L-99-024, Responds to NRC RAI Re Conversion to ITS for Chapters 3.4, 3.5,3.6,3.7,3.9 & 5.0,per 990419-20 Meetings with NRC
| ML20209B808 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/30/1999 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20209B813 | List: |
| References | |
| NEL-99-0240, NEL-99-240, NUDOCS 9907080148 | |
| Download: ML20209B808 (30) | |
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1 D:n Mor:y S:uth:m Nucle:r Vice President Operating C mpany.inc.
Farley Project Post Othce Box 1295 Birmingham, Alabama 35201 Tel 205 992.5131 June 30, 1999 SOUTHERN COMPANY Energy to Serve lbur Wbrid"
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Docket Nos.:
50-348 NEL-99-0240 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Response to Request for Additional Information - Chapter 4.0 i
and NRC Staff Comments Related to Previous Responses to Requests for Additional Information Related to Conversion to the Lmoroved Technical Snecifications - Chanters 3.4. 3.5. 3.6. 3.7. 3.9 and 5.0 Ladies and Gentlemen:
By letters dated March 12,1998 and April 24,1998, Southern Nuclear Operating Company l
(SNC) submitted the Farley Nuclear Plant (FNP) - specific Improved Technical Specifications (ITS) conversion documentation packages in accordance with 10 CFR 50.90. The April 24,1998 letter, which submitted the Clean-Typed copies of the FNP ITS, included an attachment which provided hard copies of changes to the original submittal to correct minor editorial errors and inconsistencies within the package. By letter dated August 20,1998, SNC submitted an electronic copy of the Discussion of Changes (DOCS) and Significant Hazards Evaluations (SHEs) associated with the ITS conversion. Included with that letter were hard copies of changes to the original submittal to correct additional minor editorial errors and inconsistencies within the package. By letter dated November 20,1998, SNC submitted responses to a Request for Additional Information (RAI) for Chapters 3.6 and 5.0. By letter dated February 20,1999, SNC submitted responses to a RAI for Chapter 3.4. By letters (2) dated April 30,1999, SNC submitted responses to RAls for Chapters 3.1,3.2,3.5. 3.7,3.8, and 3.9. By [[letter::L-99-021, Forwards Response to RAI Re Conversion to ITSs for Chapter 3.3.Attachment II Includes Proposed Revs to Previously Submitted LAR Re Rais,Grouped by RAI number.Clean-typed Copies of Affected ITS Pages Not Included|letter dated May 28, 1999]], SNC submitted responses to a RAI for Chapter 3.3. Included with the above responses
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were hard copies of changes to the original submittal to reflect the SNC responses to the RAIs.
'f NRC E-mail dated June 10,1999 requested SNC provide additional information for Chapter 4.0.
During meetings held with the NRC on April 19-20,1999, the staff stated that it was not necessary to provide mark-ups of the Current Technical Specifications (CTS) in responses to
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RAls. Therefore, the attached pages do not contain CTS mark-ups. Attachment I provides the SNC responses to the RAI for Chapter 4.0 and revised responses to NRC RAI questions discussed in those meetings and subsequent conference calls. Attachment II includes proposed l
responses and revisions to the previously submitted license amendment request related to these RAls, grouped by RAI number.
9907000148 990630 PDR ADOCK 05000348 P
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Page 2 U.S. Nuclear Regulatory Commission In response to this RAI, some changes to the SiiEs were required. As denoted in 10 CFR j
50.92(c), SNC has determined the proposed changes to the FNP Technical Specifications do not involve a significant hazards consideration. The revised SIIEs are included in Attaciunent II.
SNC has also determined that the proposed changes will not significantly afTect the quality of the human environment. A copy of the proposed changes has been sent to Dr. D. E. Williams, the Alabama State Designee, in accordance with 10 CFR 50.91(b)(i).
I Clean-typed copies of the affected ITS pages are not included. A complete clean-typed copy of the FNP ITS will be re-submitted at the end of the NRC review process.
Mr. D. N. Morey states that he is a Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company and that, to the best of his knowledge and belief, the facts set forth in this letter and attachments are true.
If there are any questions, please advise.
Respectfully submitted, SOUTliERN NUCLEAR OPERATING COMPANY W'W Dave Morey l
Sworn to andsubscribed before me this 80 ay of
, 1999 W&L bnde W Notary Public My Commission Erpires:
AUQu Y l,000l
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WAS/marITSRAI_6. DOC l
Attachments I.
SNC Responses and Revised Responses to NRC Requests for Additional Information Related to Conversion to the Improved Technical Specifications - Chapters 3.4,3.5, 3.6, 3.7, 3.9,4.0 and 5.0.
II.
SNC Responses and Revised Responses to NRC Requests for Additional Information Related to Conversion to the Improved Technical Specifications, Chapters 3.4,3.5,3.6, 3.7,3.9,4.0 and 5.0 - Associated Package Changes Grouped by RAI Number cc:
See next page.
l 3
Page 3 U. S. Nuclear Regulatory Commission ec:
Southern Nuclear Ooeratina Comoany j
Mr. L. M. Stinson, General Manager - Farley U. S. Nuclear Regulatorv Commission. Washinaton. D. C.
Mr. L. M. Padovan, Licensing Project Manager - Farley U. S. Nuclear Regulatory Commission. Recion 11 j
Mr. L. A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley Alabama Deoartment of Public Health Dr. D. E. Williamson, State Health Officer-1 4
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k ATTACHMENT I SNC Responses and' Revised Responses to NRC Requests for AdditionalInformation Related to Conversion to the Improved Technical Specifications Chapters 3.4,3.5,3.6,3.7,3.9,4.0 and 5.0 1
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SNC Revised Response to NRC RAI Related to Chapter 3.4
' ITS 3.4.5 RCS Looos - MODE 3 NRC Question:
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ITS SR 3.4.5.2 requires verification of SG secondary side water levels are 2 74%
-(wide range) for RCS loops. The CTS is 10 % while the STS has a bracketed 17%.
This is a Beyond Scope issue.
SNC Response:
The Bases for SR 3.4.5.2 discusses the potential impact of not maintaining the minimum level in the steam generators. The basis for the level is to ensure that the steam generator tubes remain covered, thereby ensuring that the associated loop is capable of providing the heat sink for the removal of the decay heat. The CTS Bases do not provide the basis
-. for the 10% steam generator level. As part of the conversion to the ITS, Farley requested that Westinghouse perform an evaluation to determine the level necessary to meet the basis stated for SR 3.4.5.2 (as well as SR 3.4.6.2). A bounding level, for the applicable MODES, to ensure that the steam generator tubes remain covered was determined to be 74% wide range steam generator level.
As stated on the second page of Enclosure 4, " Bracketed information in the STS is confirmed to be applicable to FNP or replaced with information that is applicable to FNP or deleted as appropriate. In cases where previously NRC approved CTS information is used to replace the generic STS information in brackets, nojustification for altering the bracketed STS information is provided. The basis for all such changes to the STS is to maintain the current FNP licensing basis as specified in the CTS." As this change replaces the bracketed information with information that is different from both the current licensing basis and the STS, a JFD should have been included in Enclosure 5 for this change. It was inadvertently missed due to the fact that it was replacing bracketed information. A JFD for this change has been added to the package. While this change differs from the current licensing basis, it requires a greater steam generator level based on a Farley specific evaluation to ensure that the steam generator tubes remain covered, and is therefore a more conservative position for plant operation.
SNC Revised Response:
I During the development of the Licensing package for conversion to the Improved Technical Specifications (ITS), calculations were performed to determine the steam generator level necessary to ensure an operable steam generator (tubes covered). A
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and 5, was determined and incorporated into the ITS submittal. During procedure review for implementation of the ITS, an issue was discovered related to the Steam Generator level associated with SR 3.4.5.2. It was noted that the adoption of this conservative number would result in operation outside of the normal operating band for steam generator level in Mode 3. Due to the fact that the narrow range SG level transmitters are Page l of 4 j
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SNC Revised Response to NRC RAI Related to Chapter 3.4 hot calibrated and the wide range SG level transmitters are cold calibrated, a single level value related to a single indicator cannot be used through all Modes. The change in temperature and density of the fluid introduces significant inaccuracies in the indicated level as the actual fluid conditions deviate from the conditions at which the transmitters were calibrated. For example, if the cold-calibrated. wide range indication is held constant at the level required to ensure that the steam generator tubes remain covered, as temperature increases the actual level will increase. The actual level, and the hot-calibrated narrow range level indication, will exceed the high level turbine trip setpoint as temperatures increase toward normal operating temperatures in Mode 3. As a result, additional calculations were performed to determine a more accurate number to use for Steam Generator level in Mode 3. It was determined that a 28% narrow range level requirement will ensure operability of the Steam Generators, while at the same time ensuring that plant systems will not be challenged due to instrument uncertainties associated with wide range levels and Mode 3 plant conditions. The associated package changes are included in Attachment II.
ITS 3.4.9 Pressurizer NRC Question:
- 11. CTS 3.4.4 shows a pressurizer water volume at 868 cubic feet with a corresponding
% indicated level. ITS 3.4.9 adopted the STS. The numbers in your analysis assumption are the numbers that should be used. Also note that the 92% shown in the STS is 92% ofinstrument span.
Additional Comment: The Staff s position is that the value associated with Pressurizer level in the TS should be consistent with the initial conditions assumed in licensing basis analyses. They stated that Westinghouse uses 62%, not 92% in their analyses. The Staff suggested that SNC consider TSB-017 in responding to this question.
SNC Response:
The Bases for the STS state the following: "The intent of the LCO is to ensure that a steam bubble exists in the pressurizer prior to power operation to minimize the consequence of potential overpressure transients. The presence of a steam bubble is consistent with analytical assumptions." The intent is not to preserve the safety analysis assumption for pressurizer level at the initiation of an accident. Specific numbers in safety analyses vary depending upon which accident is being analyzed. In some cases, a
._ __high level is more limiting.. In some cases, p low level is more limiting <..The NRC. _._ _ _.._ _
s-approved change to NUREG-1431, TSTF-162, clarifies that the LCO limit requirement is to ensure that operation is within the safety analyses assumption of ensuring that a steam bubble exists in the pressurizer. Setting the TS level at the value associated with the Pressurizer Water Level-High trip setpoint satisfies the requirement to maintain a bubble in the pressurizer and corresponds to the basis originally agreed upon during the Page 2 of 4
SNC Revised Response to NRC RAI Related to Chapter 3.4 development of the NUREG. A review of the bases for LCO 3.4.9 was performed to ensure applicability to Farley. The intent of this LCO, as stated in the NUREG, is applicable to Farley and has been adopted.
SNC Revised Response:
SNC hu contacted Westinghouse concerning the basis for the value associated with the Pressurizer level. Westinghouse confirmed that the purpose of the value in the STS was to ensure that a steam bubble exists in the Pressurizer, consistent with the discussion in the Bases. Both the CTS value of 63.5% Pressurizer level and the STS bracketed value of 92% Pressurizer level meet that criteria. Based on further discussions with the Staff, SNC has revised the submittal to maintain the current licensing basis for Farley of 63.5%
Pressurizer level. The associated package changes are included in Attachment II.
ITS 3.4.11 Pressurizer Power Ooerated Relief Valves fPORVs)
NRC_ Question:
- 14. ITS SR 3.4.11.2 NOTE is not clear. Reword to eliminate ambiguity.
Additional Comment: The Staff recommended TSTF-288, Rev.1 for approval.
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TSTF-288, Rev.1 addresses this issue. The StatTsuggested that SNC consider i
incorporating TSTF-288, Rev.1.
SNC Response:
The note and associated discussions have been deleted.
SNC Revised Response:
SNC has incorporated TSTF-288, Rev.1 into the Farley ITS, with the exception of the changes to STS Bases page 3.4-52 which conflict with Inserts 1 and 2 and the discussion ofLCO 3.0.4 in Chapter 3.0. In addition, Inserts 1 and 2 have been modified consistent with the Farley current licensing basis. As stated in the TSTF, the bracketed material is only applicable on a plant-specific basis and is not included in the Farley ITS. The associated package changes are included in Attachment II.
ITS 3.4.14 RCS Pressure Isolation Valve fPIV) Leakage
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NRC Question:
- 23. - STS 3.4.14 Required Action A contains a NOTE concerning the requirements for valves used to isolate one or more leaking PIVs which includes a specific requirement that the valve "must have been verified to meet SR 3.4.14.1." In ITS Page 3 of 4
SNC Revised Response to NRC RAI Related to Chapter 3.4 3.4.14 Required Action A, this requirement is deleted which is a deviation from the STS. JFD #1a discusses this deletion but the deletion appears to be a cresentation preference hat is not sufficientlyjustified as a STS deviation by JFD #1a because there is no plant specific design difference necessitating this deletion. Provide justification for deleting "have been verified to meet SR 3.4.14.1" from the NOTE in ITS 3.4.14 Required Action A based upon plant specific design or include this requirement in ITS 3.4.14 Required Action A.
A dditional Comment: The Staff stated that there were two options that SNC could m: 1) SNC can maintain the current licensing basis which would include isolating the flow path by the use of two valves within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and removing the STS LCO 3.0.4 allowance for this LCO to ascend through Modes with inoperable components, or 2) SNC can revise the note for Required Action A to state that the valves used to meet the Required Action "must be verified to meet SR 3.4.14.1" and to describe in the Bases the methodology used if this option is implemented at power.
SNC Response:
The FNP CTS Actions and the corresponding STS Actions require that the valves used to meet the Action isolate the PIV leakage. Failure to isolate the leakage (meet the requirements of the LCO) is failure to meet the Action requirements of Condition A and would result in entry into Condition B and a plant shutdown. Since in orc'.r to meet the isolation Actions, the leakage must be monitored after the isolation valves are closed and verified to be within the LCO limits, the isolation capability of the valves is verified when they are actually used to meet the Action requirements. Compliance with the CTS and ITS Action Requirements are adequate to address PIV leakage and the deletion of the STS requirement to only use valves previously tested per SR 3.4.14.1 is consistent with the FNP current licensing basis.
SNC Revised Response:
SNC will adopt option 2. The methodology will be that of SR 3.4.13.1 (RCS water inventory balance) with the leakage limits of SR 3.4.14.1 applied. The associated package changes are included in Attachment II.
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l Chapter 3.5
SNC Revised Response to NRC RAI Related to Chapter 3.5 NRC Question:
3.5.1-1 CTS 3.5.1, Applicability and Footnote
- DOC 12-M ITS 3.5.1 Applicability Note JD 1 The Applicability of CTS 3.5.1 is Modes 1,2, and 3 with pressurizer pressure above the
. P-11 setpoint (2000 psig). The ITS 3.5.1 Applicability is revised to Modes 1,2, and 3, with RCS pressure > 1000 psig. In addition, a Note has been added to the Applicability for ITS 3.5.1-which states that, in Mode 3, above 1000 psig, the accumulators may be isolated for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose ofisolation valve testing per SR 3.4.14.1.
DOC 6-M and JD 1 both state that the Note is necessitated by the adoption of the 1000 psig applicability limit.
Comment: The staff does not recall seeing the exception outlined in the ITS Applicability Note proposed in any previous amendment requests for conversions to the Westinghouse STS. Please describe the plant-specific conditions and methods for testing the accumulator isolation valves in accordance with SR 3.4.14.1 that require the proposed exception. What is unique about Farley's testing of these isolation valves that requires an exception to the Applicability that no other Westinghouse licensee has requested to date?
If you conclude and the staff agrees, based on the additional information you provide, that the Note is 'necessary, the staff suggests that it be revised to read, "In Mode 3, with pressurizer pressure > 1000 psig, the accumulators may be isolated for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1." The staff suggests this revision of the Note to: (1) clarify what is meant by "above 1000 psig," and (2) make the wording of the note consistent with a sin. ' tr note in ITS 3.5.2.
Additional Comment: The original response to RAI 3.5.1-1 did not include a DOC to describe the change from referencing pressurizer pressure to referencing RCS pressure in the statement of the LCO. Please provide the appropriate DOC. In addition, this change is consistent with TSTF-117 and does not need to be treated as a deviation from the STS.
SNC Response:
The valves nearer the RCS (inboard) are tested by using RCS pressure to seat the valves.
Those nearer the accumulators (outboard) are tested by using one of the accumulators as a pressure source via a test piping arrangement. Testing is normally scheduled for performance at about 1600 psig RCS pressure following outages.
.g to metal seating arrangement and are manufactured by Copes Vulcan. Accumulator isolation valve testing at FNP has shown these valves to require a relatively high differential pressure to consistently fully seat. Operational considerations require that valve' leakage be extremely low to prevent accumulator level increase and boron dilution. The plant design is such that Page1of2
SNC Revised Response to NRC RAI Related to Chapter 3.5
- .ccumulator level and boron concentration change control is very limited. Attempts to test these valves at less than 1000 psi RCS pressure have had mixed results while testing at higher pressures has typically been more successful. To perform this test to the level required for operational considerations the exception is needed. Although a formal survey of other utilities has not been done, indications are that the model of accumulator discharge check valve used at FNP is not commonly used for this application.
The wording of the note has been changed to be consistent with the intent of the Staff suggestion. The references to pressurizer pressure in LCO 3.5.1 are revised to indicate RCS pressure. This deviation to the STS is required because the FNP pressurizer pressure narrow range instrumentation range is 1700 to 2500 psig. The RCS wide-range O to 3000 psig pressure instruments are used for monitoring system pressure below I
approximately 1700 psig. The appropriate changes to the package are attached.
SNC Revised Response:
DOCS 4-L and 6-M have been revised to reflect the revision of references from pressurizer pressure to RCS pressure. TSTF-ll7 has been incorporated and JD-3 deleted, f
In addition, the Note under the Applicability section has been revised. SNC determined that due to the design of Farley, testing of the accumulator check valves will also atTect parameters in the accumulator in addition to the isolation valves. Because no separate testing system exists in the design of the Farley accumulators, a differential pressure must be established between accumulators to test the check valves. As the accumulators are interconnected during the test, and parameters such as pressure and level will be adjusted, sometimes going outside of the normal band for OPERABILITY, a revision to the existing Note is required. The revised Note and associated package changes are included in Attachment II.
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3.6.1-3 DOC 3/4.6.1.6-2A DOC 3/4.6.1.6-ILA CTS 3.6.1.6 ACTIONS ITS 3.6.1 ACTIONS and Associated Bases The CTS markup of CTS 3/4.6.1.6 shows that the entire specification except for the shutdown requirement of the ACTION statement as being relocated to a licensee controlled document (DOC 3/4.6.1.6 - 1 LA). This relocation designation for CTS 3.6.1.6 ACTIONS is incorrect. While the shutdown portion of the ACTION statement is correctly marked up to indicate it becomes ITS 3.6.1 ACTION B and justified as an Administrative change, the allowed outage time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not appropriately marked up orjustified. The correct change would show that the allowed outage time portion of CTS 3.6.1.6 ACTIONS is not relocated out of TS but is incorporated into ITS 3.6.1 ACTION A. Thus 1:ie 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed outage time is changed to a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time, and the change is considered as More Restrictive rather than Less Restrictive - Generic I
(LA). See Comment Number 3.6.1-4,3.6.1-5 and 3.6.1-6 Comment: Revise the CTS markup of CTS 3.6.1.6 ACTIONS to show that it has been reformatted and changed to conform to ITS 3.6.1 ACTIONS and provide the appropriate discussion and justification for the More Restrictive change in the Completion Times. See Comment Numbers 3.6.1-4,3.6.1-5, and 3.6.1-6.
i Additional Comment: The order of the new Condition is inconsistent with standard STS format. Revise the response to reverse Conditions A and B.
SNC Response:
The completion time ofI hour in Condition A ofITS LCO 3.6.1 is applied once containment is determined to be inoperable. This occurs when the definition of operability for containment is no longer met or it is known that a required surveillance cannot be met. As stated in DOC 3/4.6.1.6-ILA, the requirements for containment structural integrity are maintained in the surveillance requirements of LCO 3.6.1,
" Containment" as SR 3.6.1.1 and SR 3.6.1.2 and in Specification 5.5.6, " Pre-Stressed Concrete Containment Tendon Surveillance Program" in the Administrative Controls section of the STS. With regard to the above staff comment, Condition A ofITS LCO 3.6.1 would not be entered until surveillance 3.6.1.2 was not met. SR 3.6.1.2 continues to be met until the requirements of the Pre-Stressed Concrete Containment Tendon Surveillance Program are not met. CTS 3/4.6.1.6 contains actions which allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
.+- to restore any non-conforming conditions before requiring a plant shutdown to Mode 5:=---
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In the conversion documentation, the markups show that the details of CTS 3/4.6.1.6, including the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restoration time and specific guidance for performing the required surveillances are moved into a program outside of the Technical Specifications similar to the existing programs for ASME Inservice Testing and Containment Leakage. Therefore, Page 1 of 7
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SNC Revised Response ts NRC RAI Related to Chapter 3.6 the 24-hour allowance for restoration of the structural integrity of containment to within the limits currently contained in CTS 3/4.6.1.6 would be maintained within the Pre-i Stressed Concrete Containment Tendon Surveillance Program. As such, SR 3.6.1.2 would continue to be met until the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance to correct any non-conforming i
condition had been exceeded within the Pre-Stressed Concrete Containment Tendon Surveillance Program. However, based on discussions with the NRC staff, in order to preclude misinterpretations a new Condition has been added to ITS LCO 3.6.1 to incorporate the CTS licensing basis.
l SNC Revised Response:
Conditions A and B have been reversed to make them consistent with the standard STS format. The associated package changes are included in Attachment II.
NRC Question:
3.6.1-7 JFD 1 j
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ITS SR 3.6.1.1 and Associated Bases CTS 4.6.1.1.c and 3/4.6.1.2 require leak rate testing in accordance with the Containment Leakage Rate Testing Program which is based on the requirements of10 CFR 50 Appendix J, Option B. STS SR 3.6.1.1 requires the visual examination and leakage rate testing be performed in accordance with 10 CFR 50 Appendix J as modified by approved exemptions. ITS SR 3.6.1.1 modifies STS SR 3.6.1.1 to conform to CTS 4.6.1.1.c and 3/4.6.1.2 as modified in the CTS markup. The STS is based on Appendix J, Option A while the CTS and ITS are based on Appendix J, Option B. Changes to the STS with regards to Option A versus Option B are covered by a letter from Mr. Christopher I. Grimes to Mr.
David J. Modeen NEI, dated 11/2/95 and TSTF-52. While the ITS SR 3.6.1.1 differences from STS SR 3.6.1.1 are in conformance with the letter and TSTF 52 as modified by staff comments, the changes to the ITS Bases as well as ITS 3.6.2 and ITS 3.6.3 and their Associated Bases are not in conformance with the letter, TSTF-52 as modified by the staff and the CTS. See Comment Numbers 3.6.2-3, and 3.6.3-8. Comment: Licensee should revise its submittal to conform to the 11/2/95 letter and TSTF-52 modified by the staff. See Comment Numbers 3.6.2-3 and 3.6.3-8.
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STS Bases page B 3.6-9 which start with the words,"following an outage...."
These two inserts were not approved in TSTF-52, Rev.1. Revise the response to delete these two inserts.
Page 2 0f 7 l
SNC Revised Respsnse to NRC RAI Related to Chapter 3.6 SNC Response:
A comparison of the submitted package with TSTF-52, Rev. I was performed. The main differences noted related to discussing a " design basis LOCA" vs. a "DB A." The changes necessary to incorporate TSTF-52, Rev.1 into Chapter 3.6 were made to the Bases except on page B 3.6-6,in the BACKGROUND Section ofITS 3.6.3. The first paragraph, as marked, is correct but incomplete. Additional DB As release radioactive materialinto the containment. This is discussed in the Bases ofTS 3.6.3 on page B 3.6-22 in the APPLICABLE SAFETY ANALYSES Section. Therefore, the discussions of DB As in the first and third paragraphs of the B ACKGROUND Section were left as originally submitted.
SNC Revised Response:
The response has been revised by the deletion of the two inserts on STS Bases page B 3.6-9. The associated package changes are included in Attachment II.
NRC Question:
3.6.3-8 DOC 3/4.6.1.7 - 8M DOC 3/4.6.1.7 - 6L DOC 3/4.6.1.7 - 10LA JFD PSE CTS 3.6.1.7 ACTIONS b and c CTS 4.6.1.7.2 and 4.6.1.7.3 STS ACTION E and Associated Bases STS SR 3.6.3.7 and Associated Bases ITS 3.6.3 ACTION D and Associated Bases ITS SR 3.6.3.7 and Associated Bases STS 3.6.3 ACTION E and SR 3.6.3.7 were developed from NUREG-0452, the old Westinghouse STS. In NUREG-0452 the surveillances for containment purge valves with resilient seals specified a leakage rate per valve which was an exemption from 10 CFR 50 Appendix J since Appendix J did not specify individual valve leakage only overall or combined valve leakage. If the leakage rate was exceeded, it was an indication ofimminent gross seal failure; thus the ACTIONS required that the valve leakage be restored to within limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the plant was shutdown. These requirements have been in NUREG-0452 since at least 1981. In the improved STS (NUREG-1431) the SR was maintained (surveillance based on individual valves), however, the specific limit was relocated since it was an approved exemption to 10 CFR 50 Appendix J. The
-' ~ - --ACTIONS required when valve-leakage was exceeded were modified to allow "
i continued indefinite operation provided that the penetration flow path was isolated and that the leakage when tested on a periodic basis through the isolation device (valve, blind flange, etc.) did not exceed the specified purge valve leakage limit. In the CTS the ACTIONS (CTS 3.6.L7 ACTIONS b and c) and the Page 3 of 7
SNC Revised Response to NRC RAI Related to Chapter 3.6 surveillances (CTS 4.6.1.7.2 and 4.6.1.7.3), which were implemented by Amendments 74 for Unit I and 66 for Unit 2, dated November 16,1987, are entirely different from the ACTIONS and surveillances in NUREG- 0452. The ACTIONS and surveillances are based on penetration leakage not valve leakage both individual and combined and operations can continue almost indefinitely as long as the combined Type B and C leakage limits and/or the individual penetration leakage is not exceeded. In addition, the leakage does not have to be restored to within limits until the next shutdown and only if the individual penetration leakage is exceeded. Based on the above discussion and the safety evaluation associated with Amendments 74 for Unit I and 66 for Unit 2, the staff j
believes that Farley has two options available with regards to containment purge valve with resilient seal leakage. They are cs follows:
1.
Farley can use STS ACTION E and SR 3.6.3.7 and their associated Bases.
However, it would be with the understanding and commitment that the leakage rate to be determined would be on an individual valve basis. This would be a major change from current licensing basis and could be considered as a beyond scope of review item depending on the specified individual valve leakage limit and how it was determined. See Comment Number 3.6.3-14.
2.
Modify STS ACTION E and/or SR 3.6.3.7 and their associated Bases to reflect current licensing bases as specified in CTS ACTIONS b and c and CTS 4.6.1.7.2 and 4.6.1.7.3. In this option CTS ACTION c would have to be retained in the ITS since it was part of the original bases which allowed Farley to deviate from the requirements specified in NUREG-0452. In addition, the leakage limits specified in CTS 4.6.1.7.2,4.6.1.7.3.a and 4.6.1.7.3.b all need to be specified in one place either in the Bases or in the Containment Leakage Rate Program. See Comment Number 3.6.3-14.
Comment: Revise the CTS /ITS markup accordingly and provide any additional discussion and justification as required. See Comment Number 3.6.3-14 Additional Comment: The Staff maintains that CTS ACTION c and the associated surveillance requirement 4.6.1.7.3.b have to be retained in the ITS since they were part of the original bases which allowed Farley to deviate from the requirements specified in NUREG-0452. Revise the response to maintain CTS ACTION c in the ITS.
SNC Response:
The package has been changed to reflect the current licensing basis with respect to the
-. -limit on purge valve penetration leakage as opposed to individual purgevalve leakage.
The associated DOC and JD have reflected the changes.
The CTS action statement c provides specific actions for the purge supply and exhaust penetration leakage limit verified by CTS surveillance 4.6.1.7.3.b. CTS surveillance j
Page 4 of 7
SNC Revised Response to NRC RAI Related to Chapter 3.6 l
4.6.1.7.3.b requires verification of purge supply and exhaust penetration leakage to be less than or equal to 0.05 L., The conservative CTS leakage limit of 0.05 L. for penetrations with purge supply and exhaust valves with resilient seals is unrelated to the requirements of10 CFR 50, Appendix J and is not required to be in the TS by 10 CFR 50 Appendix J or by 10 CFR 36 but was included in the CTS as a commitment to the NRC in response to issues related to the use of resilient seals in the purge valves. As a commitment not directly required by regulations, relocation to the Technical Requirements Manual (TRM) is appropriate.
I In the STS, the allowable leakage from these penetrations is controlled by the total Type.
B and C leakage limit (0.6 L.) and ultimately by the overall containment leakage limit (1.0 L.). Both of these 10 CFR 50 Appendix J limits are specified in the Containment j
Leakage Rate Testing Program in the administrative controls section of the TS. As such,
)
the removal of the 0.05 L. limit for penetrations with purge supply and exhaust valves with resilient seals from the CTS is consistent with the requirements contained in the STS and acceptable considering the governing total Type B and C and overall containment leakage limits which remain in the TS. This leakage limit, associated actions, and surveillance requirements are moved from the CTS to the TRM.' Reliance on requirements contained in the TRM is acceptable since changes to the requirements in the TRM will be controlled in accordance with the 10 CFR 50.59 process.
SNC Revised Response:
While SNC maintains that relocation to the TRM would be acceptable, SNC agrees to l
retain the CTS requirements in ACTION c and the associated surveillance requirement l
4.6.1.7.3.b in the ITS CTS ACTION c will become newITS Condition F and CTS surveillance 4.6.1.7.3.b will be incorporated into ITS SR 3.6.3.5 with the value for penetration leakage incorporated into the Containment Leakage Rate Program. The l
package has been revised to reflect this addition to the STS for Farley. The associated I
package changes are included in Attachment II.
NRC Question:
3.6.3-20 JFD E STS B3.6.3 Bases - E.1, E.2 and E.3 STS B3.6.3 Bases - SR 3.6.3.1 STS B3.6.3 Bases - SR 3.6.3.7 STS B3.6.3 Bases - REFERENCES ITS B3.6.3 Bases - D.1, D.2 and D.3 ITS B3.6.3 Bases - SR 3.6.3.1 ITS B3.6.3 Bases - SR 3.6.3.5 I
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STS B3.6.3 Bases - E.1, E.2 and E.3 and SR 3.6.3.7 refer to "NRC initiative, Generic Issue B-20" while ITS B3.6.3 Bases - SR 3.6.3.1 refers to "NRC initiative Generic Issue B-24." The ITS markup for ITS B3.6.3 Bases D.1, D.2, Page 5 of 7 c
SNC Revised Response to NRC RAI Related to Chapter 3.6 and D.3, SR 3.6.3.1, SR 3.6.3.5 and REFERENCES deletes these items based on JFD E which states that SCS cannot verify B-20 and B-24, therefore it is an STS error. B-20 and B-24 are valid generic issues-B-20 is Multiplant Action (MPA)
B020 " Containment Leakage Due to Seal Deterioration" and B-24 is MPA B024
" Venting and Purging Containment while At Full Power and Effect ofLOCA."
In fact, MPA-24 is referenced in the Safety Evaluation implementing t
Amendments 74 for Unit I and 66 for Unit 2. Comment: Revise the ITS markup for ITS B3.6.3 Bases - D.1, D.2 and D.3, SR 3.6.3.1, SR 3.6.3.5 and REFERENCES to include MPA B-20 and B-24 or provide a discussion and justification to show that they are not applicable to Farley.
Additional Comment: The Staff suggested alternative wording to clarify that the references were the basis for the generic resolution of the generic issues and do not reflect Farley specific commitments. The following wording was suggested as an example: "The normal Frequency for SR 3.6.1.7,184 days, was established as part of the generic resolution by the NRC Staff of Generic Issue B-20 (Ref. 3)."
SNC Response:
Copies ofMPA B-20 and B-24 cannot be located. In discussions with the NRC staff, the staff stated that MPA B-20 was incorporated into MPA B-24. References to MPA B-20 and B-24 have been identified in correspondence between Alabama Power Company (APCo) and the NRC, including the above referenced safety evaluation. In NRC to APCo letter dated August 5,1981, the staff discmed Generic Issue B-20 in Enclosure 1, which is an amplification of position B.4 of Pm 4 Technical Position (BTP) CSB 6-4, and recommended the addition of provisions e Ae Technical Specifications to test active purge vent systems once every 3 months and passive purge systems once every 6 months.
In NRC to APCo letter dated June 19,1986, which discussed the Farley Technical Specifications, the staff stated the following: "We consider this action will close out the Multiplant Action B-24 Technical Specifications...." In NRC to APCo letter dated June 19,1986, which issued amendments 74 for Unit I and 66 for Unit 2, the staff stated in the Safety Evaluation: "The staff found that the purge / vent systems at Farley 1 and 2 met the systems design and performance criteria as set forth in Branch Technical Position CSB 6-
'4, NUREG-0737, Item II.E.4.2, and the guidance developed as part ofMulti-Plant Action B-24." Based on review of the above correspondence, it is believed that Farley is currently in compliance with the frequency of purge system leakage rate testing discussed in MPA B-20 and B-24 and will continue to be in compliance after conversion to the ITS.
However, without copies of the referenced documents, the Bases for FNP ITS 3.6.3 will not be revised to incorporate the references to MPA B-20 and B-24 related to the frequency of purge system leakage rate testing.
= ~ - SNC Revised Response:-
SNC will incorporate references to MPA B-20 and B-24 related to the frequency of purge system leakage rate testing in the Bases for FNP ITS 3.6.3. These references will be similar to the Staffs suggestion, clarifying that they do not reflect Farley specific Page 6 of 7
SNC Revised Response to NRC RAI Related to Chapter 3.6 commitments but are the basis for NRC Staffgeneric resolutions. The associated package changes are included in Attachment II.
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SNC Revised Resptnse to NRC RAI Related to Cnapter 3.7 3.7.1.1 ITS LCO 3.7.1 DOC 4L and 4aL NRC Comment:
In the licensee's submittal, these changes are identified as a beyond-scope change. The licensee should review approved TSTF-235 Rev. I and revise, as appropriate, the ITS and the Bases before review of this issue is pursued as a beyond-scope change.
i Additional Comment:
An error was identified in the submitted response to RAI 3.7.1.1. In the adoption of TSTF-235, Rev.1, a typographical error was made on INSERT LL (page'186a). The second portion of Condition C indicates that action should be taken with "One or more steam generators with 2 4 MSSVs OPERABLE." The second portion of Condition C should indicate that action be taken with "One or more steam generators with 2 4 MSSVs inoperable."
SNC Response:
SNC has adopted TSTF-235, Rev.1, in the Farley ITS. Associated changes to the package are attached.
SNC Revised Response:
SNC has revised the submittal consistent with the above Staff comment. The associated package changes to the Chapter 3.7 package are included in Attachment II.
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SNC Revised Respanse to NRC RAI Related to Chapter 3.9 ITS 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level NRC Question: -
3.
STS 3.9.5, ACTION, A.4 requires closing all containment penetrations providing direct access from containment atmosphere to outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i
when RHR loop requirements are not met. ITS 3.9.4, ACTION, A.4 requires placing containment penetrations in the status described in LCO 3.9.3
" Containment Penetrations" within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, under the same conditions. TSTF-197 was rejected and has been superseded by TSTF 197, Rev.1. Revise Action A.4 Accordingly.
Additional Comment: TSTF-197, Rev. I was not approved. The Staffrequested that it be modified. It has been superceded by TSTF 197, Rev. 2. Revise the package accordingly.
SNC Response:
- TSTF-197, Rev.1, has been incorporated into the Farley ITS conversion package.
SNC Revised Response:
TSTF-197, Rev. 2, has been incorporated into the Farley ITS conversion package.
ITS 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level NRC Question:
4.
Same comment as above for TSTF 197.
Additional Comment: TSTF-197, Rev. I was not approved. The Staffrequested that it be modified. It has been superceded by TSTF 197, Rev. 2. Revise the package accordingly.
SNC Response:
TSTF-197, Rev.1, has been incorporated into the Farley ITS conversion package.
SNC Revised Response:
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4.0-1 ITS 4.3.1 Criticality JD6 STS 4.3.1.1.fis revised by the deletion of"the NRC approved [ specific document containing the analytical methods, title, date, or specific configuration or figure].", and the addition ofInsert D.
Comment: The revised 4.3.1.1.f should retain the wording "NRC approved" preceeding " Figures 4.3-2 through 4.3-5".
SNC Response:
The words "the NRC approved" have been added to the Farley ITS Section 4.3, part 4.3.1.1.f, consistent with the STS. The associated package changes are included in Attachment II.-
NRC Question:
4.0-2 ITS 4.3.1 Criticality JD7 STS 4.3.1.1 is revised by the addition ofInsert F.
Comment: 4.3.1.1 is not revised by the addition ofInsert F as 4.3.1.1.g,instead, it is revised by the addition ofinsert E, and insert F is added to STS 4.3.1.2 as 4.3.1.2.a and 4.3.1.2.b. Correct the documentation of JD 7, SNC Response:
JD-7 has been revised to reference Insert E. The associated package changes are includedin AttachmentII.
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j 1.
A series of conference calls were held with the Staff to discuss the implementation
- of the Reactor Coolant Pump Flywheel Inspection Program for Farley. The final resolution requested by the Staff was for Farley to adopt the statement of the program utilized by Vogtle in their amended ITS (NRC SER dated July 21,1998).
SNC Response:
)
1 SNC has revised the submittal consistent with the above Staff request. The associated 4
changes to the Chapter 5.0 package are included in Attachment II.
ITS 5.5.17 s
Containment Leakage Rate Testine Procram f
NRC Comment:
2.
In NRC RAI 3.6.3-8 for CTS Chapter 3.6, the NRC Staff required that leakage testing requirements for individual containment purge penetrations be maintained in the ITS in addition to the overall limits for Type B and C testing per Appendix J, Option B testing.
1 SNC Response:
The actual limits are contained in the Containment Leakage Rate Testing Program in the ITS. As part of the revised response to RAI 3.6.3-8, the limit for individual containment purge penetration leakage has been incorporated into the Containment Leakage Rate Testing Program. SNC has revised the submittal consistent with the above Staff request.
The associated changes to the Chapter 5.0 package are included in Attachment II.
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SNC Responses and Revised Responses to NRC Requests for Additional Information Related to Conversion to the Improved Technical Specifications Chapters 3.4, 3.5, 3.6, 3.7, 3.9, 4.0 and 5.0 Associated Package Changes Grouped by RAI Number
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