CNL-15-128, Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 5 of 8

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Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 5 of 8
ML15176A688
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 06/19/2015
From:
Tennessee Valley Authority
To:
Office of Nuclear Reactor Regulation
Shared Package
ML15176A678 List:
References
SQN-TS-11-10, CNL-15-128
Download: ML15176A688 (499)


Text

Sequoyah ITS Conversion Database Page 1 of 2 Licensee Response/NRC Response/NRC Question Closure Id 365 NRC Question KAB064 Number Select NRC Response Application Attachment 1

Attachment 2

Response Page 10 of the Calculation B87 140924 017, Revision 9, dated September Statement 24, 2104 notes 4 and 4a state:

4. An Ab = +/-0.5% of value is a requirement of the calculation.

Applies to 27DAT, 27DBT, 27DCT.

4a. Per Requirement 1, the acceptance band for calibration of the under voltage relays shall be set equal to the Re (+/-0.2% of value) per the requirements of TSTF-493. However, an acceptance band of 0.5% of value is conservatively retained from previous revisions in the calculation of the accuracy values for determination of the Allowable Values. Applies to 271A, 271B, 272A, 272B.

Ab493 = Re = +/-0.2% of value (for use in calibration tolerances) Ab =

+/-0.5% of value (for use in accuracy calculations)

With regard to note 4 on sheet 10 of the calculation, please clarify the definition of An Ab. These are two different terms per the definitions and abbreviations on sheet 4 of the calculation. The note also states that it is a requirement of the calculation. NRC staff understands that the calculation is normally performed to find the requirements for the setting. Please explain what is meant by is the requirement of the calculation.

Note 4a states Ab493 to be +/-0.2% but uses an Ab of +/-0.5%. The statement that an acceptance band of 0.5% of value is conservatively retained is confusing. Regulatory Information Summary (RIS) 2006-17 provides the guidance to calculate the as-left value. RIS 2006-17 in part states, the setting tolerance band is less than or equal to the square root of the sum of the squares of reference accuracy, measurement and test equipment, and readability uncertainties. Please explain how you meet the guidance contained in RIS 2006-17.

Response

10/6/2014 6:00 PM Date/Time Closure Statement Question Closure Date Notification Scott Bowman Michelle Conner Khadijah Hemphill Andrew Hon https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=365 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 2 Lynn Mynatt Ray Schiele Roger Scott Added By Kristy Bucholtz Date Added 10/6/2014 1:00 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=365 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 3 Licensee Response/NRC Response/NRC Question Closure Id 377 NRC Question KAB064 Number Select Licensee Response Application Attachment 1

Attachment 2

Response

Statement The following information is provided concerning the Staffs comments to the response to RAI KAB064 and SQN calculation B87 140924 017, Revision 9.

1. With regard to note 4 on sheet 10 of the calculation, please clarify the definition of An Ab. These are two different terms per the definitions and abbreviations on sheet 4 of the calculation.

Response

As it pertains to Note 4, An Ab, is not defining two different terms. An is the beginning article for the sentence. Note 4 explains that for calculation B87 140924 017, Revision 9, the term Ab (Acceptance band

- the range of values around the correct value determined to be acceptable without recalibration) equals +/- 0.5% of value.

2. The note also states that it is a requirement of the calculation. NRC staff understands that the calculation is normally performed to find the requirements for the setting. Please explain what is meant by is the requirement of the calculation.

Response

Note 4 applies to relays 27DAT, 27DBT, and 27DCT.

These relays are not within the scope of TSTF-493.

Therefore, for these instruments, an acceptance band https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=377 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 3 (Ab) = 0.5% must be used for the as-left tolerance and will be documented in an output configuration control document.

3. Note 4a states Ab493 to be +/-0.2% but uses an Ab of

+/-0.5%. The statement that an acceptance band of 0.5% of value is conservatively retained is confusing.

Response

The calculation was revised to incorporate the requirements of TSTF-493. The term Ab493 was added to determine the acceptance band for the As-Left tolerance to comply with TSTF-493. Ab493 is smaller than Ab, which produces a tighter As-Left tolerance. An Ab of +/-0.5% is maintained in the calculation for the determination of the Allowable Value. A larger Ab is more conservative for the determination of the Allowable Value.

4. Regulatory Information Summary (RIS) 2006-17 provides the guidance to calculate the as-left value.

RIS 2006-17 in part states, the setting tolerance band is less than or equal to the square root of the sum of the squares of reference accuracy, measurement and test equipment, and readability uncertainties. Please explain how you meet the guidance contained in RIS 2006-17.

Response

TVA is adopting NUREG-1431, Revision 4, which incorporates TSTF-493. TSTF-493 satisfies the guidance of RIS 2006-17 when calculating As-Left tolerances.

For the As-Left tolerances, TVA is setting the tolerance band equal to the square root sum of the squares (SRSS) of the reference accuracy and measuring and test equipment (M&TE) inaccuracies. In some calculation, the M&TE inaccuracies may be set to zero, which will cause a tighter As-Left tolerance.

Response

10/22/2014 2:30 PM Date/Time Closure Statement https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=377 1/13/2015

Sequoyah ITS Conversion Database Page 3 of 3 Question Closure Date Notification Scott Bowman Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Scott Bowman Date Added 10/22/2014 1:28 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=377 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 3 Licensee Response/NRC Response/NRC Question Closure Id 389 NRC Question KAB064 Number Select NRC Response Application Attachment 1

Attachment 2

Response

Statement

1. Item 1 of the licensee response stated the term AnAb is not defining two different terms. Sheet 4 of calculation 27DAT, Revision 9, defines An as the normal accuracy of the device and Ab as the acceptance band - the range of values around the correct value determined to be acceptable without recalibration. These are clearly two separate terms with different meanings. The use of term AnAb is confusing.

Response to Item 1 also stated the term Ab (Acceptance band

- the range of values around the correct value determined to be acceptable without recalibration) equals +/-0.5% of value. Staff agrees with the use of term Ab for this explanation as cited by the licensee. Licensee explanation used the correct term Ab as opposed to AnAb. Please update the calculation to avoid any confusion. If the licensee plans to continue the use of term AnAb then it must be clearly defined as a single term in the definitions and abbreviations section of the calculation.

Further please clarify that the term Ab - acceptance band without calibration is the same as the term as-left used in TSTF-493. If not clarify the difference and provide the values of the term as-left.

2. The licensee response that acceptance band for TSTF-493, Ab493 =+/-0.2% is acceptable as used in the calculation.

However, the term acceptance band for allowable value is confusing. The licensee seems to be discussing the tolerance for allowable value or setting tolerance for allowable value. Since the allowable value has to consider drift and should avoid unnecessary reportability, its value will be somewhat larger than Ab, the acceptance band around the setpoint that is acceptable without recalibration.

https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=389 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 3 Hence the use of Ab = +/-0.5% for allowable value allowance or tolerance is confusing and undesirable. Staff suggests the calculation be changed to prevent confusion by using terms Ab with two different meanings.

3. TSTF-493 Notes 1 and 2 have not been added to the technical specifications (TS) to address as-left and as-found values.

Please add these notes to the technical specifications. If these notes are detailed in another document then reference the appropriate documents in the TS affected pages. Also please provide the wording of the notes and the values for as-left and as-found terms for staff review. These notes apply to reactor coolant pump (RCP) undervoltage loops (271A, 271B, 272A, and 272B) and 6.9kV shutdown board loss of voltage relays (27TS1A, 27TS1B, 27TS1C).

4. Definitions of Westinghouse methodology related terms have not been included in the definitions and abbreviations section of the calculation. These definitions should be included for completeness and to prevent confusion while reviewing the calculation.
5. Calculation sheets 4 and 5 provide definitions and abbreviations. However, these sheets do not include terms Lan and Las which are used on sheet 39 of the calculation.

Also the term Afc used on sheets 23 and 39 of the calculation is not included in the definitions and abbreviations section.

Please define these and any other abbreviations that have been used but not included in the definitions and abbreviations section of the calculation.

6. On page 23A of the calculation it is stated that An=As where An is the normal accuracy and As is the seismic accuracy as defined on sheet 4 of the calculation. Normally these two values are different unless the seismic accuracy is negligible. If the seismic accuracy is negligible then it should be so stated in the calculation. If not, the two values will be different. Please clarify.

Response

12/1/2014 6:00 PM Date/Time Closure Statement Question Closure Date https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=389 1/13/2015

Sequoyah ITS Conversion Database Page 3 of 3 Notification Scott Bowman Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Kristy Bucholtz Date Added 12/1/2014 1:01 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=389 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 4 Licensee Response/NRC Response/NRC Question Closure Id 402 NRC Question KAB064 Number Select Licensee Response Application Attachment 1

Attachment 2

Response

Statement 1. Calculation 27DAT does not use the term, AnAb anywhere. In three different locations, the calculation uses the phrase, An Ab, not AnAb (difference is the space between An and Ab). In all three of these cases, the An is the article used in a sentence (such as a, an, or the), not the defined calculation term, An. The only calculation term used in those three instances is the term, Ab. Therefore, since the calculation does not use any term designated as AnAb no clarifications are required in the definitions or abbreviations.

Ab - acceptance band without calibration is the same as the term as-left used in TSTF-493, and is defined in Branch Technical Instruction, BTI-EEB-TI-28, R10. BTI-EEB-TI-28, R10 has a specific section that addresses TSTF-493 requirements and indicates that the As-Left Tolerance is equivalent to the Acceptance Band (Ab).

Calculation 27DAT, provided in the SQN response to KAB064 on October 21, 2014, as Attachment 2, contains the value of Ab493 for the relays subject to TSTF-493 requirements.

2. There is no term acceptance band for allowable value used in calculation 27DAT. Note 4a on page 37 of the pdf (sheet 10) of calculation 27DAT states that, the acceptance band [As-Left] for calibration of the under voltage relays shall be set to the Re

(+/-0.2% of value) per the requirements of TSTF-493. Where Re (repeatability inaccuracy) = Ab493. This value of +/-0.2% is more conservative than the previous value of +/-0.5%, in that it establishes a tighter tolerance for the As-Left acceptance criteria.

Therefore, in terms of the calibration tolerances, the tighter requirements for TSTF-493 (Ab493) are used.

Note 4a further states that an acceptance band of 0.5% of value is conservatively retained from previous revisions in the calculation https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=402 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 4 of the accuracy values for determination of the Allowable Values [emphasis added]. Note 4a shows, Ab = +/-0.5% of value (for use in accuracy calculations). These statements do not mean that the Allowable Value (AV) is equal to the acceptance band of

+/-0.5%.

The inclusion of any Ab term within the accuracy computation addresses the fact that uncertainty can be imparted to the measurement of the process in question due to the calibration process, including tolerances. When used in an accuracy computation that will be used for the derivation of Setpoints and Allowable Values, the use of a larger uncertainty term is conservative because it produces more separation between the Analytical Limit and the Setpoint. Therefore, the larger Ab term is retained for use in the accuracy computations for the Setpoint and Allowable Value. Although the Setpoint is specifically defined in another calculation, the summary of results on pdf page 88, of calculation 27DAT, (Sheet 49B) shows that margin exists between the Setpoint and Analytical Limit, considering the uncertainty values (An and As), which include the larger Ab value within their computations, as shown on pdf page 59 (Sheet 23A) of the calculation. In calculation 27DAT, page 80 of the pdf (Sheet 40) shows the computation of the Allowable Value using accuracies, based on the larger Ab. Therefore, Ab has been conservatively applied to the accuracy computations for the Setpoint and Allowable Value, and Ab493 has been conservatively applied for the calibration tolerances.

3. As part of the original submittal for the ITS LAR, the two TSTF-493 notes were included in ITS LCO 3.3.1 and 3.3.2 for all the functions that are required to meet TSTF-493. The functions have the (b) and (c) footnote annotation and the footnotes are at the bottom of the pages. See page 5 and 6 of Attachment 1 to the initial response for RAI KAB064 for examples of these footnotes.

As stated in footnote (c), the methodologies used to determine the as-found and as-left values are specified in UFSAR Section 7.1.2.

An FSAR change is currently in progress to provide the methodologies in Section 7.1.2 of the UFSAR, and will be complete prior to implementation of ITS. In calculation 27DAT, the As-Found value for the RCP Undervoltage loops (271A, 271B, 272A, and 272B) is found on page 59 of the pdf (Sheet 23A), As-Found is equal to +/-0.57% of setpoint. The As-Left value for the RCP Undervoltage loops is found on page 22 of the pdf (Sheet 2), the As-Left is equal to +/-0.2% of value. The As-Found value for the https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=402 1/13/2015

Sequoyah ITS Conversion Database Page 3 of 4 6.9kV shutdown board loss of voltage relays (27TS1A, 27TS1B, 27TS1C) is found on page 60 of the pdf (Sheet 24), As-Found is equal to +/-1.926% of setpoint. The As-Left value for the 6.9kV shutdown board loss of voltage relays is found on page 22 of the pdf (Sheet 2), As-Left is equal to +/-0.5% of value.

As-Found and As-Left values are controlled through Setpoint and Scaling Documents (SSDs). SSDs serve as the design output document to transmit the requirements to site organizations to ensure values assessed in the safety analyses and/or other design documents relative to instrument setpoints, scaling and calibration are in fact incorporated in the plant as assessed in the relevant design documents. Changes to As-Found and/or As-Left values require a Design Change to be processed via the Engineering Change Process. The As-Found and As-Left values listed in the SSDs are incorporated into Surveillance Instructions (SIs) that are performed to verify Technical Specification Surveillance Requirements. The SIs are annotated with requirements to evaluate setpoints found outside the As-Found tolerances to verify the channel is functioning as required before returning the channel to service. Additionally, this condition will be entered into the Corrective Action Program. The SIs also require that an instrument channel shall be declared inoperable if it cannot be reset to within the As-Left tolerance.

4. TVAs process for describing the method for determining the acceptability of setpoints for nuclear safety-related and Technical Specification instrumentation channels is governed by Branch Technical Instruction BTI-EEB-TI-28. Calculations under the scope of this instruction must follow the process outlined in this branch instruction. In some cases, terms and definitions are specified within the individual calculations; however, any term not specified within the individual calculation is defined in this branch technical instruction.
5. The terms LAn and LAs are defined in BTI-EEB-TI-28, Revision
10. LAn is the Normal Loop Accuracy. LAs is the Post-Seismic Loop Accuracy. The term Afc is defined on pdf page 59 (sheet 23A) of calculation 27DAT. Afc is the Acceptable-As-Found, Component. Therefore, Afc is the As-Found value for a particular component. BTI-EEB-TI-28, Rev. 10 also defines the term Afc as the Acceptable As-Found - Component.
6. Calculation 27DAT defines the term Se on pdf page 27 (Sheet 5) https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=402 1/13/2015

Sequoyah ITS Conversion Database Page 4 of 4 as the inaccuracy following a seismic event. Page 36 of the pdf (Sheet 9) states that Se is negligible and refers to Note 6, which is on pdf page 38 (Sheet 11). Note 6 states that post-seismic effects are negligible for this solid state relay. Therefore, An = As, as stated in the calculation.

Response

12/16/2014 5:00 PM Date/Time Closure Statement Question Closure Date Notification Scott Bowman Kristy Bucholtz Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Scott Bowman Date Added 12/16/2014 3:58 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=402 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 403 NRC Question KAB064 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 12/18/2014 Date Notification Scott Bowman Kristy Bucholtz Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 12/18/2014 2:31 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=403 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 2 ITS NRC Questions Id 174 NRC Question KAB065 Number Category Technical ITS Section 3.3 ITS Number 3.3.1 DOC Number JFD Number JFD Bases Number Page Number(s)

NRC Reviewer Rob Elliott Supervisor Technical Gursharan Singh Branch POC Conf Call N

Requested NRC Question Request for additional information regarding Sequoyah RCP Underfrequency Relays setpoint calculation number SQN-EEB-MS-T128-0076, Rev. 5 The above calculation was provided in support of the Sequoyah ITS request. Staff requests the following clarifications with request to this calculation:

1. Note 2 on page number 14 of the calculation justifies the use of a drift value of +/-0.553 Hz. In its letter dated June 3, 1994 ABB stated that ABB Type 81 Frequency Relay employs a very stable crystal controlled oscillator as frequency reference. It further stated that a drift value of 0.01 Hz over a period of 22.5 months will be very conservative. ABB also stated that the suggested drift of high magnitude suggested by the licensee would be indicative of a defective relay. In note 2, the licensee states that the drift value of

+/-0.553 Hz is highly conservative.

Please note that using a high drift value will mask the potential degrading of the instrument. The deviation in the as-found value should be based on regulatory information summary (RIS) 2006-17.

The deviation number selected should be high enough to prevent unwanted excursions beyond the allowable value while it should be low enough to detect potential degradation of the instrument. The licensee is requested to use a drift number using the guidance of https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=174 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 2 RIS 2006-17.

2. Please note that ISA RP-67.04.02 recommends that the accuracy of measurement and test equipment should be four times better than the accuracy of the instrument that is being calibrated. IEEE-498 also recommends this accuracy. The staff notes that the accuracy of the underfrequency relay is 0.008Hz whereas the accuracy of the test instrument is 0.05 Hz. The selection of a calibration instrument that is six times more inaccurate than the instrument being calibrated is highly undesirable and must be justified within the calculation.

Attach File 1 Attach File 2 Issue Date 7/15/2014 Added By Kristy Bucholtz Date Modified Modified By Date Added 7/15/2014 8:18 AM Notification Scott Bowman Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=174 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 2 Licensee Response/NRC Response/NRC Question Closure Id 371 NRC Question KAB065 Number Select Licensee Response Application Attachment Attachment 1 to KAB065.pdf (81KB) 1 Attachment Attachment 2 SQN-EEB-MS-T128-0076.pdf (14MB) 2 Response In response to KAB065, the RCP Underfrequency Relay setpoint calculation Statement number SQN-EEB-MS-TI28-0076, Revision 5, has been revised . Revision 7 revises the RCP underfrequency relay drift, Measuring and Test Equipment (M&TE) values and calibration tolerances to support the implementation of TSTF-493. Due to the reference accuracy of the relay greatly exceeding the minimum increment available to adjust the relay setpoint, the drift value is considered negligible and set to zero. This drift value has been factored into the as-found value calculation using the square root sum of the squares method based on TSTF-493, Revision 4, which incorporated the guidance of RIS 2006-17. The accuracy of the M&TE is also addressed in the calculation revision. The reference accuracy of the underfrequency relay is +/-0.008 Hz, the M&TE will be at least as accurate as the underfrequency relay. TVAs standard program and processes dictates that the calibration standards shall have an accuracy of at least four times the required accuracy of the M&TE being calibrated. When it is not possible to have a 4:1 ratio, standards shall have an accuracy that ensures that the plant equipment being calibrated will be within its required tolerances. The basis for acceptance of standards with accuracies less than four times that of the M&TE will be documented and authorized by the responsible TVA management. With the calibration requirements of the M&TE to a reference standard and the M&TE being at least as accurate as the underfrequency relay, which is in compliance with TVAs calibration program, the as-found values will be low enough to detect potential degradation of the instrument.

Additionally, as a result of the increased accuracy used in calculation, SQN-EEB-MS-TI28-0076, Revision 7, the Allowable Value (AV) for ITS Table 3.3.1-1, Function 12 (Underfrequency RCPs), on pages 120 and 152 of Enclosure 2, Volume 8, will be revised. The AV will be revised from the originally proposed value of 56.3 Hz to 56.973 Hz. Corresponding changes will be made to the CTS markups for CTS Table 2.2-1, Functional Unit 16 (Underfrequency-Reactor Coolant Pumps), and Discussion of Change M24.

See Attachment 1 for the draft revised CTS and ISTS markups.

See Attachment 2 for the revised calculation for the RCP underfrequency relays.

Response

10/16/2014 2:05 PM Date/Time https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=371 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 2 Closure Statement Question Closure Date Notification Scott Bowman Kristy Bucholtz Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Scott Bowman Date Added 10/16/2014 1:05 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=371 1/13/2015

Enclosure 2, Volume 8, Rev. 0, Page 22 of 1148 ITS A01 ITS 3.3.1 Table 3.3.1-1 TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT ALLOWABLE VALUES

14. Deleted 56.973 11 15. Undervoltage-Reactor 5022 volts-each bus 4739 volts-each bus Coolant Pumps 57.0 M24 56.3 12
16. Underfrequency-Reactor 56.0 Hz - each bus 55.9 Hz - each bus Coolant Pumps 14
17. Turbine Trip A. Low Trip System 45 psig 39.5 psig Pressure B. Turbine Stop Valve 1% open 1% open Closure 15 18. Safety Injection Input Not Applicable Not Applicable from ESF 16.a 19. Intermediate Range Neutron 1 x 10-4% of RATED THERMAL 6 x 10-5% of RATED Flux - (P-6) Enable Block POWER THERMAL POWER Source Range Reactor Trip LA07 16.e 20. Power Range Neutron Flux 10% of RATED THERMAL 12.4% of RATED (not P-10) Input to Low Power POWER THERMAL POWER Reactor Trips Block P-7 A21 September 20, 2007 SEQUOYAH - UNIT 1 2-6a Amendment No. 16, 85, 136, 141, 307, 310, 316 Page 18 of 47 Enclosure 2, Volume 8, Rev. 0, Page 22 of 1148

Enclosure 2, Volume 8, Rev. 0, Page 46 of 1148 ITS A01 ITS 3.3.1 Table 3.3.1-1 TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT ALLOWABLE VALUES

b. RCS Loop T Equivalent to Power > 50% RTP Coincident with Steam Generator Water 15.0% of narrow range 14.4% of narrow range Level -- Low-Low (Adverse) instrument span instrument span and Containment Pressure (EAM) 0.5 psig 0.6 psig or Steam Generator Water 10.7% of narrow range 10.1% of narrow range Level -- Low-Low (EAM) instrument span instrument
14. Deleted 56.973 11 15. Undervoltage-Reactor 5022 volts-each bus 4739 volts - each bus Coolant Pumps M24 57.0 56.3 12 16. Underfrequency-Reactor 56.0 Hz - each bus 55.9 Hz - each bus Coolant Pumps 14 17. Turbine Trip 14.a A. Low Trip System 45 psig 39.5 psig Pressure 14.b B. Turbine Stop Valve 1% open > 1% open Closure 15
18. Safety Injection Input from Not Applicable Not Applicable ESF September 20, 2007 SEQUOYAH - UNIT 2 2-7 Amendment Nos. 7, 76, 132, 296, 299, 306 Page 42 of 47 Enclosure 2, Volume 8, Rev. 0, Page 46 of 1148

Enclosure 2, Volume 8, Rev. 0, Page 76 of 1148 DISCUSSION OF CHANGES ITS 3.3.1, REACTOR TRIP SYSTEM (RTS) INSTRUMENTATION requirements to ensure that the automatic protective action will correct the abnormal situation before a safety limit is exceeded. This change is consistent with TSTF-493 Option A. This change is considered a more restrictive change because additional requirements have been added to Surveillance Requirements.

56.973 M24 CTS Table 2.2-1 for Functional Unit 16 (Underfrequency-Reactor Coolant Pumps) lists the Nominal Trip Setpoint as 56.0 Hz - each bus, and the Allowable Value as 55.9 Hz - each bus. ITS Table 3.3.1-1 for Function 12 (Underfrequency RCPs) lists the Nominal Trip Setpoint as 57.0 Hz and the Allowable Value as 56.3 Hz. This changes the CTS by increasing the Nominal Trip Setpoint and the Allowable Value for the Underfrequency RCP reactor trip.

Additionally, TVA is proposing to The purpose of the Underfrequency RCP reactor trip is to ensure that protection change the is provided against violating the DNBR limit due to a loss of flow in two or more Allowable Value RCS loops from a major network frequency disturbance. TVA has determined based on TVA's that to provide adequate protection changes to the Underfrequency RCP Nominal Trip Setpoint and the Allowable Value are needed. This change was value revised calculations previously proposed in SQN license amendment request TVA-SQN-TS-02-01, necessary to Revision 1 (ADAMS Accession No. 042430467) but later withdrawn in TVA-SQN-support the TS-02-01, Revision 2 (ADAMS Accession No. ML061990303) pending resolution implementation of of issues with TSTF-493. In Revision 2 TVA stated that a new TS amendment TSTF-493. request would be submitted to the NRC once TSTF-493 receives NRC approval.

As TSTF-493 has been approved by the NRC and is being adopted under this conversion, TVA is proposing to change the setpoints to those proposed in the previous submittal. This change is acceptable because the revised Allowable Value and Nominal Trip Setpoint continue to provide assurance that the safety limit for the underfrequency reactor trip function is not impacted. In addition, this change ensures instrument uncertainties have been included in the as-found tolerance calculations in a manner that is acceptable and the surveillance Note requirements also ensure that there will be a reasonable expectation that these instruments will perform their safety function if required. This change is designated as more restrictive because more stringent acceptance requirements are being applied in the ITS than were applied in the CTS.

These changes are Nominal Trip S RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) The proposed change removes all designated periodic Surveillance Frequencies from CTS 4.3.1.1.1, as addressed in CTS Table 4.3-1, CTS 4.3.1.1.2, and CTS 4.3.1.1.3, and places the Frequencies under licensee control in accordance with a new program, the Surveillance Frequency Control Program.

ITS 3.3.1 Surveillance Requirements require similar Surveillances and, except for special or conditional frequencies stated in the individual surveillance, specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Sequoyah Unit 1 and Unit 2 Page 25 of 45 Enclosure 2, Volume 8, Rev. 0, Page 76 of 1148

Enclosure 2, Volume 8, Rev. 0, Page 120 of 1148 CTS RTS Instrumentation (Without Setpoint Control Program) 1 3.3.1A 56.973 Table 3.3-1 Table 3.3.1-1 (page 4 of 8)

Table 4.3-1 Reactor Trip System Instrumentation APPLICABLE MODES (l)

OR OTHER [NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP 3 FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT]

12 56.3 57.0 (g) 2 3 17 13. Underfrequency 1 [3] per bus K SR 3.3.1.9 (b)(c) [57.1] Hz [57.5] Hz Table 2.2-1 RCPs SR 3.3.1.10 1 SR 3.3.1.16 9 Function 16 14 13 14 14. Steam Generator 1,2 [4 per SG] E SR 3.3.1.1(b)(c) [30.4]% [32.3]%

Table 2.2-1 (SG) Water Level - SR 3.3.1.7 (b)(c)

Function 13 Low Low INSERT 4 SR 3.3.1.10 SR 3.3.1.16 14 INSERT 5

15. SG Water Level - 1,2 2 per SG E SR 3.3.1.1(b)(c) [30.4]% [32.3]%

Low SR 3.3.1.7 (b)(c) 2 SR 3.3.1.10 SR 3.3.1.16 Coincident with 1,2 2 per SG E SR 3.3.1.1(b)(c) [42.5]% full [40]% full Steam SR 3.3.1.7 (b)(c) steam flow at steam flow at Flow/Feedwater Flow SR 3.3.1.10 RTP RTP SR 3.3.1.16 Mismatch 14 18 2 Table 2.2-1 16. Turbine Trip h Function 17 L 39.5 45 (j) (b)(c)

a. Low Fluid Oil 1 3 N SR 3.3.1.10 [750] psig [800] psig 2 3 Pressure SR 3.3.1.15 13 (j)
b. Turbine Stop 1 4 N SR 3.3.1.10 [1]% open [1]% open 2 3 Valve Closure SR 3.3.1.15 2 M

15 13 19 17. Safety Injection (SI) 1,2 2 trains O SR 3.3.1.14 NA NA 2 Table 2.2-1 Input from Function 18 12 Engineered Safety Feature Actuation System (ESFAS)

DOC M22 (b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

DOC M23 (c) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The NTSP and the methodologies used to determine the as-found and 10 as-left tolerances are specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR 3

by reference]. UFSAR Section 7.1.2 DOC L02 (g) Above the P-7 (Low Power Reactor Trips Block) interlock.

h Note ** (j) Above the P-9 (Power Range Neutron Flux) interlock. 2


REVIEWERS NOTE--------------------------------------------------------------------------------------

(l) Unit specific implementations may contain only Allowable Value depending on Setpoint Study methodology used by the unit. 4 Sequoyah Unit 1 Amendment XXX Westinghouse STS 3.3.1A-20 Rev. 4.0 2 1 Enclosure 2, Volume 8, Rev. 0, Page 120 of 1148

Enclosure 2, Volume 8, Rev. 0, Page 152 of 1148 CTS RTS Instrumentation (Without Setpoint Control Program) 1 3.3.1A 56.973 Table 3.3-1 Table 3.3.1-1 (page 4 of 8)

Table 4.3-1 Reactor Trip System Instrumentation APPLICABLE MODES (l)

OR OTHER [NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP 3 FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT]

12 56.3 57.0 (g) 2 3 17 13. Underfrequency 1 [3] per bus K SR 3.3.1.9 (b)(c) [57.1] Hz [57.5] Hz Table 2.2-1 RCPs SR 3.3.1.10 1 SR 3.3.1.16 9 Function 16 14 13 14 14. Steam Generator 1,2 [4 per SG] E SR 3.3.1.1(b)(c) [30.4]% [32.3]%

Table 2.2-1 (SG) Water Level - SR 3.3.1.7 (b)(c)

Function 13 Low Low INSERT 4 SR 3.3.1.10 SR 3.3.1.16 14 INSERT 5

15. SG Water Level - 1,2 2 per SG E SR 3.3.1.1(b)(c) [30.4]% [32.3]%

Low SR 3.3.1.7 (b)(c) 2 SR 3.3.1.10 SR 3.3.1.16 Coincident with 1,2 2 per SG E SR 3.3.1.1(b)(c) [42.5]% full [40]% full Steam SR 3.3.1.7 (b)(c) steam flow at steam flow at Flow/Feedwater Flow SR 3.3.1.10 RTP RTP SR 3.3.1.16 Mismatch 14 18 2 Table 2.2-1 16. Turbine Trip h Function 17 L 39.5 45 (j) (b)(c)

a. Low Fluid Oil 1 3 N SR 3.3.1.10 [750] psig [800] psig 2 3 Pressure SR 3.3.1.15 13 (j)
b. Turbine Stop 1 4 N SR 3.3.1.10 [1]% open [1]% open 2 3 Valve Closure SR 3.3.1.15 2 M

15 13 19 17. Safety Injection (SI) 1,2 2 trains O SR 3.3.1.14 NA NA 2 Table 2.2-1 Input from Function 18 12 Engineered Safety Feature Actuation System (ESFAS)

DOC M22 (b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

DOC M23 (c) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The NTSP and the methodologies used to determine the as-found and 10 as-left tolerances are specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR 3

by reference]. UFSAR Section 7.1.2 DOC L02 (g) Above the P-7 (Low Power Reactor Trips Block) interlock.

h Note ** (j) Above the P-9 (Power Range Neutron Flux) interlock. 2


REVIEWERS NOTE--------------------------------------------------------------------------------------

(l) Unit specific implementations may contain only Allowable Value depending on Setpoint Study methodology used by the unit. 4 Sequoyah Unit 2 Amendment XXX Westinghouse STS 3.3.1A-20 Rev. 4.0 2 1 Enclosure 2, Volume 8, Rev. 0, Page 152 of 1148

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 379 NRC Question KAB065 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 11/19/2014 Date Notification Scott Bowman Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Kristy Bucholtz Date Added 11/19/2014 6:48 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=379 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 388 NRC Question KAB065 Number Select NRC Response Application Attachment 1

Attachment 2

Response

Statement KAB-065 was inadvertantly closed. Please respond to the following response:

Calculation number SQN-EEB-MS-TI28-0076, Revision 7 was provided as part for response to RAI KAB-065 along with the affected TS pages. Please note that TSTF-493 notes pertaining to as-left and as-found values were not included as part of the TS changes. Please add these notes to the technical specifications. If these notes are detailed in another document then reference the appropriate documents in the TS affected pages. Also please provide the wording of the notes and the values for as-left and as-found terms for staff review.

Response

12/1/2014 6:00 PM Date/Time Closure Statement Question Closure Date Notification Scott Bowman Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Kristy Bucholtz Date Added 12/1/2014 12:58 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=388 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 3 Licensee Response/NRC Response/NRC Question Closure Id 400 NRC Question KAB065 Number Select Licensee Response Application Attachment 1

Attachment 2

Response

Statement By response dated October 16, 2014, SQN responded to RAI KAB065. As part of the response, SQN provided Attachments 1 and 2. Attachment 1 contained a portion of ITS Table 3.3.1-1. The table reflects that SR 3.3.1.10 is required for ITS 3.3.1, Function 12 (Underfrequency RCP). SR 3.3.1.10 (Perform CHANNEL CALIBRATION) has two associated footnotes, (b) and (c). Footnote (b) states, If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. Footnote (c) states, The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and as-left tolerances are specified in UFSAR Section 7.1.2. An FSAR change is currently in progress to provide the methodologies in Section 7.1.2 of the UFSAR, and will be complete prior to implementation of ITS.

Attachment 2 (previously provided), is calculation SQN-EEB-MS-TI28-0076, Revision 7, Demonstrated Accuracy Calculation RCP UNDERFREQUENCY RELAYS. The As-Found and As-Left values are located on pdf page 53 https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=400 1/13/2015

Sequoyah ITS Conversion Database Page 2 of 3 (sheet 26) of Attachment 2. The As-Found (Afc) value is

+/-0.011 Hz. The calculation for the As-Found value is performed on pdf page 49 (sheet 24). The As-Left (Ab) value is +/-0.011 Hz. The calculation for the As-Left value is performed on pdf page 38 (sheet 16A).

As-Found and As-Left values are controlled through Setpoint and Scaling Documents (SSDs). SSDs serve as the design output document to transmit the requirements to site organizations to ensure values assessed in the safety analyses and/or other design documents relative to instrument setpoints, scaling and calibration are in fact incorporated in the plant as assessed in the relevant design documents. Changes to As-Found and/or As-Left values require a Design Change to be processed via the Engineering Change Process. The As-Found and As-Left values listed in the SSDs are incorporated into Surveillance Instructions (SIs) that are performed to verify Technical Specification Surveillance Requirements. The SIs are annotated with requirements to evaluate setpoints found outside the As-Found tolerances to verify the channel is functioning as required before returning the channel to service. Additionally, this condition will be entered into the Corrective Action Program. The SIs also require that an instrument channel shall be declared inoperable if it cannot be reset to within the As-Left tolerance.

Response

12/16/2014 2:00 PM Date/Time Closure Statement Question Closure Date Notification Scott Bowman Kristy Bucholtz Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Scott Bowman https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=400 1/13/2015

Sequoyah ITS Conversion Database Page 3 of 3 Date Added 12/16/2014 12:56 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=400 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 404 NRC Question KAB065 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 12/18/2014 Date Notification Scott Bowman Kristy Bucholtz Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 12/18/2014 2:32 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=404 1/13/2015

Sequoyah ITS Conversion Database Page 1 of 3 ITS NRC Questions Id 190 NRC Question KAB066 Number Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s)

NRC Reviewer Roger Pederson Supervisor Technical Mark Blumberg Branch POC Conf Call N

Requested NRC RAI ARCB2-1 (in response to KAB-044)

Question In a letter dated November 7, 2013 (ADAMS Accession No. ML13246A358),

the NRC informed the Technical Specifications Task Force of concerns that the NRC staff had recently identified during a review of plant-specific license amendments requesting adoption of three travelers including traveler TSTF-51, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations.

TSTF-51 states, in part, that The addition of the term recently associated with handling irradiated fuel in all of the containment function Technical Specification requirements is only applicable to those licensees who have demonstrated by analysis that after sufficient radioactive decay has occurred, off-site doses resulting from a fuel handling accident remain below the Standard Review Plan limits (well within 10CFR100). [or 10 CFR 50.67]

NUREG-0800, Standard Review Plan (SRP) 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, July 2000 (ADAMS Accession No. ML003734190), states, in part, that https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=190 04/27/2015

Sequoyah ITS Conversion Database Page 2 of 3 The models, assumptions, and parameter inputs used by the licensee should be reviewed to ensure that the conservative design basis assumptions outlined in RG-1.183 have been incorporated.

Appendix B of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000 (ADAMS Accession No. ML003716792),

Regulatory Position 1.1 states:

The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or weight of a dropped fuel assembly...

After reviewing the information submitted by the licensee to adopt changes to the Improved Technical Specifications (ITS) (that incorporated TSTF-51),

the Nuclear Regulatory Commission (NRC) Staff is concerned that the licensee has not provided an analysis that will provide the NRC Staff reasonable assurance that the fuel handling accident (FHA) doses remain within regulatory limits (i.e. when to reference to irradiated fuel and Mode 6 are removed from the APPLICABILITY of several technical specifications and the words suspend all operations involving movement of fuel within the spent fuel pit or crane operations with loads over the spent fuel pit are removed from ACTION statements). The analysis provided in Calculation LTR-CRA-02-219, Revision 1, Radiological Consequences of Fuel Handling Accidents for the Sequoyah Nuclear Plant, Units 1 and 2, does not appear to address this scenario and therefore, does not justify the proposed changes.

For the proposed change please provide an FHA analysis that evaluates the dropping of loads allowed over irradiated fuel assemblies (i.e. sources, new fuel, tools, reactivity control components) onto irradiated fuel assemblies. The analysis should only credit those safety systems required to be operable as required by technical specification. Provide the inputs, assumptions and methodology used, and the results. Provide a justification for any assumptions made. Although it is not required the staff has found it more efficient if the licensees calculation is provided. A calculation may not need to be performed if Sequoyah chooses to limit the movement of loads over irradiated fuel prior to the decay time assumed in the accident analysis. If this option is chosen, please provide the appropriate licensing changes.

Attach File 1

Attach File 2

Issue Date 9/30/2014 Added By Khadijah Hemphill Date Modified https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=190 04/27/2015

Sequoyah ITS Conversion Database Page 3 of 3 Modified By Date Added 9/30/2014 4:00 PM Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=190 04/27/2015

Sequoyah ITS Conversion Database Page 1 of 5 Licensee Response/NRC Response/NRC Question Closure Id 413 NRC Question KAB066 Number Select Licensee Response Application Attachment Attachment 1 for KAB066 12_17_2014.pdf (1MB) 1 Attachment Attachment 2 for KAB066 12_18_2014.pdf (2MB) 2

Response

Statement In response to RAI KAB066, the following information is provided.

The SQN fuel handling accident (FHA) dose consequences analysis is based on damage to an irradiated fuel assembly that has met a decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> and a decontamination factor (DCF) of 200 that is applied to the overall iodine inventory release to the pool. The SQN ITS license amendment request, as submitted, does not provide a specific technical specification to verify that fuel assemblies decay for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to movement, and does not ensure that irradiated fuel assemblies in the spent fuel pool are covered by at least 23 feet of water, at all times.

Therefore, the following changes are proposed for the SQN ITS.

CTS 3.9.3, Decay Time, will be retained in ITS as ITS 3.9.8, Decay Time. CTS 3.9.3 Applicability will be revised to, During CORE ALTERATIONS. The Frequency for CTS 4.9.3 will be revised to, Prior to CORE ALTERATIONS. Discussion of Change (DOC)

M01, as well as DOC M01 indicators, will be added to the submittal to justify the changes to the CTS Mode of Applicability and Frequency. As a result of the addition of ITS 3.9.8, the following changes will be necessary:

1. The CTS markups will be revised. (Pages 232 and 233 of Enclosure 2, Volume 14)
2. The Discussion of Changes Section will be retitled, Discussion of Changes ITS 3.9.8, Decay Time. DOCs A01 and M01 will be added to this section as Inserts 1 and 2.

DOC LA01 will be revised, as shown in Insert 3. (Page 234 of Enclosure 2, Volume 14)

3. ITS 3.9.8 and the Bases for ITS 3.9.8 will be added to the submittal. (Insert 4 located after the inserts for CTS 3.9.3 Discussion of Changes in Enclosure 2, Volume 14) https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=413 04/27/2015

Sequoyah ITS Conversion Database Page 2 of 5 The CTS definition for CORE ALTERATION (CTS 1.9 CORE ALTERATION) will be retained in the ITS 1.1 Definitions Section (pages 55 and 85 of Enclosure 2, Volume 3). A new Justification for Deviation (JFD) 7, as well as JFD 7 indicators, will be added to justify the change to the ISTS. As a result of the addition of the definition for CORE ALTERATION, the following changes will be necessary:

1. The CTS markups will be revised. (Pages 7 and 24 of Enclosure 2, Volume 3)
2. DOC A06 will be revised to remove CORE ALTERATION from the list of deleted CTS definitions. (Page 45 of Enclosure 2, Volume 3)
3. JFD 7 will be added to the Justification for Deviations ITS 1.0, Use and Application. (Page 114 of Enclosure 2, Volume 3)

ITS 3.7.13 (ISTS 3.7.15), on pages 513 and 514 of Enclosure 2, Volume 12, will be revised to change the Mode of Applicability and add ITS 3.7.13 Required Action A.2. The Mode of Applicability will be revised to, Whenever irradiated fuel assemblies are in the spent fuel pool. ITS 3.7.13 ACTION A will be revised to include Required Action A.2. ITS 3.7.13 Required Action A.2 will require restoration of the spent fuel pool level to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the spent fuel pool water level is less than 23 feet.

Additionally, ITS 3.7.13 ACTION A will be revised so that the NOTE, LCO 3.0.3 is not applicable, applies to both ITS Required Action A.1 and A.2. JFDs 4 and 5 will be added to the Justification for Deviations Section to justify the change to the Mode of Applicability and the addition of Required Action A.2. The changes to ITS 3.7.13 Mode of Applicability and ACTION A reflect SQNs current licensing basis as reflected in CTS 3.9.11. As a result of the revisions described above, the following changes will be necessary:

1. The CTS markups will be revised. (Pages 487 and 498 of Enclosure 2, Volume 12)
2. DOC L01, associated with changes to the CTS Mode of Applicability and the action to restore the spent fuel pool water level, will be deleted, as well as DOC L01 indicators.

(Pages 487, 498, and 510 of Enclosure 2, Volume 12)

3. The ISTS markups will be revised, as discussed above, and JFD 4 and 5 indicators will be added. (Pages 513 and 514 of Enclosure 2, Volume 12) https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=413 04/27/2015

Sequoyah ITS Conversion Database Page 3 of 5

4. JFDs 4 and 5 will be added to the Justification for Deviations Section. (Page 515 of Enclosure 2, Volume 12)
5. The ISTS 3.7.15 Bases will be revised to align with changes made to the Specification. (Pages 518 and 521 of Enclosure 2, Volume 12)
6. JFDs 6 and 7, as well as JFD 6 and 7 indicators, will be added to the Justification for Deviations Bases Section.

(Pages 518, 521, and 523 of Enclosure 2, Volume 12)

Additionally, ITS LCO 3.0.3 Bases, on pages 45 and 60 of Enclosure 2, Volume 5, will be revised. The Bases for LCO 3.0.3 describes exceptions to LCO 3.0.3 and provides ITS LCO 3.7.13 as an example. Because of the changes described above to ITS 3.7.13, the example in the Bases for LCO 3.0.3 has been revised to align with changes made to ITS 3.7.13 Specification.

The changes to ITS 3.9.8 and the addition of the definition for CORE ALTERATION provide an explicit requirement that the decay time of the reactor be greater than or equal to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to commencing of CORE ALTERATIONS. In a letter dated November 7, 2013, (ADAMS Accession No. ML13246A358), the NRC stated a concern with CORE ALTERATIONS prior to the assumed decay time. Specifically, the NRC's concerns were associated with related changes with the following Technical Specification Task Force (TSTF) changes:

1. TSTF-51, Revision 2, "Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations,"

approved on November 1, 1999 (ADAMS Accession No. ML993190284), and

2. TSTF-471, Revision 1, "Eliminate Use of Term Core Alterations in Actions and Notes," approved on December 7, 2006 (ADAMS Accession No. ML062860320).

In this letter the NRC stated, "The NRC staff is concerned that a dropped source, fuel assembly, or component (or any other item allowed to be moved by CORE ALTERATIONS) could damage or break a fuel assembly creating a radioactive source term.

Additionally, a dropped source, component, or fuel assembly could add reactivity if it is dropped over or in the vicinity of other fuel. Therefore, SQN will limit CORE ALTERATIONS to a decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=413 04/27/2015

Sequoyah ITS Conversion Database Page 4 of 5 In addition, as established in the "Radiological Consequences of Fuel Handling Accidents for the Sequoyah Nuclear Plant Unit 1 and 2, LTR-CRA-02-219 Revision 2," the minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to CORE ALTERATIONS, in conjunction with the requirements of LCO 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS)," LCO 3.7.11, "Control Room Air-Conditioning System (CRACS)," LCO 3.8.2, "AC Sources-Shutdown," LCO 3.8.10, "Distribution Systems-Shutdown," LCO 3.9.1, "Boron Concentration," LCO 3.9.3, "Nuclear Instrumentation," LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level," and LCO 3.9.7, "Refueling Cavity Water Level," ensures that the release of fission product radioactivity from a FHA at SQN results in doses that are within the requirments of 10 CFR 50.67 and Regulatory Position C.4.4 of Regulatory Guide 1.183.

The change to ITS 3.7.13 ensures that the DCF (200) used in the radiological consequences of a FHA at SQN remain valid. ITS 3.7.13 will ensure that there is 23 feet of water above the top of the irradiated fuel assemblies stored in the racks in the spent fuel pool.

See Attachment 1 for draft changes associated with ITS 3.9.8 and the inclusion of the definition of CORE ALTERATIONS in ITS 1.1.

See Attachment 2 for draft changes associated with ITS 3.7.13 and the Bases for ITS LCO 3.0.3.

Response

12/30/2014 9:25 PM Date/Time Closure Statement Question Closure Date Notification Mark Blumberg Scott Bowman Kristy Bucholtz Margaret Chernoff Michelle Conner Robert Elliott Ravinder Grover Matthew Hardgrove Khadijah Hemphill Andrew Hon Lynn Mynatt Amrit Patel Ray Schiele https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=413 04/27/2015

Sequoyah ITS Conversion Database Page 5 of 5 Added By Michelle Conner Date Added 12/30/2014 8:26 PM Date 1/5/2015 9:49 AM Modified Modified By Scott Bowman https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=413 04/27/2015

ITS Enclosure 2, Volume 3, Rev. 0, Page 7 of 117 Chapter 1.0 DEFINITIONS OPERATIONAL A04 CHANNEL CHANNEL FUNCTIONAL TEST (COT)

OPERATIONAL A01 TEST COT 1.6 A CHANNEL FUNCTIONAL TEST shall be:

or actual L01

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions. A04 INSERT 3
b. Bistable channels - the injection of a simulated signal into the sensor to verify A05 OPERABILITY including alarm and/or trip functions.
c. Digital channels - the injection of a simulated signal into the channel as close to the sensor A04 input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:
1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.

A06

b. All equipment hatches are closed and sealed.
c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
d. The containment leakage rates are within the limits of Specification 4.6.1.1.c,
e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE, and
f. Secondary containment bypass leakage is within the limits of Specification 3.6.3.

CONTROLLED LEAKAGE A07 1.8 This definition has been deleted. stet CORE CORE ALTERATION A01 ALTERATION 1.9 CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the A06 vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. S CORE OPERATING (COLR) parameter LIMITS CORE OPERATING LIMIT REPORT REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall A01 be determined for each reload cycle in accordance with Specification 6.9.1.14. Unit operation within these cycle operating limits is addressed in individual specifications.

specific 5.6.3. Plant parameter April 13, 2009 SEQUOYAH - UNIT 1 1-2 Amendment No. 12, 71, 130, 141, 155 176, 201, 203, 259, 323 Page 3 of 37 Enclosure 2, Volume 3, Rev. 0, Page 7 of 117

ITS Enclosure 2, Volume 3, Rev. 0, Page 24 of 117 Chapter 1.0 DEFINITIONS A04 OPERATIONAL CHANNEL CHANNEL FUNCTIONAL TEST (COT)

OPERATIONAL COT A01 TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be: or actual L01

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions. A04 INSERT 3
b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY A05 including alarm and/or trip functions.
c. Digital channels - the injection of a simulated signal into the channel as close to the sensor A04 input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:
1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as A06 permitted by Specification 3.6.3.
b. All equipment hatches are closed and sealed.
c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
d. The containment leakage rates are within the limits of Specification 4.6.1.1.c,
e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE, and
f. Secondary containment bypass leakage is within the limits of Specification 3.6.3.

CONTROLLED LEAKAGE A07 1.8 This definition has been deleted. stet CORE CORE ALTERATION ALTERATION A01 1.9 CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the A06 vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING CORE OPERATING LIMITS REPORT (COLR)

LIMITS parameter REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be A01 determined for each reload cycle in accordance with Specification 6.9.1.14. Unit operation within these cycle operating limits is addressed in individual specifications.

specific 5.6.3. Plant parameter April 13, 2009 SEQUOYAH - UNIT 2 1-2 Amendment Nos. 63, 117, 132, 146, 167, 191, 193, 250, 315 Page 20 of 37 Enclosure 2, Volume 3, Rev. 0, Page 24 of 117

Enclosure 2, Volume 3, Rev. 0, Page 45 of 117 DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS to the CHANNEL FUNCTIONAL TEST for digital channels was consistent with the existing channel functional test definition and therefore acceptable.

These changes are designated as administrative because they do not result in a technical change to the Technical Specifications.

A05 CTS Section 1.0 includes a CHANNEL FUNCTIONAL TEST definition for bistable channels. The definition of CHANNEL FUNCTIONAL TEST for bistable channels requires "the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions. However, this CTS definition is essentially duplicative of the TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT) definition. ITS Section 1.1 does not include this definition, since the requirements for bistable channels are covered by the TADOT definition.

This change is acceptable because the TADOT definition adequately covers bistable channels, and does not impose any new requirements or alter any existing requirements. This change is categorized as administrative because the bistable portion of the definition is duplicative of the TADOT definition.

A06 CTS Section 1.0 includes the following definitions:

  • CONTAINMENT INTEGRITY
  • GASEOUS RADWASTE TREATMENT SYSTEM
  • PURGE - PURGING
  • SITE BOUNDARY
  • UNRESTRICTED AREA
  • VENTILATION EXHAUST TREATMENT SYSTEM
  • VENTING
  • E - AVERAGE DISINTEGRATION ENERGY
  • CORE ALTERATION The ITS does not use this terminology and ITS Section 1.1 does not contain these definitions.

These changes are acceptable because the terms are not used as defined terms in the ITS. Discussions of any technical changes related to the deletion of these terms are included in the DOCs for the CTS sections in which the terms are used. These changes are designated as administrative because they eliminate defined terms that are no longer used.

A07 CTS Section 1.0 shows the following definitions as being deleted:

  • CONTROLLED LEAKAGE
  • MEMBER(S) OF THE PUBLIC
  • REPORTABLE EVENT
  • SOLIDIFICATION
  • SOURCE CHECK Sequoyah Unit 1 and 2 Page 4 of 11 Enclosure 2, Volume 3, Rev. 0, Page 45 of 117

Enclosure 2, Volume 3, Rev. 0, Page 55 of 117 Definitions CTS 1.1 1.1 Definitions 1.6 CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal TEST (COT) into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

1.9 CORE ALTERATION 7 1.10 CORE OPERATING LIMITS The COLR is the unit specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle.

These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.3.

Plant operation within these limits is addressed in individual Specifications.

1.11 DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be TSTF-490 those listed in [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or those listed in Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity"]. TSTF-INSERT 1 490

- AVERAGE shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and TSTF-490 gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > [15] minutes, making up at least 95% of the total noniodine activity in the coolant.

CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 1.1-2 Rev. 4.0 1 Enclosure 2, Volume 3, Rev. 0, Page 55 of 117

Enclosure 2, Volume 3, Rev. 0, Page 85 of 117 Definitions CTS 1.1 1.1 Definitions 1.6 CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal TEST (COT) into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

1.9 CORE ALTERATION 7 1.10 CORE OPERATING LIMITS The COLR is the unit specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle.

These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.3.

Plant operation within these limits is addressed in individual Specifications.

1.11 DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be TSTF-490 those listed in [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or those listed in Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity"]. TSTF-INSERT 1 490

- AVERAGE shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and TSTF-490 gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > [15] minutes, making up at least 95% of the total noniodine activity in the coolant.

CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 1.1-2 Rev. 4.0 1 Enclosure 2, Volume 3, Rev. 0, Page 85 of 117

Enclosure 2, Volume 3, Rev. 0, Page 114 of 117 JUSTIFICATION FOR DEVIATIONS ITS 1.0, USE AND APPLICATION

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.
3. Typographical error is corrected. The proper section for Surveillance Requirement (SR)

Applicability is Section 3.0.

4. These punctuation corrections have been made consistent with the Writers Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
5. Typographical error is corrected.
6. The ISTS definition of Shutdown Margin states in part, "However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation." The CTS definition of Shutdown Margin does not contain this allowance, therefore the ITS does not include this allowance. This is acceptable since the information is changed to reflect the current licensing basis.
7. The ISTS does not contain a definition for CORE ALTERATION. The CTS definition for CORE ALTERATION has been included in ITS. This change is acceptable because the information reflects the current licensing basis.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 3, Rev. 0, Page 114 of 117

Enclosure 2, Volume 14, Rev. 0, Page 2 of 236 LIST OF ATTACHMENTS

1. ITS 3.9.1 - Boron Concentration
2. ITS 3.9.2 - Unborated Water Source Isolation Valves
3. ITS 3.9.3 - Nuclear Instrumentation
4. ITS 3.9.4 - Containment Penetrations
5. ITS 3.9.5 - Residual Heat Removal (RHR) and Coolant Circulation - High Water Level
6. ITS 3.9.6 - Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level
7. ITS 3.9.7 - Refueling Cavity Water Level
8. Relocated/Deleted Current Technical Specifications (CTS)

ITS 3.9.8 - Decay Time Enclosure 2, Volume 14, Rev. 0, Page 2 of 236

, Volume 14, Rev. 0, Page 229 of 236 ATTACHMENT 8 ITS 3.9.8, DECAY TIME RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 14, Rev. 0, Page 229 of 236

, Volume 14, Rev. 0, Page 230 of 236 CTS 3/4.9.3, DECAY TIME , Volume 14, Rev. 0, Page 230 of 236

, Volume 14, Rev. 0, Page 231 of 236 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 231 of 236

Enclosure 2, Volume 14, Rev. 0, Page 232 of 236 ITS A01 CTS 3/4.9.3 REFUELING OPERATIONS 3/4 9.3 DECAY TIME LIMITING CONDITION FOR OPERATION STET LCO 3.9.8 3.9.3 The reactor shall be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

APPLICABILITY:

APPLICABILITY During movement or irradiated fuel in the reactor pressure vessel.

M01 ACTION: APPLICABILITY: During CORE ALTERATIONS STET ACTION A With the reactor subcritical for less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

LA01 M01 Suspend CORE ALTERATIONS SURVEILLANCE REQUIREMENTS STET SR 3.9.8.1 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

M01 Prior to CORE ALTERATIONS SEQUOYAH - UNIT 1 3/4 9-3 Page 1 of 2 Enclosure 2, Volume 14, Rev. 0, Page 232 of 236

Enclosure 2, Volume 14, Rev. 0, Page 233 of 236 ITS A01 CTS 3/4.9.3 REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION STET LCO 3.9.8 3.9.3 The reactor shall be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.

APPLICABILITY APPLICABILITY: During CORE ALTERATIONS M01 ACTION:

STET ACTION A With the reactor subcritical for less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

LA01 Suspend CORE ALTERATIONS M01 SURVEILLANCE REQUIREMENT STET SR 3.9.8.1 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

Prior to CORE ALTERATIONS M01 SEQUOYAH - UNIT 2 3/4 9-4 Page 2 of 2 Enclosure 2, Volume 14, Rev. 0, Page 233 of 236

Enclosure 2, Volume 14, Rev. 0, Page 234 of 236 DISCUSSION OF CHANGES ITS 3.9.8 CTS 3/4.9.3, DECAY TIME ADMINISTRATIVE CHANGES INSERT 1 None MORE RESTRICTIVE CHANGES INSERT 2 None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES INSERT 3 LA01 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, or ISI Program) CTS 3.9.3 requires the reactor to be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during movement or irradiated fuel in the reactor pressure vessel. ITS 3.9 does not include the requirement for decay time. This changes the CTS by moving the explicit decay time requirements from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS LCO 3.9.3 is to ensure that sufficient time has elapsed to allow radioactive decay of the short-lived fission products in the irradiated fuel consistent with the assumptions used in the fuel handling accident analysis. This change is acceptable because the removed information will be adequately controlled in the TRM. Changes to the TRM are controlled by the provisions of 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as less restrictive removal of detail change because a requirement is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 14, Rev. 0, Page 234 of 236

INSERT 1 A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

INSERT 2 M01 CTS LCO 3.9.3 Applicability is, During movement of [SIC for SQN Unit 1] irradiated fuel in the reactor pressure vessel. CTS 3.9.3 ACTION requires, in part, suspending all operations involving the movement of irradiated fuel in the reactor pressure vessel, when the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time is not met. CTS Surveillance Requirement 4.9.3 states, The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verification of the date and time of subcriticality prior to the movement of irradiated fuel in the reactor pressure vessel. ITS LCO 3.9.8 Applicability is, During CORE ALTERATIONS. ITS 3.9.8 Required Action A.1 requires the suspension of CORE ALTERATIONS. ITS SR 3.9.8.1 requires verification that the reactor has been subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> with a Frequency of prior to CORE ALTERATIONS (See DOC LA01 for a discussion concerning the removal of the requirement to verify subcriticality by date and time). This changes the CTS Applicability, ACTION, and Surveillance Requirement by replacing the phrase, during movement of irradiated fuel in the reactor pressure vessel, with the phrase, during CORE ALTERATIONS.

These changes provide an explicit requirement that the decay time of the reactor be greater than or equal to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to commencing CORE ALTERATIONS. In a letter dated November 7, 2013, (ADAMS Accession No. ML13246A358), the NRC stated a concern with CORE ALTERATIONS prior to the assumed decay time. Specifically, the Staff's concerns were associated with related changes with the following Technical Specification Task Force (TSTF) changes:

1. TSTF-51, Revision 2, "Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations," approved on November 1, 1999 (ADAMS Accession No. ML993190284), and
2. TSTF-471, Revision 1, "Eliminate Use of Term Core Alterations in Actions and Notes,"

approved on December 7, 2006 (ADAMS Accession No. ML062860320).

In this letter the NRC stated, "The NRC staff is concerned that a dropped source, fuel assembly, or component (or any other item allowed to be moved by CORE ALTERATIONS) could damage or break a fuel assembly creating a radioactive source term. Additionally, a dropped source, component, or fuel assembly could add reactivity if it is dropped over or in the vicinity of other fuel. Therefore, SQN will limit both the movement of irradiated fuel assemblies in the reactor pressure vessel and CORE ALTERATIONS to a decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. This change is designated as more restrictive because the Applicability of the Specification has been expanded.

INSERT 3 LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS Surveillance Requirement 4.9.3 states that, The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verification of the date and time of subcriticality prior to the movement of irradiated fuel in the reactor pressure vessel. ITS SR 3.9.8.1 states, Verify the reactor has been subcritical for > 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. ITS SR 3.9.8.1 does not contain the details on the methods of verification of subcriticality. This changes the CTS by moving details on methods of verification of subcriticality to the ITS 3.9.8 Bases. Additionally, the Frequency of prior to movement of irradiated fuel in the reactor pressure vessel, is being changed to, Prior to CORE ALTERATIONS. This change is discussed in Discussion of Change (DOC) M01.

The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to determine that reactor has been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to commencing CORE ALTERATIONS. This change is acceptable, because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because details for meeting Technical Specification requirements are being removed from the Technical Specifications to the ITS Bases. the

INSERT 4 CTS Decay Time 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Decay Time 3.9.3 LCO 3.9.8 The reactor shall be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

Applicability APPLICABILITY: During CORE ALTERATIONS.

M01 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION M01 A. Reactor subcritical for A.1 Suspend CORE Immediately

< 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. ALTERATIONS.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.3 M01 SR 3.9.8.1 Verify the reactor has been subcritical for 100 Prior to CORE hours. ALTERATIONS SEQUOYAH - UNIT 1 3.9.8-1 Amendment XXX

INSERT 4 (continued)

CTS Decay Time 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Decay Time 3.9.3 LCO 3.9.8 The reactor shall be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

Applicability APPLICABILITY: During CORE ALTERATIONS.

M01 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION M01 A. Reactor subcritical for A.1 Suspend CORE Immediately

< 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. ALTERATIONS.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.3 M01 SR 3.9.8.1 Verify the reactor has been subcritical for 100 Prior to CORE hours. ALTERATIONS SEQUOYAH - UNIT 2 3.9.8-1 Amendment XXX

INSERT 4 (continued)

JUSTIFICATION FOR DEVIATIONS ITS 3.9.8, DECAY TIME

1. None.

Sequoyah Unit 1 and Unit 2 Page 1 of 1

INSERT 4 (continued)

Decay Time B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Decay Time BASES BACKGROUND The primary purpose of the decay time requirement is to ensure that the fission product inventories assumed in the fuel handling accident analysis are met. As soon as the reactor is subcritical, the quantity of fission products in the core decreases as the fission products undergo natural radioactive decay. As long as the reactor remains subcritical, this decrease will continue and the radiation levels will also decrease.

APPLICABLE The fuel handling accident is the postulated event of concern in MODE 6 SAFETY during fuel handling operations (Ref. 1). It establishes the minimum ANALYSES decay time. It is assumed that all of the fuel rods in the equivalent of one fuel assembly are damaged to the extent that all the gap activity in the rods is released. The damaged fuel assembly is assumed to be the assembly with the highest fission product inventory. The fission product inventories are those assumed to be present 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the reactor becomes subcritical.

The decay time satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that the reactor be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to commencing CORE ALTERATIONS. The requirement to be subcritical for greater than or equal to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> ensures that the fission product radioactivity has undergone natural radioactive decay and that the consequences of a fuel handling accident will be within the bounds of the safety analysis.

APPLICABILITY This LCO applies during CORE ALTERATIONS, since the potential for a release of fission products exists.

ACTIONS A.1 With the reactor subcritical for less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, there shall be no operations involving CORE ALTERATIONS. This will preclude a fuel handling accident with fuel containing more fission product radioactivity than assumed in the safety analysis.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.

SEQUOYAH - UNIT 1 B 3.9.8-1 Revision XXX

INSERT 4 (continued)

Decay Time B 3.9.8 BASES SURVEILLANCE SR 3.9.8.1 REQUIREMENTS Prior to CORE ALTERATIONS, the reactor must be determined to be subcritical for greater than or equal to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verifying the date and time that the reactor achieved subcritical conditions.

REFERENCES 1. UFSAR, Section 15.5.6.

SEQUOYAH - UNIT 1 B 3.9.8-2 Revision XXX

INSERT 4 (continued)

Decay Time B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Decay Time BASES BACKGROUND The primary purpose of the decay time requirement is to ensure that the fission product inventories assumed in the fuel handling accident analysis are met. As soon as the reactor is subcritical, the quantity of fission products in the core decreases as the fission products undergo natural radioactive decay. As long as the reactor remains subcritical, this decrease will continue and the radiation levels will also decrease.

APPLICABLE The fuel handling accident is the postulated event of concern in MODE 6 SAFETY during fuel handling operations (Ref. 1). It establishes the minimum ANALYSES decay time. It is assumed that all of the fuel rods in the equivalent of one fuel assembly are damaged to the extent that all the gap activity in the rods is released. The damaged fuel assembly is assumed to be the assembly with the highest fission product inventory. The fission product inventories are those assumed to be present 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the reactor becomes subcritical.

The decay time satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that the reactor be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to commencing CORE ALTERATIONS. The requirement to be subcritical for greater than or equal to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> ensures that the fission product radioactivity has undergone natural radioactive decay and that the consequences of a fuel handling accident will be within the bounds of the safety analysis.

APPLICABILITY This LCO applies during CORE ALTERATIONS, since the potential for a release of fission products exists.

ACTIONS A.1 With the reactor subcritical for less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, there shall be no operations involving CORE ALTERATIONS. This will preclude a fuel handling accident with fuel containing more fission product radioactivity than assumed in the safety analysis.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.

SEQUOYAH - UNIT 2 B 3.9.8-1 Revision XXX

INSERT 4 (continued)

Decay Time B 3.9.8 BASES SURVEILLANCE SR 3.9.8.1 REQUIREMENTS Prior to CORE ALTERATIONS, the reactor must be determined to be subcritical for greater than or equal to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verifying the date and time that the reactor achieved subcritical conditions.

REFERENCES 1. UFSAR, Section 15.5.6.

SEQUOYAH - UNIT 2 B 3.9.8-2 Revision XXX

INSERT 4 (continued)

JUSTIFICATION FOR DEVIATIONS ITS 3.9.8 BASES, DECAY TIME

1. None.

Sequoyah Unit 1 and Unit 2 Page 1 of 1

Enclosure 2, Volume 14, Rev. 0, Page 236 of 236 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.8 CTS 3/4.9.3, DECAY TIME There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 14, Rev. 0, Page 236 of 236

Enclosure 2, Volume 12, Rev. 0, Page 487 of 704 ITS A01 ITS 3.7.13 REFUELING OPERATIONS 3/4.9.11 SPENT FUEL PIT WATER LEVEL POOL A01 LIMITING CONDITION FOR OPERATION LCO 3.7.13 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

During movement of irradiated fuel assemblies in the spent fuel pool. L01 Applicability APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pit.

irradiated L02 pool A01 ACTION: stet immediately A02 ACTION A With the requirements of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within L02 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The provisions of Specification 3.0.3 are not applicable.

ACTION A stet L01 Note pool SURVEILLANCE REQUIREMENTS A01 SR 3.7.13.1 4.9.11 The water level in the spent fuel pit shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pit. L01 pool A01 in accordance with the Surveillance stet LA01 Frequency Control Program SEQUOYAH - UNIT 1 3/4 9-11 Page 1 of 22 Enclosure 2, Volume 12, Rev. 0, Page 487 of 704

Enclosure 2, Volume 12, Rev. 0, Page 498 of 704 ITS A01 ITS 3.7.13 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL-SPENT FUEL PIT POOL A01 LIMITING CONDITION FOR OPERATION LCO 3.7.13 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

During movement of irradiated fuel assemblies in the spent fuel pool. L01 Applicability APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pit. A01 irradiated L02 pool ACTION: stet immediately A02 ACTION A With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit L02 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The provisions of Specification 3.0.3 are not applicable.

ACTION A stet L01 Note pool SURVEILLANCE REQUIREMENTS A01 SR 3.7.13.1 4.9.11 The water level in the spent fuel pit shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pit. L01 pool A01 stet in accordance with the Surveillance LA01 Frequency Control Program SEQUOYAH - UNIT 2 3/4 9-13 Page 12 of 22 Enclosure 2, Volume 12, Rev. 0, Page 498 of 704

Enclosure 2, Volume 12, Rev. 0, Page 510 of 704 DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequency is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.9.11 Applicability states "Whenever irradiated fuel assemblies are in the spent fuel pit." CTS SR 4.9.11 requires the water level in the spent fuel pit to be verified every 7 days when Not Used irradiated fuel assemblies are in the spent fuel pit. ITS 3.7.13 is applicable "During movement of irradiated fuel assemblies in the spent fuel pool." ITS SR 3.0.1 requires ITS SR 3.7.13.1 to be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. In addition, since the Applicability is now limited to when irradiated fuel is being moved, the CTS ACTION to "restore water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after movement of fuel has been suspended" has also been deleted. This changes the CTS by restricting the Applicability of the spent fuel pool water level Specification and performance of the Surveillance to when there is a potential for a fuel handling accident, i.e., during the movement of irradiated fuel assemblies in the spent fuel pool.

The purpose of CTS 3.9.11 is to ensure that the minimum spent fuel pit water level assumption in the fuel handling accident analysis is met. This change is acceptable because the requirements continue to ensure that the conditions assumed in the safety analyses and licensing basis are maintained. The SQN fuel handling accident analysis (outside containment) assumes that a single fuel assembly is damaged. A key assumption in the analysis is that there is 23 feet of water over the damaged assembly, as this depth is directly related to the cleanup of the fission products before release from the spent fuel pool. A fuel handling accident is only assumed to occur when an irradiated fuel assembly is being moved. Therefore, ITS 3.7.13 imposes controls on minimum spent fuel pool water level only during the movement of irradiated fuel assemblies in the spent fuel pool. ITS 4.3.2 specifies the requirement that the spent fuel pool be designed and maintained to prevent inadvertent draining of the pool below elevation 722. This change is designated as less restrictive because the ITS LCO requirements are applicable in fewer operating conditions than in the CTS.

Sequoyah Unit 1 and Unit 2 Page 2 of 3 Enclosure 2, Volume 12, Rev. 0, Page 510 of 704

Enclosure 2, Volume 12, Rev. 0, Page 511 of 704 DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL L02 (Category 4 - Relaxation of Required Action) CTS 3.9.11 ACTION states that when the spent fuel pit water level is not met, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas. ITS 3.7.13 Required Action A.1 states that when spent fuel pool water level is not within limits, immediately suspend movement of irradiated fuel assemblies in the spent fuel pool. This changes the CTS by deleting the requirements to suspend movement of new fuel and to suspend crane operation over the spent fuel storage areas.

The purpose of the CTS 3.9.11 ACTION is to preclude a fuel handling accident from occurring when the initial conditions for that accident are not met. A fuel handling accident is only assumed to occur when an irradiated fuel assembly is being moved. ITS 3.7.13 ACTION A continues to require suspending movement of irradiated fuel. However, damaging a fuel assembly which has not been irradiated has no significant radiological effects and is not assumed in the fuel handling accident analysis. Therefore, stopping the handling of fuel assemblies which have not been irradiated when the spent fuel pool water level is less than the limit is not required.

The dropping of loads onto fuel assemblies in the spent fuel pool is not an initiator that is assumed in the fuel handling accident analysis. The movement of heavy loads is addressed by NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and Generic Letter 81-07. In the closeout of Generic Letter 81-07, the NRC concluded that restrictions on heavy loads over the spent fuel pool need not be included in the Technical Specifications. Therefore, these activities are not restricted in the Technical Specifications when the spent fuel pool water level is not within limit. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Sequoyah Unit 1 and Unit 2 Page 3 of 3 Enclosure 2, Volume 12, Rev. 0, Page 511 of 704

Enclosure 2, Volume 12, Rev. 0, Page 513 of 704 CTS Fuel Storage Pool Water Level 1 Spent 3.7.15 13 3.7 PLANT SYSTEMS Spent 3.7.15 Fuel Storage Pool Water Level 1 13 spent 3.9.11 LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated 1 13 fuel assemblies seated in the storage racks.

Whenever are spent APPLICABILITY APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool. 4 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Spent ACTION A. Fuel storage pool water A.1 --------------NOTE-------------- 1 5 level not within limit. LCO 3.0.3 is not applicable.

A.1 Suspend movement of Immediately 5 irradiated fuel assemblies in the fuel storage pool. 1 Insert 1 spent 5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY spent SR 4.9.11 SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft [ 7 days 1 13 above the top of the irradiated fuel assemblies seated in the storage racks. OR In accordance 2 with the Surveillance Frequency Control Program ]

13 Amendment XXX SEQUOYAH UNIT 1 Westinghouse STS 3.7.15-1 Rev. 4.0 3 1 Enclosure 2, Volume 12, Rev. 0, Page 513 of 704

5 INSERT 1 AND A.2 Restore the spent fuel pool 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> water level to within limit.

Enclosure 2, Volume 12, Rev. 0, Page 514 of 704 CTS Fuel Storage Pool Water Level 1 Spent 3.7.15 13 3.7 PLANT SYSTEMS Spent 3.7.15 Fuel Storage Pool Water Level 1 13 spent 3.9.11 LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated 1 13 fuel assemblies seated in the storage racks.

Whenever are spent APPLICABILITY APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool. 4 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Spent ACTION A. Fuel storage pool water A.1 --------------NOTE-------------- 1 5 level not within limit. LCO 3.0.3 is not applicable.

A.1 Suspend movement of Immediately 5 irradiated fuel assemblies in the fuel storage pool. 1 Insert 1 spent 5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY spent SR 4.9.11 SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft [ 7 days 1 13 above the top of the irradiated fuel assemblies seated in the storage racks. OR In accordance 2 with the Surveillance Frequency Control Program ]

13 Amendment XXX SEQUOYAH UNIT 2 Westinghouse STS 3.7.15-1 Rev. 4.0 3 1 Enclosure 2, Volume 12, Rev. 0, Page 514 of 704

5 INSERT 1 AND A.2 Restore the spent fuel pool 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> water level to within limit.

Enclosure 2, Volume 12, Rev. 0, Page 515 of 704 JUSTIFICATION FOR DEVIATIONS ITS 3.7.13, SPENT FUEL STORAGE POOL WATER LEVEL

1. Sequoyah Nuclear Plant (SQN) design does not include ISTS 3.7.12, "Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)" and ISTS 3.7.14, "Penetration Room Exhaust Air Cleanup System (PREACS)." Therefore, ISTS 3.7.15 has been renumbered as ITS 3.7.13. Additionally, the title "Fuel Storage Pool Water Level" has been changed to "Spent Fuel Pool Water Level."
2. ISTS SR 3.7.15.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.
3. Changes are made (additions, deletions, and/or changes) to the ISTS which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
4. ISTS 3.7.15 Applicability is "During movement of irradiated fuel assemblies in the fuel storage pool." ITS 3.7.13 Applicability is "Whenever irradiated fuel assemblies are in the spent fuel pool." The change in the Mode of Applicability from "During movement of irradiated fuel assemblies in the fuel storage pool," to "Whenever irradiated fuel assemblies are in the spent fuel pool," is acceptable because this reflects SQN's current licensing basis as reflected in CTS 3.9.11.
5. ISTS 3.7.15 has a single Required Action, A.1, "Suspend movement of irradiated fuel assemblies in the spent fuel storage pool." ITS 3.7.13 adds an additional Required Action, A.2, "Restore spent fuel pool water level to within limit." This additional Required Action is acceptable because it reflects SQN's current licensing basis as reflected in CTS 3.9.11 Actions. With the addition of ITS 3.7.13 Required Action A.2, the NOTE for Condition A becomes applicable to both ITS 3.7.13 Required Action A.1 and A.2.

Therefore, the "A.1" designator has been moved down to the Required Action and the NOTE expanded to address the entire column of Condition A, Required Actions. As discussed in the ISTS Bases for LCO 3.0.3, ISTS LCO 3.7.15 (ITS LCO 3.7.13) can be applicable in any or all MODES. If the LCO and Required Actions of ISTS LCO 3.7.15 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 12, Rev. 0, Page 515 of 704

Enclosure 2, Volume 12, Rev. 0, Page 518 of 704 Fuel Storage Pool Water Level 1 Spent B 3.7.15 13 BASES whenever are 6

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool, since the potential for a release of fission products 1 exists.

ACTIONS A.1 and A.2 6 The Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does 6 not apply.

Actions are When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel storage pool water level is lower than the required level, the 1

movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not spent fuel pool level preclude movement of a fuel assembly to a safe position.

is not within the limit If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel Insert 2 assemblies is not sufficient reason to require a reactor shutdown. 7 SURVEILLANCE SR 3.7.15.1 or restore spent fuel pool Required Actions are 1 REQUIREMENTS 13 level to within the limit This SR verifies sufficient fuel storage pool water is available in the event spent 1 of a fuel handling accident. The water level in the fuel storage pool must be checked periodically. [ The 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience. 4 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.


REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 5 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.


]

spent During refueling operations, the level in the fuel storage pool is in 1 equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.

SEQUOYAH UNIT 1 13 Revision XXX Westinghouse STS B 3.7.15-2 Rev. 4.0 2 1 Enclosure 2, Volume 12, Rev. 0, Page 518 of 704

7 INSERT 2 With the spent fuel pool water level less than 23 feet above the top of irradiated fuel assemblies seated in storage racks, the assumptions of iodine decontamination factors following a fuel handling accident cannot be met.

Required Action A.2 requires the restoration of the spent fuel pool water level to the minimum required level to preserve the assumptions of the fuel handling accident analysis (Ref. 3). The completion time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is considered sufficient to correct minor problems and restore the water level.

assumption in the Completion Time design basis fuel handling accident analysis The design basis fuel handling accident assumes the drop and damage of an irradiated fuel assembly; however, there are other potential failure mechanisms of the irradiated fuel in the spent fuel pool that could result in the release of fission product gases, which are bounded by the design basis fuel handling accident. As a result, with

Enclosure 2, Volume 12, Rev. 0, Page 521 of 704 Fuel Storage Pool Water Level 1 Spent B 3.7.15 13 BASES whenever are 6

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool, since the potential for a release of fission products 1 exists.

ACTIONS A.1 and A.2 6 The Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does 6 not apply.

Actions are When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel storage pool water level is lower than the required level, the 1

movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position.

spent fuel pool level is not within the limit If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel Insert 2 assemblies is not sufficient reason to require a reactor shutdown. 7 SURVEILLANCE SR 3.7.15.1 or restore spent fuel pool Required Actions are 1 REQUIREMENTS 13 level to within the limit This SR verifies sufficient fuel storage pool water is available in the event spent 1 of a fuel handling accident. The water level in the fuel storage pool must be checked periodically. [ The 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience. 4 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.


REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 5 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.


]

spent During refueling operations, the level in the fuel storage pool is in 1 equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.

SEQUOYAH UNIT 2 13 Revision XXX Westinghouse STS B 3.7.15-2 Rev. 4.0 2 1 Enclosure 2, Volume 12, Rev. 0, Page 521 of 704

7 INSERT 2 With the spent fuel pool water level less than 23 feet above the top of irradiated fuel assemblies seated in storage racks, the assumptions of iodine decontamination factors following a fuel handling accident cannot be met.

Required Action A.2 requires the restoration of the spent fuel pool water level to the minimum required level to preserve the assumptions of the fuel handling accident analysis (Ref. 3). The completion time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is considered sufficient to correct minor problems and restore the water level.

assumption in the Completion Time design basis fuel handling accident analysis The design basis fuel handling accident assumes the drop and damage of an irradiated fuel assembly; however, there are other potential failure mechanisms of the irradiated fuel in the spent fuel pool that could result in the release of fission product gases, which are bounded by the design basis fuel handling accident. As a result, with

Enclosure 2, Volume 12, Rev. 0, Page 523 of 704 JUSTIFICATION FOR DEVIATIONS ITS 3.7.13 BASES, FUEL STORAGE POOL WATER LEVEL

1. Sequoyah Nuclear Plant (SQN) design does not include ISTS B 3.7.12, "Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)" and ISTS B 3.7.14, "Penetration Room Exhaust Air Cleanup System (PREACS)." Therefore, ISTS B 3.7.15, "Fuel Storage Pool Water Level" has been renumbered as ITS B 3.7.13, "Fuel Storage Pool Water Level."
2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The ISTS Bases contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.
4. ISTS SR 3.7.15.1 (ITS SR 3.7.13.1) provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Therefore, the Frequency for ITS SR 3.7.13.1 is accordance with the Surveillance Frequency Control Program.
5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.
6. Changes are made to be consistent with changes made to the Specification.
7. ISTS 3.7.15 has a single Required Action, A.1, "Suspend movement of irradiated fuel assemblies in the spent fuel pool. ITS 3.7.13 adds an additional Required Action A.2, "Restore spent fuel pool water level to within limit." Therefore, the ISTS Bases have been revised to include a discussion concerning ITS 3.7.13 Required Action A.2.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 12, Rev. 0, Page 523 of 704

Enclosure 2, Volume 5, Rev. 0, Page 45 of 90 LCO Applicability B 3.0 BASES LCO 3.0.3 (continued)

MODE 4 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Whenever Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate are Spent remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.15, "Fuel Storage Pool Water Level." LCO 3.7.15 13 has an Applicability of "During movement of irradiated fuel assemblies in spent the fuel storage pool." Therefore, this LCO can be applicable in any or all and 1 13 MODES. If the LCO and the Required Actions of LCO 3.7.15 are not met "Restore while in MODE 1, 2, or 3, there is no safety benefit to be gained by s spent fuel placing the unit in a shutdown condition. The Required Action of pool water 13 LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the level to spent fuel storage pool" is the appropriate Required Action to complete in lieu of within limit" the actions of LCO 3.0.3. These exceptions are addressed in the are individual Specifications.

LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

Sequoyah Unit 1 Revision XXX Westinghouse STS B 3.0-4 Rev. 4.0 1 Enclosure 2, Volume 5, Rev. 0, Page 45 of 90

Enclosure 2, Volume 5, Rev. 0, Page 66 of 90 LCO Applicability B 3.0 BASES LCO 3.0.3 (continued)

MODE 4 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Whenever Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example are Spent of this is in LCO 3.7.15, "Fuel Storage Pool Water Level." LCO 3.7.15 13 has an Applicability of "During movement of irradiated fuel assemblies in spent the fuel storage pool." Therefore, this LCO can be applicable in any or all 1

and MODES. If the LCO and the Required Actions of LCO 3.7.15 are not met 13 "Restore while in MODE 1, 2, or 3, there is no safety benefit to be gained by s

spent fuel placing the unit in a shutdown condition. The Required Action of pool water 13 LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the level to spent fuel storage pool" is the appropriate Required Action to complete in lieu of within limit" the actions of LCO 3.0.3. These exceptions are addressed in the are individual Specifications.

LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

Sequoyah Unit 2 Revision XXX Westinghouse STS B 3.0-4 Rev. 4.0 1 Enclosure 2, Volume 5, Rev. 0, Page 66 of 90

Sequoyah ITS Conversion Database Page 1 of 2 Licensee Response/NRC Response/NRC Question Closure Id 419 NRC Question KAB066 Number Select Licensee Response Application Attachment Attachment 1 supplement for KAB066 01_12_2015 - Copy.pdf (2MB) 1 Attachment 2

Response

Statement This response supplements the response to RAI KAB066. During review, it was identified that Attachment 2 to the response for RAI KAB066 required additional revisions. Specifically, ITS 3.7.13 (ISTS 3.7.15), on pages 513 and 514 of Enclosure 2, Volume 12, will be revised to retain current licensing basis in ITS 3.7.13 ACTION A. ITS 3.7.13 Required Action A.1 will be revised to retain CTS 3.9.11 ACTION to suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas. Justification for Deviations (JFDs) 3 and 5 will be revised to justify the changes to ISTS 3.7.15 (ITS 3.7.13) ACTION A to reflect SQNs current licensing basis. As a result of the revisions described above, the following changes will be necessary:

1. The CTS 3.9.11 markups will be revised. (Pages 487 and 498 of Enclosure 2, Volume 12)
2. Discussion of Change (DOC) L02, associated with changes made to CTS 3.9.11 ACTION regarding restrictions on the movement of new fuel and the use of crane operation with loads over the spent fuel pool, will be deleted, as well as the DOC L02 indicators. (pages 487, 498 and 511 of Enclosure 2, Volume 12)
3. The ISTS 3.7.15 markups will be revised, as discussed above, and JFD 3 and 5 indicators will be revised. (Pages 513 and 514 of Enclosure 2, Volume 12)
4. The ISTS 3.7.15 Bases will be revised to align with changes made to the Specification, and JFD 6 indicators will be added to justify the changes.

(Pages 518 and 521 of Enclosure 2, Volume 12)

5. JFD 7 will be revised in the Justification for Deviations ITS 3.7.13 Bases Section. (Pages 518, 521, and 523 of Enclosure 2, Volume 12) https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=419 04/27/2015

Sequoyah ITS Conversion Database Page 2 of 2 Additionally, ITS LCO 3.0.3 Bases, on pages 45 and 66 of Enclosure 2, Volume 5, will be revised. The Bases for LCO 3.0.3 describes exceptions to LCO 3.0.3 and provides ITS LCO 3.7.13 as an example. Because of the changes described above to ITS 3.7.13, the example in the Bases for LCO 3.0.3 will be revised to align with changes made to the ITS 3.7.13 Specification.

See Attachment 1 for the draft revised changes associated with ITS 3.7.13 and the Bases for ITS LCO 3.0.3.

Response

1/14/2015 6:25 AM Date/Time Closure Statement Question Closure Date Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Scott Bowman Date Added 1/14/2015 5:24 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=419 04/27/2015

Enclosure 2, Volume 12, Rev. 0, Page 487 of 704 ITS A01 ITS 3.7.13 REFUELING OPERATIONS 3/4.9.11 SPENT FUEL PIT WATER LEVEL POOL A01 LIMITING CONDITION FOR OPERATION LCO 3.7.13 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

During movement of irradiated fuel assemblies in the spent fuel pool. L01 Applicability APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pit.

irradiated L02 pool A01 ACTION: stet immediately A02 ACTION A With the requirements of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within L02 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The provisions of Specification 3.0.3 are not applicable.

ACTION A stet stet L01 Note pool SURVEILLANCE REQUIREMENTS A01 SR 3.7.13.1 4.9.11 The water level in the spent fuel pit shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pit. L01 pool A01 in accordance with the Surveillance stet LA01 Frequency Control Program SEQUOYAH - UNIT 1 3/4 9-11 Page 1 of 22 Enclosure 2, Volume 12, Rev. 0, Page 487 of 704

Enclosure 2, Volume 12, Rev. 0, Page 498 of 704 ITS A01 ITS 3.7.13 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL-SPENT FUEL PIT POOL A01 LIMITING CONDITION FOR OPERATION LCO 3.7.13 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

During movement of irradiated fuel assemblies in the spent fuel pool. L01 Applicability APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pit. A01 irradiated L02 pool ACTION: stet immediately A02 ACTION A With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit L02 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The provisions of Specification 3.0.3 are not applicable.

ACTION A stet stet L01 Note pool SURVEILLANCE REQUIREMENTS A01 SR 3.7.13.1 4.9.11 The water level in the spent fuel pit shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pit. L01 pool A01 stet in accordance with the Surveillance LA01 Frequency Control Program SEQUOYAH - UNIT 2 3/4 9-13 Page 12 of 22 Enclosure 2, Volume 12, Rev. 0, Page 498 of 704

Enclosure 2, Volume 12, Rev. 0, Page 510 of 704 DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequency is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.9.11 Applicability states "Whenever irradiated fuel assemblies are in the spent fuel pit." CTS SR 4.9.11 requires the water level in the spent fuel pit to be verified every 7 days when Not Used irradiated fuel assemblies are in the spent fuel pit. ITS 3.7.13 is applicable "During movement of irradiated fuel assemblies in the spent fuel pool." ITS SR 3.0.1 requires ITS SR 3.7.13.1 to be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. In addition, since the Applicability is now limited to when irradiated fuel is being moved, the CTS ACTION to "restore water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after movement of fuel has been suspended" has also been deleted. This changes the CTS by restricting the Applicability of the spent fuel pool water level Specification and performance of the Surveillance to when there is a potential for a fuel handling accident, i.e., during the movement of irradiated fuel assemblies in the spent fuel pool.

The purpose of CTS 3.9.11 is to ensure that the minimum spent fuel pit water level assumption in the fuel handling accident analysis is met. This change is acceptable because the requirements continue to ensure that the conditions assumed in the safety analyses and licensing basis are maintained. The SQN fuel handling accident analysis (outside containment) assumes that a single fuel assembly is damaged. A key assumption in the analysis is that there is 23 feet of water over the damaged assembly, as this depth is directly related to the cleanup of the fission products before release from the spent fuel pool. A fuel handling accident is only assumed to occur when an irradiated fuel assembly is being moved. Therefore, ITS 3.7.13 imposes controls on minimum spent fuel pool water level only during the movement of irradiated fuel assemblies in the spent fuel pool. ITS 4.3.2 specifies the requirement that the spent fuel pool be designed and maintained to prevent inadvertent draining of the pool below elevation 722. This change is designated as less restrictive because the ITS LCO requirements are applicable in fewer operating conditions than in the CTS.

Sequoyah Unit 1 and Unit 2 Page 2 of 3 Enclosure 2, Volume 12, Rev. 0, Page 510 of 704

Enclosure 2, Volume 12, Rev. 0, Page 511 of 704 DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL L02 (Category 4 - Relaxation of Required Action) CTS 3.9.11 ACTION states that when the spent fuel pit water level is not met, suspend all movement of fuel Not Used assemblies and crane operations with loads in the fuel storage areas. ITS 3.7.13 Required Action A.1 states that when spent fuel pool water level is not within limits, immediately suspend movement of irradiated fuel assemblies in the spent fuel pool. This changes the CTS by deleting the requirements to suspend movement of new fuel and to suspend crane operation over the spent fuel storage areas.

The purpose of the CTS 3.9.11 ACTION is to preclude a fuel handling accident from occurring when the initial conditions for that accident are not met. A fuel handling accident is only assumed to occur when an irradiated fuel assembly is being moved. ITS 3.7.13 ACTION A continues to require suspending movement of irradiated fuel. However, damaging a fuel assembly which has not been irradiated has no significant radiological effects and is not assumed in the fuel handling accident analysis. Therefore, stopping the handling of fuel assemblies which have not been irradiated when the spent fuel pool water level is less than the limit is not required.

The dropping of loads onto fuel assemblies in the spent fuel pool is not an initiator that is assumed in the fuel handling accident analysis. The movement of heavy loads is addressed by NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and Generic Letter 81-07. In the closeout of Generic Letter 81-07, the NRC concluded that restrictions on heavy loads over the spent fuel pool need not be included in the Technical Specifications. Therefore, these activities are not restricted in the Technical Specifications when the spent fuel pool water level is not within limit. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Sequoyah Unit 1 and Unit 2 Page 3 of 3 Enclosure 2, Volume 12, Rev. 0, Page 511 of 704

Enclosure 2, Volume 12, Rev. 0, Page 513 of 704 CTS Fuel Storage Pool Water Level 1 Spent 3.7.15 13 3.7 PLANT SYSTEMS Spent 3.7.15 Fuel Storage Pool Water Level 1 13 spent 3.9.11 LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated 1 13 fuel assemblies seated in the storage racks.

Whenever are spent APPLICABILITY APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool. 4 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Spent ACTION A. Fuel storage pool water A.1 --------------NOTE-------------- 1 5 level not within limit. LCO 3.0.3 is not applicable.

all A.1 Suspend movement of Immediately 5 irradiated fuel assemblies in and crane operations with loads 3 the fuel storage pool. 1 Insert 1 spent 5 stet areas SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY spent SR 4.9.11 SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft [ 7 days 1 13 above the top of the irradiated fuel assemblies seated in the storage racks. OR In accordance 2 with the Surveillance Frequency Control Program ]

13 Amendment XXX SEQUOYAH UNIT 1 Westinghouse STS 3.7.15-1 Rev. 4.0 3 1 Enclosure 2, Volume 12, Rev. 0, Page 513 of 704

5 INSERT 1 AND A.2 Restore spent fuel pool 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> water level to within limit.

Enclosure 2, Volume 12, Rev. 0, Page 514 of 704 CTS Fuel Storage Pool Water Level 1 Spent 3.7.15 13 3.7 PLANT SYSTEMS Spent 3.7.15 Fuel Storage Pool Water Level 1 13 spent 3.9.11 LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated 1 13 fuel assemblies seated in the storage racks.

Whenever are spent APPLICABILITY APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool. 4 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Spent ACTION A. Fuel storage pool water A.1 --------------NOTE-------------- 1 5 level not within limit. LCO 3.0.3 is not applicable.

all A.1 Suspend movement of Immediately 5 irradiated fuel assemblies in and crane operations with loads 3 the fuel storage pool. 1 Insert 1 spent 5 stet areas SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY spent SR 4.9.11 SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft [ 7 days 1 13 above the top of the irradiated fuel assemblies seated in the storage racks. OR In accordance 2 with the Surveillance Frequency Control Program ]

13 Amendment XXX SEQUOYAH UNIT 2 Westinghouse STS 3.7.15-1 Rev. 4.0 3 1 Enclosure 2, Volume 12, Rev. 0, Page 514 of 704

5 INSERT 1 AND A.2 Restore spent fuel pool 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> water level to within limit.

Enclosure 2, Volume 12, Rev. 0, Page 515 of 704 JUSTIFICATION FOR DEVIATIONS ITS 3.7.13, SPENT FUEL STORAGE POOL WATER LEVEL

1. Sequoyah Nuclear Plant (SQN) design does not include ISTS 3.7.12, "Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)" and ISTS 3.7.14, "Penetration Room Exhaust Air Cleanup System (PREACS)." Therefore, ISTS 3.7.15 has been renumbered as ITS 3.7.13. Additionally, the title "Fuel Storage Pool Water Level" has been changed to "Spent Fuel Pool Water Level."
2. ISTS SR 3.7.15.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.
3. Changes are made (additions, deletions, and/or changes) to the ISTS which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
4. ISTS 3.7.15 Applicability is "During movement of irradiated fuel assemblies in the fuel storage pool." ITS 3.7.13 Applicability is "Whenever irradiated fuel assemblies are in the spent fuel pool." The change in the Mode of Applicability from "During movement of irradiated fuel assemblies in the fuel storage pool," to "Whenever irradiated fuel assemblies are in the spent fuel pool," is acceptable because this reflects SQN's current licensing basis as reflected in CTS 3.9.11.
5. ISTS 3.7.15 has a single Required Action, A.1. ITS 3.7.13 adds an additional Required Action, A.2, "Restore spent fuel pool water level to within limit." This additional Required Action is acceptable because it reflects SQN's current licensing basis as reflected in CTS 3.9.11 Actions. With the addition of ITS 3.7.13 Required Action A.2, the NOTE for Condition A becomes applicable to both ITS 3.7.13 Required Action A.1 and A.2. Therefore, the "A.1" designator has been moved down to the Required Action and the NOTE expanded to address the entire column of Condition A, Required Actions. As discussed in the ISTS Bases for LCO 3.0.3, ISTS LCO 3.7.15 (ITS LCO 3.7.13) can be applicable in any or all MODES. If the LCO and Required Actions of ISTS LCO 3.7.15 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition.

ISTS 3.7.15 Required Action A.1 is, "Suspend movement of irradiated fuel assemblies in the spent fuel storage pool." ITS 3.7.13 Required Action A.1 is, "Suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas." The change in Required Action A.1 from, "Suspend movement of irradiated fuel assemblies in the spent fuel storage pool," to "Suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas," is acceptable because this reflects SQN's current licensing basis as reflected in CTS 3.9.11.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 12, Rev. 0, Page 515 of 704

Enclosure 2, Volume 12, Rev. 0, Page 518 of 704 Fuel Storage Pool Water Level 1 Spent B 3.7.15 13 BASES whenever are 6

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool, since the potential for a release of fission products 1 exists.

ACTIONS A.1 and A.2 6 The Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does 6 not apply.

Actions are When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel storage pool water level is lower than the required level, the stet 1

and crane operations with loads movement of irradiated fuel assemblies in the fuel storage pool is areas 6 immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not or crane load 6 the spent fuel pool water preclude movement of a fuel assembly to a safe position.

level is not within the limit If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 Required Actions are would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor 6 operations. Therefore, inability to suspend movement of irradiated fuel Insert 2 assemblies is not sufficient reason to require a reactor shutdown. 7 SURVEILLANCE SR 3.7.15.1 1 REQUIREMENTS 13 This SR verifies sufficient fuel storage pool water is available in the event

, suspend crane spent 1 of a fuel handling accident. The water level in the fuel storage pool must operations with loads, be checked periodically. [ The 7 day Frequency is appropriate because or restore spent fuel the volume in the pool is normally stable. Water level changes are pool water level to controlled by plant procedures and are acceptable based on operating within the limit experience. 4 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.


REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 5 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.


]

spent During refueling operations, the level in the fuel storage pool is in 1 equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.

SEQUOYAH UNIT 1 13 Revision XXX Westinghouse STS B 3.7.15-2 Rev. 4.0 2 1 Enclosure 2, Volume 12, Rev. 0, Page 518 of 704

7 INSERT 2 The design basis fuel handling accident assumes the drop and damage of an irradiated fuel assembly; however, there are other potential failure mechanisms of the irradiated fuel in the spent fuel pool that could result in the release of fission product gases, which are bounded by the design basis fuel handling accident. As a result, with the spent fuel pool water level less than 23 feet above the top of irradiated fuel assemblies seated in storage racks, the iodine decontamination factor assumption in the design basis fuel handling accident analysis cannot be met.

Required Action A.2 requires the restoration of the spent fuel pool water level to the minimum required level to preserve the assumptions of the fuel handling accident analysis (Ref. 3).

The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is considered sufficient to correct minor problems and restore the water level.

Enclosure 2, Volume 12, Rev. 0, Page 521 of 704 Fuel Storage Pool Water Level 1 Spent B 3.7.15 13 BASES whenever are 6

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool, since the potential for a release of fission products 1 exists.

ACTIONS A.1 and A.2 6 The Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does 6 not apply.

Actions are When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel storage pool water level is lower than the required level, the stet 1

and crane operations with loads movement of irradiated fuel assemblies in the fuel storage pool is areas 6 immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not or crane load 6 the spent fuel pool water preclude movement of a fuel assembly to a safe position.

level is not within the limit If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 Required Actions are would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor 6 operations. Therefore, inability to suspend movement of irradiated fuel Insert 2 assemblies is not sufficient reason to require a reactor shutdown. 7 SURVEILLANCE SR 3.7.15.1 1 REQUIREMENTS 13 This SR verifies sufficient fuel storage pool water is available in the event

, suspend crane spent 1 of a fuel handling accident. The water level in the fuel storage pool must operations with loads, be checked periodically. [ The 7 day Frequency is appropriate because or restore spent fuel the volume in the pool is normally stable. Water level changes are pool water level to controlled by plant procedures and are acceptable based on operating within the limit experience. 4 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.


REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 5 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.


]

spent During refueling operations, the level in the fuel storage pool is in 1 equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.

SEQUOYAH UNIT 2 13 Revision XXX Westinghouse STS B 3.7.15-2 Rev. 4.0 2 1 Enclosure 2, Volume 12, Rev. 0, Page 521 of 704

7 INSERT 2 The design basis fuel handling accident assumes the drop and damage of an irradiated fuel assembly; however, there are other potential failure mechanisms of the irradiated fuel in the spent fuel pool that could result in the release of fission product gases, which are bounded by the design basis fuel handling accident. As a result, with the spent fuel pool water level less than 23 feet above the top of irradiated fuel assemblies seated in storage racks, the iodine decontamination factor assumption in the design basis fuel handling accident analysis cannot be met.

Required Action A.2 requires the restoration of the spent fuel pool water level to the minimum required level to preserve the assumptions of the fuel handling accident analysis (Ref. 3).

The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is considered sufficient to correct minor problems and restore the water level.

Enclosure 2, Volume 12, Rev. 0, Page 523 of 704 JUSTIFICATION FOR DEVIATIONS ITS 3.7.13 BASES, FUEL STORAGE POOL WATER LEVEL

1. Sequoyah Nuclear Plant (SQN) design does not include ISTS B 3.7.12, "Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)" and ISTS B 3.7.14, "Penetration Room Exhaust Air Cleanup System (PREACS)." Therefore, ISTS B 3.7.15, "Fuel Storage Pool Water Level" has been renumbered as ITS B 3.7.13, "Fuel Storage Pool Water Level."
2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The ISTS Bases contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.
4. ISTS SR 3.7.15.1 (ITS SR 3.7.13.1) provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Therefore, the Frequency for ITS SR 3.7.13.1 is accordance with the Surveillance Frequency Control Program.
5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.
6. Changes are made to be consistent with changes made to the Specification.
7. ISTS 3.7.15 has a single Required Action, A.1. ITS 3.7.13 adds an additional Required Action A.2, "Restore spent fuel pool water level to within limit." Therefore, the ISTS Bases have been revised to include a discussion concerning ITS 3.7.13 Required Action A.2.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 12, Rev. 0, Page 523 of 704

Enclosure 2, Volume 5, Rev. 0, Page 45 of 90 LCO Applicability B 3.0 BASES LCO 3.0.3 (continued)

MODE 4 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Whenever Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate are Spent remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.15, "Fuel Storage Pool Water Level." LCO 3.7.15 13 has an Applicability of "During movement of irradiated fuel assemblies in spent the fuel storage pool." Therefore, this LCO can be applicable in any or all 1

13 MODES. If the LCO and the Required Actions of LCO 3.7.15 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by s all placing the unit in a shutdown condition. The Required Action of 13 LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the areas 5 spent fuel storage pool" is the appropriate Required Action to complete in lieu of s stet the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications. and crane operations with loads LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows and "Restore spent placing the unit in a MODE or other specified condition stated in that fuel pool water level Applicability (e.g., the Applicability desired to be entered) when unit to within limit" are conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

Sequoyah Unit 1 Revision XXX Westinghouse STS B 3.0-4 Rev. 4.0 1 Enclosure 2, Volume 5, Rev. 0, Page 45 of 90

Enclosure 2, Volume 5, Rev. 0, Page 66 of 90 LCO Applicability B 3.0 BASES LCO 3.0.3 (continued)

MODE 4 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Whenever Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example are Spent of this is in LCO 3.7.15, "Fuel Storage Pool Water Level." LCO 3.7.15 13 has an Applicability of "During movement of irradiated fuel assemblies in spent the fuel storage pool." Therefore, this LCO can be applicable in any or all 1

13 MODES. If the LCO and the Required Actions of LCO 3.7.15 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by s

all placing the unit in a shutdown condition. The Required Action of 13 LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the areas 5 spent fuel storage pool" is the appropriate Required Action to complete in lieu of s stet the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications. and crane operations with loads LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows and "Restore spent placing the unit in a MODE or other specified condition stated in that fuel pool water level Applicability (e.g., the Applicability desired to be entered) when unit to within limit" are conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

Sequoyah Unit 2 Revision XXX Westinghouse STS B 3.0-4 Rev. 4.0 1 Enclosure 2, Volume 5, Rev. 0, Page 66 of 90

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 437 NRC Question KAB066 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This RAI is being closed at this time. However, the follow-up RAIs for KAB066 have been posted under MHC003 and MHC004.

Question Closure 4/23/2015 Date Notification Mark Blumberg Scott Bowman Margaret Chernoff Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 4/23/2015 11:26 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=437 04/27/2015

Sequoyah ITS Conversion Database Page 1 of 2 ITS NRC Questions Id 191 NRC Question KAB067 Number Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s)

NRC Reviewer Roger Pederson Supervisor Technical Mark Blumberg Branch POC Conf Call N

Requested NRC Question RAI ARCB/SCVB2-2 (in response to KAB-044)

Calculation LTR-CRA-02-219, Revision 1 assumes a mixing volume for the fuel handling accident in containment that is 10 times higher (325,500 versus 32,550 cubic feet) than was credited in license amendment 288/278 (Unit 1/Unit 2) (ADAMS Accession No. ML033070057), but does not justify the proposed change.

Appendix B of RG 1.183, Regulatory Position 5.5 states:

Credit for dilution or mixing of the activity released from the reactor cavity by natural or forced convection inside the containment may be considered on a case-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

The calculation does not address the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=191 04/27/2015

Sequoyah ITS Conversion Database Page 2 of 2 surface of the reactor cavity and the exhaust plenums. Please provide a justification of why the revised mixing volume is appropriate.

Attach File 1

Attach File 2

Issue Date 9/30/2014 Added By Khadijah Hemphill Date Modified Modified By Date Added 9/30/2014 4:02 PM Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=191 04/27/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 405 NRC Question KAB067 Number Select Licensee Response Application Attachment 1

Attachment 2

Response

Statement In response to RAI KAB-067, Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev 4.0 (SQN-TS-11-10) - Supplement 1 was submitted to the NRC for review on December 16, 2014.Attachment 1 of the supplement contains LTR-CRA-02-219, Revision 2, Radiological Consequences of Fuel Handling Accidents for the Sequoyah Nuclear Plant Units 1 and 2, which revised the containment mixing volume assumption such that for a Fuel Handling Accident (FHA) inside the containment, the containment mixing volume assumption has been deleted and the activity released from the damaged fuel not retained in the water pool is assumed to be released linearly from the pool to the environment within two hours.No credit is taken for mixing in the containment volume.

Response

12/18/2014 4:00 PM Date/Time Closure Statement Question Closure Date Notification Mark Blumberg Scott Bowman Kristy Bucholtz Margaret Chernoff Michelle Conner Robert Elliott Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Michelle Conner Date Added 12/18/2014 3:04 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=405 04/27/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 438 NRC Question KAB067 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 4/23/2015 Date Notification Mark Blumberg Scott Bowman Margaret Chernoff Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 4/23/2015 11:27 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=438 04/27/2015

Sequoyah ITS Conversion Database Page 1 of 2 ITS NRC Questions Id 192 NRC Question KAB068 Number Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s)

NRC Reviewer Roger Pederson Supervisor Technical Mark Blumberg Branch POC Conf Call N

Requested NRC RAI ARCB2-3 (in response to KAB-044)

Question Enclosure 2, Volume 14, Rev. 0, page 135 of 236 proposed making the following modification:

Fuel handling accidents, analyzed in Reference 3, include dropping a single irradiated fuel assembly.

The justification provided is given below:

Changes are made (additions, deletions, and/or changes) to the ISTS

[or ITS] that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

The staff is concerned that the above justification conflicts with the staffs assessment of Sequoyahs licensing bases. The dropping of heavy objects appears to be in the licensing bases and the need to consider heavy objects is addressed in the SRP and RG 1.183, as discussed above (RAI ARCB2-1 or KAB066). Please modify the justification to address the NRC staffs concern, or replace the text proposed to be removed.

Attach File 1 Attach File 2 Issue Date 9/30/2014 https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=192 3/16/2015

Sequoyah ITS Conversion Database Page 2 of 2 Added By Khadijah Hemphill Date Modified Modified By Date Added 9/30/2014 4:03 PM Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=192 3/16/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 380 NRC Question KAB068 Number Select Licensee Response Application Attachment Attachment 1 for RAI KAB068.pdf (31KB) 1 Attachment 2

Response

Statement In response to KAB068, the ITS 3.9.4 Bases Applicable Safety Analyses Section, on pages 124 and 135 of Enclosure 2, Volume 14, will be revised to retain the words and handling tool or a heavy object onto other irradiated fuel assemblies. The Justification for Deviations 2 indicator will be deleted because the previously deleted text will be retained.

See Attachment 1 for the draft revised ITS 3.9.4 Bases.

Response

11/24/2014 4:55 AM Date/Time Closure Statement Question Closure Date Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Scott Bowman Date Added 11/24/2014 3:55 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=380 3/16/2015

Enclosure 2, Volume 14, Rev. 0, Page 124 of 236 Containment Penetrations B 3.9.4 BASES BACKGROUND (continued)

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits.

The Containment Purge and Exhaust System includes two subsystems.

The normal subsystem includes a 42 inch purge penetration and a 42 inch exhaust penetration. The second subsystem, a minipurge system, includes an 8 inch purge penetration and an 8 inch exhaust penetration. During MODES 1, 2, 3, and 4, the two valves in each of the normal purge and exhaust penetrations are secured in the closed position. The two valves in each of the two minipurge penetrations can be opened intermittently, but are closed automatically by the Engineered Safety Features Actuation System (ESFAS). Neither of the subsystems is subject to a Specification in MODE 5.

In MODE 6, large air exchangers are necessary to conduct refueling operations. The normal 42 inch purge system is used for this purpose, 2 and all four valves are closed by the ESFAS in accordance with LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation."

[ The minipurge system remains operational in MODE 6, and all four valves are also closed by the ESFAS.

[or]

The minipurge system is not used in MODE 6. All four 8 inch valves are secured in the closed position. ]

INSERT 1 The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at (either open or least one side. Isolation may be achieved by an OPERABLE automatic closed) isolation valve, or by a manual isolation valve, blind flange, or equivalent. 2 Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during [recently] irradiated 1 fuel movements (Ref. 1).

APPLICABLE During movement of irradiated fuel assemblies within containment, the SAFETY most severe radiological consequences result from a fuel handling ANALYSES accident [involving handling recently irradiated fuel]. The fuel handling 1 accident is a postulated event that involves damage to irradiated fuel stet (Ref. 2). Fuel handling accidents, analyzed in Reference 3, include dropping a single irradiated fuel assembly and handling tool or a heavy 2 object onto other irradiated fuel assemblies. The requirements of LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum SEQUOYAH UNIT 1 Revision XXX Westinghouse STS B 3.9.4-2 Rev. 4.0 2 Enclosure 2, Volume 14, Rev. 0, Page 124 of 236

Enclosure 2, Volume 14, Rev. 0, Page 135 of 236 Containment Penetrations B 3.9.4 BASES BACKGROUND (continued)

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits.

The Containment Purge and Exhaust System includes two subsystems.

The normal subsystem includes a 42 inch purge penetration and a 42 inch exhaust penetration. The second subsystem, a minipurge system, includes an 8 inch purge penetration and an 8 inch exhaust penetration. During MODES 1, 2, 3, and 4, the two valves in each of the normal purge and exhaust penetrations are secured in the closed position. The two valves in each of the two minipurge penetrations can be opened intermittently, but are closed automatically by the Engineered Safety Features Actuation System (ESFAS). Neither of the subsystems is subject to a Specification in MODE 5.

In MODE 6, large air exchangers are necessary to conduct refueling operations. The normal 42 inch purge system is used for this purpose, 2 and all four valves are closed by the ESFAS in accordance with LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation."

[ The minipurge system remains operational in MODE 6, and all four valves are also closed by the ESFAS.

[or]

The minipurge system is not used in MODE 6. All four 8 inch valves are secured in the closed position. ]

INSERT 1 The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at (either open or least one side. Isolation may be achieved by an OPERABLE automatic closed) isolation valve, or by a manual isolation valve, blind flange, or equivalent. 2 Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during [recently] irradiated 1 fuel movements (Ref. 1).

APPLICABLE During movement of irradiated fuel assemblies within containment, the SAFETY most severe radiological consequences result from a fuel handling ANALYSES accident [involving handling recently irradiated fuel]. The fuel handling 1 accident is a postulated event that involves damage to irradiated fuel stet (Ref. 2). Fuel handling accidents, analyzed in Reference 3, include dropping a single irradiated fuel assembly and handling tool or a heavy 2 object onto other irradiated fuel assemblies. The requirements of LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum SEQUOYAH UNIT 2 Revision XXX Westinghouse STS B 3.9.4-2 Rev. 4.0 2 Enclosure 2, Volume 14, Rev. 0, Page 135 of 236

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 406 NRC Question KAB068 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 12/19/2014 Date Notification Mark Blumberg Scott Bowman Margaret Chernoff Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 12/19/2014 7:21 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=406 3/16/2015

Sequoyah ITS Conversion Database Page 1 of 2 ITS NRC Questions Id 194 NRC Question KAB070 Number Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s)

NRC Reviewer Roger Pederson Supervisor Technical Mark Blumberg Branch POC Conf Call N

Requested NRC RAI ARCB2-5/SCVB2-5 (in response to KAB-044)

Question The proposed changes discussed in the previous question (ARCB2-4) impact the pressure in adjacent areas to the control room envelope. With these systems inoperable there are no technical specification controls to assure that the systems will function. Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, dated May 2003 (ADAMS Accession No. ML031490664) states:

Any test to determine CRE [control room envelope] integrity should be performed while the CRE, its associated ventilation systems, and the ventilation systems located in, traversing, or serving areas adjacent to the CRE are functioning in a manner that reflects CRE inleakage when these ventilation systems are operating in response to a particular challenge.

and, In addition to the above, CRE testing should be performed when changes are made to the structures, systems, components, and procedures that could impact CRE integrity. The structures, systems, and components could be within the envelope itself or could serve or be within areas adjacent to the envelope. Additional testing may be warranted if the conditions associated with a particular challenge result in a change in operating mode, alignment, or response that could result in a new limiting condition. Testing should be https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=194 05/07/2015

Sequoyah ITS Conversion Database Page 2 of 2 commensurate with the type and degree of modification or repair that has been made. For some changes, a new baseline test may be required.

Please provide a justification for the control room unfiltered inleakage assumed for the FHA in the unit with the inoperable ABGTS and/or ABGTS actuation equipment. The changes requested may have an impact on the unfiltered inleakage of a common shared control room, therefore, impacting both units.

Please provide a justification for the control room unfiltered inleakage values assumed for any applicable design basis accident in one unit with an inoperable ABGTS and/or ABGTS actuation instrumentation in opposite unit.

Attach File 1

Attach File 2

Issue Date 9/30/2014 Added By Khadijah Hemphill Date Modified Modified By Date Added 9/30/2014 4:06 PM Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=194 05/07/2015

Sequoyah ITS Conversion Database Page 1 of 2 Licensee Response/NRC Response/NRC Question Closure Id 416 NRC Question KAB070 Number Select Licensee Response Application Attachment KAB070 Attachment 1.pdf (188KB) 1 Attachment 2

Response

Statement An inoperable Auxiliary Building Gas Treatment System (ABGTS) or inoperable ABGTS actuation instrumentation does not affect operation of the Control Room Emergency Ventilation System (CREVS) or alter the main control room (MCR) unfiltered inleakage assumptions in the Sequoyah Nuclear Plant (SQN) design basis accident (DBA) analyses. The only DBA that is postulated to occur when ABGTS and the associated actuation instrumentation are not required by Technical Specifications is a fuel handling accident (FHA) when both units are shutdown with average reactor coolant temperature at or below 200°F. Current Technical Specifications (CTS) and the proposed Improved Technical Specifications (ITS) require the common ABGTS and the associated actuation instrumentation to be Operable when either unit is in Mode 1, 2, 3, or 4.

As indicated in SQN Updated Final Safety Analysis Report (UFSAR), Table 15.5.6-1, Parameters Used in Fuel Handling Accident Analyses, 51 cfm unfiltered MCR inleakage is assumed following the swap to the CREVS. This unfiltered air flow value is consistent with the value listed in Table 2, Fuel Handling Accident Assumptions, of the revised Calculation LTR-CRA-02-219, Revision 2, Radiological Consequences of Fuel Handling Accidents for the Sequoyah Nuclear Plant Units 1 and 2.

CTS 6.17.c of the Control Room Envelope (CRE) Habitability Program requires, in part, determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the frequencies specified in Sections C.1 and C.2 of Regulatory Guide (RG) 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, May 2003. The most recently recorded CRE inleakage during performance of the SQN CREVS tracer gas test was 22 +/-9 cfm for https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=416 05/07/2015

Sequoyah ITS Conversion Database Page 2 of 2 Train A and 15 +/-7 for Train B; i.e., a 39% margin and a 50% margin for Train A and Train B, respectively, when including the cfm error) between the measured CRE inleakage and the unfiltered MCR inleakage assumption in the design basis FHA. ITS 5.5.16.c (CTS 6.17.c) will continue to require determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and frequencies specified in RG 1.197.

As discussed in Section 6.4.1.2 of the SQN UFSAR; during operation in the emergency mode, the CREVS maintains a positive pressure of at least 1/8 inch water gauge in the CRE relative to the outside atmosphere and a slightly positive pressure relative to adjoining spaces. The CRE boundary consists of the walls, ceiling, and floor in the areas adjacent to the MCR. Adjacent areas are those areas between the ABSCE and the MCR envelope including the Unit 1 and 2 cable spreading rooms, the stairwells, and the 6900 V shutdown board rooms. These adjacent areas (see Attachment 1) are not in the auxiliary building secondary containment enclosure and therefore, are not affected by ABGTS operation. As such, an inoperable ABGTS or inoperable ABGTS actuation instrumentation will not affect operation of the CREVS or alter the MCR unfiltered inleakage assumptions in any SQN DBA analysis, including the FHA.

Response

1/4/2015 8:20 PM Date/Time Closure Statement Question Closure Date Notification Mark Blumberg Scott Bowman Kristy Bucholtz Margaret Chernoff Michelle Conner Robert Elliott Matthew Hamm Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Michelle Conner Date Added 1/4/2015 7:20 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=416 05/07/2015

Control Room Envelope ABSCE Boundary

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 442 NRC Question KAB070 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 4/29/2015 Date Notification Mark Blumberg Margaret Chernoff Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 4/29/2015 10:06 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=442 05/07/2015

Sequoyah ITS Conversion Database Page 1 of 2 ITS NRC Questions Id 195 NRC Question KAB071 Number Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s)

NRC Reviewer Roger Pedersen Supervisor Technical Mark Blumberg Branch POC Conf Call N

Requested NRC RAI ARCB2-7 (in response to KAB-044)

Question The proposed technical specification changes allow the containment building airlocks and penetrations to be open and the containment ventilation isolation instrumentation to be non-operational after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of fuel decay time (ITS 3.9.4 and ITS 3.3.6, respectively). Previously, only the containment equipment door was allowed to be open during the movement of fuel after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of fuel decay time (TS 3.9.4).

Presumably, Calculation LTR-CRA-02-219, Revision 0, assumed that all releases from a FHA within containment would be through the containment equipment door after 30 seconds. Calculation LTR-CRA-02-219, Revision 1, uses the same X/Q value (1.80E-3 sec/m3) to model containment releases from the now open containment building airlocks and penetrations after 300 seconds.

Please justify that the 1.80E-3 sec/m3 X/Q value used in Calculation LTR-CRA-02-219, Revision 1, to model FHA releases inside containment beyond 300 seconds bounds all potential containment release pathways (such as the containment equipment door, airlocks, and penetrations) or provide a revised limiting X/Q value for all containment release pathways. If a revised limiting X/Q value is provided, please update the doses analysis in Calculation LTR-CRA-02-219 to utilize the revised X/Q value.

Attach File 1 Attach File 2 https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=195 05/07/2015

Sequoyah ITS Conversion Database Page 2 of 2 Issue Date 9/30/2014 Added By Khadijah Hemphill Date Modified Modified By Date Added 9/30/2014 4:07 PM Notification Mark Blumberg Scott Bowman Kristy Bucholtz Michelle Conner Ravinder Grover Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=195 05/07/2015

Sequoyah ITS Conversion Database Page 1 of 2 Licensee Response/NRC Response/NRC Question Closure Id 417 NRC Question KAB071 Number Select Licensee Response Application Attachment 1

Attachment 2

Response TVA provided a revised limiting atmospheric dispersion factor, /Q value, Statement for all containment release pathways in a letter; "Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev 4.0 (SQN-TS-11-10) - Supplement 1," dated December 16, 2014 (ADAMS Accession No. ML14350B364). Attachment 1 of the supplement provided the revised Westinghouse report, LTR-CRA-02-219 Revision 2:

Radiological Consequences of Fuel Handling Accidents for the Sequoyah Nuclear Plant Units 1 and 2.

The analysis demonstrates that a design basis fuel handling accident (FHA) whether inside containment or in the Auxiliary Building (AB) have the same offsite dose consequence since the accident occurring in different locations does not change the amount of activity released over the two-hour period. The AB vent stack release point provides the limiting release point for the Main Control Room (MCR) dose because of less atmospheric dispersion and more severe MCR dose consequences than a containment purge release. As stated in Section 2.1 of LTR-CRA-02-219, based on a review of containment penetrations as potential point source release locations, the AB vent stack is the limiting location for the calculation of MCR doses based on its proximity to the MCR air intake locations.

Therefore, any release from containment would be bounded by the release from the AB vent stack. The revised /Q value of 2.56E-3 sec/m3 is the limiting atmospheric dispersion value for calculating the radiological consequences to personnel in the MCR following a design basis FHA at Sequoyah Nuclear Plant, Units 1 and 2.

Response

1/4/2015 8:30 PM Date/Time Closure Statement Question Closure Date Notification Mark Blumberg Scott Bowman Kristy Bucholtz Margaret Chernoff Michelle Conner Robert Elliott https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=417 05/07/2015

Sequoyah ITS Conversion Database Page 2 of 2 Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Added By Michelle Conner Date Added 1/4/2015 7:31 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=417 05/07/2015

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 447 NRC Question KAB071 Number Select Application NRC Question Closure Attachment 1 Attachment 2

Response

Statement

Response

Date/Time Closure Statement This question is closed and no further information is required at this time to draft the Safety Evaluation.

Question Closure 5/6/2015 Date Notification Mark Blumberg Margaret Chernoff Michelle Conner Khadijah Hemphill Andrew Hon Lynn Mynatt Ray Schiele Roger Scott Added By Khadijah Hemphill Date Added 5/6/2015 4:32 PM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=447 05/07/2015

Sequoyah ITS Conversion Database Page 1 of 1 ITS NRC Questions Id 6 NRC Question KNH-001 Number Category Technical ITS Section TSTF-425 - PRA ITS Number DOC Number JFD Number JFD Bases Number Page Number (s)

NRC Reviewer Rob Elliott Supervisor Technical Jonathan Evans Branch POC Conf Call N

Requested NRC Question In Enclosure 10, the proposed changes indicate that TSTF-425, Revision 3 is included in the change. However, the Sequoyah Nuclear Plant Submittal does not appear to include the documentation regarding the probabilistic risk assessment technical adequacy consistent with the guidance in NEI 04-10. Please provide documentation.

Attach File 1 Attach File 2 Issue Date 2/4/2014 Added By Khadijah Hemphill Date Modified Modified By Date Added 2/4/2014 4:04 PM Notification Michelle Conner Khadijah Hemphill Ray Schiele Gerald Waig https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=6 8/21/2014

Sequoyah ITS Conversion Database Page 1 of 1 Licensee Response/NRC Response/NRC Question Closure Id 5 NRC Question KNH-001 Number Select Licensee Response Application Attachment Cover Letter from Westinghouse LTR-RAM-II-11-010.pdf (94KB) 1 Attachment Appendix A NUC-SQN-MEB-MDN-000-2010-0200 REV 1.pdf (485KB) 2

Response

Statement Enclosure 10 for TSTF-425 refers to Reference 7, Westinghouse LTR-RAM-II-11-010, "RG 1.200 PRA Peer Review Against the ASME/ANS PRA Standard Requirements for the Sequoyah Nuclear Plant Probabilistic Risk Assessment," dated March 18, 2011, and Reference 8, NUC-SQN-MEB-MDN-000-000-2010-0200 Revision 1, "SQN Probabilistic Risk Assessment - Summary Document."

Attachment 1 is the cover letter for Reference 7. Attachment 2 is Appendix A, Resolution of F&Os, for Reference 8.

Additionally, Sequoyah Nuclear Plant, Units 1 and 2 - License Renewal Application, Attachments C, D and E, Part 8 of 8 (ML13024A010) contains Attachment E.1, Evaluation of SQN PRA Model. This Attachment begins on page 95 of the .pdf file.

Response

2/5/2014 12:30 PM Date/Time Closure Statement Question Closure Date Notification Scott Bowman Michelle Conner Robert Elliott Khadijah Hemphill Lynn Mynatt Lisa Regner Ray Schiele Roger Scott Gerald Waig Added By Scott Bowman Date Added 2/5/2014 11:24 AM Date Modified Modified By https://members.excelservices.com/rai/index.php?requestType=areaItemPrint&itemId=5 8/21/2014

Date: March 18, 2011 To: Tom Zachariah cc: Paul Hijeck, David Finnicum From: David E. McCoy Ext: 205-664-3020 Fax: 860-731-2498 Your ref: NA Our ref: LTR-RAM-II-11-010

Subject:

RG 1.200 PRA Peer Review Against the ASME/ANS PRA Standard Requirements for the Sequoyah Nuclear Plant Probabilistic Risk Assessment Attached is the final report documenting the results of the full scope Regulatory Guide (RG) 1.200 peer review for the Sequoyah Nuclear Plant (SQN) Probabilistic Risk Assessment (PRA). This peer review process was performed under Task PA-RMSC-0386.

Please transmit this report to Tennessee Valley Authority for their use.

This report is proprietary to Tennessee Valley Authority because it contains plant-specific information for the Sequoyah Nuclear Plant. This report cannot be released to anyone outside Tennessee Valley Authority without their express written permission.

Questions may be referred to the undersigned.

Regards, David E. McCoy*

Author Risk Applications & Methods II

  • Electronically Approved Records are authenticated in the Electronic Document Management System

Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-1

Subject:

PLANT PROBABILISTIC RISK ASSESSMENT -

SUMMARY

NOTEBOOK Appendix A - Resolution of F&Os Finding Level F&Os F&O Number F&O Details 14 MDN00000020100203 does not document an assessment of the impact of flooding events on existing HFEs carried over from the internal events scenario used to represent the flooding event.

(This F&O originated from SR IFQUA6)

Associated SR(s)

IFQUA6 Basis for Significance Although this is a documentation issue, it is important to an understanding of the results and to show that the technical element is satisfied.

Possible Resolution Document a process for assessment of the impact of the flood scenarios on existing HFEs from the internal events PRA sequences used to represent the flood scenarios.

EPRI 1019194 Section 7.3 describes the types of HFE adjustments to be considered.

Response

To address human actions and their modification due to flooding events Section 9.3 was added to the document.

Section 9.3 addresses the changes to the human actions in the model by accounting for:

1. Human actions that are influenced by HRA actions, these are events that occur within an hour of flood initiation.
2. Human actions that are failed due flooding.

Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-2

Subject:

PLANT PROBABILISTIC RISK ASSESSMENT -

SUMMARY

NOTEBOOK F&O Number F&O Details 17 Dependency analysis was performed for the postinitiator HEPs using the EPRI HRA Calculator. However, several issues were identified including:

1) Use of the same cue for two actions can result in conservative dependency values. For example, the use of the same cue for actions HARR1 and AFWOP3 resulted in complete dependency between the actions. However, review of the cues indicated that the cue for AFWOP3 should be different than that for HARR1.
2) Inconsistent entry of the timing information creates results that may appear invalid. For example, the timing entries for actions HARR2 and AFWOP3 make it appear that core damage as a result of failure of HARR2 would occur before the cue for AFWOP3 is received. Discussion revealed that the Tsw for HARR2 is based on the time at which the RWST would empty rather than core damage as stated in the HRA Calculator.
3) Inclusion of screening HFEs in the dependency analysis can result in errors. The screening HEPs do not have information that is necessary for the dependency analysis (e.g., timing inputs). This can result in the wrong event being treated as the independent event in the combination. For example, review of dependency combination 41 shows that the dependency analysis treats HACD1 as the first or independent HFE in the combination and AFWOP5 as following HACD1. This results in a joint HEP of 1.0 based on complete dependency. However, the description of HFE HACD1, Perform cooldown with main feedwater following AFW failure, indicates that AFWOP5 should be the first event. This would result in a joint HEP of 2.9E03.
4) The dependency level of the cognitive recoveries were not entered in the HRA Calculator database for the postinitiators. This requires manual entry by the analyst and does not default to the recommended dependence level. Failure to enter this information may underestimate or overestimate the HEP depending on the applicable dependence level.

Some of these items were corrected during the review but they are documented in an F&O due to the need to evaluate the extent of the condition.

(This F&O originated from SR HRG7)

Associated SR(s)

HRG7 Basis for Significance Incorrect assignment of cues, timing, and resource entries can result in incorrect dependency analysis results.

Possible Resolution

Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-3

Subject:

PLANT PROBABILISTIC RISK ASSESSMENT -

SUMMARY

NOTEBOOK

1) Review the cues, timing assumptions, and resource requirements for significant HEPs to ensure that the factors are correctly assessed in the dependency analysis.
2) For combinations where the timing indicates Tsw is reached for the first action before the cue for the second is received, document the basis for acceptability of the dependent combination.
3) Ensure all information affecting the dependency analysis is entered into the HRA Calculator for the screening HFEs to ensure they are treated correctly in the dependency analysis.
4) Ensure that the dependence level between cognitive actions and applicable recoveries is set in the HRA Calculator database.

Response

1. Cue for AFWOP3 has been updated to correct cue. Review has been performed for all remaining actions to determine if any additional cues need to be updated. This review verified the accuracy of HRA cues and updated six of the identified cues.
2. The end point for Tsw is an irreversible damage state. For HARR2, this irreversible damage state is the loss of all ECCS pumps when the RWST is depleted and autoswap has failed. This is the correct irreversible damage state as the operator does not have until core damage to perform that action if the pumps fail when their suction source runs dry. The dependency analysis was reviewed for overlapping timeframes.
3. Screening value HEPs were removed from the database if there values were set to 1.0. The HEPs that were originally in the model were no longer required and were deleted from the fault tree.
4. This has been corrected for all of the actions in the SQN HRA.

Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-4

Subject:

PLANT PROBABILISTIC RISK ASSESSMENT -

SUMMARY

NOTEBOOK F&O Number F&O Details 18 MDN00000020100203 Section 9.5 only addresses quantification and results for CDF. There is no discussion of LERF for the flooding scenarios or documentation indicating that the flood scenarios were reviewed to determine if they would have an impact on the Level 2 CETs. The linked fault tree model should have the capability to produce LERF results, but this had not been done at the time of the review. In addition, there was no discussion in the Level 2 Notebook (MDN00000020100206) that indicates the results include the internal flood scenarios.

(This F&O originated from SR IFQUA10)

Associated SR(s)

IFQUA10 Basis for Significance No LERF results for internal flooding scenarios was provided for review.

Possible Resolution

1) Document a review of the top events in the Level 2 model to confirm that there are no unique flooding impacts that affect the CETs.
2) Document the LERF results for the internal flood scenarios similar to the results for other initiators in Section 11 of MDN00000020100206. This can be done in the flood notebook or the Level 2 notebook, but, if done in the Level 2 notebook this should be referenced in the Internal Flooding Analysis notebook.

Response

The internal flooding calculation was revised to add Section 10 (Results Analysis for Large Early Release Frequency).

Section 10.1 addresses the eighteen questions concerning LERF and their impact.

Section 10.3 and 10.4 address the LERF results due to flooding To address the additional information the following Appendices were added to the model:

Appendix Q Significant Cutset Review for Large Early Release Appendix R NonSignificant Cutset Review for Large Early Release Appendix S Importance Reports for Large Early Release

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NOTEBOOK F&O Number F&O Details 110 MDN00000020100206 Section 5.6 notes that credit was taken for scrubbing of releases from a ruptured SG. However, the technical justification for this credit needs to be strengthened. The current basis compares the zero power collapsed level to the top of the SG tubes. However, ES3.1, PostSGTR Cooldown Using Backfill allows the level in the ruptured SG to be between 20% narrow range and 75%

narrow range during the cooldown (Step 7). The expected levels during SGTR recovery should be used to justify the scrubbing credit.

It also appears that the analysis implicitly assumes that if FW will be applied to the ruptured SG if FW is available. No consideration of operator failure to provide FW flow to the ruptured generator is included in the analysis.

(This F&O originated from SR LEC4)

Associated SR(s)

LEC4 LEC13 LEE3 Basis for Significance The technical basis of the credit for scrubbing of SGTR releases does not consider the levels allowed in the EOPs.

Possible Resolution Revise the justification in MDN00000020100206 Section 5.6 to include consideration of the SG levels maintained during recovery using the applicable EOPs.

Response

The documentation has been updated to include a discussion of the water levels above the steam generator tubes during tube rupture recovery actions. These water levels (between 4.7 and 9.8 feet) should be sufficient to take credit for fission product scrubbing. This analysis assumes that the operator is successful in providing feedwater flow to the ruptured steam generator.

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SUMMARY

NOTEBOOK F&O Number F&O Details 111 The total LERF is compared with other Westinghouse 4loop plants and with other Ice Condenser plants. However, there is no comparison at the level of significant contributors or plant damage states. Without the contributor information, it is not really possible to determine how similar the LERF results are to other plants.

(This F&O originated from SR LEF2)

Associated SR(s)

LEF2 Basis for Significance There is no review of the contributors to LERF with the results for similar plants to ensure that plantspecific modeling choices have not skewed the results.

Possible Resolution Document a comparison of the LERF results to plants of similar design at the significant contributor and PDS levels (similar to Tables 111 and 113).

Response

The documentation has been updated to include comparisons by initiating event for several other PWRs in Table 117.

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SUMMARY

NOTEBOOK F&O Number F&O Details 114 Demand data is obtained directly from the plant process computer for most components, as described in Section 7.3 of the data notebook (MDN0000002010 0202). The status change information from the computer is filtered and used to determine the number of demands.

The use of automatic data collection, however, means that start and run events that occur in all modes of operation are included. In addition, postmaintenance test starts are also included in the data set. This is identified as a source of uncertainty in the sensitivities and uncertainties notebook (MDN00000020100209) and a specific set of sensitivity studies were performed that assumed that various numbers of successful starts were invalid. The results show that the impact on CDF is relatively small, unless the number of successful starts is overestimated by a large amount. However, this SR is explicit in its requirement to not count post maintenance test events.

(This F&O originated from SR DAC6)

Associated SR(s)

DAC6 Basis for Significance This is considered to be a finding since a specific technical requirement of the SR is not met.

Possible Resolution To comply with this SR, the postmaintenance test starts (following a component failure) should be removed. It would also be more correct to also screen out component demands that occur during shutdown periods (e.g., by filtering out data based on the date of the event).

Response

The work orders for the components that were credited for success in the data analysis were reviewed to discover the number of post maintenance tests that were performed on the components. Table 15 was added to document the number of post maintenance tests that were removed from the analysis.

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SUMMARY

NOTEBOOK F&O Number F&O Details 115 The super initiator "general transient" may overlook certain differences among its contributors. For example, the impact of specific IEs like LOSP and Loss of DC that may prevent PORV operation and challenge the Pressurizer Safeties do not appear to be captured.

In addition, failure to provide a separate event tree for SBO may overestimate the success of power recovery by not addressing the operation of systems such as charging and AFW following power recovery.

(This F&O originated from SR ASA10)

Associated SR(s)

ASA10 ASB1 SCB3 Basis for Significance The accident sequences do not contain sufficient detail to capture important system requirements and required operator interactions for all initiating events.

Possible Resolution

1) Subdivide the General Transients event tree to better represent the unique challenges presented by specific initiating events (e.g., Transient with Loss of PCS, Transient with PCS Available, LOSP) or document how those challenges are addressed in the top logic model.
2) Modify the existing event sequence and/or linked fault tree to ensure that the challenge to the Pressurizer Safeties is captured for initiating events that would prevent the PORVs from opening.
3) Explicitly model the SBO sequences to ensure that the necessary mitigating systems are addressed following power recovery.

Response

GTRAN was restructured to address this comment. The tree was updated to explicitly ask demand for PORVs and Safeties

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NOTEBOOK F&O Number F&O Details 119 It was noted that HFE HAPRZ (discussed in Section 6.8 and Section 7.2) is not calculated using HRA Calculator. This event seems to have been carried over from the Watts Bar analysis and is treated as basic event U1_L2_NOTRCSDEPNOSBO.

In addition, although Section 6.8 says that the No RCS Dep branch is set to a value of 1 for SBO cases, the value of basic event U1_L2_NOTRCSDEPSBO in the provided MASTERL2.CAF fault tree was set to 0.9995. This also appears to be a carryover from Watts Bar.

(This F&O originated from SR LEC7)

Associated SR(s)

LEC7 Basis for Significance The HFEs for intentional depressurization needs to be evaluated to determine their applicability to SQN.

Possible Resolution

1) Include HAPRZ in the HRA analysis for SQN or justify the applicability of the Watts Bar value and provide an appropriate reference to the source.
2) Verify that the proper value of basic event U1_L2_NOTRCSDEPSBO is being used in the quantification.

Response

The current analysis has been updated to change the value of failure to depressurize the RCS during SBO scenarios to 1.0 (assumed failure) in the model. The basic event HAPRZ, which represents failure to depressurize for nonSBO scenarios, uses a value of 0.1 for failure to depressurize, which was taken from WCAP 16341P, revision 0.

The level 2 event trees also uses the compliment to this action called HAPRZSUC which has a probability of 0.9.

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SUMMARY

NOTEBOOK F&O Number F&O Details 21 Section 7.0 of the Initiating Events Analysis observes a decreasing trend in initiator frequency in the more recent generic data sources. However, there is no comparison of the SQN results against the generic results nor an explanation of any significant differences.

(This F&O originated from SR IEC12)

Associated SR(s)

IEC12 Basis for Significance The current evaluation of the initiator frequency results does not compare SQN results to the generic frequency results.

Possible Resolution The section 7.0 discussion could be expanded to include a comparison of the SQN results to the generic results and an explanation of any significant differences.

Response

Added text to Initiating events notebook that compares Sequoyah initiator frequencies to generic industry data.

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SUMMARY

NOTEBOOK F&O Number F&O Details 23 Section 4.3.1 of the Data Analysis notebook discusses the basic event probability model methodology. Generic data sources selected for use are applicable for SQN.

For those components which had a failure during the analysis time period (1/1/03 11/30/09), the distributions are updated via the Bayesian update program built into CAFTA program. However, the intent of this supporting requirement is to assure realistic parameter estimates are calculated for SIGNIFICANT basic events based on relevant generic and plantspecific evidence, not just those for which failures have occurred. Where no failures have occurred, use of the generic data may be conservative since it includes failures from potentially less reliable components across the industry.

(This F&O originated from SR DAD1)

Associated SR(s)

DAD1 DAD3 Basis for Significance Using potentially conservative failure rates for significant components can skew the risk results. Both generic and plantspecific experience should be considered for the significant basic events.

Possible Resolution Consider performing a Bayesian update for all significant basic events not just those for which failures have occurred.

Response

Significant contributors that were not Bayesian updated were identified as:

BATFR Battery Fails to Operate BUSFR Bus Fails to Operate CBKFO Circuit Breaker Fails to Open FNSFD Standby fan fails to start HXRPL Heat Exchanger (River Water) Plugs or Fouls MOCXC Motor Operated Valve Transfers Closed POEFR ERCW pumps fail to run PSRFR RHR pumps fail to run STRPL Strainers plug

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SUMMARY

NOTEBOOK TSCPL Traveling water screens plug XRFR Transformer fails to operate These events were Bayesian updated using plant specific data. The notebook has been updated to reflect these additional updates.

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SUMMARY

NOTEBOOK F&O Number F&O Details 24 Appendix F of the Data Analysis notebook provides graphs that show the prior and posterior distributions. Table 19 lists generic and Bayesianupdated mean values, along with a ratio of the posterior to prior mean value. However, there are no conclusions drawn about whether or not the posterior distributions are reasonable given the relative weight of evidence provided by the prior and the plantspecific data. (Note: the statement that "There are no significant differences between the industry data from NUREG/CR6928 and the posterior distributions for the SQN failure rates" in section 11.0 is not judged to be sufficient. For example, the ratio of the posterior to prior mean for the AHUFR type code in Table 19 is 10.6. For type code LSTFR, the ratio is 4.3. The significance of these differences should be discussed.)

(This F&O originated from SR DAD4)

Associated SR(s)

DAD4 DAE2 Basis for Significance The reasonableness check needs to assess whether the Bayesian updates yield expected results given the relative weight of evidence provided by the prior and the plantspecific data.

Possible Resolution Discuss the observed differences in the prior and posterior distributions and draw conclusions on the significance associated with those differences.

Response

The posterior distributions were validated using the following process. Using a Monte Carlo simulation, the posterior distributions were samples to see the probability of having a recurrence in the number of events observed in the data window given the number of successes in the data window. If the mean value was within 0.05 to 0.95 the resultant distribution was used within the model.

Appendix F was rewritten to address this analysis as well as to present the prior, posterior, and plant specific distributions.

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SUMMARY

NOTEBOOK F&O Number F&O Details 25 The method from NUREG/CR6823 is used to Bayesianupdate a Jeffreys noninformative prior distribution with plantspecific experience. However, there is no comparison of the posterior means to plantspecific means. (See the last sentence in NUREG/CR6823, section 6.7.1.2.)

(This F&O originated from SR DAD4)

Associated SR(s)

DAD4 DAE2 Basis for Significance A reasonableness check should be performed to assure the Bayesianupdated maintenance unavailabilites yield expected results when compared to plantspecific mean values given the amount of plantspecific data.

Possible Resolution Compare the Bayesianupdated maintenance unavailabilites to plantspecific mean values, discuss the observed differences and draw conclusions on the significance associated with those differences.

Response

The fundamental assumption used in the Bayesian update process described in the Data Analysis notebook for unavailability calculations is that there is no prior information from which to Bayesian update. Therefore, the methodology used was to use a Jefferys noninformative prior (0.5) as the foundation for the update process.

All of the available data that was used was from plant specific data collection, therefore the posterior mean and plant specific mean are directly correlated. The following assumption was added to Section 3.0 to address the noninformative prior.

"For unavailability calculations, a Jefferys noninformative prior was used as there was no informative prior information available."

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SUMMARY

NOTEBOOK F&O Number F&O Details 28 The importance of components and basic events are identified in sections 5.1 and 5.7 of the Accident Sequence notebook, respectively. However, documentation that determined the importance results make logical sense could not be identified.

(This F&O originated from SR QUD7)

Associated SR(s)

QUD7 Basis for Significance Multiple reviews of the model solution results yielded model changes, as documented in Table 7.01 and Appendix F of the Quantification notebook.

Importance measures are calculated in section 5.7; however, these need to be evaluated in light of the model solution results. In other words, do the importance measure reports yield the expected results?

Possible Resolution Document an evaluation of the importance measure results in light of the CDF results.

Response

A review of the importance of components and basic events has been performed to determine that they make logical sense. The review shows that the risk significant components are consistent with the model results and limitations. Significant contributors include basic events associated with diesels, ERCW, Component Cooling, RHR, Atmospheric Relief Valves (ARVs) and Air Compressors. In SQN, failure of the auxiliary control air headers impacts the ARVs that are needed to cooldown/depressurize in LOCA scenarios since the condenser is unavailable from a Phase B isolation. The emergency diesel, ERCW, RCP breakers, and RHR are important since their failure result in scenarios involving SBO and RCP seal LOCAs.

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SUMMARY

NOTEBOOK F&O Number F&O Details 31 Section 4.5, The calculation above provides that the containment hole size must lie between a 1 inch equivalent path and a 4 inch path. Therefore, it is acceptable to use the NRC value of 2 inches. Based on the statement, the 1 equivalent hole should have been considered.

(This F&O originated from SR LED7)

Associated SR(s)

LED7 Basis for Significance It is unknown what the applicable break size is between 1" and 4", therefore the conservative approach is to use 1".

Possible Resolution Perform detailed analysis to ensure the use of the 2" equivalent hole is allowable or use 1" and include the additional penetrations in the containment isolation analysis.

Response

Section 4.4 discusses the reasoning for concluding that the 2 hole size is acceptable for use in the Sequoyah level 2 analysis. The reference shows that the release rate corresponding to a 1771 scfm rate would be represented by a vent line diameter greater than 1 and slightly less than 2. Because the point corresponding to 1771 scfm at 19 psig (which is half of the assumed severe containment challenge pressure) is only slightly below the 2 contour line shown in Reference 33, and there is conservatism built into both the assumed containment failure pressure and the assumed leak rates at that pressure, it is judged appropriate to use 2 as the bounding value for a large leak rate.

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SUMMARY

NOTEBOOK F&O Number F&O Details 37 Several areas were identified that need additional discussion with respect to the Success Criteria Analysis. For example:

1) The differences between plant response to a pipebreak SLOCA and a consequential PORV LOCA are not fully discussed. Given the differences in break location, there should be some discussion in the Success Criteria Notebook of why the pipebreak SLOCA analyses bound the consequential PORV LOCA. In addition, while there is a discussion in the TH Notebook comparing the values of some key parameters for the pipebreak SLOCA and the consequential PORV LOCA, this does not fully explore differences in plant response that may affect the success criteria.
2) There needs to be more discussion of why the 480 gpm per pump RCP Seal leaks are included in the Medium LOCA (MLOCA) grouping. It is stated in Section 4.4.10 of the TH Notebook that the 480 gpm seal LOCA meets the MLOCA requirement of not requiring AFW for accident mitigation, but there is no documentation of success criteria analyses that support this statement.
3) The basis for assuming a SGTR flow of 700 gpm in Section 7.2.10 of the TH Notebook needs to be discussed in more detail than simply noting that no historic SGTR has been of the magnitude of a doubleended guillotine rupture of a SG tube.
4) The LOCA analysis is limited to the upper and lower end of the break range for each class. TH analysis at the middle of the break range within the Large, Medium, and Small LOCA categories may provide insights that have not been revealed by the upper and lower end of the break. For instance, it is not clear if sequence MLOCA011 can be a success path for a break in the 3 to 5 inch range.

(This F&O originated from SR SCB3)

Associated SR(s)

SCB3 Basis for Significance There is a lack of discussion regarding how these items were treated in the success criteria analysis.

Possible Resolution Expand the discussion of the noted items in the Success Criteria documentation.

Response

1) The small LOCA events assume that he break occurs low within the physical structure of the RCS. These breaks will always have a higher deltaP value than

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SUMMARY

NOTEBOOK those of breaks at the top of the RCS (PORV LOCA). Due to the additional pressure and other thermohydraulic characteristics the success criteria is bounding for the SLOCA cases.

2) The 480 gpm seal LOCA is now grouped as a SLOCA. This requires the use of AFW for successful accident mitigation.
3) The value of 700 gpm was used as an attempt to bound the analysis. The selection of 700 gpm was done to assure that the analysis was realistic in nature, but conservative as well.
4) The MLOCA event tree has been restructured to require successful injection of the CLAs this is to assure that any break size within the MLOCA range can be successfully mitigated after failure of the CVCS system to inject.

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SUMMARY

NOTEBOOK F&O Number F&O Details 39 All mitigation strategies credited in the accident sequence model when the high pressure recirculation has failed are not prescribed by the corresponding EOPs. In other words, the mitigation credit in the event tree model has no basis. This issue has been selfidentified by the SQN PRA staff and a corrective action report has been written for the EOP group to resolve this issue. At this stage the PRA group "firmly" believes that the EOP will be modified, not the model. Thus it is a tracking issue.

(This F&O originated from SR SCA3)

Associated SR(s)

SCA3 Basis for Significance There is a CR written by the TVA.

Possible Resolution Tack the CR and ensure that the procedures are modified or that the model is changed to reflect the asoperated plant.

Response

EOP revisions were approved at the SQN PORC meeting on May 6th 2011.

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SUMMARY

NOTEBOOK F&O Number F&O Details 313 Section 4.4.2 of the TH Notebook (MDN0000002010205) discusses the use of MAAP for LLOCA in the cold leg. The conclusion is that the large LOCA (LLOCA) limitations are not applicable to break sizes < 10 inches. The reference used for this is a MAAP training lecture. Use of MAAP to model the injection phase of the LLOCA needs additional justification with reference to the applicable technical documents.

(This F&O originated from SR SCB4)

Associated SR(s)

SCB4 Basis for Significance MAAP is known to have difficulty modeling the initial phase of the LLOCA events.

Possible Resolution Use RELAP or other alternative codes to analyze the initial phase of the LLOCA or provide a more comprehensive justification for the use of MAAP which includes benchmarking against other codes.

Response

The limitation noted for MAAP are for the larger end of the LLOCA spectrum per EPRI TR1020236. The success criteria for the large LOCA was consistent with and largely derived from the SQN design basis analysis and SAR. While this does lead to conservative results in the LLOCA event tree, the expenditure of additional resources for the further refinement using additional codes such as RELAP is not warranted, given that LLOCA events are not risk significant in the SQN model. The low importance of the LLOCA sequences is consistent with other PWRs in the industry.

The MAAP analysis for the LLOCA events were used mostly as confirmation of the event trees based on the SQN SAR and for timing of HRA events. Specifically for the HRA events, MAAP was only used to determine depletion of the RWST and long term time to core damage based on failure of hot leg recirculation. Both of these cases are significantly past the initial stages of a LLOCA where MAAP is noted to lack the thermal hydraulic detail required to evaluate the initial blowdown (EPRITR1020236).

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SUMMARY

NOTEBOOK F&O Number F&O Details 314 Several documentation issues were noted in the Success Criteria and TH Notebooks.

Specifically,

1) Figures 760 and 761 of the TH Notebook (MDN0000002010205) need to be replaced with updated results.
2) The discussion of accident sequence node LPH in Section 7.3.1 of the TH Notebook (MDN0000002010205) states that The time for switchover to hot leg recirculation is specified in the EOP E1 as 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the initiation of a large LOCA (Reference 4, Step 31c). In the paragraph immediately below this statement, the calculation of the time available for recovery from a failure of recirculation uses a switchover time of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Discussion with TVA personnel indicated that the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> value was copied from the WBN notebook. The actual time specified in the SQN procedures is 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
3) Table 713 of the TH Notebook (MDN0000002010205) does not include success path ISLM014 as shown in Figure 6.410 of the Accident Sequence Notebook (MDN00000020100201). In addition, success path ISLM017 in Table 713 of the TH Notebook is not shown in Figure 6.410 of the Accident Sequence Notebook.
4) Section 4.4.11 of the TH Notebook (MDN0000002010205) discusses the classification of a Stuck Open PORV as a small LOCA. The basis needs to be provided.

(This F&O originated from SR SCC1)

Associated SR(s)

SCC1 Basis for Significance The documentation needs to match the current analysis.

Possible Resolution

1) Replace figures 760 and 761 with the correct figures.
2) Revise the text to use the correct information for SQN.
3) Ensure the sequence designations in Table 713 of the TH Notebook (MDN000 0002010205) match those in Figure 6.410 of the Accident Sequence Notebook (MDN00000020100201).
4) Justify the classification of the Stuck open PORV as a SLOCA.

Response

1) Figures 760 and 761 were revised in the TH calculation MDN0000002010205.

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SUMMARY

NOTEBOOK In the original MAAP runs, the SG ARVs were opened at 30 minutes, this dropped pressure in the RCS. Opening of the SG ARV was not credited in the event tree for the sequences evaluated in figures 760 and 761. This is applicable to the WBN TH analysis as well.

2) The TH Notebook was revised to be consistent with EOI E1 step 22. The correct time of switching over to Hot Leg Recirculation of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> was included in Section 7.3.1 of the TH Notebook.
3) Table 713 and Figures 6.410 were revised to be consistent.
4) Additional information was included in section 4.4.11 of the TH notebook to justify the classification of a Stuck Open PORV. This information includes a comparison of core damage timing and mass/energy release rates through a SOPORV and SLOCA.

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SUMMARY

NOTEBOOK F&O Number F&O Details 319 Section 7.2 of the HRA Notebook (MDN00000020100204) does not explicitly discuss how the required and available manpower is addressed in the analysis.

Manpower requirements are included in the operator interview checklist as item 37.

However, it is not clear how this information was used in the development of the HEPs since some instances were observed where the operator interview responses were not used in the HRA calculator (see HFE HARR1).

(This F&O originated from SR HRH2)

Associated SR(s)

HRH2 Basis for Significance Some accident scenarios can require more manpower than others and this is not discussed.

Possible Resolution Add a discussion of how the manpower requirements are accounted for in the HRA, especially for those HFEs which require local actions.

Response

A discussion of the required and available manpower to perform the actions and equipment manipulations was documented in sections 7.1 and 7.2 of the HRA notebook. Also, HARR1 was revised to match the operator interview for the manpower requirements.

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SUMMARY

NOTEBOOK F&O Number F&O Details 320 Several issues related to the TH analyses used to support the HRA were identified.

Specifically,

1) Some time windows are buried in MAAP output files which are not included in the TH Notebook and take time to review. For example, the time window for AFWOP5 is not easily available.
2) TH Notebook MDN0000002010205 Section 7.3.3 discusses the actions required following a failure of high pressure recirculation. The required action related to failure of the automatic recirculation alignment (HARR1) has two big pieces. The first is to stop the pump to avoid pump damage. If the pumps are damaged, high pressure recirculation can't be successful. The time window is short for this action and is related to RWST depletion. If the pumps are stopped on time the next action is to manually establish recirculation. The time window for that action is based on the RCS inventory depletion which is, relatively speaking, much longer.

If HP recirculation is not successful, the RCS is depressurized to facilitate low pressure recirculation (AFWOP3). These two actions (HP recirculation and RCS depressurization and establish LP injection/recirculation) are for the same mitigation function. Therefore, it is unclear why there are big differences between the time windows for these two actions. In addition, the HRA Calculator input for these actions appears to be different from the descriptions in Section 7.3.3 of MDN0000002010205.

3) The use of bounding analyses for the HFEs results in non sequence specific timing information in the HRA. For example, HARR1 is used in the accident sequences after AFWS success in SSBO and SSBI accident sequences. However, the timing window of HARR1 is based on the medium LOCA and it is conservative for these sequences.

(This F&O originated from SR HRI1)

Associated SR(s)

HRI1 Basis for Significance The documentation of the TH analyses performed to support the HRA is difficult to trace and in some cases contains conflicting information.

Possible Resolution

1) Revise Table 81 in the TH Notebook (MDN0000002010205) to include additional information such as the time to core damage and a reference to the applicable TH cases.

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SUMMARY

NOTEBOOK

2) Review the results of TH cases supporting the HRA to ensure reasonable consistency of time windows for different actions with the same purpose.
3) Refine the timing analysis as necessary to ensure the results are realistic and represent the accident sequence(s) in which the actions are used.

Response

1) TH notebook revised - all HRA timing in Table 8.1
2) All TH result cases were reviewed to ensure that the time windows in use were consistent between different actions with the same purposes.
3) As stated in the details of the F&O, the analysis used is conservative. The timing analysis is for the most time limiting break for which the action is applied. This conservative timing selection addresses all potential scenarios/break sizes and would only reduce HEP and add additional margin to the analysis. This is considered to be appropriate due to the ranges of break sizes included in the broad bands of initiating event groupings.

Evaluation of the recovery of additional margin from developing lower HEP individual analyses for each application of HARR1 will be completed in future revisions of the SQN PRA model.

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SUMMARY

NOTEBOOK F&O Number F&O Details 325 Several documentation issues were noted. For example:

1) Sequences ISLM008 and ISLM017 were deleted from the ISLOCA event tree.

However, there is no discussion of why this was done.

2) Paragraphs in section 6.4.7 need to be revised. Specifically, the first sentence in the first paragraph on page 62, starting with "If the temperature of the RCS is 557°F and dropping, the steam dumps, S/G PORVs and blowdown isolation valves are closed." needs to be finished. There is the "if" but no "then." It is also unclear how this sentence is related to the accident sequence event tree or the following statements in the paragraph related to the PORVs.

The second paragraph on page 62 has grammatical errors (e.g., the possibility of have a RCP Seal LOCA).

3) The discussion of manual control rod insertion following ATWS in section 7.9 needs to be revised to reflect the intent to remove credit for this action from the model.

(This F&O originated from SR ASC1)

Associated SR(s)

ASC1 Basis for Significance Inconsistencies in the documentation can affect maintenance and update of the model.

Possible Resolution

1) Add a discussion explaining why sequences ISLM008 and ISLM017 are not used or renumber the sequences to ensure there are no gaps in the numbering. Also, ensure all related documents (e.g., the SC and TH notebooks) are revised for consistency.
2) Review sections 6.4.7 and 7.9 and revise, as needed, to ensure that the discussion reflects the accident sequence models.

Response

1) The sequences were not renumbered following the latest update to the event trees. The numbering scheme will be updated in the next revision of the notebook.
2) The grammatically errors noted have been updated and revised.
3) The ATWS discussion of MRI has been updated to state that only the mechanical binding of the control rods or the failure of the automatic control system are

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SUMMARY

NOTEBOOK modeled.

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SUMMARY

NOTEBOOK F&O Number F&O Details 43 Nonwater flood sources are excluded on the basis of Assumption 11 of the notebook. However, the Standard states (in Note 1 for this SR) that nonwater sources should be considered, A more detailed basis for excluding these sources should be developed to meet the requirements of this SR.

(This F&O originated from SR IFSOA1)

Associated SR(s)

IFSOA1 Basis for Significance This is considered to be a finding since the requirements of the SR have not been fully met.

Possible Resolution Update the analysis to consider nonwater sources, or better justify why the flooding impacts of these nonwater sources are not significant and hence do not require evaluation.

Response

Assumption 11 was reworded to:

All sources of fluid within the plant were analyzed for flooding considerations.

However, the glycol system is the only system which could have an impact on the flooding analysis. All other sources such a resin did not have enough volume to cause impact to plant operation. The glycol system also has a minimum volume, but the location of the piping, in the control rod drive rooms, causes system to be a source of spray initiating events.

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SUMMARY

NOTEBOOK F&O Number F&O Details 47 No discussion of sources of uncertainty associated with the flooding initiating events is currently provided in the flooding notebook (MDN00000020100203). It is noted that the notebook includes documentation of sources of uncertainty for other portions of the flooding analysis. Sources of model uncertainty for internal flooding are also documented in MDN00000020100209, Uncertainty and Sensitivity Analysis; however, again flood initiator uncertainties are not discussed. If no uncertainties are identified for the flood initiator frequency evaluation, then the notebook should state this to be consistent with the approach used for the IFPP, IPSO, and IFSN tasks.

(This F&O originated from SR IFEVB3)

Associated SR(s)

IFEVB3 Basis for Significance This is considered to be a finding since the requirements of this SR are not met Possible Resolution Provide an assessment of sources of modeling uncertainty for the flood initiator frequency determination.

Response

Section 8.8 was added to the Internal Flooding Notebook with the following:

The internal flooding frequency calculation has several different uncertainties associated with the calculation. The current model uses a summation of three different frequencies, passive pipe break failures, human induced floods, and maintenance induced flooding. Each of these flooding events has its own inherent uncertainties.

For passive pipe break failures rates have been given an uncertainty parameter as presented in Section 8.5. The impact of these uncertainties can be treated by the use of a random sampling Monte Carlo process as discussed in Section 10.1.

Human induced flooding events present another difficult challenge. The use of the HRA Calculator program from Scientech creates an assumed uncertainty term for any HRA action. Since the human induced flooding events is a combination of both pre initiating event and post initiating event, each portion has an independent uncertainty term. The HRA Calculation program also arbitrarily assigns an uncertainty term to HRA actions based on the calculated probabilities, see the HRA Calculation for more information on the uncertainty parameters (Reference 68). The other fundamental issue that is presented in human induced flooding events is the location of work. Depending on where the actual work is being performed in a flood area,

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SUMMARY

NOTEBOOK isolation could be a concern as the next available valve could be in an inaccessible area. Additionally, there are no detailed procedures to address having a flood occur during a maintenance event.

Maintenance induced flooding events also present a level of uncertainty. The three main inputs to the calculation of this frequency, failure rate of an MOV, mission time, and frequency of the activity all introduce some level of uncertainty into the calculation. The large internal rupture of an MOV is assumed in NUREG/CR6928 to be a factor of 0.02 less than that of a small internal leak on an MOV (Reference 104),

as there has been no actual large internal rupture events in the industry. The mission time is also assumed based on a seven day repair interval, this number could potentially be greater than that if the component is not covered by an Technical Specification or, more likely, less than the assumed seven day repair time. The final area of uncertainty is the frequency of the activity. Most of the procedures reviewed in Appendix J have frequencies as well as conditions. These conditions could cause the actual maintenance activity to occur more times than the frequency noted in the procedure.

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SUMMARY

NOTEBOOK F&O Number F&O Details 411 While the PRA model considers the possibility of two PORVs being blocked at the same time, there does not appear to have been an investigation of whether coincident maintenance can occur in the various SQN systems (or if coincident inter system maintenance can occur). Therefore this SR is not met.

It was also observed that the PORV blocking basis events noted above did not appear to be documented in either the data notebook or the appropriate system notebook.

(This F&O originated from SR DAC14)

Associated SR(s)

DAC14 Basis for Significance This is a finding since the technical requirements of the SR are not met.

Possible Resolution A study should be conducted to determine if coincident maintenance conditions can occur. If so, the system models may need to be modified and additional basic events to represent the coincident maintenance states would need to be added. If it is determined that no coincident maintenance can occur, then this should be documented in the data notebook or within the system notebooks.

Documentation for the calculation of the time that either one or both PORVs can be blocked should also be added to either the system notebook or the data notebook.

Response

The following was added to the data analysis notebook to address coincident maintenance:

Coincident maintenance is scheduling maintenance where multiple SSCs are out of service at the same time. Specifically components on the same train, RHR train A and SI train A for example, being out of service for maintenance at the same time. The Outage and Site Scheduling Directive Manual 1.0 (Reference 28) dictates that:

Twelve (12) week schedule by FEG groups ensures that within a train week, no two (2) accident mitigating devices are removed from service at the same time [i.e., A train Residual Heat Removal (RHR) is not removed from service at the same time as A train Containment Spray.]

This requirement is further discussed in the Outage and Site Scheduling Directive Manual 4.7 (Reference 29) which states that any systems important to PRA that are unavailable at the same time must meet the requires of the plant risk matrix.

Normally maintenance on any systems important to the PRA is not scheduled at the same time. If it is these instances are extremely rare and the current model does not

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SUMMARY

NOTEBOOK exclude coincident maintenance events from appearing in a single cutset. Therefore the probability of having coincident maintenance events is extremely rare and accounted for during the normal cutset processing.

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SUMMARY

NOTEBOOK F&O Number F&O Details 52 Some HFEs are set to a value of 0.0 for quantification. For example, HACI1 and HAAE1 are recovery actions for automatic signals ANDed with the signal logic.

However, the HRA analysis sets the HEP probability to 0.0 based on an analysis that the operator action is not required. This screening approach, combined with the model structure, removes the auto actuation contribution to mitigating system failure during quantification.

(This F&O originated from SR HRG1)

Associated SR(s)

HRG1 Basis for Significance Screening HFEs using a value of 0.0 remove the auto actuation hardware failure contributions in the quantification results.

Possible Resolution Revise the model to remove noncredited operator recovery actions from the linked fault tree or set all noncredited events to TRUE during quantification.

Response

For those events where 0.0s were used in the model the fault tree was updated to remove the events so that the conflict concerning an AND gate and a zero event will not longer be encountered during normal quantification.

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SUMMARY

NOTEBOOK F&O Number F&O Details 62 The justification for excluding plant data prior to July 2002 in the calculation of plant specific IE frequencies is not documented well enough to support IEC2.

(This F&O originated from SR IED2)

Associated SR(s)

IEC2 IED2 Basis for Significance A justification was provided during the review, but it is not documented in the notebook.

Possible Resolution Include a justification for excluding data prior to July 2002 in the IE notebook.

Response

Added discussion to notebook stating that date range was adequate to get a good sample of plant data without going too far back and including events that occurred when the plant may have had different procedures and operating practices

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SUMMARY

NOTEBOOK F&O Number F&O Details 63 The alignment flags in the ERCW system are not fully implemented to represent the system alignment within the Initiating event portion of the tree. For example, the gates under U0_AEX_G006 should contain flags to indicate which pump is running and which two pumps are not, so that the two non running pumps would have considerations for failures to start.

(This F&O originated from SR IEA6)

Associated SR(s)

IEA6 IEC10 Basis for Significance The current fault tree configuration does not properly account for the system alignment.

Possible Resolution Include the alignment flags in the indicated and similar gates. Review the remainder of the tree to ensure that the alignments are properly identified.

Response

The current flag alignment for ERCW has been revised so, for the baseline model, without setting a specific configuration, the flag files were set to the respective time in each configuration to that a probability is now used not a true or false value.

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SUMMARY

NOTEBOOK F&O Number F&O Details 65 The support system initiating event trees for the most part include provisions for common cause failures and routine system alignments. There are some discrepancies in the modeling of common cause failures in the ERCW and CCS models that require attention, however. For example:

1) While a common cause event for all 3 of the 1A, 1B, and CS pumps failing to run exists, there are not events for the 1A and CS pumps or the 1B and CS pumps.
2) The structure of the ERCW tree is such that pump common cause failures could result in a pump failing due to an independent failure as well as a common cause failure in a single cutset. (See gate U0_AEX_G001)
3) The common cause initiating event group U0_ERW08POEFRI is not valid, since it is entirely based on 8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> exposure time for all the components. The common cause failure frequencies are therefore overestimated. The CCS tree uses a different approach than the ERCW tree for common cause initiating events. An alternate approach is also given in EPRI reports 1013490 and 1016741.

(This F&O originated from SR IEA6)

Associated SR(s)

IEA6 IEC10 SYB3 Basis for Significance Support system initiating event failures are inconsistently applied among the support system models, and may be giving incorrect results.

Possible Resolution Review initiating event common cause events and select a consistent modeling approach among the support system initiating event models.

Response

With respect to the common cause failure of the CCS pumps:

The common cause failure of the 1A and the CS pump or the 1B and the CS pump would not meet the requirements to cause an initiating event for the CCS system.

Only failure of the A train would cause the plant to have to trip as the loads on the common train are not required for operation at power. Therefore only the common cause failure of all three pumps is modeled in the fault tree.

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SUMMARY

NOTEBOOK With respect to the common cause failure events from the ERCW fault tree:

The common cause failure events in the ERCW system where common cause failure and independent failures show up in the same cutset present a minimal and conservative impact.

With respect the common cause calculation of basic events:

The common cause failure rates for ERCW pumps failing to run and CCS pumps failing to run were revised based on the EPRI document 1013490 using the discussion presented on page 58. The assumptions and calculation of these basic events is noted in Appendix B of each calculation.

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SUMMARY

NOTEBOOK F&O Number F&O Details 66 Section 5 of the IE notebook shows a Bayesian process was used to combine plant specific and generic data. However, LOCA frequencies from NUREG1829 were also updated with plant specific data. Since the frequencies in NUREG1829 were based on expert judgment and not actual industry data, and it is not expected that a plant would experience such an event, it does not seem appropriate to use the Bayesian update process for these events. The update did not appear to significantly alter the IE frequencies, however, so there is little impact on CDF.

(This F&O originated from SR IEC4)

Associated SR(s)

IEC4 Basis for Significance Frequencies in NUREG1829 were based on expert judgment and not actual industry data, and it is not expected that a plant would experience such an event, it does not seem appropriate to use the Bayesian update process for these events.

Possible Resolution Use the frequencies derived from NUREG1829 without Bayesian updating with plant data.

Response

The frequencies presented in NUREG1829 represent the best estimates available at that time. There is no restriction on updating an expert solicitation, as the update process will only serve as to better estimate the actual failure rate for the initialing events.

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SUMMARY

NOTEBOOK F&O Number F&O Details 67 Section 6 of the Initiating Events Analysis, the associated system notebooks, and the HRA notebook document the use of plantspecific information in the assessment and quantification of recovery actions where available, in a manner consistent with the applicable HR SRs.

An issue was noted with the ERCW initiating event tree. Event HAAEIE "Operator Fails to Start ERCW Pump (Initiating Event)" has been set to zero based on an analysis that found one pump was sufficient to cool plant loads, so if one of the two running pumps trips, operator action is not required to start another pump. Operator action to start a standby pump would be required, however, if flow was to be lost from both running pumps. The current model essentially assumes a successful operator action to start both of those pumps.

(This F&O originated from SR IEC11)

Associated SR(s)

IEC11 Basis for Significance The operator action HAAEIE has inappropriately been assumed to be 100%

successful.

Possible Resolution Reevaluate the failure rate for operator action HAAEIE, given the revised requirements of the ERCW system with regards to causing an initiating event.

Response

The ERCW initiating event model has been updated.

Calculation CNNUCSQNMEBMDQ00006720000095 revised the existing success criteria used in the initiating event model. The results of the calculation that as long as the containment spray heat exchangers were not in service, the maximum required flow on the ERCW system would be roughly 9,000 gallons. This is within the design flow rate of 10,000 gallons per minute from one ERCW pump. Due to the change in the success criteria, the initiating event model was update to requiring the failure of two running ERCW pumps as well as failure of both standby ERCW pumps to start.

The HRA action HAAEIE was added to the model under the appropriate failure to start gate, no longer under an AND gate.

Additionally, the fault tree logic in question was update so that failure to start takes into account the failure of operation action HAAEIE.

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SUMMARY

NOTEBOOK F&O Number F&O Details 610 Tables 42 and 43 of MDN00000020100209 contain a list of modeling assumptions and their impact on the PRA model. However, the majority of items in Table 43 have an impact of Unknown. Classification of model impact for these assumptions is necessary to meet this SR.

(This F&O originated from SR QUE4)

Associated SR(s)

QUE4 QUF4 Basis for Significance The SR requires identification of the impact of identified assumptions on the model.

Possible Resolution Provide an evaluation of an impact of the items listed as Unknown in table 43.

Response

The Uncertainty and Sensitivity Analysis calculation has been updated in the following ways:

Text concerning the discussion of Unknown impacts and performing a respective uncertainty analysis was removed from Section 5.0.

Table 43 was updated to remove the column "Model Impact" and the column "Comments" was updated to "Model Impacts and Comments" and expanded.

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SUMMARY

NOTEBOOK F&O Number F&O Details 612 From the results presented in sections 5.2 and 5.7 of MDN00000020100208, it can be inferred that the definition of significant basic event and significant accident sequence are consistent with those listed in Part 2 of the standard. This is not explicitly stated in the documentation, however. The definition of significant cutset is not provided, nor does the 100 cutset list provided in the documentation imply that the part 2 definition was used, as the 100 cutsets do not represent 95% of the risk.

(This F&O originated from SR QUF6)

Associated SR(s)

QUF6 Basis for Significance Documentation of the definition of "significant" is required by the SR.

Possible Resolution Provide a definition of significant cutset, significant sequence, and significant basic event in the documentation.

Response

The documented definition in Section 12.2 of the ASME/ANS combined standard was added to the quantification calculation.

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SUMMARY

NOTEBOOK Suggestion Level F&Os F&O Number F&O Details 11 RG 1.200 Revision 2 documents a qualified acceptance of this SR. The NRC resolution states that to meet Capability Category II, the impacts of floodinduced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category III of this requirement) must be qualitatively assessed using conservative assumptions.

(This F&O originated from SR IFSNA6)

Associated SR(s)

IFSNA6 Basis for Significance This is an enhancement required to satisfy the RG 1.200 qualification.

Possible Resolution Document a qualitative assessment of the impacts of jet impingement, pipe whip, humidity, condensation, temperature and other floodinduced mechanisms that are not explicitly modeled.

Response

The analysis was changed so that all components within a flood area are failed on initiation regardless of the equipment qualifications or other HELB mitigation features.

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SUMMARY

NOTEBOOK F&O Number F&O Details 12 All components were assumed failed if subjected to submergence. MDN000000 20100203 Appendix F documents whether components were spray vulnerable.

Discussion with the responsible analyst revealed that factors considered in spray vulnerability determinations included shielding, sealing, and equipment qualification records. However, this is not documented in the notebook.

(This F&O originated from SR IFSNA7)

Associated SR(s)

IFSNA7 IFSNB2 Basis for Significance The technical requirement is met, but the documentation could be enhanced.

Possible Resolution Document the basis for determining whether a component is "spray vulnerable" in MDN00000020100203.

Response

The following text was added to Appendix F:

During the walkdowns spray vulnerability was determined by observations of the components. For MOVs, an obvious seal with water proofing had to be observed to determine if the component was vulnerable to spray.

After the walkdowns were complete a comparison to the EQ database was done to observe if any components which were seen during the walkdowns and noted as being vulnerable to spray actual were environmentally qualified. Those components were removed from the spray analysis as well as Appendix G and Appendix H.

Tables were also added to Appendix F to show those components that are currently listed as environmentally qualified in the MAXIMO database.

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SUMMARY

NOTEBOOK F&O Number F&O Details 13 The water depth required to cause failure of doors is documented in the Appendix F walkdown sheets, but the derivation of this value is not documented.

(This F&O originated from SR IFSNA9)

Associated SR(s)

IFSNA9 IFSNB2 Basis for Significance Discussion with the responsible analyst revealed that holdup of water by doors was not credited in the final analysis. Therefore this is a documentation issue that does not affect the results.

Possible Resolution Document the basis for the water depth required to cause door failure used in Appendix F.

Response

Door failure height calculations were performed post walkdown. The failure heights were broken into two different sections. If the door was observed to be a fire door, the calculation of failure height used the HELB analysis, if the door was wire mesh or a non fire door then the failure height was assumed to be 0 feet.

The height of water necessary to fail a door is calculated based on if the height will exceed the actual height of the door. For calculations where the height of water is less than the height of the door, the following equation can be used:

2 Where p is the failure pressure and is the specific weight of water.

For those calculations where the height of water would exceed the actual height of the door, a different equation must be used. The door is now considered to be a completely submerged surface, so the following equation can be used:

2 Where p is the failure pressure, is the specific weight of water, and hdoor is the actual height of the door.

For the purposes of Appendix F, the failure pressure of all fire doors was taken from the HELB analysis, which states that a pressure of 1.5 psid will cause the doors to fail.

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SUMMARY

NOTEBOOK Using the equations above, the failure height of water is calculated to be 6.92 feet.

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SUMMARY

NOTEBOOK F&O Number F&O Details 15 Documentation of the scenario impacts needs to be strengthened in some areas. For example:

1) In MDN00000020100203 Section 9.4.3.3 for flood area 734.0A13 it is stated that "the flood frequency for those events that impact both ACAS compressors will be 1/3 of the original frequency." No basis is provided for this statement.
2) Section 7.3 makes the general statement that for the Turbine Building "flood originated in any level would propagate freely to the basement of the building without any hindrance." This same assumption was applied in partitioning Auxiliary Building area 930.0A1, but this is not documented.

(This F&O originated from SR IFSNA6)

Associated SR(s)

IFSNA6 IFSNB2 Basis for Significance The technical process is acceptable, but the documentation could be strengthened.

Possible Resolution Document assumptions pertinent to the flood scenarios such as those noted in the F&O description. This could be done in the Notebook or the FRANX database.

Response

The text quoted in the analysis has been updated. A further walkdown of the plant was done to reflect piping that could impact both the ACAS air compressors. Further walkdowns conducted in April 2011 showed there was no piping that was within twenty feet of both air compressors skids on the refuel floor. Therefore the section of the notebook addressing such issues was removed.

Additionally, the breaking apart of 690 into zones has been enhanced in the documentation and is discussed in the plant partitioning section. The documentation of the flooding scenarios and the flooding analysis now includes the partitions as part of the analysis, not as a change after review of the cutsets.

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SUMMARY

NOTEBOOK F&O Number F&O Details 16 MDN00000020100204 Section 7.3 states that a reasonableness check was performed and generally describes the factors considered. However, it would be helpful to provide tables that grouped HEPs by the relevant factors to support the conclusions reached.

(This F&O originated from SR HRG6)

Associated SR(s)

HRG6 Basis for Significance The check was performed, but the results are not documented in a way that supports verification of the conclusions.

Possible Resolution Include a table that groups the HFEs by the complexity to support the conclusion that "HFEs that are more complex have higher failure probabilities than simple actions."

Similarly, a table that groups the HEPs by the available time would support the conclusion that "HFEs with shorter time windows have higher failure probabilities due to factors including insufficient time to credit review from the STA and negative performance shaping factors (e.g., high stress levels). These factors are not included in the referenced Table 102.

Response

Tables have been added to HRA notebook section 7.3 for the complexity and time margin comparisons completed for the SQN Rev 5 PRA and HRA model update.

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SUMMARY

NOTEBOOK F&O Number F&O Details 19 The CCS system success criteria is modeled as dependent on the temperature of the ERCW system. MDN00007020100217 Appendix D documents the derivation of the probability for flag event FLG0070_ERCW_TEMP_GT_70 (FLG_0024SUMMER in the linked fault tree). However, the temperature data is presented in a graphical format rather than a tabular format. A tabular format would make it easier for a reviewer to perform a validation of the data.

(This F&O originated from SR SYA10)

Associated SR(s)

SYA10 Basis for Significance This is a documentation issue not affecting the technical quality of the model.

Possible Resolution Provide the data used to derive FLG0070_ERCW_TEMP_GT_70 in a tabular format.

Response

The data used to create the graph in the CCS notebook is 111,210 cells long. This data is not feasible to be presented as part of the system notebook.

An electronic copy will be kept with the notebook and will be available to any person wishing to review the data.

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SUMMARY

NOTEBOOK F&O Number F&O Details 112 Some sources of uncertainty which are characterized as having an impact on LERF are not analyzed using sensitivity analysis (e.g., core melt arrest in vessel and modeling of the ARFs).

In addition, Section 5.14.10 states that Although postcore damage human actions such as intentional depressurization of the RCS are modeled as realistically as possible, the uncertainty related to these actions is addressed in sensitivity studies.

The notebook references the sensitivity and uncertainty analysis notebook as the location for these studies. However, MDN00000020100209 performed the uncertainty by setting the values of HEPs in the model to their 5th and 95th percentile values. This does not address uncertainty in the value of the HFE for intentional depressurization of the RCS which is modeled as part of the U1_L2_NOTRCSDEPNOSBO basic event.

(This F&O originated from SR LEF3)

Associated SR(s)

LEF3 Basis for Significance This is a completeness issue which would enhance the analysis.

Possible Resolution Add sensitivity analyses to address these additional sources of uncertainty in the LERF results.

Response

The HFE for intentional depressurization of the RCS was changed from U1_L2_NOTRCSDEPNOSBO to HAPRZ in the model, and was included in the uncertainty study for the HFEs.

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SUMMARY

NOTEBOOK F&O Number F&O Details 113 The operator actions reflected in the event trees, and the sequence specific timing and dependencies that are traceable to the HRA for these actions are discussed in the HRA notebook. The operator actions modeled for each sequence are listed as a separate subsection in MDN00000020100201. However, it is suggested that a summary discussion of operator actions affecting the accident sequences, including a discussion of the top events impacted, be included in the AS notebook.

(This F&O originated from SR ASA4)

Associated SR(s)

ASA4 ASC2 HRG4 Basis for Significance Documentation enhancement. The links can be identified through cross reference to the system, success criteria, and HRA notebooks.

Possible Resolution Provide a cross reference for each operator action to the affected event tree top event in the Accident Sequence Notebook.

Response

The operator actions are discussed along with each of the top events for all event trees in the accident sequence notebook.

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NOTEBOOK F&O Number F&O Details 117 Calculation Type 1 is used for mission time events and Type 2 is used for basic events where the probability is based on periodic tests. The CAFTA users manual states that "it is better to use the more precise formulas of calculation types 3, 4, 5 or 6. This is especially important if you are using larger numbers (e.g., t > .05), or if you will be doing uncertainty analysis." (see CAFTA users manual Tables 62a and 62b and the text below Table 62b.)

(This F&O originated from SR QUE3)

Associated SR(s)

QUE3 Basis for Significance The current method yields a valid approximation of the basic event probabilities, but does not represent the recommended practice for CAFTA.

Possible Resolution Use the more precise formulas for the time dependent basic event probability calculations.

Response

The current model does not have any events where the lambda*t value approaches 0.05.

The models used will have their calculation types updated to 3 and 5 where appropriate when any update is to be performed.

For the current model, no events generated a random value greater than 1.0 for the uncertainty graphs.

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SUMMARY

NOTEBOOK F&O Number F&O Details 118 The system notebooks typically state that high energy line break (HELB) is considered in the Internal Flooding Analysis (e.g., ERCW Notebook Section 3.4.7.2.6). However, assumption 3.1 of the IF Notebook states that Additional failure modes; jet impingement, pipe whip, humidity, condensation, and temperatureinduced failures are outside the scope of this analysis.

(This F&O originated from SR SYC2)

Associated SR(s)

SYC2 Basis for Significance The treatment of HELB is not clearly documented in the system notebooks.

Possible Resolution Modify the statements in the system notebooks to clearly state that HELB is not treated and to provide a justification for this.

Response

The assumption was removed from the IF notebook. The HELB events are now addressed in the IF analysis.

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SUMMARY

NOTEBOOK F&O Number F&O Details 120 The SQN PRA considered Early containment failure as well as Late containment failure and basemat melt through. After containment failure, there is no additional equipment nor human action credited to mitigate the consequences.

There is also no evidence that a review was performed to determine if crediting operation of additional equipment or human actions after containment failure would reduce LERF.

(This F&O originated from SR LEC11)

Associated SR(s)

LEC11 LEC12 Basis for Significance Recommendation for meeting the requirements for Capability Category II/III.

Possible Resolution Document a review of the LERF results to determine if credit for equipment operation after containment failure or additional operator action credit would be effective in reducing LERF.

Response

There are no additional actions or equipment currently credited in the level 2 analysis to mitigate the consequences of a release after containment failure. This results in somewhat conservative results. A review has been performed to determine if crediting additional equipment or crediting additional human actions could result in a LERF reduction. An action identified during this review involves crediting manually closing the RCP seal water return outboard isolation valve following core damage in the event that it fails to close on demand. A sensitivity study was performed to determine the effect of this action using various assumed failure probabilities (see Section 12.7), although the feasibility of implementing the action has not been studied in detail.

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SUMMARY

NOTEBOOK F&O Number F&O Details 121 A thorough list of references is documented with each system notebook. However, the reference revision level is not always included (see the Diesel Generator and RPS notebooks, for example.)

(This F&O originated from SR SYA2)

Associated SR(s)

SYA2 SYC2 Basis for Significance The supporting requirement is met, but the documentation could be enhanced.

Possible Resolution Provide the applicable revision level for each reference to improves traceability of the source documents.

Response

Reference levels were left off of the references in all system notebooks consistent with the TVA practices for PRA calculations.

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NOTEBOOK F&O Number F&O Details 22 Per discussion with the SQN PRA data analyst, maintenance is generally performed on a train basis rather than across redundant components. Where redundant maintenance is permissible, e.g., the ERCW system, which has 8 pumps, the fault tree allows for the generation of cut sets that have multiple pumps in maintenance.

However, it would be helpful to document a verification that simultaneous unavailability of redundant equipment is not how work is planned.

(This F&O originated from SR SYA20)

Associated SR(s)

SYA20 SYC2 Basis for Significance Since maintenance on redundant equipment is not modeled as a planned event, documentation should be provided or referenced that describes how maintenance is planned/coordinated. This assures that any maintenance dependencies are not overlooked.

Possible Resolution Document the maintenance approach taken on redundant equipment and any impact on the PRA model.

Response

Coincident unavailability is now discussed in Section 7.4.4 of the data analysis notebook.

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NOTEBOOK F&O Number F&O Details 26 No modifications to plant design or operating practices were identified that lead to a condition where past data are no longer representative of current performance.

Thus limiting the use of old data was not required. However, for completeness, it is suggested that the data analysis document the consideration of this supporting requirement.

(This F&O originated from SR DAD8)

Associated SR(s)

DAD8 DAE2 Basis for Significance No documentation addressing this supporting requirement was identified.

Possible Resolution Document a consideration of modifications to plant design or operating practices that could lead to a condition where past data are no longer representative of current performance in the Data Analysis notebook.

Response

Section 7.2.1 was added to the data analysis notebook to address plant design changes.

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NOTEBOOK F&O Number F&O Details 27 The data analysis aligns well with the PRA Standard requirements and is generally welldocumented. Adding a 'roadmap' to the PRA Standard data SRs as was done elsewhere in the PRA documentation would enhance the performance of PRA applications, upgrades, and peer review.

(This F&O originated from SR DAE1)

Associated SR(s)

DAE1 Basis for Significance Adding a 'roadmap' to the PRA Standard data SRs would enhance the performance of PRA applications, upgrades, and peer review.

Possible Resolution In the Data Analysis notebook, add a 'roadmap' to the PRA Standard data SRs.

Response

Appendix I was added to the notebook to address the ASME/ANS standard sections.

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SUMMARY

NOTEBOOK F&O Number F&O Details 34 The success criteria description needs to include the boundary conditions such as RCS pressure. In general, it is not clear what the condition is that allows the SI pumps to operate.

(This F&O originated from SR ASA3)

Associated SR(s)

ASA3 ASB2 SCA3 Basis for Significance Documentation of required conditions permitting some equipment to operate is not provided.

Possible Resolution Provide more detailed discussion of the success criteria and mitigation system operating characteristics (e.g. pressure, flow rate) and how the conditions are achieved. For instance, the SI pump injection pressure and how the pressure is achieved in the accident sequence (i.e., by opening pressurizer or SG PORVs) should be discussed.

Response

All boundary conditions are listed in either the parameter file or the input decks electronic copies of these are available on request.

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SUMMARY

NOTEBOOK F&O Number F&O Details 310 The EOPs associated with a specific accident sequence success path are not identified in the Thermal Hydraulic Analysis or the Success Criteria Notebook. These are also not explicitly discussed in the Accident Sequence Notebook.

(This F&O originated from SR ASA5)

Associated SR(s)

ASA5 Basis for Significance There is no discussion relating the Emergency Operating Procedure (EOP) with accident progression in AS notebook.

Possible Resolution Provide discussions relating EOPs to the accident sequence and top events ordering.

Response

The EOP steps are now incorporated into the accident sequence notebook.

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NOTEBOOK Appendix A Resolution of F&Os Facts and Observations Summary - Suggestion F&Os F&O Number F&O Details 311 Use of the design basis for certain success criteria may result in conservative modeling. For example:

1) Section 4.4.3 of the TH Notebook (MDN0000002010205) discusses the MAAP 4.0.7 limitations which prevent use of MAAP for determining the number of accumulators required for Large LOCA (LLOCA) success. The use of the design basis assumption that 3 of 3 intact loop accumulators are required is likely to be conservative.
2) Section 4.4.7 of the TH Notebook (MDN0000002010205) discusses the number of lines needed for the Emergency Core Cooling System (ECCS). Based on MAAP limitations, the conclusion is that the current analyses only support ECCS flow through all intact lines. This conclusion is likely to be conservative for some sequences.

(This F&O originated from SR SCB1)

Associated SR(s)

SCB1 Basis for Significance The use of design basis success criteria for the accumulators and for the required number of injection paths may be conservative.

Possible Resolution Perform PRA specific analysis using an alternative code to determine if success can be achieved with fewer than 3 accumulators or with flow to fewer ECCS injection paths.

Response

MAAP currently is the consensus model of choice for analysis supporting the development of the PRA model. The use of other codes does not facilitate the development of the PRA model. The current success criteria of 3 of 3 CLAs will be retained in the model.

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SUMMARY

NOTEBOOK F&O Number F&O Details 322 It is stated that the impacts of the initiating event on mitigation systems are captured in the top events. However, there is no discussion of these impacts in the accident sequence notebook.

(This F&O originated from SR ASB1)

Associated SR(s)

ASB1 Basis for Significance There is no documentation found that explicitly describes the dependencies between the mitigation systems and the initiating events.

Possible Resolution Discuss the impact of initiating events on individual mitigation systems under each top event. Alternatively, provide an initiating event to mitigating system dependency matrix.

Response

An initiating event impact table was added to the Success Criteria Notebook

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SUMMARY

NOTEBOOK F&O Number F&O Details 323 The impact of the phenomenological conditions created by the accident progression is not discussed in the accident sequence notebook.

(This F&O originated from SR ASB3)

Associated SR(s)

ASB3 Basis for Significance There is no discussion of phenomenological conditions in the AS notebook.

However, the environmental conditions affecting equipment operation is captured in the system analysis notebooks.

Possible Resolution Add a discussion of the phenomenological conditions created by the accident sequence and their impact on the credited mitigation equipment.

Response

The current phenomenological conditions, initiator impact, are discussed within each system notebook.

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SUMMARY

NOTEBOOK F&O Number F&O Details 324 The intersystem dependencies are embedded in the accident sequences, but there is no explicit discussion of these dependencies.

(This F&O originated from SR ASB5)

Associated SR(s)

ASB5 Basis for Significance There is no intersystem dependency discussion in the AS notebook.

Possible Resolution

1) Add an explicit discussion of the intersystem dependency to the discussion of each accident sequence, or
2) Include a system dependency matrix in the AS Notebook to illustrate the dependencies.

Response

A system dependency matrix has been included within the SC notebook.

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SUMMARY

NOTEBOOK F&O Number F&O Details 41 Section 5.2 of the Internal Flooding notebook (MDN00000020100203) considers flood areas in the buildings of both units, and includes all common buildings. At the building level, the text discusses whether the building contains shared equipment; however, the text and tables do not indicate which specific flood areas can impact both units. It would be helpful to enhance the documentation to indicate which flood areas have multiunit impacts.

Similarly, the discussion of food sources should attempt to identify sources with multiunit impacts.

(This F&O originated from SR IFPPA3)

Associated SR(s)

IFPPA3 IFSOA2 Basis for Significance This is a suggestion since it pertains solely to enhancement of the documentation of the flood area partitioning and flood source identification process. The flood analysis itself correctly addresses multiunit impacts.

Possible Resolution Include (in the text of section 5.2 or within the tables of included areas) indication of what areas have multiunit impacts. Include similar documentation in section 6.1 for flood sources.

Response

All areas currently analyzed that contain ERCW, CCS, HPFP, RCW or any other infinite source of water are addressed in Section 5.2 The tables provided list all areas of the plant including those where there are and are not multiunit impacts.

No changes were made to the internal flooding document Section 5.2 or 6.1

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NOTEBOOK F&O Number F&O Details 46 The SQN flooding analysis has addressed some, but not all of the requirements for Category II/III. EPRI flooding data based on generic industry experience is used for flood initiating events due to pipe ruptures. Plantspecific data that might influence the pipe failure data (e.g., material condition of the fluid systems and water hammer experience) are not considered. However, a review of plantspecific maintenance induced flooding events was performed (Appendix G of the flooding notebook) and was considered in the calculation of maintenance flooding frequency. To fully meet Category II/III, an assessment should be made of plant material condition and water hammer experience) and whether plant conditions warrant any adjustments to the generic flood frequencies that are used.

(This F&O originated from SR IFEVA6)

Associated SR(s)

IFEVA6 Basis for Significance This is judged to be a suggestion, since the response to this F&O will most likely only impact documentation. Also, the analysis meets the Category I requirements, which may be sufficient for most applications Possible Resolution Review plantspecific experience pertaining to plant material condition and water hammer and document the results of the review in the flooding notebook.

Response

A review of plant specific flooding events was performed in Appendix G. This data was incorporated into the analysis for initiating event frequencies.

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NOTEBOOK F&O Number F&O Details 48 Dependency between preinitiator events was determined to not be applicable due to the large amount of time between test and maintenance events between various system trains and the use of different crews to perform each train's activities.

Dependence between preinitiators and postinitiators is also not appropriate. The rationale for not considering dependency for preinitiators seems appropriate.

However, several inconsistencies were noted in the HRA notebook documentation (MDN00000020100204) concerning the preinitiator dependency treatment.

Various HRA calculator entries for the preinitiator events (in Appendix B) indicate that dependency between events is to be considered (see for example, event SHEEMC_4). The introductory material in Appendix F contains some statements indicating preinitiator dependency will be considered, and other statements explaining why dependency between these events is not expected. These inconsistencies should be corrected.

(This F&O originated from SR HRD5)

Associated SR(s)

HRD5 Basis for Significance This is considered to be a suggestion since it pertains to correcting documentation errors. The underlying analyses themselves are correct and will not be impacted by these errors.

Possible Resolution Correct the documentation errors in the HRA notebook as noted.

Response

The documentation errors in Appendix B of the HRA notebook were corrected.

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SUMMARY

NOTEBOOK F&O Number F&O Details 49 Appropriate generic data sources appear to be used in the SN PRA, as documented in the data analysis notebook (MDN00000020100202). Component failure rates are taken primarily from NURG/CR6928 (with other sources used in cases in which data for specific component types are not available). Common cause data is obtained from recent NRC (INEL) and PWROG data sources. Offsite power recovery data is obtained from NUREG/CR6890. Component recovery is not used. Table 2 and Appendix A describe the boundaries assumed for each major component type.

The SQN PRA makes use of generic unavailability data from NUREG/CR6928 for components for which plantspecific data is unavailable (as noted in Table 8). It is assumed (see Assumption 1 in the data notebook) that all generic data is applicable to SQN; however, since this SR requires that the consistency of the SQN practices and philosophies be checked against the generic data source assumptions, additional documentation needs to be provided to better meet the requirements of this SR. It is recognized that assumption 1 is listed as an important uncertainty and is discussed in the Uncertainties notebook(MDN00000020100209). However, since the unique attributes concerning the use of generic unavailability data are not discussed, adding an additional assumption item for this issue may be appropriate.

(This F&O originated from SR DAC1)

Associated SR(s)

DAC1 DAE2 Basis for Significance This is considered to be a suggestion as it pertains primarily to a documentation enhancement. The use of generic unavailability estimates for some plant components is probably acceptable; however documentation of the basis for accepting this data as appropriate to SQN is required.

Possible Resolution Enhance the documentation in section 6.2 to better describe the acceptability of the generic estimates for SQN. Consideration should be given to specifically identifying this generic data use as an important assumption in the Uncertainties notebook (MDN00000020100209) as well. That notebook has an overall item concerning the use of generic data; however, a specific item for the use of generic unavailability data could also be added.

Response

A discussion of the component boundaries and maintenance practices was added to section 6.2.

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SUMMARY

NOTEBOOK F&O Number F&O Details 410 Failure data records are obtained from the plant's Cause Determination and Evaluation (CDE) records that are prepared by system engineers in response to failure events. The guidance for CDE development in plant procedure SPP6.6 describes bases for failures, discusses degraded conditions, and notes that Technical Specification failures or operability issues are not automatically Maintenance Rule functional failures (or PRA failures).

The CDE records are also then reviewed by the PRA staff to determine if a PRA failure has occurred. The CDEs that were used in the data analysis are included in Appendix D of the data notebook (MDN00000020100202).

Because there are several DAC SRs that specify requirements for the data collection and analysis process, it is suggested that the data analysis documentation be enhanced to specifically note these requirements and how they are met, especially since the other plant procedures do not specifically state these requirements (since the procedures are for system engineers and other nonPRA personnel).

(This F&O originated from SR DAC4)

Associated SR(s)

DAC4 DAC5 DAC11 DAC12 DAC13 DAE2 Basis for Significance This is a suggestion since it pertains to enhancing the documentation to place all of the data analysis ground rules within the data notebook for clarity.

Possible Resolution Enhance the data analysis notebook to specifically list the data collection requirements for DAC4, C5, C6, C11, C12, and C13.

Response

DAC4 Functional failures are determined based on the system engineer and maintenance rule expert panel. These determinations are outlined in SPP6.6, and are only made by a qualified individual.

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SUMMARY

NOTEBOOK DAC5 There was an identified event where multiple repeat failures occurred within the same time. Each of these events was assigned a specific CDE, however as noted in the documentation of CDE 1615 the three events were all assigned to one failure event in the PRA model.

DAC11 Unavailability is defined in the maintenance rule technical instruction TI4. The definition presented states that unavailability is only counted while at power (mode 1), additionally in the definition, unavailability is credited when the component would not be able to perform its designed function.

DAC12 The definition of the component boundaries for tracking unavailability are documented in TI4. For frontline systems only front line impacts are assigned to that system. If the ERCW header or other multisystem impact components are unavailable then the unavailability is tracked at that level.

DAC13 For all significant unavailabilities, start and finish times are accurately documented in the maintenance rule spreadsheets.