ML20085M881
| ML20085M881 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/20/1995 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20085M885 | List: |
| References | |
| NUDOCS 9506290391 | |
| Download: ML20085M881 (17) | |
Text
Ctation Support Departrnent
'e .
' PECO ENERGY is"J=J=L,,
965 Chesterbrook Boulevard Wayne, PA 19087-5691 June 20,1995 Docket Nos. 50-352 50453 Ucense Nos. NPF-39 NPF-85 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Umerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 95-06-0, and Request For a LGS Unit 1,10 CFR 50, Appendix J Exemption Gentlemen:
PECO Energy Company (PECO) is submitting Technical Specifications (TS) Change Request No. 95-06-0, in accordance with 10 CFR 50.90, requesting a change to TS (i.e.,
Appendix A) of Operating Uconse Nos. NPF-39, and NPF-85 for Umerick Generating Station (LGS), Units 1 and 2. In addition, PECO is requesting a one-time (i.e., temporary) Unit 1 schedular exemption from 10 CFR 50, Appendix J, " Primary Containment Leakage Testing for Water-Cooled Power Reactors," in accordance with 10 CFR 50.12.
The purpose of the proposed Technical Specifications (TS) changes are to remove the surveillance interval text for the Appendix J, Type A test (Integrated Leak Rate Test or ILRT), and Drywell-to-Suppression Chamber (bypass) leakage test specified in TS Surveillance Requirements (SRs) 4.6.1.2.a. 4.6.1.2.b, and 4.6.2.1.e. Elimination of the specific Type A test interval text from TS would allow future changes to the interval by exercising an Appendix J exemption without requiring additional TS changes. Similarly, the bypass test interval is a detail of the test method and is adequately addressed in implementing procedures. Portions of existing TS are permitted to be relocated to licensee controlled documents, if deemed marginal to operational safety, allowing the operator to concentrate on only those TS which are important to the safe operation of the plant. The test interval details are marginal to operational safety, and are repeated in 10 CFR 50, Appendix J, and implementing procedures; therefore, these details are proposed to be removed from TS.
PECO is also requesting a LGS Unit 1,10 CFR 50, Appendix J exemption to permit the Type A test to be extended from the sixth, Unit 1, refueling outage to the seventh, Unit 1, refueling outage. The exemption is being pursued due to the impending Appendix J rule change to permanently increase the interval to once overy ten years (option B). The rulemaking is part of the NRC initiative to elitninate requirements that are marginal to safety. Granting the one-time exemption would allow PECO to obtain outage schedule benefits resulting in cost savings and reduced worker radiation exposure prior to the Appendix J, option B, rulemaking.
The NRC has granted a similar request for Consolidated Edison Company of New York, Indian Point Unit 2, by letter dated March 17,1995.
Information Supporting this TS Change and Exemption Request are contained in Attachment 1 to this letter. The proposed replacement TS pages for the Unit 1 and Unit 2 TS are contained ir.
9506290391 950620 PDR ADDCK 05000352 l p ,
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' June 20,1995 Page 2 Attachment 2. The TS change in*ormation is being submitted under affirmation, and the required affidavit is enclosed.
We request that if approved and granted, the TS Change and Exemptbn Request be issued by ,
October 1,1995, and become effective within 30 days of issuance in order to eliminate ,
committed costs associated with pre-outage expenses associated with the performance of the 10 CFR 50, Appendix J, Type A test during the upcoming Unit 1 refueling outage.
If you have any questions, please do not hesitate to contact us.
Very truly yours, b lAst G. A. Hunger, Jr.,
Director-Licensing Enclosure, Attachments ,
I cc: T. T. Martin, Administrator, Region I, USNRC (w/ enclosure, attachments)
N. S. Perry, USNRC Senior Resident inspector, LGS (w/ enclosure, attachments)
R. R. Janati, PA Bureau of Radiation Protection (w/ enclosure, attachments) j i
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COMMONWEALTH OF PENNSYLVANIA :
- ss.
COUNTY OF CHESTER :
W. H. Smith, Ill, being first duty sworn, deposer and says: That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the enclosed Technical Specifications Change Request .;
t No. 95-06-0 ' Elimination of 10 CFR 50, Appendix J, Type A Test Interval Text from TS," for Limerick Generating Station, Unit 1 and Unit 2, Facility Operating License Nos. NPF-39 and NPF-85, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the ,
best of his knowledge, information and belief.
l AHf '
Vice Presid Subscribed and sworn to ;
before me this 20 bay l l
of 1995. ;
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Nota Public Notanal Seal Lou Skrocki. Notary Public rtn Twp., Chester County a My ission Expires May 17.1999 Marvear Pennsytvana 4mxiton d tbtims
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i ATTACHMENT 1 t
-i UMERICK GENERATING STATION UNITS 1 AND 2 f 1
Docket Nos. !'
50-352 50-353 i i
Ucense Nos.
' Units 1 and 2 Technical Specifications Change Request No. 95464 'I and Unit 1,10 CFR 50, Appendix J, Exemption Request" Information Supporting Changes - 13 pages e t
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' DISCUSSION AND DESCRIPTION OF THE PROPOSED CHANGES PECO Energy Company (PECO) is requesting Technical Specifications (TS) changes, and a one-time i (i.e., temporary) schedular exemption from the requirements of 10 CFR 50, Appendix J.
I The purpose of the proposed TS changes are to remove the surveillance interval text for the 10 CFR 50, Appendix J, Type A test (Integrated Leak Rate Test or ILRT), and Drywell-to-Suppression Chamber (bypass) leakage test specified in TS SurveiHance Requirements (SR) 4.6.1.2.a 4.6.1.2.b, and 4.6.2.1.e.
Elimination of the specific Type A test interval text from TS would allow future changes to the interval by exercising an Appendix J exemption without requiring additional TS changes. Similarly, the bypass test .
Interval is a detau of the test method and is adequately addressed in implementing procedures. NUREG !
- 1433 " Standard Technical Specifications for General Electric Plants, BWR/4," issued on September 28, !
1992, provides the standard format and content of TS to improve operational safety, make the TS
]
requirements more understandable, and lessen the administrative burdens imposed by current TS. ;
Portions of existing TS are permitted to be relocated to licensee controlled documents, if deemed marginal to operational safety, allowing the operator to concentrate on only those TS which are important to the safe operation of the plant. The test laterval details are marginal to operational safety, and are repeated in 10 CFR 50, Appendix J, and implementing procedures; therefore, these details are proposed to be removed from TS.
LGS TS SR 4.6.1.2.a is currently worded as follows.
"Three Type A Overall Integrated Containment Leakage Rate Tests shall be conducted at 40 + /.
10 month intervals during shutdown at P.,44.0 psig, during each 10-year service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.""
Note ** applies to Unit 1 TS and states:
'The interval between the second and third Overall Integrated Leakage Rate tests of the first 10-year service period will be extended to the sixth Unit 1 refueling outage. As a result, the duration of the first 10-year service period will be ex;cnded to the end of the sixth Unit I refueling outage."
The following is the proposed TS SR 4.6.1.2.a requirement.
" Type A overall Integrated Containment Leakage Rate tests shall be conducted at P., in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions."
LGS TS SR 4.6.1.2.b is currently worded as follows.
"If any periodic Type A test falls to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests faH to meet 0.75 L,, a 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 following is the proposed TS SR 4.6.1.2.b requirement.
- 1f any periodic Type A test falls, or if two ccnsecutive Type A tests fail, a Type A test shall be performed in accordance with 10 CFR 50, Appendix J as modified by approved exemptions."
LGS TS SR 4M.2.1.e is currently worded as follows. %
"Drywell-to-suppression chamber bypass leak tests shah be conducted at 40 +/- 10 month intervals to coincide with the ILRT at an initial differential pressure of 4 psi and verifying that the
Page 2 of 13 A//k calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test falls to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fall to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed.'
The following is the proposed TS SR 4.6.2.1.e requirement.
"Drywell-to-suppression chamber bypass leak tests shall be conducted to coincide with the Type A test (in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions) at an initial differential pressure of 4 psi and verifying that the A//k calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test falls to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed."
In addition to the proposed TS changes to TS page 3/4 6-14, a change to SR 4.6.2.1.f was made to correct a typographical error. Specifically, the reference to the bypass test Specification was changed from 4.6.2.1.d to 4.6.2.1.0.
PECO is also seeking a LGS Unit 1,10 CFR 50, Append!x J exemption to permit the Type A test to be extended from the sixth, Unit 1, refueling outage to the seventh, Unit 1, refueling outage. The exemption is being pursued due to the impending Appendix J change to permanently increase the interval to once overy ten years (option B), and the NRC's guidance provided in a memorandum " Guidance for Requesting Exemptions to the Requirements for Type A Testing in 10 CFR 50, Appendix J," from Mr.
Clyde Shiraki - NRC Project Manager, to all Project Directors, dated February 10,1995.
TS SAFETY ASSESSMENT The failure effects which are potentially created by the proposed Technical Specification changes are considered in the following section. The accidents which are potentially negatively impacted by the proposed change are any Loss of Coolant Accident (LOCA) inside primary containment, with or without offsite power available, described in the LGS Updated Final Safety Analysis Report (UFSAR) Section 15.6.5.
The proposed TS changes will eliminate the interval details from TS which will allow changes to the 10 CFR 50, Appendix J, Type A test by exercising an Appendix J exemption without requiring additional TS changes. The proposed TS changes will eliminate the drywell-to-suppression chamber bypass leakage (bypass) test interval detaFs from TS and align the bypass interval with the Type A test interval.
NUREG - 1433 " Standard Technical Specifications for General Electric Plants, BWR/4," issued on September 28,1992, provides the standard format and content of TS to improve operational safety, make the TS requirements more understandable, and lessen the administrative burdens imposed by current TS. Portions of existing TS are permitted to be relocated to licensee controlled documents, if deemed marginal to operational safety, allowing the operator to concentrate on only those TS which are important to the safe operation of the plant. The LGS Type A test, and bypass test interval details are marginal to operational safety, and are repeated in 10 CFR 50, Appendix J, and implementing procedures; therefore, these details are proposed to be removed from TS.
The Type A test is performed to determine that the total leakage from containment does not exceed the maximum allowable leakage rats (La) at a calculated peak containment internal
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1 Page 3 of 13 pressure (Pa), as defined in 10 CFR 50, Appendix J. The containment limits fission product leakage during and following Design Bases Events in accordance with the requirements described in 10 CFR 100. The primary containment is not considered to be an accident initiator, it is an accident mitigator. There are no physical or operational changes to the containment structure, system or components being made as a result of the proposed changes. Primary containment leakage rate requirements TS 4.6.1.2.a. and 4.6.1.2.b are presented as a supporting SR for Primary Containment OPERABILITY. The TS SR interval details are also found in 10 CFR 50, Appendix J. The Appendix J regulations require licensee compliance, can not be revised by the licensee, and will be addressed by direct reference in TS. Therefore, details of the regulation currently found in TS are repetitious and removing them relieves the burden of future unnecessary TS change requests where future changes are managed through the exemption process. Furthermore, approved exemptions to the regulations, and exceptions presented within the regulations themselves, are also details which are presented adequately without repeating the details within the TS. Future changes to LGS Implementing procedures will be made in accordance with the existing provisions of 10 CFR 50.59 and will be subject to the change control process in the Administrative Controls Section 6.0 of TS.
The drywell-to-suppression chamber bypass test is performed to confirm that the total leakage from the drywell-to-suppression chamber does not exceed the allowable leakage area. Any excess leakage could result in excessive containment pressures, since steam flow into the airspace would bypass the heat sink capabilities of the suppression pool. The suppression chamber is not considered to be an accident initiator, it is an accident mitigator. There are no physical or operational changes to the drywell-to-suppression chamber diaphragm floor, suppression system or its components being made as a result of the proposed changes. The bypass leakage rate requirement TS '. 6.2.1.e is presented as a supporting SR for Suppression Chamber OPERABILITY, The current SR interval was NRC approved and amended to coincide with the 10 CFR 50, Appendix J, Type A test (ILRT). The basis of the bypass test interval was established and documented in the NRC Safety Evaluation dated February 17,1994 supporting Unit 1 TS amendment 68, and Unit 2 TS amendment 31. The bypass test also measures leakages through active (vacuum breakers) and passive components (concrete and welded steel) similar to those of the Type A test. These active components will be tested, in accordance with TS SR 4.6.2.1.f. during each refueling outage where the Type A test is not performed and the passive components have been demonstrated.001 to have time-dependant failures. The SR interval is adequately presented in the test implementing procedure, and will be addressed by direct reference to coincide with the Type A test interval in TS.
Information Sucoortina a Findina of No Sianificant Hazards Consideration We have concluded that the proposed changes to the LGS Unit 1 and Unit 2 TS to eliminate the specific 10 CFR 50, Appendix J, Type A test interval from TS does not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.
- 1. The orocosed TS chances do not involve a slanificant increase in the orobabt!ity or conseauences of an accident oreviousiv evaluated.
The primary containment and the suppression chamber are not considered to be accident initiators, they are accident mitigators. There are no physical or operational changes to the containment or suppression structure, system or components being !
made as a result of the proposed changes. These changes will not impose different requirements and adequate control of information will be maintained. These TS changes l will not alter assumptions made in the safety analysis and licensing basis. Therefore, the proposed TS changes to eliminate the details of the test intervals will not increase the probability or consequences of an accident previously evaluated.
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- 2. Jbe orocosed TS chances do not create the oossibility of a new or different kind of accident from any accident oreviously evalugtgl t ,
The proposed changes remove the specific surveillance test interval text from TS and address the interval by direct reference to the applicable regulation. The proposed TS changes do not make any physical or operational changes to existing plant systems or components. Furthermore, the primary containment and suppression chamber act as accident mitigators not initiators. Therefore, the possibility of a new or different kind of accident than from any accident previously evaluated is not introduced.
- 3. The crocosed TS chanaes do not involve a slanificant reduction in a marain of safety.
LGS TS Bases 3/4 6.1.2 state that surveillance testing is consistent with 10 CFR 50, Appendix J and does not specify a SR test interval. TS Bases 3/4 6.2, describing the bypass test does not specify a SR test interval. However, the NRC Safety Evaluation related to amendment Nos. 68 (Unit 1) and 31 (Unit 2) concluded that it is acceptable for the drywell-to-suppression chamber test frequency to coincide with the 10 CFR 50, Appendix J, Type A test, since individual vacuum breaker leakage tests are an acceptable alternative to an integrated suppression pool bypass test during outages for which a Type A containment Integrated leak rate test is not conducted. The attemative bypass test requiremer.t TS SR 4.6.2.1.f, is not affected by these changes.
The Type A test, and bypass SR test intervals are adequately presented in the test implementing procedures, and TS will directly reference 10 CFR 50, Appendix J, for the appropriate test interval.
Therefore, the proposed TS changes do not involve a significant reduction in a margin of safety.
Conclusion The Plant Operations and Review Committee and the Nuclear Review Board have reviewed these proposed changes to the LGS Unit 1 and 2 TS, and have concluded that the changes do not involve an unreviewed safety question, and will not endanger the health and safety of the public.
EXEMPTION REQUEST PECO Energy Company (PECO) in accordance with 10 CFR 50.12, requests a one-time exemption (i.e.,
temporary) from the requirements of 10 CFR 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," for Limerick Generating Station (LGS), Unit 1, Operating License No. NPF-39. This requested exemption, along with PECO's Technical Specifications Change Request No. 95-06-0, will allow a one-time test interval extension for the 10 CFR 50, Appendix J, Type A ,
l test (containment integrated leak rate test) from the current scheduled 62 months to approximately 89 months.
The Type A test is performed to determine that the total leakage from containment does not exceed the maximum allowable leakage rate (La) at a calculated peak containment intomal pressure (Pa), as defined in 10 CFR 50, Appendix J. The containment limits fission product leakage during and following Design Bases Events in accordance with the requirements described in 10 CFR 100. There are no physical, testing or operational changes to the containment structure, system or components being made as a i result of the proposed exemption.
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'10 CFR 50.54(o) requires that the primary reactor containments for water cooled power reactors shall be l subject to the requirements set forth in 10 CFR 50, Appendix J. Append!x J, section Ill.D.1.(a) states, 1
'After the preoperational leaka0e rate tests, a set of three Type A tests shall be performed, at I approximately equal intervals during each 10-year service period. The third test of each set shall be l conducted when the plant is shutdown for the 10-year plant inservice inspections.'
The purpose for requesting this exemption is anticipatory to NRC Rulemaking invoMng 10 CFR 50, Appendix J, (i.e. Option B), whereby the NRC is considering permanently increasing the interval between i Type A tests from three time; in 10 years to once in 10 years. The NRC is conducting an expedited rulemaking effort to revise Appendix J as part of the initiative to eliminate requirements that are marginal to safety. This effort is discussed in SECY-94-036, " Staff Plans for revising 10 CFR 50, Appendix J,
' Containment Leakage Testing
- and Handling Exemption Requests," dated February 17,1994 and SECY-94-090, " Institutionalization of Continuing Program for Regulatory improvement," dated March 31,1994.
The Rufe change is currently expected to be approved no sooner than October of 1995. In a memorandum to all NRR Project Directors, dated February 10,1995, the NRC provided guidance for requesting exemptions to the requirements for Type A testing applicable to plants that have refueling outages scheduled near the anticipated approval date of the Rule and that would not have ample time to imptoment the new regulation and make usc of the relaxed requirements. This exemption request is consistent with the NRC memorandum, whereby the Type A test exemption is temporary and involves the Type A test only (i.e., will not change the Type B or C tests). The proposed changes are limited to that aspect of compliance with Appendix J which will provide a significant hardship without a corresponding increase in safety. This was the basis for the earlier PECO exemption request which was granted by the NRC on February 16,1994 temporarily increase the LGS Unit 1 Type A test interval from 40 + /- 10 months to the end of the sixth, Unit 1, refueling outage (approximately 62 months).
The benefit of not performing the Type A and bypass tests would be a reduction in refueling outage duration and personnel radiation exposure.
Test Intervals The proposed Type A test interval change will increase the interval approximately 39 months beyond the original 40 +/- 10 month surveillance test interval (39 mos.+ 50 mos. = 89 mos.) or 27 months beyond the currently approved Unit 1 exemption. In addition, the Appendix J ten year service period (Aug 1984 to Aug 1994) will be extended by approximately 44 months.
10 CFR 50 Appendix J states "After the preoperational leakage rate test, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections."
This Appendix J exemption will supersede an earlier exemption granted by NRC letter dated February I 16,1994 which allowed the third (post-operational) Type A test to coincide with the 10-year plant 1 inservice inspection (ISI) currently scheduled for the sixth, Unit 1, refueling outage. Also, the drywell-to- l suppression chamber bypass test TS interval was permanently amended by letter dated February 17 j 1994 to allow the test to coincide with the Type A test, and is therefore also currently scheduled for the sixth, Unit 1, refueling outage. The bypass test was last performed in January 1992, and is proposed to be extended approximately 25 months beyond the current TS specified interval (i.e.,40 + /- 10 months).
This proposed exemption requests a one-time (i.e., temporary) schedular change to defer the Type A test one more fuel cycle to be performed during the seventh, Unit 1, refueling outage scheduled for April 1998, and therefore, defer the bypass test to the same interval.
JUSTIFICATION FOR EXEMPTION PECO Energy is requesting a one-time (i.e., temporary) schedular exemption from 10 CFR 50, Appendix J, section Ill.D.1(a) which establishes the periodic test schedule for Type A tests. Specifically, we are
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requesting exemption from the requirement that three (post-operational) Type A tests be performed at approximately equal intervals during the 10-year service period, and where the third test coincides with the end of the 10-year ISI interval, and that this exemption supersede the February 16,1994 exemption.
RAlls For Specific Exemption l
Pursuant to 10 CFR 50.12(a)(2), the NRC wiling 1 consider granting an exemption to a regulation requirement of 10 CFR Part 50, unless special circumstances are present. This exemption request meets 1 the special circumstances of paragraphs (a)(2)(li), (a)(2)(iii), and (a)(2)(v) of 10 CFR 50.12. This l exemption request demonstrates that the underlying purpose of the rule will be achieved, compliance i would result in undue hardship, and the exemption will provide only temporary relief. In accordance l with 10 CFR 50.12(a)(1) the granting of this requested exemption will not present undue risk to the l health and safety of the public and is consistent with the common defense and security.
Description of Containment As described in LGS UFSAR Section 3.8.1, the LGS Unit 1 and Unit 2 containments are the Mark ll design, each having a conical drywell located over a cylindrical suppression chamber. These two sections are comprised of a structurally integrated, reinforced concrete pressure vessel, lined with welded steel plate which is anchored to the concrete slab by structural steel beams embedded in the ;
concrete, and provided with a steel domed head for closure et the top of the drywell. !
l The thickness of the reinforced concrete base foundation is eight feet, the side walls are six feet - two l Inches, the drywell-to-suppression chamber diaphragm slab is reinforced concrete three and a half feet thick, and the welded stool domed head is one and a half to four inches thick.
l Each containment penetration consists of a pipe sleeve with an annular ring welded to it. The ring is embedded in the concrete wall, and provides an anchorage for the penetration to resist normal operating and accident loads. The pipe sleeve is also welded to the containment liner plate to provide a leak-tight penetration.
Discussion of Chanae Factors affecting leak tightness of containment may be categorized as 1) active components which are leak rate tested by 10 CFR 50, Appendix J, Type B and C leak tests, and 2) passive components which constitute the containment structure and are tested during the Type A leak test (primary containment and drywell-to-suppression chamber tests are performed together).
The purpose of containment leak testing is to detect any containment leakages resulting from active or passive failures in the containment isolation boundaries before an accident occurs.
Active Comoonents The major leakage paths are composed of penetration seals, containment isolation valves, and gross containment component failure.
Penetration seal leakage consists of alriock door seals, doors or hatches with resilient seals or gaskets except for seal welded doors, penetrations whose design incorporates reslilent seals, gaskets, or seal compounds, piping penetrations fitted with expansion bellows, and electrical penetrations fitted with flexible metal seal assemblies. Type B testing is performed on these types of components and the Type B testing will not be affected by the exemption request. !
Containment isolation valves provide either a potential or direct connection between the inside )
and the outside atmospheres of the primary containment under normal operation, are required l to close automatically upon receipt of a containment isolation signal in response to controls ;
intended to affect containment isolation and may be required to operate intermittently under post
Page 7 of 13 accident conditions. Leakages through these valves can be caused by leaking valve seats, isolation valve closure failures, or failure to return a penetration to its norma!!y closed condition.
For all these initiating events, except valve alignment, this type of leakage is detectible by Type C local leak rate testing. Following any maintenance on a Containment Isolation Valve, post maintenance testing (PMT) is performed followed by an independent verification to ensure proper valve alignment. Type C testing is performed on these types of valves and the Type C testing will not be affected by the exemption request.
LGS has a comprehensive Primary Containment Testing Program. The program has actively tracked and trended Type B and C test results, and PECO Energy has been able to assess from test data, confirmed by repairs, that the majority of containment leakages were the result of (active) Type B and C penetrations. Exhibit A contains excerpts from the LGS Annual Assessment Report detalling program data and results since 1990 (the last time an ILRT was performed). In addition, maintenance of Type B and C components has been performed which has decreased overall valve leakage within the scope of the fo; lowing LGS programs: Predictive Maintenance Program; Nuclear Plant Reliability Data System (NPRDS); inservice Testing Program; Reliability Centered Maintenance Program; and the Chronic System Problem Program.
A gross containment failure as a result of a low probability event is detected only by a Type A test. 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 degradation of active components.
The Type A tested boundary includes the containment structure and the piping from the containment penetration to the outboard primary containment isolation valve. The piping includes some flanged connections which are not tested via the Type B or C testing programs.
The piping system with flanged connections is part of the Containment Atmosphere Control (CAC) system. The risk of gasket degradation is judged to be negligible based on the fact that they are mechanical joints which are seismically supported and operate at low pressure and temperature conditions. Under these service conditions gaskets are not considered to degrade, and are judged to last the life of the plant unless the joint is subjected to maintenance. The other components in containment which will go untested are the containment structure itself and ]
small instrumentation lines. Time-based degradation of any of the instrumentation lines would i most likely be identified by faulty instrument indication or during instrument calibrations. In l examining the potential for a time-dependent failure mechanism which could cause a significant !
degradation of the containment structure, it is concluded that the risk, if any, of such a j mechanism is small when considering the design requirements established for the containment structure.
Passive Structure Two mechanisms could adversely affect the passive structural capability of containment. The first is deterioration of the structure due to pressure, temperature, radiation, chemical or other l such effects. Secondly, modifications can be made to the structure which, if not controlled, could leave the structure with reduced capability.
In the absence of actual accident conditions, structural deterioration is a gradual phenomenon occurring over a period of time (designs consider the life of the plant) well in excess of the proposed interval extension. Other than accident conditions, the only other pressure challenge
- to containment is the Type A test itself. LGS Unit 1 has not experienced an accident nor containment challenges that would adversely impact its structural integrity or leak tightness. j 10 CFR 50, Appendix J, and TS require an inspection of accessible interior and exterior surfaces i of the containment structure and components to be performed prior to the Type A test to identify any evidence of structural deterioration which taay affect either the containment structure !
or its leak tightness. LGS has not identified any evidence of structural deterioration that would j
Page 8 of 13 adversely impact its structural integrity or leak tightness.
Modifications that would alter the passive containment structure are infrequent and would receive extensive review to ensure containment leak tightness capabilities are not diminished.
The LGS design change and 10 CFR 50.59 programs have been demonstrated to be effective in providing a high quality oversight of such safety significant modificationc. PECO has previously provided assurances, as pat of the earlier 1994 granted extension, that no modifications impacting the primary containment were performed since the last Type A test (November 1990).
This justification also assumes that any activities performed during the sixth, Unit 1, refueling outage with the potential to impact the primary containment leakage rate will be tested to demonstrate conclusively that there would be no negative impact to the primary containment leakage.
Historical LGS Tvoe A Testina Results The intent of the Type A test is to determine that the total leakage from containment does not exceed the maximum allowable leakage rate (La) as specified in TS, the UFSAR, and 10 CFR 50 Appendix J. The LGS design maximum allowable leakage rate measured in weight percent /24 hours at the peak accident pressure of 44 psig is 0.5% wt/ day. TS and Appendix J require this value to be 75% of La or 0.375 %wt/ day to allow for deterioration of leakage paths between tests (i.e. 40 +/- 10 months). The following review of the LGS Unit 1, Type A test data is presented to evaluate the risk of time-dependent failures.
The first Type A test performed was during pre-operational testing in August 1984. The test resulted in a total time leakage rate of 0.255 % wt/ day.
The first, post-operational, Type A test was performed in August 1987. The initial Type A test period leakage rate stabilized at approximately 1.0 % wt/ day which failed to meet the TS criteria of less than 0.75 La. An investigation revealed the cause of the excessive leakage, which were packing leaks on nine Containment Atmosphere Control (CAC) System valves. The failure of the packing glands was attributed to Modification 980 which replaced bearings on purge and vent valves. Local leak rate testing (LLRT) did not reveal these leaks since the packing was not included in the test boundary. The CAC valves' configuration was modified (Mod 5730) so that the LLRT would identify any packing leaks in the future. :n addition, Mod 5730 reviewed all inboard PCIVs outside containment and modified valve configurations if necessary to ensure that the packing would be subject to local leak rate test pressure.
The CAC valves were repaired, which brought leakage rates down significantly and tha Type A test was started again and successfully passed with an as-left leakage of .178 % wt/ day. The investigation following the test failure confirmed that the initial Type A test failure was an activity-based failure and not a time-based failure.
The results of the second post-operational, Type A test performed in November 1990 was .334
% wt/ day.
The test results of the three tests do not indicate any trend of containment structure degradation, since the Type A test results over the first two intervals went down and then went back up. The majority of leakage can be attributed to Type B and C penetrations since the containment structure was not altered in any fashion, and this is consistent with the industry data presented by NUMARC at the NRC Workshop Session on Appendix J Containment Integrated Testing held April 27,1993. In addition, Draft NUREG 1493,
- Performance-Based l Containment Leak Test Program
- shows that industry experience indicates that over 95% of the l failures associated with Type A tests are found to be due to Type B and C tested penetrations. 1 Therefore, continu9d overall leak-tightness of the active containment components can be l assured by the LGS, Type B and C testing program. l l
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Risk imoact Assessment l The proposed exemption introduces the possibility that primary containment leakage in excess l of the allowable would go undetected during this proposed 39 month extension of the Type A test interval. From a risk standpoint, the purpose of the Appendix J testing is to determine that the total leakage from primary containment does not exceed the maximum allowable leakage rate, La, as specified in the Limerick Generating Station (LGS) Facility Operating License. This ;
primary containment maximum allowable leakage rate provides an input assumption to the '
calculation required to ensure that the maximum potential offsite dose during a design basis accident, defined in Section 15.6.5 of the LGS Updated Final Safety Analysis Report (UFSAR),
does not result in a dose in excess of that specified in 10 CFR 100. Such leakages could occur through containment penetrations, airlocks, or containment structural faults. The containment structure is passive. Under normal operating conditions, there is no significant environmental or operational stress present that could contribute to its degradation.
Postulated containment failures under severe accident conditions are primarily due to effects associated with severe accidents. Such effects were considered as part of the LGS Individual l Plant Examination (IPE) Volume 2, dated July 1992 (Doc 50 352/50-353). Although the review concluded that the consequences of an accident are not increased, the LGS Units 1 and 2 IPE was reviewed to assess the impact of exceeding the allowable leakage rate, if for some reason a non-mechanistic type of failure were to occur. The IPE evaluated the effect of various j containment leakage sizes under different scenarios. The IPE concluded that a containment l leakage rate of 35 % wt/ day would represent less than a 5 % increase in risk to the public of l being exposed to radiation. This analysis was based on a study performed at Oak Ridge National Laboratory which evaluated the impact of leakage rates on public risk. As stated earlier, the current LGS La is 0.5 % wt/ day which is significantly less than the 35 % wt/ day described in the IPE.
Based on information provided in draft NUREG -1493, the increased risk of population dose attributed to extending the Type A test interval would be extremely small. Draft NUREG-1493 includes the results of a sensitivity study performed to explore the risk impact of several alternative leak rate test schedules. Alternative 4 from this study examines relaxing the Type A testing frequency from 3 tests in ten years to 1 test in ten years. The draft NUREG concludes that the increase in population exposure risk to those in the vicinity of the representative plants ranged from 0.02% to 0.14%. This very low impact on risk is attributable to; the effectiveness of Type B and C tests in identifying potential leak paths (over 95%); a low likelihood of Type A test identified leakages in excess of 2 times the allowable; and the insensitivity of risk to containment leak rate (i e., no discemible increase in population dose risk with containment leak rates 100 times greater than currently allowed). The draft NUREG conclusion is identified as "an imperceptible increase in risk."
Based on a review of activities, we have concluded that there have not been any alterations or challenges to the Unit 1 primary containment since the last Type A test, nor will there be any future maintenance activities during the proposed extended test interval that will adversely affect Unit 1 primary containment leakage rates without implementation of strict administrative controls that require the performance of an individual local leak rate test.
The above review found that the risk of a non-detectable increase of Unit 1 primary containment leakage is considered to be negligible due to the conclusion that the LGS Type B and C Testing Program has been effective in identifying most of the leakage and the testing program will continue to be performed through-out the proposed extended Type A test interval. The allowable leakage for Type B and C testing is 0.6 La (0.300 % wt/ day) which when compared to the Type A test acceptance criteria 0.75 La (0.375 % wt/ day) is a significant amount of the allowable leakage. Finally, the review assessed the results of previous Type A tests and we concluded that the only failure mechanisms which have been detected during the past tests are activity-based (active) and that there is no indication of time-dependent (passive) failures that
W Page 10 of 13 would not be identified during the performance cf Type B and Type C testing. Therefore, we have concluded that the proposed extended test interval would not result in a non-detectable Unit 1 primary containment leakage rate in excess of the LGS allowable value (i.e.,0.5% wt/ day) established by 10 CFR 50, Appendix J.
Based on the above technical justification we request a one-time exemption of the requirements of 10 CFR 50, Appendix J, section Ill.D.1(a), in accordance with three of the criteria of 10 CFR 50.12.
pasis For Exemption in accordance with 10 CFR 50.12(a)(1) the granting of this requested exemption will not present undue risk to the health and safety of the public and is consistent with the common defense and security.
The NRC may, upon application, grant exemptions from the requirements of 10 CFR 50, where special circumstances are present. 10 CFR 50.12(a)(2)(li) defines such a circumstance where, ' Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule;.." The undertying purpose of 10 CFR 50, Appendix J, section lil.D.1(a) is to establish and maintain a level of confidence that any primary containment leakage, during a hypothetical design basis accident, will remain less than or equal to the maximum allowable value, La, established by 10 CFR 50, Appendix J, by performing periodic Type A testing. The current approved schedule requires that the third, post-operational (or fourth), Unit 1, Type A test be performed during the sixth refueling outage scheduled for January 1996. This is not necessary to achieve the underlying purpose of the rule, since there is technical justification based on testing history and the structural capability of the containment, and there will be a continual presence of a comprehensive LGS Type B and C test and maintenance program. The above technical justification supports the conclusion that the requested schedular exemption to defer the third, post-operational, Unit 1, Type A test, an additional 27 months (until the seventh. Unit 1 refueling outage) will maintain the same level of confidence that any Unit 1 primary containment leakage will remain less than or equal to the maximum allowable leakage rate value, La, during the proposed extension.
10 CFR 50.12(a)(2)(lii) states that the NRC may grant exemptions from requirements of 10 CFR 50 where, "Compilance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated;.." The current LGS Unit 1, Type A test schedule established by 10 CFR 50, Appendix J, section Ill.D.1.a and modified through exemptions will require that the Type A test be performed during the sixth refueling outage scheduled for January 1996. This current test schedule will result in unnecessary additional personnel radiation exposure in order to perform the test and unnecessary Gustified above) costs associated with performing the test and an increase in the refueling outage length of approximately two days.
10 CFR 50.12(a)(2)(v) states that the NRC may grant exemptions from requirements of 10 CFR 50 where, "The exemption would provide only temporary relief from the applicable regulation and the licensee or
- applicant has made good faith efforts to comply with the regulation..." The requested exemption would provide only temporary reflef pertaining only to the third, post-operational, Unit 1, Type A test.
INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT An Environmental Assessment is not required for the Technical Specifications changes proposed by this request because the requested changes to the LGS, Units 1 and 2, TS confirm to the criteria for
' actions eligible for categorical exclusion," as specified in 10 CFR 51.22(c)(9). The requested TS changes wl!I have no impact on the environment. The proposed TS changes do not involve a Significant Hazards Consideration as discussed in the preceding Safety Assessment section. The proposed ,
- changes do not involve a significant change in the types or significant increase in the amounts of any l effluent that may be released offsite. In addition, the proposed TS changes do not involve a significant increase in Individual or cumulative occupational radiation exposure.
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With respect to the requested exemption for LGS, Unit 1, the following information is provided to )
support an Environmental Assessment. l l
Identification of Proposed Action
\
l The proposed action is to grant an exemption from 10 CFR 50, Appendix J, Section Ill.D.1.(a) which requires a set of three Type A tests (i.e., Containment Integrated Leakage Rate Test (CILRT)) to be j performed at approximately equal intervals during each 10-year service period and specifies that the '
third test of each set shall be conducted when the plant is shut down for the performance of the 10-year plant inservice inspection (ISI). This one-time exemption would allow the third, post-operational, Unit 1, Type A test to be performed during the seventh Unit I refueling outage scheduled to begin no later than April 1998, approximately 89 months after the last Unit 1 test. Furthermore, this one-time exemption
, would result in the third Type A test not being performed at an interval approximately equal to previous
! intervals (i.e., approximately 40 months) during the first 10-year service period, the third Unit 1 Type A l test would be performed beyond (i.e.,44 months) the end of the first 10-year service period, and the Type A test would not coincide with the 10-year ISI outage.
l The Need for the Proposed Action l The requested exemption is needed because the requirements of 10 CFR 50, Appendix J, and l the current Unit 1 Type A test schedule, would require that the 10 CFR 50, Appendix J, Type A test be performed during the sixth refueling outage. By exempting the test from its current frequency, PECO Energy would be able to obtain outage schedule benefits resulting in cost savings and reduced worker i radiation exposure in anticipation of the performance based rulemaking (Appendix J, option B) without being unduly penalized due to the LGS fuel cycle schedule. The rulemaking is part of the NRC initiative to eliminate requirements that are marginal to safety.
EnvironmentalImpacts of the Proposed Action The requested exemption would not significantly increase the probability of exceeding the maximum allowable value of expected primary containment leakage (i.e., La. established by 10 CFR 50, Appendix J), during a hypothetical design basis accident; therefore, the primary containment integrity would be maintained. Although the requirements in 10 CFR 50, Appendix J, section Ill.D.1(a) state that three Type A tests shall be performed in each 10-year service period and at approximately equal intervals during that service period, we have concluded that performing the third, post-operational, Type A test of first 10-year service period approximately 89 months after the second Type A test, approximately 44 months after the end of the first 10-year service period, and approximately 27 months beyond the 10-year ISI outage, would meet the underlying purpose of the rule, and that any primary containment leakage during a hypothetical design basis accident will remain less than the maximum allowable leakage rate value, La, established by 10 CFR 50, Appendix J. This determination was made based on the following assessment.
There is the potential that containment degradation could remain undetected during the proposed survelliance Interval extension and result in the containment leakage exceeding the allowable value assumed in the safety analysis. However, the potential prirnary containment degradation mechanisms due to activity-based and time-dependent causes have been reviewed. This review l concluded that there has not been any alterations or challenges to primary containment since the last Type A test, nor will there be any future modification or maintenance activities during the proposed extended test interval which will adversely affect primary containment leakage ratos without implementation of strict administrative controls that require the performance of individual local leak rate testing. This review found that the risk of a non-detectable increase of primary containment leakage is I considered to be neg!!gible due to the conclusion that 10 CFR 50, Appendix J Type B and Type C i testing will identify most of the containment leakage and the comprehensive LGS Type B and Type C l testing and maintenance programs will continue to be conducted through-out the proposed extended i test interval. There are no physical or operational changes to any plant equipment associated wkh the l
t
1 Page 12 of 13 proposed exemption to increase the surveillance interval of the Type A test. Therefore, the probability of a malfunction of equipment is not increased. Finally, this review assessed the results of previous Unit 1 Type A test results and concluded that the only failure mechanisms which have been detected during the past Type A tests are activity-based and that there is no indication of time-dependent failures that would not be identified during the performance of Type B and Type C tests. Therefore, we have concluded that the proposed extended test interval would not result in a non-detectable Unit 1 primary containment leakage rate in excess of the allowable value (i.e.,0.5% wt/ day) established by 10 CFR 50, Appendix J.
Accordingly, the consequences of an accident would not be increased, that is, the post-accident radiological releases would not be greater than previously determined. The requested exemption would i not affect plant radiological effluents. Therefore, there are no significant radiological environmental Impacts associated with the requested exemption. With regard to potential non-radiological impacts, the l requested exemption involves a one-time schedular change to surveillance and testing requirements. It !
does not affect non-radiological plant effluents and has no other environmental impact. l Alternative to the Proposed Action l Since we have concluded that there is no significant environmental impact associated with the requested exemption, any alternatives would have either no or greater environmental Impact.
The principal attemative would be to deny the requested exemption which would require the !
performance of a 10 CFR 50, Appendix J, Type A test during the Unit 1, sixth refueling outage. This would not reduce the environmental impact attributed to the facility as compared to the impact of granting the requested exemption.
Alternative Use of Resources This proposed exemption does not involve the use of any resources not previously considered in connection with the Nuclear Regulatory Commission's Final Environmental Statement dated, April 1984, ,
related to the operation of the Umerick Generating Station, Unit 1 and Unit 2. ll Information Supporting a Finding of No Significant impact We have concluded, based on the preceding environmental assessment, that the proposed action will not have a significant effect on the quality of the human environment; therefore, an environmental Impact statement for the requested exemption would not be required.
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Page 13 of 13 EXHIBIT A 1994 LGS Annual Summary Assessment Report Excerpts V. Analysis of Barriers to Fission Product Release and Plant Health Indicators Hiah Pressure LLRT Leakaae !
M M M M M Unit 1 (gpm) 0.78 0.78 4.88 1.03 1.20 Unit-2 (gpm) 1.34 1.00 2.67 0.54 0.54 I
This indicator quantifies the total leakage through the 27, high-low pressure interface valves on each unit.
In 1994, the total measured leakage rate on Unit-1 was slightly higher than in 1993. The Unit-2 leakage rate I remained the same as last year's value, because none of the valves were worked or required testing. The l total quantity of this indicator is very low compared to the Tech. Spec. requirement of 27 gpm/ unit. This l barrier to fission product release continues to be well maintained.
Percentaae of all As-Found LLRTs that Fall M M M M M 6.5% 8.0% 3.2% 2.5% 2.9%
Tne percentage of as-found test fallures remains consistently low, and within a normal expected band based on Umerick experience.
Total Intearated Leak Rate. Tvoe B & C Tests M M M M M Unit-1 .35L, .45L, .41L, .40L, .39L, Unit 2 .43L, .36L, .40L, .27L, .26L, Performance of the primey containment isolation valve barrier, as measured by local leak rate tests, continues to remain at a satidactory level. Improvement was made on both units, due to corrective valve maintenance decreasing overalt valve leakage, thus increasing the effectiveness of this barrier to fission product release. The total integrated leak rate is being successfully managed below the 0.60L, Technical Specification action ilmit.