View source for NL-04-1001, Edwin 1. Hatch Nuclear Plant Unit 1 and 2 - Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves

{{Adams
| number = ML042390316
| issue date = 08/23/2004
| title = Edwin 1. Hatch Nuclear Plant Unit 1 and 2 - Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves
| author name = Sumner H
| author affiliation = Southern Nuclear Operating Co, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000321, 05000366
| license number = DPR-057, NPF-005
| contact person = 
| case reference number = NL-04-1001
| document type = Letter type:NL, Technical Specification, Amendment
| page count = 25
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter:H. L Sumner, Jr.          Southern Nuclear Vice President            Operating Company. Inc.
Hatch Project              Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERNh COMPANY August 23, 2004                                                            Energy to Serve Your World' Docket Nos.:       50-321                                                   NL-04-1001 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin 1. Hatch Nuclear Plant Unit I and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Ladies and Gentlemen:
In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby proposes to amend the Plant Hatch Units I and 2 Technical Specifications, Appendix A to operating licenses DPR-57 and NPF-5, respectively. The proposed amendment revises the Surveillance Requirements for certain containment purge valves. The amendment request would replace requirements for valve seat replacement with a requirement to perform an Appendix J leakage rate test of the valves at a frequency of at least once every 30 months. This amendment request is based on a comprehensive analysis performed by the BWR Owners' Group (BWROG) which concluded that extensive testing of large containment isolation valves using resilient seats has revealed no indication of aging induced seat degradation.
Enclosure I provides a description and justification for the proposed change, a figure of the containment vent and purge valve configuration and two tables of containment valve leak rate test data. Enclosure 2 provides the significant hazards evaluation as well as the environmental evaluation. Enclosure 3 contains the marked up and published Technical Specifications and Bases pages.
Precedent licensing submittals have been approved by NRC for Southern Nuclear Operating Company, Vogtle Electric Generating Plant; Entergy Operations Inc., Grand Gulf Nuclear Station; and Wisconsin Electric Power Company, Point Beach Plant. These precedents are further discussed in Enclosure 1.
SNC requests approval of the proposed license amendments by September 1, 2005. The proposed changes would be implemented within 60 days of issuance of the amendment.
(Affirmation and signature are on the following page).
710 l a;

U. S. Nuclear Regulatory Commission NL-04-1001 Page 2 Mr. H. L. Sumner, Jr. states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.
This letter contains no NRC commitments. If you have any questions, please advise.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY I 7~ 2"          C>14441~
H. L. Sumner, Jr.
Swrn to andsub                              _3 day of
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==Enclosures:==
: 1. Description and justification of change
: 2. No Significant Hazards and Environmental Evaluation
: 3. Changed Technical Specifications and Bases Pages cc:    Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch RType: CHA02.004 U. S. Nuclear Regulatorv Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch State of Georgia Mr. L. C. Barrett, Commissioner - Department of Natural Resources

Enclosure 1 Edwin I. Hatch Nuclear Plant, Units I and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change

Enclosure 1 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change Description The proposal would revise Technical Specifications for containment purge valves with resilient seats, by eliminating the Unit I SR 3.6.1.3.11 and Unit 2 SR 3.6.1.3.12 requirements for replacing the valve seats.
Instead, an Appendix J Leakage Rate Test would be performed every 30 months (i.e., every refueling outage). This change is based on an analysis performed by the BWR Owners' Group (BWROG) which concluded that extensive testing of large containment isolation valves using resilient seats has revealed no indication of aging induced seat degradation. The Plant Hatch Unit 1 and 2 specific data and the BWROG generic data are included in the Justification.
Justification

===Background===
The Primary Containment is vented to maintain pressure during reactor heat-up and normal operations.
The containment atmosphere is also purged in preparation for personnel entries. The 18 inch purge valves discussed in Unit 1 SR 3.6.1.3.11 and Unit 2 SR 3.6.1.3.12 are the containment isolation valves for the vent and purge system. There are two fast-acting air operated butterfly valves in series for each penetration (Figure 1). These valves are normally closed and fail closed. They may be opened for inerting, de-inerting, venting and purging the containment atmosphere. The 18 inch vent lines exhaust to the Standby Gas Treatment system and include a two inch bypass for pressure control during normal operations (Reference 1).
The Technical Specifications currently require that the resilient seats in the primary containment purge valves be replaced every 24 months to provide additional assurance of operability. This requirement derives from earlier concerns that this type of seat had the potential to degrade in a shorter time period than other seat types. This concern was based on the importance of maintaining the penetration leak tight (due to the direct path between primary containment and the environment).
As a result of reports of unsatisfactory performance of resilient seats in butterfly-type isolation valves due to seat deterioration, the Nuclear Regulatory Commission (NRC) established Generic Issue B-20 "Containment Leakage Due to Seal Deterioration" (Reference 2) to study this issue and propose a regulatory resolution of the problem. IE Circular 77-11, "Leakage of Containment Isolation Valves with Resilient Seals," (Reference 3) reported difficulty with satisfying leak test requirements on large containment isolation butterfly valves with resilient seats. IE Circular 77-11 reported that: "The cause of the excessive leakage has been determined to be either general degradation of the resiliency characteristic of the seal, cold temperatures and the associated hardening of the seal, or a combination of the two," and that ". . .examination of the resilient valve seat material indicated that the material had hardened and lost some resiliency and showed signs of wear due to valve cycling. Exposure to various environmental conditions such as humidity and temperature have also, in some cases, apparently accelerated the degradation or changed the performance characteristics of the seating material." Among the actions taken by licensees at that time, to assure continued nearly leak-tight behavior, was to reduce the time between leakage tests.
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Enclosure I Edwin I. Hatch Nuclear Plant, Units I and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change In response to NRC concerns about containment leakage due to seat deterioration, Hatch committed to replace the resilient seats in the 18 inch vent and purge valves every refueling outage (Reference 4). The NRC found that the seat replacement would assure early detection of aging deterioration of the resilient seat material (Reference 5).
Precedent The proposed change is consistent with several NRC approved precedent submittals:
On December 27, 1994, SNC 's Vogtle Electric Generating Plant (VEGP) Unit I and Unit 2 requested an amendment to the leak test frequency for containment purge valves with resilient seats (Reference 6).
During the period 1987 to 1994 these valves had been routinely tested every three months and no indication of seat degradation had been observed. Based on these results, SNC concluded that seat degradation does not occur and that the history of the valves supported a test interval, in accordance with 10 CFR 50 Appendix J, of once per refueling outage (24 months).
On July 7, 1995, the NRC issued an amendment revising the VEGP Units I and 2 frequency of conducting leak testing of containment purge valves with seats made of resilient material, from every three months to each refueling outage (Reference 7). In its Safety Evaluation for Vogtle, the NRC staff states that: "Operating experience has shown that for well maintained butterfly valves with resilient seats, used at suitable environmental and operating conditions the 24-month Appendix J leakage rate test interval is sufficiently frequent. Accordingly the staff will approve a reduced testing frequency if supported by plant specific data (i.e., history of test results)."
On September 19, 1996, Entergy Operations Inc. submitted a revision (Reference 8), to a proposed amendment (References 9 and 10) to the Grand Gulf Nuclear Station (GGNS) Operating License to perform leakage rate testing for each primary containment purge valve with resilient seats at a frequency of 36 months with at least 2 pairs (of 4 pairs) of valves tested every 18 months; and once within 92 days test all remaining purge valves, if any purge valve fails to meet its acceptance criteria. During a period of 14 years the valves had been routinely tested on an augmented testing frequency (every 184 days and once within 92 days after opening a valve) with no indication of seat degradation.
On October 18, 1996, the NRC issued an amendment revising the GGNS Unit I Operating License frequency of leakage rate testing for each containment purge isolation valve with resilient seats (Reference 11). In the Safety Evaluation for Entergy's Grand Gulf Nuclear Station Unit 1 it was stated that: "The NRC has recently approved changes in testing frequencies for containment purge valves with resilient seats at the Vogtle Electric Generating Plant (Docket Nos. 50-424 and 50-425)." Leakage rate testing of primary containment purge valves with resilient seats is now required every 36 months with at least two pairs of valves tested every 18 months. If any purge valve fails to meet its acceptance criteria, all remaining purge valves must be tested within 92 days, (unless previously tested within 92 days of the failure).
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Enclosure 1 Edwin I. Hatch Nuclear Plant, Units 1and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change By application dated May 29, 1996, (Reference 12), and supplemented by a letter dated August 20, 1996 (Reference 8) Wisconsin Electrical Power Company's Point Beach Plant (PBNP) Units 1 and 2 requested allowance for an up to 30 month test interval for containment purge supply and exhaust valves using resilient seats. The previous test interval required testing at six (6) month intervals. Leakage test results and the maintenance history of the valves from 1992 to 1996 supported extension of the test interval to up to 30 months.
In the letter dated August 20, 1996, (Reference 13) PBNP stated that: "We have reviewed leakage test results and the maintenance history of these valves from 1992 to present in order to assess the reliability of the valves and support extension of the test interval to up to 30 months. Of the 36 leakage tests performed, nine per penetration, there have been no failures when compared to Technical Specification and Appendix J limits. In addition no valve has exceeded our administrative leakage limit of 2000 sccm."
The NRC issued an amendment to the PBNP Units I and 2 operating licenses, to permit performance of leakage rate tests on containment purge supply and exhaust valves at a 30-month interval (Reference 14).
Other Regulatory Documents NUREG-1493 "Performance-Based Containment Leak-Test Program," September, 1995 (Reference 15) evaluated the impact of changing the frequency of Appendix J tests on overall reactor risk. The report concluded that the frequency of Type C Tests (Valve LLRT) could be reduced without significant impact on reactor accident risk. In addition, the report found that increasing the allowable leakage rate by one or two orders of magnitude would not significantly impact the estimates for population doses in the event of an accident.
The NRC, in Regulatory Guide 1.163, approved 10 CFR 50 Appendix J, Option B and referenced NEI 94-01 "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J" as to methods acceptable to the NRC staff for compliance and implementation.
NEI 94-01 set the rules for establishing test frequencies for Type C tests of valves. Under 10 CFR 50 Appendix J, valves were required to be leak tested every refueling outage. NEI 94-01 permitted the test frequency to be extended up to 120 months based on good performance. Good performance was defined as two consecutive successful as-found leak tests performed with an elapsed time between the first and second test of at least 24 months or nominal test interval.
In Regulatory Guide 1.163, the NRC took exception to the 120 months test frequency and stated that any test frequency beyond 60 months would first require NRC approval. Additionally, the NRC stated that the frequency for main steam and feedwater isolation valves in BWR's and containment purge and vent valves in PWR's and BWR's, should be limited to 30 months, with consideration given to operating experience and safety significance.
The NRC also referenced ANSI/ANS 56.8-1994. Section 3.3.4 in their discussion on the leak testing frequency of purge and vent valves. ANSI/ANS 56.8-1994, Section 3.3.4, gives a test frequency of 30 months plus float.
Under Option B, if a valve fails its as-found test it shall be tested at a frequency of at least once per 30 months until good performance has been established. In the case of purge and vent valves the frequency would stay at 30 months.
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Enclosure I Edwin I. Hatch Nuclear Plant, Units I and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change In summary, 10 CFR 50 Appendix J, Option B presently allows purge and vent valves to be tested at an interval of up to once every 30 months. If a purge or vent valve were to fail its as-found leak test, the testing periodicity would remain the same at every 30 months.
The 30 month frequency proposed by this license amendment request matches the frequency specified in 10 CFR 50, Appendix J, Option B.
The BWROG has performed an analysis based on extensive historical test data and maintenance information (summarized in Technical Analysis). The analysis identified no documentation that attributed test failures to aging degradation of the resiliency characteristics of the seat, consistent with the precedents cited above.
Technical Analysis Generic Data A review of industry operating experience for containment isolation butterfly valves with resilient seats demonstrates that increased testing is not required on a permanent basis, and that the history of these valves supports a test interval in accordance with 10 CFR 50, Appendix J, Option B, i.e.
up to a 30 month test interval. A survey was performed by the BWROG to identify performance history for the containment purge valves (butterfly valves) with resilient seats. Table 1 identifies leakage test history at 7 different plants. A total of 2457 leak tests have been performed. The Table identifies the causes of leak test failures.
The data in Table I identified a failure rate of 4.4E-2/test. This data is bounded by the failure data discussed in NUREG-1493 (Reference 10). Table 4-7 of NUREG - 1493 indicates that for 992 Butterfly valves leak tests, there were 93 tests where leakage exceeded administrative limits. This is a failure rate of 9.4 E-2/test.
The BWROG data indicates that many of the test failures cannot be attributed to any specific condition. Of the test failures, 26% (28/107) were related to the seats and 27% (29/107) have unknown causes (documentation does not identify a cause for the test failure). Thus only 1.IE-2 failures per test (28/2457) are known to involve the seat. Of these seat related test failures, none can be definitively attributed to aging induced seat degradation.
Additionally, Appendix A of NUREG-1493 analyzed in detail the plant specific history of all containment isolation valve (Type C) tests at the Grand Gulf and North Anna stations. The analysis revealed that the failure rate in Type C tests was dominated by the time since the last maintenance was performed on the valve. The random failure rate was found to be 3.8 to 6.5 percent. However the failure rate within two outages of valve repairs was determined to be 31 to 34 percent. These failure rates were evaluated in the NUREG and found to have little increase in overall accident risk when test frequencies are extended.
The failure rate for the BWROG data in Table I is bounded by the NUREG-1493 evaluated random failure rate of 6.5 % for all seven plants except Plants 3 and 7. The Plant 3 data is based on new as-found leak rate tests after the valves were replaced in the mid-1990's. Seven of the ten LLRT failures occurred in the first refueling outage after the valve replacement. Neglecting E14

Enclosure I Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change the 16 tests from the first outage, Plant 3 experienced a failure rate of 5.4% (3 failures in 56 tests),
which is also bounded by the random failure rate evaluated in NUREG-1493.
The Plant 7 failure rate is 17.7%. Plant Hatch is Plant 7 and is discussed further under Plant Specific Data.
As was previously concluded in NUREG-1493, there is no apparent correlation between the size or manufacturer of the valve and its failure history.
Therefore, the review of BWROG experience supports the conclusion that failures by containment isolation valves, such as the containment purge valves addressed herein, do not warrant additional actions beyond the Appendix J, Option B test frequency. The data also indicates that maintenance activities actually increase the likelihood of LLRT failure.
Plant Specific Data Table 2 identifies leakage test history at Hatch for 22 similar design, resilient seat, butterfly valves. A total of 385 As-Found leakage tests have been performed with 68 tests failing to meet the administrative leakage limit. The Table identifies the causes of the 68 leak test failures. The data in Table 2 identifies a failure rate of 17.7%.
Table 2 does illustrate that 23 seat related failures and 19 failures with an unknown cause are included in the total of 68 as-found LLRT failures. However, Hatch has been replacing the valve seats on a 18 to 48 month interval for most of these valves since the mid-1980's. A review of the Hatch work order records did not identify any failures that were attributable to aging induced seat degradation.
Thus, the Hatch data is different from the other BWR experience provided in Table 1 in that the routine seat replacement interval is much shorter for Hatch than the other plants (18-48 months vs. 72-120+ months) and the failure rate is much greater (17.7% versus 0.6% to 5.4%). Therefore, the increased maintenance frequency associated with seat replacements at Plant Hatch correlates to an increased as-found LLRT failure rate.
Appendix A of NUREG-1493 documented the same result for valves which have undergone maintenance in the previous two operating cycles. The Hatch data is bounded by the failure data discussed in NUREG-1493 (i.e., random failure rate 6.5% and post-maintenance failure rate of 34%).
The history of as-found LLRT failures at Plant Hatch have resulted in some changes to the valve seat design. The original seat ring was manufactured with a glued joint from EPT material. The glued joint was replaced with a continuous seat ring. Additionally, Hatch has changed the seat material from EPT to EPDM, which has an expected life in this application of 60 years.
Therefore, the as-found LLRT data at Hatch, the as-found LLRT data at 6 other BWR's, and the material life evaluation support the conclusion that the NRC concern for seat material aging E1-5

Enclosure I Edwin I. Hatch Nuclear Plant, Units I and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change degradation does not warrant preemptive seat replacement, which is currently required every outage at Plant Hatch.
Technical Analysis Conclusion The technical justification for revising the surveillance frequency of containment purge valves with resilient seats is based on the results from a total of 2457 tests conducted from 1973 to 2003.
This data validates the frequency specified by I OCFR50 Appendix J, Option B and exhibits that the additional actions to assure function based on IE Circular 77-11 are not warranted. The Plant Hatch historical data is consistent with the data previously evaluated by the NRC.
This operating experience has shown that for well maintained butterfly valves with resilient seats, used at stable environmental and operating conditions, the 30-month Appendix J leakage rate test interval is sufficiently frequent.
References
: 1. Hatch Unit I UFSAR Section 5.2 and Unit 2 UFSAR Section 6.2 Revision 2 1.
: 2. Generic Issue B-20 "Containment Leakage Due to Seal Deterioration"
: 3. IE Circular 77-11, "Leakage of Containment Isolation Valves with Resilient Seals," dated September 6, 1977
: 4. Letter NED-83-334, from J. T. Beckham, Jr., GPC, to John Stolz, NRC, dated June 15, 1983
: 5. Letter 85-974, from John Stolz, NRC, to J. T. Beckham, GPC, dated January 16, 1984
: 6. Letter, LCV-0483, Docket Nos. 50-424, 50-425, C. K. McCoy, Georgia Power Company, to USNRC, "Vogtle Electric Generating Plant Request to Revise Technical Specification, Revision to Leak Test Frequency for Valves with Resilient Seals," dated December 27, 1994.
: 7. Letter, L. L. Wheeler, USNRC to C. K. McCoy, Georgia Power Company, "Issuance of Amendments, Vogtle Electric Generating Plant Units I and 2, (TAC Nos. M91240 and M91241)," dated July 7, 1995.
: 8. Letter, GNRO 96/00108, Docket No. 50-416, J. J. Hagan, Entergy Operations Inc., to USNRC, "Revision to Proposed Amendment to the Operating License," dated September 19, 1996.
: 9. Letter, GNRO 96/0051, Docket No. 50-415, C. R. Hutchinson, Entergy Operations Inc., to USNRC, "Grand Gulf Nuclear Station, Purge Valve Testing, Proposed Amendment to Operating License," dated May 8, 1996.
: 10. Letter, GNRO 96/0084, Docket No. 50-416, C. R. Hutchinson, Entergy Operations Inc., to USNRC, "Purge Valve Testing, Questions Concerning Proposed Amendment to the Operating License," dated July 18, 1996.
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Enclosure 1 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Description and Justification of Change
: 11. Letter, J. N. Donohew, USNRC to J. J. Hagan, Entergy Operations Inc., "Issuance of Amendment No. 128 to Facility Operating License No. NPF-29, Grand Gulf Nuclear Station, Unit 1, (TAC No. M95338)," dated October 18, 1996.
: 12. Letter, VPNPD-96-034, B. Link, Wisconsin Electric Power Company to USNRC, "Dockets 50-266 and 50-301, Technical Specification Change Request 187, Adoption of 10 CFR 50, Appendix J, Option B, Point Beach Nuclear Plants Units 1 and 2," dated May 29, 1996.
: 13. Letter, VPNPD-96-056, B. Link, Wisconsin Electric Power Company to USNRC, "Dockets 50-256 and 50-301, Supplemental Information, Technical Specifications Change Request 187, Adoption of 10 CFR 50 Appendix J, Option B, Point Beach Nuclear Plants, Units I and 2," dated August 20, 1996.
: 14. Letter, A. G. Hansen, USNRC to R. E. Link, Wisconsin Electric Power Company, "Amendment Nos. 169 and 173 to Facility Operating License Nos. DPR-24 and DPR-27, Point Beach Nuclear Plant Unit Nos. I and 2, (TAC Nos. M95668 and M95669), dated October, 1996.
: 15. NUREG-1493 "Performance-Based Containment Leak-Test Program," September, 1995.
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Enclosure 1 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Figure 1 341       34 1 8" 319       32 PURGE                                                                2 AIRN    308      307                                                    33 FRBM      ST                                  2                            3 RX     GVPR   M/U~o
                  /                                              318       32 6 VACUUM    BREAKERS E1-8

Enclosure 1 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Table 1 Failure Tvnes Plant     Total # of Total Years of Total # of Periodic Seat      1          2          3         4       5 6   7 Tests   Valve Service Failures Replacement Interval 1          726          221          4          None           0          0          0         0       0 4   0 2          123          192          3            year0                   0          0         0       0 0   3 3           72          112         10          None           2          0          0         6       0 2   0 4          874          336         17          6 year         1          5          0         0       2 3   6 5          197           48          4          None           0          0          0         4       0 0   0 6           80           24          1          None           0          0          0         0       0 0   1 7          385          550         68         1.5 year        0        23           18        3       0 5  19 TOTAL         2457          1483        107                        3        28           18        13      2 14 29 Failures per Test 107/2457       4.4E-02 Failure Type:
: 1. ACTUATOR
: 2. SEAT DAMAGE
: 3. PACKING
: 4. DEBRIS
: 5. VALVE STROKING
: 6. OTHER (non-Seat related)
: 7. UNIDENTIFIED CAUSE El-9

Enclosure 1 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Table 2 VALVE I TEST DATE NUMBER      NUMBER SIZE     RANGE    OF TESTS OF FAILURES      % FAILURES                                      Comments 6"    1977-2002     18         3                            One seat failure, two O-ring failures 18"    1977-2002     18         2                            One seat failure, one unknown cause 18"    1977-2002     18         1                            Seat failure 18"    1977-2002     18         4                            Two seat failures, one O-ring failure, one unknown cause 20"    1977-2002     18         1                            One unknown cause 20"    1977-2002     18         3                            One seat failure, debris in seat, one unknown cause 18"    1977-2002     18         0                            None 18"    1977-2002     18         9                            Three seat failures, two O-ring failures, four unknown causes 18"    1977-2002     18         4                            One O-ring failure, one seat adjustment, two unknown causes 18"    1977-2002     18         3                            Two seat failures, one unknown cause 18"    1977-2002     18         4                            One debris in seat, one disk mis-alignment, two unknown causes 6"     1980-2003     17         4                            Three seat failures, one seat adjustment 18"    1980-2003     17         3                            Two seat failures, one O-ring failure 18"    1980-2003     17         0                            None 18"    1980-2003     17         4                            One seat failure, two O-ring failures, one unknown cause 20"    1980-2003     17         4                            One seat failure, three O-ring failures 20"    1980-2003     17         2                            Two seat failures Four O-ring failures, one corrosion in seat grove, one unknown 18"    1980-2003     17         6                            cause 18"    1980-2003     17         3                            Two seat failures, one unknown cause 18"    1980-2003     17        2                             Two disk mis-alignment 18"    1980-2003     17        3                             One seat failure, one O-ring failures, one unknown cause 18"    1980-2003     17        3                             One O-ring failure, two unknown causes Totals     385        68              17.7%
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Enclosure 2 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves No Significant Hazards and Environmental Evaluation

Enclosure 2 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves No Significant Hazards and Environmental Evaluation In 10 CFR 50.92, the NRC provides the following standards to be used in determining the existence of a significant hazards consideration:
... a proposedamendment to an operatingfacilitylicensed under 50.21(b) or 50.22 orfor a testingfacility involves no significanthazards consideration, if operationof the facility in accordance with the proposed amendment would not: (1) involve a significant increase in theprobabilityor consequences of an accidentpreviously evaluated; or (2) create the possibility of a new or different kind of accidentfrom any previously evaluated; or (3) involve a significantreduction in the margin of safety.
Southern Nuclear Operating Company (SNC) has reviewed the proposed licensing amendment and concluded that the change does not involve a significant hazards consideration because (1) the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated, (2) the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated, and (3) does not involve a significant reduction in the margin of safety.
Basis for no Significant Hazards Consideration
: 1. The proposedchange does not involve a significant increase in theprobabilityor consequences of an accidentpreviously evaluated.
The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seats. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the extensive industry operating experience derived from test results has demonstrated that the resilient seat material does not experience aging degradation and cause containment isolation valves to leak. Thus, the valves will perform as assumed in the accident analyses and therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated. Further, these valves are not accident initiators, and therefore, this change does not involve a significant increase in the probability of occurrence of a previously evaluated event.
: 2. The proposed change does not create the possibility of a new or different kind of accidentfrom any previously evaluated.
The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seats. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed nor changes in methods governing normal plant operation). In particular, it does not require the valves to function in any manner other than that which is currently required. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Enclosure 2 Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves No Significant Hazards and Environmental Evaluation Basis for no Significant Hazards Consideration (continued)
: 3. The proposedchange does not involve a significant reductionin the margin of safety.
The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seats. The proposed change does not involve a significant reduction in margin of safety because it has no effect on any safety analysis bases or assumptions. It does not change the leakage acceptance criteria. Sufficient data has been collected to demonstrate that resilient seats do not experience aging degradation. Deleting the seat replacement requirement will not reduce the margin of safety provided by Technical Specifications.
For the above reasons, the margin of safety is not reduced by this proposed Technical Specifications change.
Environmental Evaluation Southern Nuclear Operating Company has evaluated the proposed changes and determined the changes do not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in the individual or cumulative occupational exposure. Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), and an environmental assessment of the proposed changes is not required.
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Enclosure 3 Edwin I. Hatch Nuclear Plant, Units I and 2 Technical Specifications Amendment Request Appendix J Leakage Rate Test for Containment Purge Valves Changed Technical Specifications and Bases Pages

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                  FREQUENCY SR 3.6.1.3.6    Verify the isolation time of each MSIV is          In accordance with 2 3 seconds and < 5 seconds.                       the Inservice Testing Program SR 3.6.1.3.7    Verify each automatic PCIV, excluding EFCVs,       24 months actuates to the isolation position on an actual or simulated isolation signal.
SR 3.6.1.3.8    Verify each reactor instrumentation line EFCV (of  24 months a representative sample) actuates to restrict flow to within limits.
SR 3.6.1.3.9    Remove and test the explosive squib from each      24 months on a shear isolation valve of the TIP system.           STAGGERED TEST BASIS SR 3.6.1.3.10   Verify leakage rate through each MSIV is           In accordance with
                 < 11.5 scfh when tested at 2 28.0 psig.            the Primary Containment Leakage Rate Testing Program SR3place 3eat              of    the 18.valve           eah Ivalve hain        MOrVile  mt    Feial fat.
SR 3.6.1.3.12   Cycle each 18 inch excess fi isolation damper      24 months to the fully closed and fully en position.
I      Deleted         I HATCH UNIT 1                                3.6-1 2                   Amendment No.

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                   FREQUENCY SR 3.6.1.3.6   Verify the isolation time of each MSIV is          In accordance with 2 3 seconds and s 5 seconds.                       the Inservice Testing Program SR 3.6.1.3.7   Verify each automatic PCIV, excluding EFCVs,       24 months actuates to the isolation position on an actual or simulated isolation signal.
SR 3.6.1.3.8   Verify each reactor instrumentation line EFCV (of  24 months a representative sample) actuates to restrict flow to within limits.
SR 3.6.1.3.9   Remove and test the explosive squib from each      24 months on a shear isolation valve of the TIP system.           STAGGERED TEST BASIS SR 3.6.1.3.10  Verify leakage rate through each MSIV is           In accordance with s 11.5 scfh when tested at a 28.0 psig.            the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.11  Deleted                                                                  I SR 3.6.1.3.12  Cycle each 18 inch excess flow isolation damper    24 months to the fully closed and fully open position.
HATCH UNIT 1                             3.6-1 2                    Amendment No.

PClVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                  FREQUENCY SR 3.6.1.3.6        Verify the isolation time of each MSIV is          In accordance with 2 3 seconds and 5 5 seconds.                       the Inservice Testing Program SR 3.6.1.3.7        Verify each automatic PCIV, excluding EFCVs,       24 months actuates to the isolation position on an actual or simulated isolation signal.
SR 3.6.1.3.8        Verify each reactor instrumentation line EFCV (of  24 months a representative sample) actuates to restrict flow to within limits.
SR 3.6.1.3.9        Remove and test the explosive squib from each      24 months on a shear isolation valve of the TIP system.           STAGGERED TEST BASIS SR 3.6.1.3.10       Verify the combined leakage rate for all           In accordance with secondary containment bypass leakage paths is      the Primary 5 0.009 La when pressurized to 2 Pa.               Containment Leakage Rate Testing Program SR 3.6.1.3.11       Verify leakage rate through each MSIV is           In accordance with 5 100 scfh, and a combined maximum                 the Primary pathway leakage 5250 scfh for all four main        Containment steam lines, when tested at 2 28.8 psig.           Leakage Rate Testing Program However, the leakage rate acceptance criteria for the first test following discovery of leakage through an MSIV not meeting the 100 scfh limit, shall be 5 11.5 scfh for that MSIV.                     Deleted SR 3.6.1.3.12                                            IQ ^hl mOUt Ahfl^Be-+~
             . W. . Ag11e_
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                                                    -a--1 SR 3.6.1.3.13       Cycle each 18 inch excess flow isolation damper    24 months to the fully closed and fully open position.
HATCH UNIT 2                                    3.6-1 2                 Amendment No. 174

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                   FREQUENCY SR 3.6.1.3.6   Verify the isolation time of each MSIV is          In accordance with 2 3 seconds and s 5 seconds.                       the Inservice Testing Program SR 3.6.1.3.7   Verify each automatic PCIV, excluding EFCVs,       24 months actuates to the isolation position on an actual or simulated isolation signal.
SR 3.6.1.3.8   Verify each reactor instrumentation line EFCV (of  24 months a representative sample) actuates to restrict flow to within limits.
SR 3.6.1.3.9   Remove and test the explosive squib from each      24 months on a shear isolation valve of the TIP system.           STAGGERED TEST BASIS SR 3.6.1.3.10  Verify the combined leakage rate for all           In accordance with secondary containment bypass leakage paths is      the Primary s 0.009 La when pressurized to 2 Pa.               Containment Leakage Rate Testing Program SR 3.6.1.3.11  Verify leakage rate through each MSIV is           In accordance with s 100 scfh, and a combined maximum                 the Primary pathway leakage 5 250 scfh for all four main       Containment steam lines, when tested at 2 28.8 psig.           Leakage Rate Testing Program However, the leakage rate acceptance criteria for the first test following discovery of leakage through an MSIV not meeting the 100 scfh limit, shall be s 11.5 scfh for that MSIV.
SR 3.6.1.3.12  Deleted                                                                  I SR 3.6.1.3.13  Cycle each 18 inch excess flow isolation damper    24 months to the fully closed and fully open position.
HATCH UNIT 2                               3.6-1 2                   Amendment No.

PCIVs B 3.6.1.3 BASES SURVEILLANCE       SR 3.6.1.3.10 REQUIREMENTS (continued)    The analyses in References 1 and 3 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be S 11.5 scfh when tested at 2 28.0 psig.
The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 6).
SR 3.6.1.3.11 hA wakoe Se--aW Of e-aGh 18 RGh -PUFc3e Va.      -    P. all_ eX..       St Fes;W-At R;ateFial                be FeWaGed e; ,--, -0A ---
                                                                                          .. th6.
                           ...  ..                   - Mr at fk^ ^^M^r+U7Ai1h9     rtqpai - -
1-1 W            .. - , leakaoe famijurni I    Deleted    1'                 he liann-e test histow =O          ;M SR 3.6.1.3.12 This SR provides assurance that the excess flow isolation dampers can close following an isolation signal. The 24 month Frequency is based on a review of the surveillance test history and Reference 8.
REFERENCES         1.        FSAR, Section 14.4.
: 2.       Technical Requirements Manual, Table T7.0-1.
: 3.        FSAR, Section 5.2.
: 4.        10 CFR 50, Appendix J, Option B.
: 5.       NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6.       Primary Containment Leakage Rate Testing Program.
: 7.       NEDO-32977-A, 'Excess Flow Check Valve Testing Relaxation."
: 8.       NRC Safety Evaluation Report for Amendment 232.
HATCH UNIT 1                             B 3.6-26                                   REVISION

PCIVs B 3.6.1.3 BASES SURVEILLANCE    SR 3.6.1.3.10 REQUIREMENTS (continued) The analyses in References 1 and 3 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be s 11.5 scfh when tested at a 28.0 psig.
The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 6).
SR 3.6.1.3.11 Deleted                                                                 I SR 3.6.1.3.12 This SR provides assurance that the excess flow isolation dampers can close following an isolation signal. The 24 month Frequency is based on a review of the surveillance test history and Reference 8.
REFERENCES      1.      FSAR, Section 14.4.
: 2. Technical Requirements Manual, Table T7.0-1.
: 3.      FSAR, Section 5.2.
: 4.      10 CFR 50, Appendix J, Option B.
: 5.      NRC No. 93-102, 'Final Policy Statement on Technical Specification Improvements,' July 23,1993.
: 6.      Primary Containment Leakage Rate Testing Program.
: 7.      NEDO-32977-A, Excess Flow Check Valve Testing Relaxation."
: 8.      NRC Safety Evaluation Report for Amendment 232.
HATCH UNIT 1                       B 3.6-26                               REVISION

PCIVs B 3.6.1.3 BASES SURVEILLANCE      SR 3.6.1.3.10 REQUIREMENTS (continued)   This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions in the radiological evaluations that form the basis of the FSAR (Ref. 3) are met. The secondary containment bypass leakage paths are: 1) main steam condensate drain, penetration 8; 2) reactor water cleanup, penetration 14; 3) equipment drain sump discharge, penetration 18;
: 4) floor drain sump discharge, penetration 19; and 5) chemical drain sump discharge, penetration 55. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 7).
SR 3.6.1.3.11 The analyses in References 1 and 4 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be s 100 scfh, and a combined maximum pathway leakage S 250 scfh for all four main steam lines when tested at 2 28.8 psig. In addition, if any MSIV exceeds the 100 scfh limit, the as left leakage shall be s 11.5 scfh for that MSIV.
The Frequency is required by the Primary Containment Leakage Rate Testing Program.
SR 3.6.1.3.12 Rhe valve seats-of e--aG-h 188 'i;G-h pume ...- I-Ke /-._pply                 .-
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                                                                          - - -!!nfni' ba s G R F"leF-''-             is
* Eru,4n n;1an ne tes                   khi.t pn   .,r.,d R efeape n      p   0 (continued)
HATCH UNIT 2                                             B 3.6-26                                                REVISION

PCIVs B 3.6.1.3 BASES SURVEILLANCE    SR 3.6.1.3.1 0 REQUIREMENTS (continued) This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions in the radiological evaluations that form the basis of the FSAR (Ref. 3) are met. The secondary containment bypass leakage paths are: 1) main steam condensate drain, penetration 8; 2) reactor water cleanup, penetration 14; 3) equipment drain sump discharge, penetration 18;
: 4) floor drain sump discharge, penetration 19; and 5) chemical drain sump discharge, penetration 55. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 7).
SR 3.6.1.3.11 The analyses in References 1 and 4 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be s 100 scfh, and a combined maximum pathway leakage s 250 scfh for all four main steam lines when tested at 2 28.8 psig. In addition, if any MSIV exceeds the 100 scfh limit, the as left leakage shall be s 11.5 scfh for that MSIV.
The Frequency is required by the Primary Containment Leakage Rate Testing Program.
SR 3.6.1.3.12 Deleted                                                                    I (continued)
HATCH UNIT 2                        B 3.6-26                                REVISION}}