ML20083M055
| ML20083M055 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/12/1995 |
| From: | Mccoy C GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20083M059 | List: |
| References | |
| LCV-0610, LCV-610, NUDOCS 9505190094 | |
| Download: ML20083M055 (14) | |
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1 Georgia Power Company
-*.,- , ,, i 140 invemess Center Parkway Pot.t Office Box 1295-
A Birmingham, Alabama 35201 Telephone 205 877 7122 l h
- c. x. uccoy .
Georgia Power ;
- &Tj " ' tre sourtiem eiectrc system i May 12, 1995
. LCV-0610
- Docket No. 50-424 50-425 :
l U. S. Nuclear Regulatory Commission l ATTN: Document Control Desk ,
Washington, D. C. 20555 VOGTLE ELECTRIC GENERATING PLANT REQUEST FOR EXEMPTION FROM 10 CFR 50, APPENDIX J '
TYPE A TEST SCHEDULE AND TECHNICAL SPECIFICATIONS CHANGE TO SUPPORT THE REOUESTED EXEMPTION Gentlemen:
Pursuant to 10 CFR 50.12(a) and 10 CFR 50.90, Georgia Power Company (GPC) hereby requests a one-time exemption from the requirement of Section III.D.l(a) of 10 CFR 50, l
Appendix J for Vogtle Electric Generating Plant (VEGP), Unit I and a permanent change to ,
the VEGP Technical Specifications that will support this and any other future Appendix J-exemptions that may be approved by the NRC.Section III.D.l(a) of 10 CFR 50, Appendix J ;
requires that "After the preoperational leakage rate tests, a set of three Type A tests shall be l performed, at approximately equal intervals during each 10-year service period. The third test i of each set shall be conducted when the plant is shutdown for the 10-year plant inservice i inspections." Specifically, GPC requests the following: ;
- 1. Exemption from the 10 CFR 50, Appendix J schedule requirement for VEGP, Unit 1 to l provide a one-time interval extension for the Type A test from the currently scheduled 36 -
months to approximately 54 months. The Unit I test is currently scheduled to be performed <
during the IR6 refueling outage in February and March of 1996. The test would be ,
rescheduled in accordance with the new criteria (Option B) of the anticipated 10 CFR 50, Appendix J rule change. j
- 2. An amendment to License No.- NPF-68 in the form of a Plant Technical Specifications !
change to address exemptions to Appendix J. The proposed Technical Specifications l change would insert the words "Except as modified by NRC approved exemptions" at the .
beginning of the first sentence of Surveillance Requirement 4.6.1.2.
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i GeorgiaPowerA U. S. Nuclear Regulatory Commission Page 2 A detailed justification for the proposed exemption and Technical Specifications change is provided in Enclosure 1 to this letter. That enclosure contains information that, in accordance with 10 CFR 50.12(a), demonstrates that the requested exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. Funhermore, it is shown that special circumstances are present which
- warrant issuance of the requested exemption. Enclosure 2 presents the bases for GPC's determination that the proposed Technical Specifications change does not involve a significant l hazards consideration, and Enclosure 3 provides page change instructions for incorpaating the proposed Technical Specifications change. A marked-up copy of the current Technical j*
Specifications page follows Enclosure 3.
The exemption request and proposed Technical Specifications change will not change the types of any effluents that may be released offsite, nor create a significant increase in individual or cumulative occupational radiation exposure. Therefore, the quality of the environment will not ;
i be significantly affected.
This one-time exemption from current 10 CFR 50, Appendix J testing requirements and the proposed conforming Technical Specifications change should allow GPC to realize the benefits of the anticipated performant based testing criteria that will be included in the NRC proposed l Appendix J rulemaking and w n pmvide a savings of approximately $1,100,000 by allowing the l Type A test to be scheduled acco . ing to the proposed revision to Appendix J. Therefore, this [
request is submitted as a Cost bueficial Licensing Action (CBLA) and complies with the NRC guidelines for consideration as a CBLA.
To allow for outage planning and scheduling, GPC requests that the proposed exemption and ;
Technical Specifications change be approved by September 30,1995 ,
In accordance with the requirements of 10 CFR 50.91, a copy of this letter and all applicable enclosures will be sent to Mr. J. D. Tanner of the Environmental Pro'ection Division of the ;
Georgia Depanment of Natural Resources.
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LCV-0610
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GeorgiaPower A U. S. Nuclear Regulatory Commission Page 3 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 and that, to the best of >
his knowledge and belief, the facts set forth in this letter and enclosures are true. ,
l GEORGIA POWER COMPANY l By:
. C. K. McC y Sworn to and Subscribed before me this N d y of ah / ,1995.
6 YOAL1I bLtd b' My Commission Expires: % /c /9i9 Notary bublic 0 CKM/KWW l
Enclosures:
[
- 1. Basis for Requests for Exemption to 10 CFR 50, Appendix J and Request to Change Plant Technical Specifications
- 2.10 CFR 50.92 Evaluation
- 3. Technical Specifications Page Change Instructions cc: Georgia Power Company Mr. J. B. Beasley, Jr. ;
Mr. M. Sheibani ,
NORMS l U. S. Nuclear Regulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager Mr. B. R. Bonser, Senior Resident Inspector, Vogtle State of Georgia Mr. J. D. Tanner, Commissioner, Department of Natural Resources l
LCV-0610
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ENCLOSURE 1
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p Vogtle Electric Generating Plant - Unit I f
Request for Exemption to 10 CFR 50,' Appendix J and Reauest to Channe Plant Technical Soecifications 1
I Basis for Exemotion Pursuant to 10 CFR 50.12(a), Georgia Power Company (GPC), holder of Facility Operating License No. NPF-68, hereby requests a specific one-time exemption to Appendix J to 10 CFR :
50,." Primary Reactor Containment Leakage Testing for Water Cooled Reactors." Specifically,'
GPC requests that an exemption to the schedular requirements of Section III.D.I'(a) of 10 CFR ;
50, Appendix J be granted that would permit a one-time extension of the interval between the Type A tests during the first 10-year service period. The extension would allow the Type A !
integrated leak rate test (ILRT) to be performed at the next refueling outage, currently 1 scheduled for September 1997, rather than during the February / March,1996 refueling outage.
The purposes of Appendix J leak test requirements , as stated in the introduction to 10 CFRL 50, Appendix J, are to assure that (a) leakage through the primary reactor containment and {
systems and components penetrating primary containment shall not exceed allowable leakage j rate values as specified in the Technical Specifications or associated bases and (b) periodic j surveillance of reactor containment penetrations and isolation valves is performed so that !
proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating primary containment. The importance of maintaining ;
containment leakage within regulatory limits is the assurance that radioactive releases resulting j from postulated design basis accidents will remain consistent with the accident analyses. This j exemption request is concerned orny with part (a)of the Appendix J stated purpose. Part (b) of !
the stated purpose applies to pmetrations and isolation valves which are tested in accordance !
with the guidance for Types B and C local leak rate tests (LLRT).-
I Backuround Information l The NRC has reviewed it's regulations in an effort to determine which regulations might be !
revised to reduce the regulatory burden on licensees without significantly impacting nuclear . l safety. That review suggested that 10 CFR 50, Appendix J could be a primary candidate for !
. reducing the regulatory burden. As a result, the NRC is currently considering a revision to 10 !
CFR 50, Appendix J. It is expected that the current NRC proposal for revising Appendix J will relax the schedule requirements for Type A ILRT and change the schedule for Types B and C LLRT to a performance based schedule.
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ENCLOSURE 1 (CONTINUED)
REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS It is anticipated that the proposed rule change will become final by the end of 1995 or early 1996. Licensees that have refueling outages ,.ior to or near the time ofissuance of'Jne mle change will not be able to obtain the necessary Technical Specifications changes and plan for implementation of the revised rule during those refueling outages. The third VEGP, Unit 1 ILRT is scheduled to be performed during the IR6 refueling outage in February and March, 1996. GPC is, therefore, requesting a one-time exemption to delay the Type A test scheduled for the IR6 refueling outage. The test will be rescheduled in accordance with the allowable performance based frequency as specified in Option B of the anticipated 10 CFR 50, Appendix J rule change. Although changes to the acceptance criteria for leak rate tests and schedules for Types B and C tests are part of the proposed rule change, an exemption and change is not currently being sought to those test requirements.
Containment The VEGP, Unit I containment consists of a prestressed reinforced concrete cylinder and hemispherical dome supponed on a flat, conventionally reinforced concrete basemat with a central cavity and instrumentation tunnel to house the reactor vessel. The inside face of the concrete is lined with steel plates welded together to form a leak tight barrier. The liner is typically 1/4 inch thick and is thickened locally around penetrations, basemat anchorages, and large brackets. The liner plate, including the thickened plate areas, is anchored to the concrete.
Leak chase channels are provided at seam welds which are inaccessible after construction.
The foundation consists of a circular basemat which is 154 feet,6 inches in diameter and 10 feet,6 inches thick. Attached to the underside of the basemat, at its periphery, is a tendon gallery 10 feet wide by approximately 9 feet,6 inches high. The gallery provides access to the vertical tendons for installation, tensioning, and inservice inspection. Vertical shafts surround each of the shell buttresses to provide access for installation, tensioning, and inservice inspection of the hoop tendons. A steel plate with a leak chase system covers the top of the basemat. A structural concrete fill slab 2 feet,9 inches thick is placed on top of the liner plate to protect the liner from damage during erection and maintenance. Anchorage ofinterior stmetures through the floor liner into the basemat is accomplished by welding cadweld connector sleeves to opposite sides of thickened sections of the liner plate.
The cylinder and dome, both 3 feet,9 inches thick, are provided with reinforcing steel for load resistance and crack control. A haunch, varying in thickness from 3 feet,9 inches to 4 feet,9 inches is provided at thejunction of the cylinder and basemat. The containment side walls are 158 feet,9 inches high and 140 feet in diameter. The containment inside volume is roughly 2,750,000 cubic feet. Additional information about the containment can be found in Chapters 3,6, and 15 of the VEGP FSAR.
The purpose of the containment is to mitigate the consequences of postulated accidents (e.g.,
loss of coolant accident) by minimizing the release of radionuclides to the environment and,
, ENCLOSURE 1 (CONTINUED)
REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS thus, help assure the health and safety of the public. ILRTs (Type A) are performed to verify the integrity of the containment system in its loss of coolant accident configuration so that the release of fission products to the environment during accident conditions does not exceed 10 CFR 100 limits. The overall leakage rate performance of the VEGP, Unit I containment as indicated by the tests performed during the first 10-year service period has been very good.
Industry experience and VEGP containment testing indicate that the minor sources ofleakage from the containment are primarily through the containment penetrations. The containment penetrations (Types B and C) will be tested at VEGP, Unit 1 during the IR6 refueling outage in February / March,1996 in accordance with the requirements of 10 CFR 50, Appendix J and the Plant Technical Specifications.
Historical Type A Testing Results 10 CFR 50, Appendix J, Section II.K, defmes the acceptable leakage limit La as, "the maximum allowable leakage rate at pressure Pa as specified for preoperational tests in the technical specifications or associated bases and as specified for periodic tests in the operating license " Pa as used at VEGP is the calculated design basis accident peak containment pressure. The VEGP, Unit 1 Type A test history since the unit has been in operation, including the preoperational Type A test providesjustification for the proposed revision to the test schedule. Including the preoperational test, three Type A tests have been performed at VEGP, Unit 1. Those tests demonstrate that a margin exists between the Type A test results and the plant Technical Specifications limit of 0.75La where La is equal to 0.20% by weight per day of containment atmosphere at a peak accident pressure of 37 psig. The tests show that VEGP, ,
Unit I has a low leakage containment and that the proposed 18 month extension will not l jeopardize the ability of the containment to maintain the leakage rate at or below the required Type A limits. The Type A tests performed in 1986,1990, and 1993 have successfully verified containment integrity. The periodic tests are short duration tests in accordance with BN-TOP-
- 1. These tests were terminated at the time acceptance criteria were met (allowing for a small margin). The margin would have been greaur if the tests had continued for a longer period of ,
time, but due to economic considerations, they were terminated afler demonstrating that the !
acceptance criteria were satisfied. l Preoperational Type A Test i
The preoperational Type A test was successfully performed in August,1986 with the following results: 1) a calculated mass point leakage of 0.02399%/ day and a 95% upper confidence level (UCL) of 0.02489%/ day and 2) a total time leakage rate of 0.02362%/ day and an UCL of 0.02643%/ day. The preoperational ILRT report was submitted to the NRC by letter dated 1 December 12,1986 (X7BC35).
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. ENCLOSURE l'(CONTINUED) !
REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J ;
AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS .
i First Periodic Type A Test '
. The first periodic Type A test was successfully completed in March,1990 with the following j results: 1) a calculated mass point leakage rate of 0.0859%/ day and an UCL of 0.0884 l Ydday, and 2) a total time leakage rate of 0.0846Ydday and an UCL of 0.1048Ydday. The ;
ILRT test report was submitted to the NRC by letter dated July 13,1990 (ELV-01887).
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Second Periodic Type A Test {
The second periodic Type A test was successfully completed in March,1993 with the.
following results: 1) a calculated mass point leakage rate of 0.0524Pdday and an UCL of l 0.0562Ydday, and 2) a total time leakage rate of 0.0452Ydday and an UCL of 0.1344Ydday. .l The ILRT test report was submitted to the NRC by letter dated June 4,1993 (LCV-0015).
Safety Sinnificance and Risk Assessment Factors affecting leak tightness of containment may be categorized as: 1) active components which are leak rate tested by Type B and _C tests and 2) passive components which constitute ;
the containment structure and are tested during the Type A test. I Active Components i
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The purpose of containment leak testing is to detect any containment leakage resulting from active or passive failures in the containment isolation boundaries before an accident occurs.
The existing Type B and C testing programs are not being modified by this request and will continue to effectively detect containment leakage caused by the degradatior. of active reactor containment components (e.g., valves) as well as sealing material within the containment penetrations. The only potential failure that would not be detected by Types B and C testing are mechanical failure of the containment shell, penetration guard pipes, or welds between pipes and the containment shell. Only these containment structural failures could possibly be . !
affected by the proposed extension in Type A testing frequency.
Industry experience indicates that 97% of the failures associated with Type A tests are found to be due to Types B and C tested penetrations (Draft NUREG 1493," Performance-Based ,
Containment Leak-Test Program") The local leak rate testing frequencies of these penetrations are not affected by this requested exemption. Therefore, continued overall leak tightness of the active containment components can be assured by the existing Type B and C testing program. At VEGP, Unit 1, there have been no ILRT failures.
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ENCLOSURE 1 (CONTINUED)
REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS Passive Structures Two mechanisms could adversely affect the passive structural capability of the containment.
The first mechanism is a gradual deterioration of the stmeture due to pressure, temperature, radiation, chemical, or other such effects. Secondly, modifications can be made to the structure which, if not carefully controlled, could leave the structure with reduced capability.
'- Absent actual accident conditions, structural deterioration is a gradual phenomenon requiring periods of time well in excess of the proposed interval extension. Other than accident conditions, the only over-pressure challenge to containment is the ILRT itself.
10 CFR 50, Appendix J, Section V. A, requires a general inspection of accessible intenor and exterior surfaces of the containment structures and components to be performed prior to each Type A test to uncover any evidence of stmetural deterioration which may affect either the containment structural integrity or leak tightness. At VEGP, Unit 1, there has been no evidence of stmetural deterioration that would impact stmetural integrity or leak tightness.
Modifications that would alter the passive containment structure are infrequent and would receive extensive review to ensure containment capabilities are not diminished. The VEGP, Unit I design change and 10 CFR 50.59 programs have been demonstrated to be effective in providing a high quality oversight of such safety significant modifications. In addition,10 CFR 50, Appendix J, Section IV.A, requires Type A testing to be performed following any major modification to the primary containment boundary. This requirement will be maintained. No such modifications have been made to the containment since the last Type A test in 1993, nor are any planned during the next refueling outage in February / March,1996.
Risk Imoact Assessment The containment structure represents the final barrier to the release of fission products following a postulated severe accident. The risk impact of containment structural leakage is measured by a pathway created for radionuclides in the event that the containment is challenged, such as a loss of coolant accident (LOCA) or severe accident. Such leakage does not create any new accident scenarios, nor does it contribute to the initiation of any accident.
From a risk standpoint, the purpose of Appendix J leak testing is to detect any containment leakage resulting from failures in the containment isolation boundary before an accident occurs.
Such leakage could be the result ofleakage through containment penetrations, through airlocks, or through containment structural faults. The Appendix J Type B and C tests will continue to detect leakage through containment isolation valves, penetrations, and airlocks.
The only potential failure which would not be detected by Types B and C testing are mechanical failures of the containment shell (i.e., degradations or modifications to the containment shell), penetration guard pipes, or welds between pipes and the containment shell.
Thus, the only potential effect of the proposed exemption to the Type A test frequency is the i
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. ~ ENCLOSURE 1 (CONTINUED)
~ REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS probability the containment structural leakage would go undetected between tests.
The containment structure is passive. Under normal operating conditions, there is no significant environmental or operational stress which could contribute to its degradation. A {
review of modifications for potential effects to the containment structure is discussed in the preceding section. Passive failures resulting in significant containment structural leakage are, ;
therefore, extremely unlikely to develop between Type A tests. No such failures have thus far, _l i
occurred at VEGP.
Postulated containment failure under severe accident conditions is primarily due to phenomenological effects associated with severe accidents. Such phenomenological effects were considered as part of the VEGP Individual Plant Evaluation (IPE). It is highly unlikely that any of the identified containment failure mechanisms for severe accidents would be ;
impacted by the proposed increase in the testing interval. .
10 CFR 50.12 Considerations ;
In accordance with 10 CFR 50.12, the Commission may grant an exemption to the requirements of the regulations of 10 CFR 50 if the exemption is authorized by law, will not ;
present an undue risk to the public health and safety, is consistent with the common defense l and security, and special circumstances am present. The special circumstances of 10 CFR 50.12(a)(2)(ii), (iii), and (vi) apply to this reqcsted exemption. :
.i Proposed Exemption is Authorized by Law ;
There is no known law that would be violated by the granting of the proposed exemption. 10 CFR 50.12 provides the basis for granting exemptions to the requirements of 10 CFR 50 l regulations. The NRC has granted similar exemptions in the past. Therefore, the exemption is ;
authorized by law. 1 Proposed Exemption Does Not Present an Undue Risk to the Public Health and Safety 10 CFR 50, Appendix J states that the purpose of the regulation is to assure that leakage $
through primary containment and systems and components penetrating containment does not exceed allowable values. as specified in the Technical Specifications or associated bases, and s ,
that proper maintenan snd repa%re performed throughout the service life of the )
containment boundary components. The ILRT history for VEGP, Unit I during the first 10 year service period inspection interval indicated that the containment structure has not experienced degradation. The NRC has conducted a detailed study ofintegrated leak rate tests performed from 1987 to 1993. That study, documented in draft NUREG-1493, determined 6-1 i
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. ENCLOSURE I (CONTINUED) )
l REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J l AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS that 97% of the leakage rate tests that exceed the acceptance criteria are identified by LLRT programs. The LLRT program at VEGP, Unit I has been successful in maintaining low Type B and C containment leakage. Since there has been no identified containment structural leakage, the LLRT program has contributed to the successful ILRTs. Therefore, as shown in the NRC study and as indicated by the VEGP, Unit I containment performance history, postponing the ILRT by one refueling cycle remains consistent with the intent of the regulation and will not present an undue risk to the public health and safety.
Consistent With the Common Defense and Security GPC interprets the term " common defense and security" as referring principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, and the protection of restricted data. The granting of the requested exemption will not affect any of those matters, and thus, the granting of the exemption is consistent with the common defense and security of the United States.
10 CFR50.12(a)(2)(ii)- Application of the Regulation is not Necessary to Achieve the Underlying Purpose of the Rule The underlying purpose of 10 CFR 50, Appendix J will still be served if a third ILRT is not conducted during the first 10-year service period. Appendix J states that the leakage test requirements provide for periodic verification by tests of the leak tight integrity of the primary reactor containment. The Appendix further states that the purpose of the tests is to assure that leakage through the primary reactor containment shall not exceed the allowable leakage rate values as specified in the Technical Specifications or associated bases.
10 CFR 50, Appendix J, Section III.D.l(a) states that a set of three periodic tests shall be ,
performed at approximately equal intervals during each 10-year period and that the third test )
shall be conducted when the plant is shutdown for the 10-year plant inservice inspections. The proposed exemption would permit delaying of the scheduled Type A test and permit performance of the Type A test after the completion of the first 10-year inservice inspection ,
interval in accordance with the schedule to be provided in the proposed revision to Appendix J. l The methodology, acceptance criteria, and Technical Specifications leakage limits for performance of the Type A test will not change.
The testing history, structural capability of the containment, and the risk assessment discussed previously establish that 1) VEGP, Unit I has had acceptable containment leakage rate test results, 2) the structural integrity of containment is assured, and 3) there is negligible risk ,
impact in changing the Type A test schedule on a one-time basis. 1 Thus, there is significant assurance that the extended interval between Type A test::in concert w . . --
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- . - ENCLOSURE I (CONTINUED) j REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J J
AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS with the Type B and C testing continue to provide periodic verification of the leak tight 1 integrity of the containment.
' 10 CFR 50.12(a)(2)(lii)- Compliance with the Regulation Would Result in Unduc Hardship or Other - i Costs that are Significantly in Excess of Those Contemplated When the Regulation was Adapted -
Postponing the ILRT for VEOP, Unit I will eliminate unnecessary testing without any .
reduction in plant safety. The ILRT typically requires two-to-three days to perform, with the ,
possibility of significant extended time requirements. Outage activities are severely impacted i
' during the preparation period prior to the ILRT and during the performance of the ILRT. A ;
cost savings can be realized by a reduction in outage time, eliminating the impact of the ILRT on other outage activities, and direct costs related to obtaining equipment and services necessary for performance of the ILRT. This proposed exemption could result in a total cost j benefit of about $1,100,000, by eliminating one ILRT.
' 10'CFR 50.12(a)(2)(vi) - Presence of Material Circumstances not considered When the Regulation was Adopted Certain material circumstances were not considered when the regulation was adopted. The benefit of time has provided experience and information that give a better perspective about ;
containment integrity. Two important material circumstances are testing history arel the development of probabilistic risk assessments (PRAs). ,
Since the promulgation of 10 CFR 50, Appendix J, in 1973, more than 20 years of nuclear i power plant operating experience has been obtained. A review ofindustry data did not find any instances where a Type A test failed to. meet Appendix J acceptance criteria as a result of a t containment structuralleak not due to initial fabrication or a plant modification. That operating history provides a significant indicator that containment structural integrity (passive s structure) is not a significant safety concern.
Plant specific PRAs were not available in 1973, and therefore, were not considered when the .;
regulation requiring compliance with Appendix J [10 CFR 50.54(o)] was adopted. Overall i plant risk due to containment leakage is relatively small given the small probability of - !
containment leakage itself. The predominant contributor to degraded containment integrity is a the phenomenological effects of a severe accident, not pre-existing containment integrity
- conditions. An assessment of the risk impact in the exemption request indicates that there is no I undue risk to the public health and safety as a result of the proposed scheduler extension of the j
. Type A test.
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I There have been no modifications to the containment structure or liner that would impact the overall containment integrity and leak tightness.
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. ENCLOSURE 1 (CONTINUED)
REQUEST FOR EXEMPTION TO 10 CFR 50, APPENDIX J AND CHANGE TO PLANT TECHNICAL SPECIFICATIONS Basis for Technical Soecifications Change Proposed Change Technical Specifications Surveillance Requirement 4.6.1.2 is changed from "The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in accordance with the criteria specified in Appendix J of 10 CFR 50 using the methods and
. provisions of ANSI N45.4-1972:" to "Except as modified by NRC approved exemptions, the containment leakage rates shall be demonstrated at the following test schedule and shall be determined in accordance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972;"
Basis for Change The proposed Technical Specifications change is requested to support the requested exemption and any future exemptions that may be approved by the NRC. . The Technical Specifications change does not change any of the criteria of 10 CFR 50, Appendix J for containment leak rate testing. It does not change any plant procedures used in performing the tests, and it does not modify any plant equipment. It only provides a mechanism within the Technical Specifications for implementing NRC approved exemptions to 10 CFR 50, Appendix J. Without the Technical Specifications change, the exemption, when approved, could not be implemented.
Since the purpose of the Technical Specifications change is to support the requested exemption to 10 CFR 50, Appendix J, the forgoing basis and justification for the proposed exemption also apply as the basis and justification for this proposed Technical Specifications change.
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ENCLOSURE 2 Vogtle Electric Generation Plant, Unit 1 Request to Revise Technical specifications Exemption to 10 CFR 50,' Appendix J Test Criteria 10 CFR 50.92 Evaluation -
Proposed Change Technical Specifications Surveillance Requirement 4.6.1.2 is changed from "The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in accordance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972:" to "Except as modified by NRC approved exemptions, the containment leakage rates shall be demonstrated at the following test schedule and shall be '
determined in accordance with the criteria specified in Appendix J of 10 CFR 50 using the .
methods and provisions of ANSI N45.4-1972:"
Evaluation Georgia Power Company (GPC) has reviewed the proposed change and determined that it does not involve a significant hazards consideration based on the following:
- 1. The change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change does not involve a change to structures, systems, or components which would affect the probability of an accident previously evaluated in the Vogtle Electric Generating Plant (VEGP) Final Safety Analysis Report (FSAR). The change only provides a mechanism for implementing exemptions to 10 CFR 50, Appendix J containment leak rate testing criteria which have been approved by the NRC.
- 2. The proposed change will not create the possibility of a new or different kind of accident from any accident previously analyzed. The amendment would not change the design, configuration, or method of plant operation. It only allows exemption to specific 10 CFR 50, Appendix J criteria as previously approved by the NRC.
- 3. Operation of VEGP, Unit 1 in accordance with the proposed change will not involve a significant reduction in the margin of safety. The proposed change would not, in itself, change a safety limit, an LCO, or a surveillance requirement on equipment required for plant operation. Before the change could be used an exemption to 10 CFR 50, Appendix J would have to be evaluated and approved by the NRC. The change only provides a way to implement NRC approved exemptions without violating the Technical Specifications.
I ENCLOSURE 3 Vogtle Electric Generating Plant, Unit 1 Request to Revise Technical Specifications: '
Exemption to 10 CFR 50, Appendix J Page Change Instructions t
The proposed change to the Vogtle Electric Generating Plant, Unit 1 Technical Specifications ;
will be incorporated as follows:
P_ age Instructions 3/4 6-2 Replace
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