ML20058M236

Application for Amend to License NPF-68,adding Footnote to TS 4.6.1.2d,extending Type C Test Interval for Valves HV-1974 (& Associated Check Valve 1-1217-U4-113),1975,1978 & 1979,consistent w/10CFR50,App J Exemption Request

ML20058M236
Person / Time
Site:	Vogtle
Issue date:	09/30/1993
From:	Mccoy C GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20058M239	List: ML20058M236 ML20058M240
References
LCV-0163, LCV-163, TAC-M87782, TAC-M87783, NUDOCS 9310040335
Download: ML20058M236 (16)

Text

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Geteg-a Power Company

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- D rm pm A anma 2!:Q1 Mohere MS E77 7:22 K.a c.x.uccoy Georgia Power

[.g((j N '" September 30,1993 mmawy3.- ,wy LCV-0163 Docket No. 50-424 TAC No. M87782 M87783 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQ'UEST AND REQUEST TO REVISE TECHNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES In accordance with the provisions of 10 CFR 50.12, Georgia Power Company (GPC) requests a one-time exemption from the requirements of 10 CFR 50, Appendix J, Section III.D.3 as they relate to the Unit I auxiliary component cooling water (ACCW) supply and return containment isolation valves.Section III.D.3 of Appendix J requires that Type C tests be performed during each reactor shutdown for refueling but in no case at intervals greater than 2 years. The proposed exemption would allow the required test interval for valves HV-1974 (and associated check valve 1-1217-U4-113),1975,1978, and 1979 to be entended from 24 months 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 Novembec L 1994. In addition, in accordance with the provisions of 10 CFR 50.91, GPC proposes to amend the Vogtle Electric Generating Plant (VEGP) Unit 1 Technical Specifications (TS), Appendix A to Operating License NPF-68 as they relate to the subject valves. The proposed amendment would affect TS 4.6.1.2d by adding a footnote that would extend the Type C test interval for the subject valves consistent with the exemption request. As with the exemption regbest, the proposed amendment would be a one-time only extension of the Type C test interval for these valves.

Georgia Power Company submits that the proposed exemption and license amendment involve exigent circumstances and respectfully requests that the NR.C process the proposed amendment under the provisions of the regulations applicable to an exigent request. These circumstances are as follows. In February 1992, a Licensing Document Change Request (LDCR) that revised table 9310040335 930930 I PDR ADOCK 05000424 '

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. U. S. Nuclear Regulatory Commission LCV-0163 Pag 6 2 6.2.4-1 of the VEGP Final Safety Analysis Report (FSAR) was processed under the provisions of l 10 CFR 50.59. This LDCR (FS92-007) revised the subject table to show the post-accident position of the ACCW supply and return containment isolation valves to be open rather than closed, and the leakage testing requirements were revised from Type C to Type A as defined in 10 CFR 50, Appendix J. As a result, the subject valves were not Type C tested during the last Unit I refueling outage in the Spring of 1993. Recently, during a followup review to revise associated documentation to correspond to LDCR FS92-007, it was determined that the Type C test requirement was inappropriately revised to Type A. The required 24-month interval specified in ,

10 CFR 50, Appendix J and TS 4.6.1.2d will expire on October 28,1993, for the subject valves, and the unit must be in Mode 5 for the test i ng to be performed. Therefore, unless the proposed exemption request and license amendment can be processed on an exigent basis, Unit I will be forced into Mode 5 in suflicient time prior to October 28,1993, so that the required testing can be performed. Unit 2 is presently in a refueling outage, and the corresponding valves have been  ;

tested iuring this outage. Therefore, the proposed exemption and license amendment are not applicable to Unit 2.

E The proposed exemption and its basis is provided in enclosure 1. The proposed license amendment and its basis is provided in enclosure 2, including a detailed discussion of the  ;

circumstances discussed above that have necessitated this request. An evaluation pursuant to 10 CFR 50.92 showing that the proposed changes do not involve significant hazards considerations is provided as enclosure 3, an environmental assessment is provided as enclosure 4, and the marked-up TS page is provided as enclosure 5. In accordance with 10 CFR 50.91, the designated state official will be sent a copy of this letter and all enclosures. ,

The Plant Review Board has reviewed and recommended approval of this request. In addition, GPC has determined that the proposed exemption and license amendment will not have a i significant effect on the environment.

Mr. C. K. McCoy states that he is a Vice President of Georgia Power Company and is authorized  :

to execute this oath on behalf of Georgia Power Company s.nd that, to the best of his knowledge and belief, the facts set forth in this letter and enclosures are true.

GEORGIA POWER COMPANY BY: [ /

" C. K. McCoy /

Sworn to and subscribed before me this 3 bay of Lytlled,1993.

Vamf.&A Notary'Public nyccmsxnmuua um l

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" U. S. Nuclear Regulatory Commission LC%0163 Pag 6 3 CKM/NJS

Enclosures:

1. Exemption Request

2. Proposed TS Amendment 3.10 CFR 50.92 Evaluation

4. Environmental Assessment

5. Marked-Up Page xc: Georgia Power Company  :

Mr. J. B. Beasley, Jr. .

Mr. M. Sheibani NORMS U. S. Nuclear Regulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR ,

Mr. B. R. Bonser, Senior Resident Inspector, Vogtle t

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State of Georgia Mr. J. D. Tanner, Commissioner, Department of Natural Resources i

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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  !

BASIS FOR PROPOSED EXEMPTION  ;

Discussion  ;

Section III.D.3 of Appendix J to 10 CFR 50 requires that Type C tests be performed at each l 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 interval as it applies to the auxiliary i componeut cooling water (ACCW) supply and return containment isolation valves HV-1974 (and i associated check valve 1-1217-U4-113),1975,1978, and 1979. The current 2-year interval is 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 forced  !

outage requiring entry into Mode 5), but no later than November 1,1994.  !

Background

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 Review  ;

Board in accordance with section 6.4.1.6 of the Vogtle Electric Generating Plant (VEGP) TS. '

On February 21,1992, LDCR FS92-007 was approved by plant management. The subject i LDCR revised table 6.2.4-1 of the VEGP Final Safety Analysis Report (FSAR), in part, with respect to the ACCW supply and return containment isolation valves. Prior to the change, table 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. However, 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 l scenarios, ACCW flow would be maintained to support operation of the RCPs. Therefore, the LDCR revised the leakage testing requirements to Type A and stated that the post-accident 7 position was open. In addition, the associated penetrations were added to FSAR table 6.2.6-1 as .

penetrations that are not vented or drained during Type A testing. As a result, these valves were  ;

not Type C tested during the Spring 1993 refueling outage. They have, however, been tested during previous outages on both units.

The basis for the LDCR was that the subject valves do not receive a containment isolation signal ,

(they are remote manually operated), and the associated penetrations are considered essential due  ;

to the desirability of maintaining cooling water to the RCPs 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 Type A testing was suflicient for these penetrations.

However, the safety evaluation failed to consider that, while the ACCW system is seismic category 1 and the hard piping is fabricated of ASME Section 111, Class 3 materials, the ,

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. ENCLOSURE 1 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY .AND RETURN CONTAINhENT ISOLATION VALVES BASIS FOR PROPOSED EXEMPTION  !

installation was in accordance ANSI B31.1 and an N-stamp was not affixed. In addition, there are some components such as motor coolers and flexible piping that are not of Class 3 materials.

Therefore, the ACCW system does not meet the ANSI standard criteria for a closed system. In consequence, the supply and return isolation valves must be considered to perform an isolation - l function and should be subject to Type C testing. This was discovered recently during a i subsequent review for the purpose of making corresponding changes to associated 1 documentation.  !

Technical Basis 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 (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 maintenance, and _;

several inspections, but there were no "as found" Type C local leak rate test (LLRT) failures after i the initial entry into Mode 4 on both units. An MWO was written against 2HV-1979 for high seat l leakage during the period between the preoperational LLRT and initial Mode 4 entry, however, i this condition has not reoccurred.

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In addition, a review of the valves' LLRT history after initial Mode 4 entry demonstrates the 3

reliability and low leakage trends of these valves. Listed below are the maximum values for both the "as found" and "as left" LLRTs performed after initial Mode 4 entry. This data was taken 'i from six refueling outages between the two units. Note that penetration 28 is the ACCW supply -

line and penetration 29 is the ACCW return line.

PENETRATION 28 PENETRATION 29 '[

MAXIMUM LEAKAGES MAXIMUM LEAKAGES ,

i 1HV-1978 = 20.5 seem 1HV-1974 = 152 scens' IHV-1979 = 40.4 sccm 1HV-1975 = 62.0 secw 2HV-1978 = 49.2 sccm 2HV-1974 = 99.6 secs

• 2HV-1979 = 90.6 scem 2HV-1975 = 136.3 se:m i,

• Includes leakage through associated 113 check valve  !

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 ofvalve HV-1974 and check valve 113 would be 2500 sccm.) These limits were not based on Appendix J, but were established based on the low leakage history of these valves and El-2 i

ENCLOSURE 1 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST ,

REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES .

BASIS FOR PROPOSED EXEMPTION 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 (0.6La) at VEGP is 228,273 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 Type C LLRT, the leakage for each of these four valves was as follows: HV-1974 - 152 sccm (this includes leakage past check valve 1-1217-U4-113 in parallel ,

with HV-1974); HV-1975 - 11.6 sccm; HV-1978 - 9.3 sccm; and HV-1979 - 11.4 sccm. In addition, it should be noted that the test pressure, Pa, was 45 psig at the time these numbers were obtained. The test pressure has since been revised to 37 psig, so it would be reasoncble to assume that the leakage would be less at the lower pressure.

During the last outage for Unit 1, maintenance was performed on HV-1979 that could have -

affected its leakage, and no LLRT was performed since it was not required by the FSAR at the time. The motor and gearbox were removed and the limit switch settings were altered, but no work was done 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 using the hand wheel until 00 (closed) is reached, and 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 hand wheel is used to ensure that the valve is seated properly after the limit switch actuates. As a i reference point, in the Spring of 1992 this type of work was performed on Unit 2 valve HV-1978 and premaintenance and post-maintenance LLRTs were performed. The premaintenance leakage I and the post-maintenance leakage was well within the leakage limits for this valve. Therefore, GPC is confident that Unit I valve HV-1979 remains leak-tight. l l

i 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.

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 intervai for these valves as proposed has a negligible impact on that probability.

Finally, the ACCW system is seismic category 1, and the hard piping is fabricated of ASME i Section III, Class 3 materials. (There are some components such as motor coolers and flexible l piping that are not fabricated of Class 3 materials.) Therefore, even though the ACCW does not l El-3

, 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 )

l BASIS FOR PROPOSED EXEMPTION meet the ANSI standard criteria for a closed system, it can be considered to be highly reliable and i there is reasonable assurance that for most events its integrity would be maintained. >

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 CFR 50.12(a)(2) are present. ,

1. The requested exemption is authorized by law.

i No law is known to exist that would preclude the activities covered by this exemption request. _

Therefore, the Commission is authorized to grant this exemption.

2. The requested exemption does not present an undue risk to the public health and safety.

l 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 ,

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.  !

3. The requested exemption will not endanger the common defense and security.

The common defense and security are not an issue in this exemption request.

4. Special circumstances are present which necessitate the request for a one-time exemption to the regulations of 10 CFR 50, Appendix J, Section Ill.D.3.

Compliance with the regulation would result in undue hardship or other costs that are i 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 costs in the form oflost revenues.

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ENCLOSURE 1 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES BASIS FOR PROPOSED EXEMPTION The exemption would provide only temporary relief from the applicable regulation, and a good faith effort has been made to comply with the regulations. All other penetrations subject to Type C testing have been faithfully tested. When the discovery was made that the LDCR inappropriately removed the Type C test requirements from the subject valves, GPC took prompt action to investigate and correct this condition.

Safety Imoact Georgia Power Company has reviewed the proposed exemption 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 the Unit 1 ACCW supply and return containment isolation valves. As such, it has no effect on the probability of any accident previously evaluated.

Funbermore, based on the past leakage test history of these valves, there is reasonable 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 abiF:y of these valves to perfonn their isolation function. Therefore, the proposed change will not involve a 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 from any accident previously evaluated. 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.

3. The proposed change does not involve a significant reduction in a margin of safety. There continues to be reasonable assurance that the subject valves will remain capable of performing their isolation function. In addition, the proposed change avoids a plant shutdown solely for the purpose of performing Type C testing of these valves.

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. 1 ENCLOSURE 2 l

VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECIINICAL SPECIFICATIONS l REGARDING AUXILIARY COMPONENT COOLING WATER )

SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES l l

BASIS FOR PROPOSED CHANGE Proposed Change The proposed amendment would revise the existing Technical Specification (TS) surveillance requirement 4.6.1.2d for Unit I by adding a footnote that would extend the surveillance interval for the next required Type C leakage test of the auxiliary component cooling water (ACCW) supply and return containment isolation valves HV-1974 (and associated check valve 1-1217-U4-113),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, .

1994. The proposed amendment would be a one-time only extension of the surveillance interval for the subject valves. The current surveillance interval expires October 28,1993.

Circumstances Surrounding the Proposed Change 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 Review Board in accordance with section 6.4.1.6 of the Vogtle Electric Generating Plant (VEGP) TS.

On February 21,1992, LDCR FS92-007 was approved by plant management. The subject LDCR revised table 6.2.4-1 of the VEGP Final Safety Analysis Report (FSAR), in part, with  ;

respect to the ACCW supply and return containment isolation valves. Prior to the change, table 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. However, 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 suppon operation of the RCPs. Therefore, 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 table 6.2.6-1 as penetrations that are not vented or drained during Type A testing. As a result, these valves were not Type C tested during the Spring 1993 refueling outage. They have, however, been tested during previous outages on both units.

The basis for the LDCR was that the subject valves do not receive a containment isolation signal (they are remote manually operated), and the associated penetrations are considered essential due to the desirability of maintaining cooling water to the RCPs 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 Type A testing was sufficient for these penetrations.  ;

1 However, the safety evaluation failed to consider that, while the ACCW system is seismic category 1 and the hard piping is fabricated of ASME Section Ill, Class 3 materials, the E2-1

ENCLOSURE 2 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL $PECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES l BASIS FOR PROPOSED CHANGE 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 materials.

Therefore, the ACCW sy., n does not meet the ANSI standard criteria for a closed system. In consequence, the supply and return isolation valves must be considered to perform an isolation function and should be subject to Type C testing. This was discovered recently during a i subsequent review for the purpose of making corresponding changes to associated documentation. -

Georgia Power Company (GPC) requests that the proposed change be processed on an exigent basis. The required 24-month surveillance interval expires on Unit 1 on October 28,1993. In the absence of the proposed relief, Unit I would have to be placed in Mode 5 sufliciently prior to October 28,1993, so that the required testing could be performed.

Basis 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 (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 maintenance, and  !

several inspections, but there were no "as found" Type C local leak rate test (LLRT) failures afler

he initial entry into Mode 4 on both units. An MWO was written against 2HV-1979 for high seat leakage during the period between the preoperational LLRT and initial Mode 4 entry, however, this condition has not reoccurred.

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, In addition, a review of the valves' LLRT history after initial Mode 4 entry demonstrates the i reliability and low leakage trends of these valves. Listed below are the maximum values for both ,

the "as found" and "as lefl" LLRTs performed afler initial Mode 4 entry. This data was taken from six refueling outages between the two units. Note that penetration 28 is the ACCW supply '

line and penetration 29 is the ACCW return line.

1 PENETRATION 28 PENETRATION 29 ,

MAXIMUM LEAKAGES MAXIMUM LEAKAGES 1HV-1978 = 20.5 scem 1HV-1974 = 152 sccm*

1HV-1979 = 40.4 sccm 1HV-1975 = 62.0 sccm 2HV-1978 = 49.2 sccm 2HV-1974 = 99.6 secm* ,

2HV-1979 = 90.6 sccm 2HV-1975 = 136.3 sccm  ;

i Includes leakage through associated 113 check valve E2-2 j l

, ENCLOSURE 2 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES BASIS FOR PROPOSED CHANGE The Inservice Inspection Program currently specifies a maximum allowable leakage of 1000 scem for each butterfly valve and 1500 sccm for the check valve. (The leakage limit for the combination of valve HV-1974 and check valve 113 would be 2500 secm.) These limits were not based on Appendix J, but were established based on the low leakage history of these valves and 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 (0.6La) at VEGP is 228,273 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:

HV-1974 - 152 secm (this includes leakage past check valve 1-1217-U4-113 in parallel with HV-1974); HV-1975 - 11.6 sccm; HV-1978 - 9.3 scem; and HV-1979 - 11.4 secm. In addition, it should be noted that the test pressure, Pa, was 45 psig at the time these numbers were obtained.

The test pressure has since been revised to 37 psig, so it would be reasonable to assume that the leakage would be less at the lower pressure.

During the last outage for Unit 1, maintenance was performed on HV-1979 that could have affected its leakage, and no LLRT was performed since it was not required by the FS AR at the time. The motor and gearbox were removed and the limit switch settings were altered, but no work was done 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 using the hand wheel until 00 (closed) is reached, and 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 hand 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-1978 and premaintenance and post-maintenance LLRTs were performed. The premaintenance leakage and the post-maintenance leakage was well within the leakage limits for this valve. Therefore, 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 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 r em for successful containment isolation.

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 probability.

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ENCLOSURE 2 (CONTINUED) i VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS  ;

REGARDING AUXILIARY COMPONENT COOLING WATER '

SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES BASIS FOR PROPOSED CHANGE Finally, the ACCW system is seismic category 1, and the hard piping is fabricated of ASME Section III, Class 3 materials. (There are some components such as motor coolers and flexible  ;

piping that are not fabricated of Class 3 materials.) Therefore, even though the ACCW does not 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.

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 leakage testing on these  :

valves is preferable to the alternative of shutting Unit I down solely for the purpose of performing Type C testing of these valves.

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.l ENCLOSURE 3 VOGTLE ELECTRIC GENERATING PLANT I

REQUEST TO REVISE TECHNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES l 10 CFR 50.92 EVALUATION Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the  !

proposed changes and has made the following determination: ,

1. The proposed changes do not involve a significant increase in the probability or consequences i of an accident previously evaluated. The proposed change is a one-time only extension of the Type C leakage 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. l Furthermore, based on the past leakage test history of these valves, there is reasonable 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 l valves to perform their isolation function. Therefore, the proposed change will not involve a 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 from any accident previously evaluated. 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.

3. The proposed change does not involve a significant reduction in a margin of safety. There continues to be reasonable assurance that the subject valves will remain capable ofperforming ,

their isolation function. In addition, the proposed change avoids a plant shutdown solely for  ;

the purpose of performing Type C testing of these valves.  ;

Conclusion Based on the preceding analysis, Georgia Power Company has determined that the proposed l change will not significantly increase the probability or consequences of any accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety. Therefore, the proposed changes meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.  ;

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ENCLOSURE 4 VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST AND REQUEST TO REVISE TECHNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES ENVIRONMENTAL ASSESSMENT Identification ofProposed Action The requested exemptioi and license amendment would grant temporary relief from the 2 " ear .

schedular requirement associated with Type C periodic local leak rate tests (LLRTs). The 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 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 Pro _ posed Action 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 leakage test requirements set forth in 10 CFR Part 50, Appendix J. Appendix J to 10 CFR Part 50, Section III.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 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 costs in the form oflost revenues. Additionally, shutdown solely for the purpose of testing the subject valves would result in an increase in occupational radiation exposure and an ,

additional transient on the plant. ,

EnvironmentalImpact ofProposed 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 components under Appendix J to 10 CFR Part 50 is intended to demonstrate that containment leakage from these components is within defined acceptable limits. These limits provide ,

infonnation used to calculate the maximum radiological consequences of a design basis accident.

Appendix J limits the combined leak rate for all penetrations and valves subject to Type B and C '

test to less than 0.6 times the maximum allowable containment leakage rate with the containment pressurized to its design limit (commonly termed "0.6 La ")-

The subject valves, associated with the auxiliary component cooling water (ACCW) system, have been Type C tested during all previous refueling outages with the exception of the Unit 1 Spring E4-1

l ENCLOSURE 4 (CONTINUED) l VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST AND REQUEST TO REVISE TECHNICAL SPECIFICATIOM REGARDING AUXILIARY COMPONE5 LC-JLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES ENVIRONMENTAL ASSESSMENT 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 seat leakage, packing leak, flange leaks, preventive maintenance, and several inspections, but their were no "as found" Type C LLRT failures after the initial entry into Mode 4 on both Units. An  ;

MWO was written against 2HV-1979 for high seat leakage during the period between the preoperational LLRT and initial Mode 4 entry; however, this condition has not reoccurred. A review of the LLRT history for the subject valves afler initial Mode 4 entry demonstrates the reliability and low leakage trends of these valves.

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.

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 probability.

Therefore, radiological releases will not differ from those determined previously, and the proposed exemption does not otherwise affect facility radiological effluent or occupational exposures. With regard to potential nonradiological impacts, the proposed action does not affect plant nonradiological efIluents and has no other nonradiological environmental impact. Therefore, Georgia Power Company (GPC) concludes there are no measurable radiological or ,

nonradiological emironmental impacts associated with the requested exemption. l l

Alternatives to Proposed Action Since GPC has concluded there is no measurable environmental impact associated with the requested action, any alternative with equal or greater emironmental impact 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 previo. the Final Emironmental Statement for Vogtle Electric Generating Plant, Units 1 . ..

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, ENCLOSURE 4 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50 APPENDIX J EXEMPTION REQUEST AND REQUEST TO REVISE TECHNICAL SPECIFICATIONS REGARDING AUXILIARY COMPONENT COOLING WATER SUPPLY AND RETURN CONTAINMENT ISOLATION VALVES l ENVIRONMENTAL ASSESSMENT l Conclusion Based on the foregoing environmental assessment, GPC concludes that the proposed action will not have a significant effect on the quality of the human environment.

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