a result of this LDCR, these valves were not Type C tested during the Vogtle Unit I spring 1993 refueling outage, although they had been j
tested during previous outages on both units.
l The licensee's basis for the LDCR was that the subject valves do not receive a containment isolation signal (they are reente manually operated), and the associated penetrations are needed to maintain coeling water to the RCPs. The l
licensee thought that the ACCW was a closed iy:t:;a because it does not communicate directly with the containment itmosphere or primary coolant.
g Thus, when approving the LDCR, the licensee had concluded that Type A testing was sufficient for these penctrations.
l However, during a recent document review, the licensee discovered that the 1
safety evaluation for the LDCR was flawed. The evaluation failed to consider i
that, while the ACCW system is seismic category 1 and the hard piping is j
fabricated of ASME Section III, Class 3 materials, the system had been i
installed in accordance with ANSI B31.1 and no N-stamp was affixed.
In i
addition, some components, such as motor coolers and flexible piping, are not l
composed of Class 3 materials. Therefore, the ACCW system does not meet the J'
ANSI standard criteria for a closed system. Consequently, the supply and return isolation valves must be considered to perfom an isolation function and should be subject to Type C testing.
2.0 EVALUATION The subject valves have been Type C tested during all previous refueling outages with the exception of the Unit I spring 1993 outage. The licensee reviewed the-maintenance work order (MWO) history of the ACCW containscat isolation vafves.
This review found MW0s for seat leakage, pe king leaks, flange leaks, preventive maintenance, and several inspections, but found no "as found" Type C local leak rate test (LLRT) failures after the initial entry into Mode 4 on either Vogtle unit.
The licensee also reviewed the LLRT history of the valves after initial Mode 4 entry and found this history to demonstrate the reliability and low leakage trends of these valves.
Listed below are the maximum values, taken from six refueling outages between the two units, for both the "as found" and "as left" LLRTs perfomed after initial Mode 4 entry. The below values indicate the
a,ee-
'd,,.
i l
l-l-
l l 5 The licensee also reviewed the LLRT history of the valves after initial Mode 4 i
entry and found this history to demonstrate the reliability and low leakage trends of these valves.
Listed below are the maximum values, taken from six refueling outages between the two units, for both the "as found' and "as left" LLRTs perfomed after initial Mode 4 entry. The below values indicate the j
" worst case" leakage.
Penetration 28 is the ACCW supply line and penetration 29 is the ACCW return line.
l i
PENETRATION 28 PENETRATION 29 f
MAXIMUM LEAKAGES MAXIltBI LEAKAGES i
1HV-1978 = 20.5 seca 1HV-1974 - 152 secm*
l IHV-1979 = 40.4 seca 1HV-1975 - 62.0 seca i
2HV-1978 = 49.2 seca 2HV-1974 = 99.6 sccm*
f 2HV-1979 - 90.6 seca 2HV-1975 = 136.3 sece j
- Includes leakage through associated check valve 1-1217-U4-113 k
The Vogtle Inservice Inspection Program currently specifies a maximus i
allowable leakage of 1000 seca for each butterfly valve and 1500 seca for the l
check valve. The leakage limit for the combination of valve INV-1974 and l
check valve 1-1217-04-113 would be 2500 seca. These limits were not based on Appendix J requirements, but were established based on the low leakage history j
of these valves and define the point at which repair would be required. The Appendix J 1eakage limit for all penetrations subject to Type B and C testing l
(0.6L at Vogtle is 228,273 seca. The current total for Type B and C test leek:g,)e at Vogtle, as of September 10, 1993, is 14,398.8 seca. As of the last LLRT, the leakags for each of these four valves was as follows: 1HV-1974 - 152 seca (this includes leakage put check valve 1-1217-U4-113 in parallel with j
1HV-1974); IHV-1975 - 11.6 seca; IHV-1976 - 9.3 seca; and 1HV-1979 - 11.4 seca. The test pressure, P,, was 45 psig at the time t*nssa numbers were obtained.
The test pressure has since been reduced to 37 psig in accordance i
with previous license Amendments 63 (Unit 1) and 42 (Unit 2), and the leakage j
wculd be less at this lower pressure.
During the last cutage for Unit 1, the licensee perfomed maintenance on 1HV-1979 that could have'affected its leakage, but perfomed no LLRT since it was not required by the FSAR at the time. The maintehance involved removal of the motor and gearbox and altering the limit switch settings, but no work was done i
that would have affected the valve seat. The standard work practice for setting limit switches on this type of soft-seated butterfly valve following this type of. maintenance is as follows:
first, the valve is manually closed I
using the hand wheel until 0* (fully closed) is reached, and the limit switch i
is set.
Theft, the limit switch is testad by manually operating the valve again.
Finally, the valve is stroked using the motor until the limit switch actuates. At this point, the hand wheel is used to ensure that the valve is j
seated properly after the limit switch actuates. As a reference point, in the i
spring of 1992 this type of wcrh was perfomed on Unit 2 valve 2HV-1978 and pre-maintenance and post-maintenance LLRTs were perfomed. The pre-and post-i maintenance leakage was well within the leakage limits for this valve.
4
i
[,*
4 i
i*
i j,
The probability of containment isolation failure following a core damage accident is modeled in the Vogtle individual plant examination (IPE) of severe j
accidents. The probability of an event that leads to core damage and a failure of the ACCW piping inside containment with a failure to isolate containment was not considered to be credible by the licensee.
In order to model a more con'servative scenario of containment isolation failure than was considered in the base case Vogtle IPE, the licensee assumed that the i
occurrence of any core damage scenario would cause a break in the ACCW flow l
path and that the operator would be required to isolate the ACCW system for successful containment isolation.
Based on a Type C test interval of 2 years, i
thefrequencyofcorepamagewithcontainmentisolation.failurewasfoundto be on the order of-10' per reactor year. The staff concurs that extending the required Type C test interval for these valves, as proposed, has a j
negligible impact on that probability.
The ACCW system is seismic category 1, and the hard piping is fabricated of l
'ASME Section III, Class 3 materials.
Some components, such as motor coolers j
and flexible piping, are not fabricated of Class 3 materials. The licensee l
concluded that, even though the ACCW does not meet the ANSI standard criteria for a closed system, it can be considered to be highly reliable and that these is reasonable assurance that for most events its integrity would be i
maintained.
The staff concurs with this conclusion.
l The NRC staff also finds that the 2-year interval requirement for Type B and C components is sufficient for timely detection of significant deterioration while, at the same time permitting the tests to be performed during plant outages.
Leak rate testing of the penetrations during shutdowns ~is preferable l
because of the lower radiation exposure to plant personnel.
Some penetrations l
cannot be tested at power.
For those penetrations that cannot be tested during power operation or for which testing at power is inadvisable, the i
increase in confidence of containment leaktight integrity following a l
successful test is slight and does not justify a plant shutdown specifically
(
to perform the tests within.the 2-year time period, considering the factors discussed cbove.
3 Based on the above evaluation, the NRC staff finds the requested one-time only i
change to TS surveillance requirement 4.6.1.2d is acceptable.
As provided in the footnote, the surveillance interval for the next required Type C leakage test of the ACCW supply and return containment isolation valves 1HV-1974 (and i
associated check valve 1-1217-U4-113), 1HV-1975, 1HV-1978, and lHV-1979, is 4
extended for-Vogtle Unit 1 to " prior to entry into Mode 4 following the next scheJuled ref~ueling outage (or the next forced outage requiring entry into I
Mode 5), but~no later than November 1,1994."
3.0 EXIGENT CIRCUMSTANCES
The licensee requested in their application dated September 30, 1993, that the proposed amendment be processed as involving exigent circumstances.
The Commission's regulation, 10 CFR 50.91(a)(6), states that an exigent circumstance exists where the Commission finds that the licensee and the Commission must act quickly and that time does not permit the Commission to f
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publish a Fadaral Reaister notice allowing 30 days for prior public comment and it also determines that the amendment involves no significant hazards considerations. The licensee proposed that the license amendment involves l
exigent circumstances in that the 24-month testing interval, as specified in 10 CFR Part 50, Appendix J, Section III.D.3, and Technical Specification i
4.6.1.2d, for the Vogtle Unit 1 ACCW supply and return containment isolation valves, will expire on October 28, 1993, thus, requiring the facility to be shut down and placed into Mode 5 prior to October 28, 1993, in order to perform the Type C test on the subject valves.
The NRC staff has reviewed the licensee's proposed amendment and finds that the licensee did not fail to use its best efforts to make a timely application and avoid creating the exigent circumstances.
In accordance with 10 CFR 50.91(a)(6)(B), the Commission issued a Federal Reaister notice dated October 12, 1993 (58 FR 52796), which proposed a finding i
of no significant hazards consideration, provided notice of an opportunity for hearing, and allowed at least two weeks from the date of the notice for prior j
public comment.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERIINATION The Commission's regulations in 10 CFR 50.92 provide that the Commission may make a final determination that a license amendment involves no significant l
hazards considerations if operation of the facility in accordance with the amendment would not:
a.
Involve a significant increase in the probability or consequences of an j
accident previously evaluated.
l The proposed change is a one-time only extension of the Type C leakage i
test interval for the Unit 1 ACCW supply and return containment isolation valves. As such, it has no effect on the probability of any accident previously evaluated. Furthemore, based on the past leakage test history of these valves, there is reasonable assurance that i
extending the test interval to no later than November 1,1994, (or the j
next forced outage that requires entry into Mode 5) will not adversely affect the ability of these valves to perfom their isolation function.
{
Therefore, the proposed change will not involve a significant increase l
in the consequences of any accident previously evaluated.
b.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
l The proposed change does not change the configuration or method of operation of any plant equipment, and no new failure modes have been defined for any plant system or component.
Furthermore, no new limiting failure has been identified as a result of the proposed change.
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Involve a significant reduction in a margin of safety.
c.
4 There continues to be reasonable assurance that the sub, ject valves will remain capable of perfoming their isolation function.
In addition, the I
proposed change avoids a plant shutdown solely for the purpose of l
performing Type C testing of these valves.
i Based on the above, the Commission has made a final determination that the l
proposed amendment involves no significant hazards considerations.
5.0 STATE CONSULTATION
j.
l In accordance with the Commission's regulations, the Georgia State official
~
was notified of the proposed issuance of the amendment. The State official l
had no comments.
l 6.0 ENVIR0fetENTAL CONSIDERATION 4
The amendment changes surveillance requirements. The NRC staff has detemined i
that the amendment involves no significant increase in the amounts, and no j
significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative i
5 occupational radiation exposure. The Commission has made a final l
determination that the amendment involves no significant hazards I
considerations. Accordingly, the amendment meets the eligibility criteria for i
categorical exclusion set forth in 10 CFR 51.22(c)(g).
Pursuant to 10 CFR Sl.22(b) no environmental impact statsmant or environmental assessment need be I
prepared in connection with the issuance of the amendment.
7!0 "' CONCLUSION The Commission has concluded, based on the considerations discussed above, j
that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such i
activities will be conducted in compliance with the Commission's regulations, j
and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
i l
Principal Contributor:
C. E. Carpenter, Jr.
I Date:
October 27, 1993 l
O e
e
a i
c NOV 171993 Docket Nos. 50-424, 50-425 License Nos.lNPF-68, NPF-81 I
Georgia Power Company.
j
. ATTN: Mr. C. K. McCoy.
Vice President Vogtle Electric Generating Plant 3-
+
P. O. Box 1295 Birmingham, AL 35201 y
Gentlemen:
SUBJECT:
(NRC INSPECTION REPORT NOS. 50-424/93-23 AND 50-425/93-23)
This refers to the inspection conducted by B. Bonser of this office on i.
September 19 - October 23, 1993. The inspection included a review of activities authorized for your Vogtle facility. At the conclusion of the inspection,-the findings were discussed with those members of your staff
. identified in the enclosed report.
i 3
i Areas examined during the inspection are; identified in the report. Within these areas, the inspection' consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of f
activities in progress.
i Based ~on the results of this. inspection, certain of your activities appeared l
to be in violation of NRC requirements, as specified in the enclosed Notice of.
l
-Violation (Notice).
You are required to respond to this letter and should follow the instructions 4.
specified in the enclosed Notice when preparing your response.
In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response'to this
' Notice, including your proposed corrective actions and the results of future f
inspections, the NRC will determine whether further NRC enforcement action is i
necessary to ensure. compliance with NRC regulatory requirements, u
f The responses directed by this letter and the enclosed Notice are not subject
-to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No.96-511.
.In accordance with 10 CFR 2.790(a), a copy of this letter and its enciosure j
will be placed in the NRC Public Document Room.
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i5 b.
Unit 1 Summary s-The unit began the period operating at 100% power and operated at full power throughout the inspection period.
c.
Unit 2 Summary The unit began the period operating in Mode 6.
Reactor defueling was completed on September 24. Reactor refueling began and Mode 6 was entered on October 2.
Mode 5 was entered on October 9.
The unit entered Mode 4 and Mode 3 on October 16.
The unit entered Mode 2 and was taken critical on October 18 and then entered Mode 1 on October 20. The unit reached 100% power on October 23.
d.
Containment and Emergency Sump Walkdowns Periodically during the Unit 2 refueling outage the inspectors performed walkdowns of the containment building.
Prior to entering Mode 4, the inspectors performed a containment walkdown to verify that no obvious loose material was present in the containment. The inspectors identified and removed several items including plastic bags and a hard hat. The inspectors also iden-tified several missing or loose fasten?*s on the emergency sump
/'~')
suction inlet strainers and brought these items to the attention
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of licensee management.
TS 4.5.2, ECCS, requires the licensee to perform a visual inspection of containment to verify that no debris is present that l
could be transported to the containment emergency sumps prior to establishing containment integrity and entering Mode 4.
TS 4.5.2 also requires the licensee to perform a visual inspection of the containment emergency sumps and verification that the suction inlet strainers are not restricted by debris.
The inspector verified, from discussions with the Operations manager who performed these containment surveillances, that any loose debris found were removed from the containment and emergency sumps. The inspector also reviewed completed surveillance procedures 14900-2, Containment Exit Inspection, and 14903-2, Containment Emergency Sump Inspection, and verified that the results of the surveillances were adequate and met all acceptance criteria.
e.
ACCW TS Exemption Request On September 27, the licensee requested an exemption and exigent TS amendment on Unit 1.
The one time exemption requested an exemption from 10 CFR 50, Appendix J requirements as they relate to the Unit 1 ACCW supply and return containment isolation valves.
Appendix J requires that type C tests be performed during each reactor shutdown for refueling but in no case at intervals greater than two years. The proposed exemption would allow the required 9l' test interval for valves 1HV-1974 (and associated check valve
~ _ _..
f 4
1-1217-U4-113)',1975,1978,1979 to be extended from 24 months to prior to entry into Mode 4 following the next scheduled refueling outage or forced outage.
The licensee also proposed to amend TS 3
l 4.6.1.2d, Containment Leakage Surveillance Requirements.
The licensee explained the basis for their request to regional and headquarters NRC staff and the process to prepare'an exigent TS amendment was initiated. The inspectors were primarily concerned with the circumstances that led to the request.
i In February 1992, plant management approved an LDCR which revised containment penetration isolation valve information in the Vogtle 4
FSAR with respect to the ACCW supply and return containment isolation valves.
Prior to the change the ACCW valves were subject to 10 CFR 50, Appendix J, Type C leakage testing requirements. The requirements were changed to only type A l
testing.
b The' basis for the 1992 LDCR was that the ACCW valves do not receive a containment isolation. signal (they are remote manually operated valves) and the associated penetrations are considered essential due to the desirability of maintaining cooling water to j
the RCPs under. post-accident conditions.
In addition, it was thought that the ACCW system was a closed system since it does not communicate directly with the containment atmosphere or primary coolant. Type A testing was, therefore, considered sufficient for these penetrations.
The safety evaluation, however, failed to consider that the ACCW system in containment does not meet standards in ANSI N271-1976, Containment Isolation Provisions for Fluid Systems, for closed systems inside containment. Therefore the subject ACCW valves should have been considered to perform an isolation function and 4
l subject to Type C testing requirements.
l Further details describing the errors in the safety evaluation, its safety impact and justification for the exemption and TS j
change are contained in correspondence to the NRC dated September 30, 1993 and in LER 50-424/93-11. Although this error in the safety evaluation appeared to the inspectors to be an isolated case, IFI 424,425/93-23-01, Review Licensee Safety Evaluations, will be opened to monitor licensee safety evaluations for similar
- mistakes, i
i f.
Notification of Unusual Event Due To Onsite Toxic Release At 4:15 am on the morning of October 7, 1993, with Unit 2 in Mode 6, chemistry department personnel injected about 30 gallons of 2
hydrazine and about 15 gallons of ammonia into the Unit 2 feedwater system. These chemicals are used to prevent corrosion in the steam generators during unit shutdown.
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