NL-15-128, Sequoyah Nuclear 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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Licensee Response/NRC Response/NRC Question Closure Id 365NRC Question Number KAB064Select Application NRC ResponseAttachment 1 Attachment 2 Response Statement Page 10 of the Calculation B87 140924 017, Revision 9, dated September 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 Date/Time 10/6/2014 6:00 PM Closure Statement Question Closure Date Notification Scott BowmanMichelle Conner Khadijah HemphillAndrew HonPage 1of 2 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Lynn MynattRay SchieleRoger ScottAdded By Kristy Bucholtz Date Added 10/6/2014 1:00 PMDate Modified Modified By Page 2of 2 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id377 NRC Question Number KAB064Select Application Licensee ResponseAttachment 1 Attachment 2 Response Statement The following information is provided concerning the Staff's 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 acceptab le 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 instrume nts, an acceptance band Page 1of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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(Ab) = 0.5% must be used fo r 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 equi pment, and readability uncertainties". Please explain how you meet the guidance contained in RIS 2006-17.
ResponseTVA 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 Date/Time 10/22/2014 2:30 PM Closure Statement Page 2of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Question Closure Date Notification Scott BowmanMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleAdded By Scott Bowman Date Added 10/22/2014 1:28 PMDate Modified Modified By Page 3of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id 389NRC Question Number KAB064Select Application NRC ResponseAttachment 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 clearl y 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 upda te 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, Ab 493 =+/-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 unnecessa ry reportability, its value will be somewhat larger than Ab, the acceptance band
around the setpoint that is ac ceptable without recalibration.Page 1of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Hence the use of Ab = +/-0.5% for allowable value allowance or tolerance is confusing and undesirable.Staff suggests the calculation be change d 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 th e 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 pum p (RCP) undervoltage loops (271A, 271B, 272A, and 272B) and 6.9kV shutdown board loss of voltage relays (27TS1A, 27TS1B, 27TS1C).4.Definitions of Westinghouse me thodology 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 definiti ons and abbreviat ions section.
Please define these and any other abbreviations that have been used but not include d 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 accur acy is negligible then it should be so stated in the calculation.If not, the two values will be different.Please clarify.Response Date/Time 12/1/2014 6:00 PM Closure Statement Question Closure Date Page 2of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Notification Scott BowmanMichelle ConnerKhadijah HemphillAndrew Hon Lynn MynattRay SchieleRoger ScottAdded By Kristy Bucholtz Date Added 12/1/2014 1:01 PMDate Modified Modified By Page 3of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id 402NRC Question Number KAB064Select Application Licensee ResponseAttachment 1 Attachment 2 Response Statement 1. Calculation 27DAT does not use the term, "AnAb"anywhere.In three different locations, the ca lculation 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 clarif ications 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.BT I-EEB-TI-28, R10 has a specific section that addresses TSTF-493 re quirements 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 re lays 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 calib ration 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 Page 1of 4 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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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 uncerta inty 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 co mputations for the Setpoint and Allowable Value.Although the Setpoi nt is specifically defined in another calculation, the summary of results on pdf page 88, of calculation 27DAT, (Sheet 49B) show s that margin exists between the Setpoint and Analytical Limi t, considering the uncertainty values (An and As), which include the larger Ab value within their computations, as shown on pdf pa ge 59 (Sheet 23A) of the calculation.In calculation 27DAT, pa ge 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 comput ations for the Setpoint and Allowable Value, and Ab 493 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 annotati on 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 fo r examples of these footnotes.
As stated in footnote (c), the me thodologies used to determine the as-found and as-left values are sp ecified 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 27 DAT, the As-Found value for the RCP Undervoltage loops (271A, 271B, 272A, and 272B) is found on page 59 of th e 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 Page 2of 4 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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6.9kV shutdown board loss of vo ltage 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 re lays 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 c ontrolled through Setpoint and Scaling Documents (SSDs).SSD s serve as the design output document to transmit the requirem ents to site organizations to ensure values assessed in the saf ety analyses and/or other design documents relative to instrument se tpoints, 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 a nnotated with requirements to evaluate setpoints found outside th e As-Found tolerances to verify the channel is functioning as requi red before returning the channel to service.Additionally, this c ondition will be entered into the Corrective Action Program.The SIs also require that an instrument channel shall be decl ared inoperable if it cannot be reset to within the As-Left tolerance.4.TVA's process for describing the method for determining the acceptability of setpoints for nuclear safety-related and Technical Specification instrumentation ch annels is governed by Branch Technical Instruction BTI-EEB-TI-28.Calculations under the scope of this instruction must follow th e 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 calc ulation is defined in this branch technical instruction. 5. The terms LAn and LAs are defi ned in BTI-EEB-TI-28, Revision 10.LAn is the Normal Loop Accuracy. LAs is the Post-Seismic Loop Accuracy.The term Afc is defi ned 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 A fc as the Acceptable As-Found - Component. 6. Calculation 27DAT defines the te rm Se on pdf page 27 (Sheet 5) Page 3of 4 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s
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as the inaccuracy following a seismic event.Page 36 of the pdf (Sheet 9) states that Se is negligib le 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 Date/Time 12/16/2014 5:00 PM Closure Statement Question Closure Date Notification Scott BowmanKristy BucholtzMichelle ConnerKhadijah HemphillAndrew Hon Lynn MynattRay SchieleAdded By Scott Bowman Date Added 12/16/2014 3:58 PMDate Modified Modified By Page 4of 4 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id403NRC Question Number KAB064Select Application NRC Question ClosureAttachment 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 Date 12/18/2014Notification Scott BowmanKristy BucholtzMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 12/18/2014 2:31 PMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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ITS NRC Questions Id174 NRC Question Number KAB065 Category TechnicalITS Section 3.3ITS Number 3.3.1 DOC Number JFD Number JFD Bases Number Page Number(s)
NRC Reviewer Supervisor Rob ElliottTechnical Branch POC Gursharan SinghConf Call Requested N NRC Question Request for additional info rmation regarding Sequoyah RCP Underfrequency Relays setpoint calculation number SQN-EEB-MS-T128-0076, Rev. 5 The above calculation wa s 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 th e calculation justifies the use of a drift value of +/-0.553 Hz.In its le tter dated June 3, 1994 ABB stated that ABB Type 81 Frequency Relay employs a very stable crystal controlled oscillator as frequency reference.It furthe r 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 licen see would be indicative of a defective relay.In note 2, the licensee states that the drift value of +/-0.553 Hz is hi ghly 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 in formation summary (RIS) 2006-17.
The deviation number selected sh ould be high enough to prevent unwanted excursions b eyond the allowable value while it should be low enough to detect potential degradation of the instrument.The licensee is requested to use a dr ift number using the guidance of Page 1of 2 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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RIS 2006-17. 2.Please note that ISA RP-67.04.02 recommends that the accuracy of measurement and test eq uipment 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 th e instrument being calibrated is highly undesirable and must be justified within the calculation.
Attach File 1 Attach File 2 Issue Date 7/15/2014Added By Kristy BucholtzDate Modified Modified By Date Added 7/15/2014 8:18 AMNotification Scott BowmanMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottPage 2of 2 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id371 NRC Question Number KAB065Select Application Licensee ResponseAttachment 1 Attachment 1 to KAB065.pdf (81KB)Attachment 2 Attachment 2 SQN-EEB-MS-T128-0076.pdf (14MB)Response Statement In response to KAB065, the RCP Unde rfrequency Relay setpoint calculation number SQN-EEB-MS-TI28-00 76, Revision 5, has been revised .Revision 7 revises the RCP unde rfrequency relay drift, Meas uring 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 us ing 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.TVA's 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 doc umented and authorized by the responsible TVA management.With the calibration require ments of the M&TE to a reference standard and the M&TE be ing at least as accurate as the underfrequency relay, which is in compliance with TVA's calibration program, the as-found values will be low enough to detect potential de gradation of the instrument.
Additionally, as a result of the increased accuracy used in calculation, SQN-EEB-MS-TI28-0076, Revision 7, the Allowable Valu e (AV) for ITS Table 3.3.1-1, Function 12 (Underfrequency RCPs), on pages 120 and 15 2 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 Date/Time 10/16/2014 2:05 PMPage 1of 2 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=371 Closure Statement Question Closure Date Notification Scott BowmanKristy BucholtzMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleAdded By Scott Bowman Date Added 10/16/2014 1:05 PMDate Modified Modified By Page 2of 2 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=371 ITS A01 ITS 3.3.1 TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRU MENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT ALLOWABLE VALUES
- 14. Deleted
- 15. Undervoltage-Reactor 5022 volts-each bus 4739 volts-each bus Coolant Pumps
- 16. Underfrequency-Reactor 56.0 Hz - each bus 55.9 Hz - each bus Coolant Pumps
- 17. Turbine Trip A. Low Trip System 45 psig 39.5 psig Pressure B. Turbine Stop Valve 1% open 1% open Closure
- 18. Safety Injection Input from ESF Not Applicable Not Applicable 19. Intermediate Range Neutron 1 x 10
-4% of RATED THERMAL 6 x 10-5% of RATED Flux - (P-6) Enable Block Source Range Reactor Trip POWER THERMAL POWER 20. Power Range Neutron Flux 10% of RATED THERMAL 12.4% of RATED (not P-10) Input to Low Power Reactor Trips Block P
-7 POWER THERMAL POWER
September 20, 2007 SEQUOYAH - UNIT 1 2-6a Amendment No. 16, 85, 136, 141, 307, 310, 316 11 12 14 15 16.a 16.e Page 18 of 47Table 3.3.1-1 LA07 A21 57.056.3 M24 ITS A01 ITS 3.3.1 TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRU MENTATION 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
- 15. Undervoltage-Reactor 5022 volts-each bus 4739 volts - each bus Coolant Pumps
- 16. Underfrequency-Reactor 56.0 Hz - each bus 55.9 Hz - each bus Coolant Pumps
- 17. Turbine Trip A. Low Trip System 45 psig 39.5 psig Pressure B. Turbine Stop Valve 1% open > 1% open Closure 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 11 12 14 14.a 14.b 15 Table 3.3.1-1 57.056.3 M24 DISCUSSION OF CHANGES ITS 3.3.1, REACTOR TRIP SYSTEM (RTS) INSTRUMENTATION Sequoyah Unit 1 and Unit 2 Page 25 of 45 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.
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.
The purpose of the Underfrequency RCP reactor trip is to ensure that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops from a major network frequency disturbance. TVA has determined that to provide adequate protection changes to the Underfrequency RCP Nominal Trip Setpoint and the Allowable Value are needed. This change was previously proposed in SQN license amendment request TVA-SQN-TS-02-01, Revision 1 (ADAMS Accession No. 042430467) but later withdrawn in TVA-SQN-TS-02-01, Revision 2 (ADAMS Accession No. ML061990303) pending resolution of issues with TSTF-493. In Revision 2 TVA stated that a new TS amendment 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.
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 RTS Instrumentation (Without Setpoint Control Program) 3.3.1 A Westinghouse STS 3.3.1 A-20 Rev. 4.0 CTS Amendment XXXSequoyah Unit 1 1 1 2Table 3.3.1-1 (page 4 of 8) Reactor Trip System Instrumentation
FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE [NOMINAL(l) TRIP SETPOINT] 13. Underfrequency RCPs 1 (g) [3] per busK SR 3.3.1.9 SR 3.3.1.10(b)(c)SR 3.3.1.16 [57.1] Hz [57.5] Hz
- 14. Steam Generator (SG) Water Level
- Low Low 1,2 [4 per SG]E SR 3.3.1.1 SR 3.3.1.7 (b)(c) SR 3.3.1.10(b)(c)SR 3.3.1.16 [30.4]% [32.3]% 15. SG Water Level
- Low 1,2 2 per SG E SR 3.3.1.1 SR 3.3.1.7 (b)(c) SR 3.3.1.10(b)(c)SR 3.3.1.16 [30.4]% [32.3]%
Coincident with Steam Flow/Feedwater Flow Mismatch 1,2 2 per SG E SR 3.3.1.1 SR 3.3.1.7 (b)(c) SR 3.3.1.10(b)(c)SR 3.3.1.16 [42.5]% full steam flow at RTP [40]% full steam flow at RTP 16. Turbine Trip
- a. Low Fluid Oil Pressure 1 (j) 3 N SR 3.3.1.10(b)(c)SR 3.3.1.15 [750] psig [800] psig b. Turbine Stop Valve Closure 1 (j) 4 N SR 3.3.1.10 SR 3.3.1.15 [1]% open [1]% open 17. Safety Injection (SI)
Input from Engineered Safety Feature Actuation System (ESFAS) 1,2 2 trains O SR 3.3.1.14 NA NA (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.
(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 t he NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveilla nce procedures (field setting) to confirm channel performance. The NTS P and the methodologies used to determine the as-found and as-left tolerances are specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference].
(g) Above the P-7 (Low Power Reactor Trips Block) interlock.
(j) Above the P-9 (Power Range Neutron Flux) interlock.
REVIEWER'S NOTE--------------------------------------------------------------------------------------
(l) Unit specific implementations may contain only Allowable Value depending on Setpoint Study methodology used by the unit.
12 13 14 15 Table 3.3-1 Table 4.3-1 INSERT 4 Note ** DOC M22 DOC M23 56.3 57.039.5 45 1 L 17 14 18 19 Table 2.2-1 Function 13 Table 2.2-1 Function 16 Table 2.2-1 Function 17 Table 2.2-1 Function 18 M h h 14 13 13 12INSERT 5 UFSAR Section 7.1.2 DOC L02 4 10 3 3 3 3 3 2 2 9 2 2 2 2 2 2 14 RTS Instrumentation (Without Setpoint Control Program) 3.3.1 A Westinghouse STS 3.3.1 A-20 Rev. 4.0 CTS Amendment XXXSequoyah Unit 2 1 1 2Table 3.3.1-1 (page 4 of 8) Reactor Trip System Instrumentation
FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE [NOMINAL(l) TRIP SETPOINT] 13. Underfrequency RCPs 1 (g) [3] per busK SR 3.3.1.9 SR 3.3.1.10(b)(c)SR 3.3.1.16 [57.1] Hz [57.5] Hz
- 14. Steam Generator (SG) Water Level
- Low Low 1,2 [4 per SG]E SR 3.3.1.1 SR 3.3.1.7 (b)(c) SR 3.3.1.10(b)(c)SR 3.3.1.16 [30.4]% [32.3]% 15. SG Water Level
- Low 1,2 2 per SG E SR 3.3.1.1 SR 3.3.1.7 (b)(c) SR 3.3.1.10(b)(c)SR 3.3.1.16 [30.4]% [32.3]%
Coincident with Steam Flow/Feedwater Flow Mismatch 1,2 2 per SG E SR 3.3.1.1 SR 3.3.1.7 (b)(c) SR 3.3.1.10(b)(c)SR 3.3.1.16 [42.5]% full steam flow at RTP [40]% full steam flow at RTP 16. Turbine Trip
- a. Low Fluid Oil Pressure 1 (j) 3 N SR 3.3.1.10(b)(c)SR 3.3.1.15 [750] psig [800] psig b. Turbine Stop Valve Closure 1 (j) 4 N SR 3.3.1.10 SR 3.3.1.15 [1]% open [1]% open 17. Safety Injection (SI)
Input from Engineered Safety Feature Actuation System (ESFAS) 1,2 2 trains O SR 3.3.1.14 NA NA (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.
(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 t he NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveilla nce procedures (field setting) to confirm channel performance. The NTS P and the methodologies used to determine the as-found and as-left tolerances are specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference].
(g) Above the P-7 (Low Power Reactor Trips Block) interlock.
(j) Above the P-9 (Power Range Neutron Flux) interlock.
REVIEWER'S NOTE--------------------------------------------------------------------------------------
(l) Unit specific implementations may contain only Allowable Value depending on Setpoint Study methodology used by the unit.
12 13 14 15 Table 3.3-1 Table 4.3-1 INSERT 4 Note ** DOC M22 DOC M23 56.3 57.039.5 45 1 L 17 14 18 19 Table 2.2-1 Function 13 Table 2.2-1 Function 16 Table 2.2-1 Function 17 Table 2.2-1 Function 18 M h h 14 13 13 12INSERT 5 UFSAR Section 7.1.2 DOC L02 4 10 3 3 3 3 3 2 2 9 2 2 2 2 2 2 14
Licensee Response/NRC Response/NRC Question Closure Id379NRC Question Number KAB065Select Application NRC Question ClosureAttachment 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 Date 11/19/2014Notification Scott BowmanMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Kristy Bucholtz Date Added 11/19/2014 6:48 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id388NRC Question Number KAB065Select Application NRC ResponseAttachment 1 Attachment 2 Response Statement KAB-065 was inadvertantly closed. Please respond to the following response:
Calculation number SQN-EE B-MS-TI28-0076, Revisi on 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 Date/Time 12/1/2014 6:00 PMClosure Statement Question Closure Date Notification Scott BowmanMichelle ConnerKhadijah HemphillAndrew HonLynn Mynatt Ray SchieleRoger ScottAdded By Kristy Bucholtz Date Added 12/1/2014 12:58 PMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=388 Licensee Response/NRC Response/NRC Question Closure Id 400NRC Question Number KAB065 Select Application Licensee ResponseAttachment 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.Attachm ent 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 th at 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, an d will be complete prior to implementation of ITS. (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 Page 1of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=400 (sheet 26) of Attachment 2.Th e 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 calc ulation 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 anal yses and/or other design documents relative to instru ment 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 Date/Time 12/16/2014 2:00 PM Closure Statement Question Closure Date Notification Scott BowmanKristy BucholtzMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleAdded By Scott BowmanPage 2of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=400 Date Added 12/16/2014 12:56 PMDate Modified Modified By Page 3of 3 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=400 Licensee Response/NRC Response/NRC Question Closure Id404NRC Question Number KAB065Select Application NRC Question ClosureAttachment 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 Date 12/18/2014Notification Scott BowmanKristy BucholtzMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 12/18/2014 2:32 PMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 1/13/201 5htt p s://members.excelservices.com/rai/index.
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ITS NRC Questions Id 190NRC Question Number KAB066Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s) NRC Reviewer Supervisor Roger PedersonTechnical Branch POC Mark Blumberg Conf Call Requested NNRC Question RAI ARCB2-1 (in response to KAB
-044)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, thatThe addition of the term "recently"associated with handling irradiated fuel in all of th e 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 (w ell 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. ML00373 4190), states, in part, thatPage 1of 3 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=190 The models, assumpti ons, 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 Ba sis Accidents at Nuclear Power Reactors," dated July 2000 (A DAMS Accession No. ML003716792), Regulatory Position 1.1 states:
The number of fuel r ods damaged during the accident should be based on a conservative analysis that considers the most limiting case.This analysis should consider pa rameters such as the weight of the dropped heavy load or weight of a dropped fuel assembly...
After reviewing the inform ation submitted by the li censee to adopt changes to the Improved Technical Specifications (ITS) (that incorp orated TSTF-51), the Nuclear Regulatory Commission (NRC) Staff is concerned that the licensee has not provided an analysi s 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 Conseq uences of Fuel Handling Accidents for the Sequoyah Nu clear 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 licensee's 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 9/30/2014Added By Khadijah HemphillDate Modified Page 2of 3 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=190 Modified By Date Added 9/30/2014 4:00 PMNotification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder GroverKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottPage 3of 3 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=190 Licensee Response/NRC Response/NRC Question Closure Id413NRC Question Number KAB066 Select Application Licensee Response Attachment 1 Attachment 1 for KAB066 12_17_2014.pdf (1MB)Attachment 2 Attachment 2 for KAB066 12_18_2014.pdf (2MB)Response Statement In response to RAI KAB066, the fo llowing information is provided.
The SQN fuel handling accide nt (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 dec ontamination factor (DCF) of 200 that is applied to the overall iodine inventory release to the pool.The SQN ITS lic ense amendment request, as submitted, does not provide a speci fic 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 change s 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 a ddition of ITS 3.9.8, the following changes will be necessary:1.The CTS markups will be revised. (Pages 232 and 233 of , Volume 14)2.The Discussion of Change s 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 , Volume 14)3.ITS 3.9.8 and the Bases for IT S 3.9.8 will be added to the submittal.(Insert 4 located after the inserts for CTS 3.9.3
Discussion of Changes in , Volume 14)Page 1of 5 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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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 , 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 irradiat ed 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,"a pplies 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 SQN's 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 , 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 , Volume 12)Page 2of 5 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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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 JF D 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 Ba ses, on pages 45 and 60 of , 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 ch anges 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 th e addition of the definition for CORE ALTERATION provide an explic it requirement that the decay time of the reactor be greater th an 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 Fu el 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 Note s," approved on December 7, 2006 (ADAMS Accessi on 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 creati ng 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 />.Page 3of 5 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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In addition, as established in th e "Radiological Consequences of Fuel Handling Accidents for the Se quoyah Nuclear Plant Unit 1 and 2, LTR-CRA-02-219 Revision 2," th e 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, "Con trol Room Emergency Ventilation System (CREVS) Actuation Instru mentation," LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS)," LCO 3.7.11, "Control Room Air-Conditioning Sy stem (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, "R esidual Heat Re moval (RHR) and Coolant Circulation-High Water Leve l," LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circ ulation-Low Water Level," and LCO 3.9.7, "Refueling Cavity Water Level," ensures that the release of fission product radioactivity fr om 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 change s associated with ITS 3.9.8 and the inclusion of the definition of CORE ALTERATIONS in ITS 1.1.See Attachment 2 for draft change s associated with ITS 3.7.13 and the Bases for ITS LCO 3.0.3.Response Date/Time 12/30/2014 9:25 PMClosure Statement Question Closure Date Notification Mark BlumbergScott BowmanKristy BucholtzMargaret ChernoffMichelle ConnerRobert ElliottRavinder GroverMatthew HardgroveKhadijah HemphillAndrew HonLynn MynattAmrit PatelRay SchielePage 4of 5 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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Added By Michelle Conner Date Added 12/30/2014 8:26 PMDate Modified 1/5/2015 9:49 AMModified By Scott BowmanPage 5of 5 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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ITS Chapter 1.0DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be: a.Analog channels - the injection of a simulated signal into the channel as clos e to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions
.b.Bistable chan nels - the injection of a simulated signal into the sensor to verifyOPERABILITY including alarm and/or trip functions.
c.Digital chann els - the injection of a simulated signal into the channel as close to the se nsorinput to the process racks as practicable to verify OPERABILITY including alarm and/ortrip 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 va lvesystem, or 2)Closed by manual valves, blind flanges, or deactivated automatic valves secure d intheir closed positions, except for valves that are open under administrative control as 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 se aling 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.
CONTROLLE D LEAKAGE 1.8 This definition has been deleted.
CORE 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 vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMIT 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 determined for each reload cycle in accordance with Specification 6.9.1.14. Unit operation within these operating limits is addressed in individual specifications. April 13, 2009 SEQUOYAH - UNIT 1 1-2 Amendment No. 12, 71, 130, 141, 155 176, 201, 203, 259, 323 A04 A04 L01 A05 A06 A07 A06 A01OPERATIONAL 5.6.3. Plant parameteror actual (COT)COT INSERT 3 A01 (COLR)CHANNEL OPERATIONAL
TEST A04 S CORE OPERATING LIMITS REPORT cycle specific parameter Page 3 of 37 Chapter 1.0 ITS DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
- 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.
- b. Bistable channels
- the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
- c. Digital channels
- the injection of a simulated signal into the channel as close to the sensor 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. b. All equipment hatches are closed and se aled. 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
1.8 This definition has been deleted. CORE 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 vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS 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 determined for each reload cycle in accordance with Specification 6.9.1.14. Unit operation within these operating limits is addressed in individual specifications.
April 13, 2009 SEQUOYAH - UNIT 2 1-2 Amendment Nos. 63, 117, 132, 146, 167, 191, 193, 250, 315 CHANNEL OPERATIONAL TEST OPERATIONAL (COT)COT or actual INSERT 3 A04 A01 L01 A04 A05 A04 A06 A07 A06 CORE OPERATING LIMITS REPORT (COLR) A01Page 20 of 37 5.6.3. Plant parameter cycle specific parameter DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 4 of 11 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 PROCESS CONTROL PROGRAM (PCP) REPORTABLE EVENT SOLIDIFICATION SOURCE CHECK
Definitions
1.1 Westinghouse
STS 1.1-2 Rev. 4.0 CTS 1 Amendment XXX SEQUOYAH UNIT 1 1.1 Definitions
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.
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.
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 those listed in [T able 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"].
- 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 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.
1.6 1.10 1.11 TSTF-490 TSTF-490 TSTF-490 INSERT 1 Definitions
1.1 Westinghouse
STS 1.1-2 Rev. 4.0 CTS 1 Amendment XXX SEQUOYAH UNIT 2 1.1 Definitions
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.
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.
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 those listed in [T able 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"].
- 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 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.
1.6 1.10 1.11 TSTF-490 TSTF-490 TSTF-490 INSERT 1 JUSTIFICATION FOR DEVIATIONS ITS 1.0, USE AND APPLICATION Sequoyah Unit 1 and 2 Page 1 of 1 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.
LIST OF ATTACHMENTS 1.ITS 3.9.1 - Boron Concentration2.ITS 3.9.2 - Unborated Water Source Isolation Valves 3.ITS 3.9.3 - Nuclear Instrumentation 4.ITS 3.9.4 - Containment Penetration s 5.ITS 3.9.5 - Residual Heat Re moval (RHR) and Coolant Circulation - High Water Level 6.ITS 3.9.6 - Residual Heat Re moval (RHR) and Coolant Circulation - Low Water L evel 7.ITS 3.9.7 - Refueling Cavity Water Level8.Relocated/Deleted Current Technical Specifications (CTS)
ATTACHMENT 8 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS
CTS 3/4.9.3, DECAY TIME
Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
REFUELING OPERATIONS 3/4 9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 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 or irradiated fuel in the reactor pressure vessel.
ACTION: 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. SURVEILLANCE REQUIREMENTS 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.
SEQUOYAH - UNIT 1 3/4 9-3 LA01 CTS 3/4.9.3 Page 1 of 2 REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 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.
ACTION: 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. SURVEILLANCE REQUIREMENT 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.
SEQUOYAH - UNIT 2 3/4 9-4 LA01 CTS 3/4.9.3 Page 2 of 2 DISCUSSION OF CHANGES CTS 3/4.9.3, DECAY TIME Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES None
MORE RESTRICTIVE CHANGES
None
RELOCATED SPECIFICATIONS
None
REMOVED DETAIL CHANGES
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
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 , t he 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.
Decay Time 3.9.8 3.9 REFUELING OPERATIONS
3.9.8 Decay Time
LCO 3.9.8 The . APPLICABILITY:
During CORE ALTERATIONS.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor subcritical for < 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. A.1 Suspend CORE ALTERATIONS. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify hours. Prior to CORE ALTERATIONS
3.9.8-1 Decay Time 3.9.8 3.9 REFUELING OPERATIONS
3.9.8 Decay Time
LCO 3.9.8 The . APPLICABILITY:
During CORE ALTERATIONS.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor subcritical for < 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. A.1 Suspend CORE ALTERATIONS. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify hours. Prior to CORE ALTERATIONS
3.9.8-1 JUSTIFICATION FOR DEVIATIONS ITS 3.9.8, DECAY TIME Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.None.
SEQUOYAH - UNIT 1 B 3.9.8-1 Revision XXX 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 3.9.8-2 Revision XXX 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 B 3.9.8-1 Revision XXX 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 B 3.9.8-2 Revision XXX 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.
JUSTIFICATION FOR DEVIATIONS ITS 3.9.8 BASES, DECAY TIME Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.None DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.9.3, DECAY TIME Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.
Page 1 of 22 REFUELING OPERATIONS 3/4.9.11 SPENT FUEL PIT WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY
- Whenever irradiated fuel assemblies are in the spent fuel pit.
ACTION: 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 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. SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the sp ent 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. SEQUOYAH - UNIT 1 3/4 9-11 ITS 3.7.13ITS LCO 3.7.13 Applicabilit y During movement of irradiated fuel assemblies in the spent fuel pool.
ACTION A ACTION A Note in accordance with the Surveillance Frequency Control Program SR 3.7.13.1 A01 LA01 L01 L02 POOL pool A01 A01 L01 immediatelyirradiated A02 L02 L01 Page 12 of 22 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL-SPENT FUEL PI T LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY
- Whenever irradiated fuel assemblies are in the spent fuel pit.
ACTION: 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 leve l 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 />. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 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.
SEQUOYAH - UNIT 2 3/4 9-13 ITS 3.7.13ITS LCO 3.7.13 Applicabilit y During movement of irradiated fuel assemblies in the spent fuel pool.
ACTION A ACTION A Note in accordance with the Surveillance Frequency Control Program SR 3.7.13.1 A01 LA01 L01 L02 POOL pool A01 A01 L01immediately irradiated A02 L02 L01 DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 2 of 3 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 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.
DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 3 of 3 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.
Fuel Storage Pool Water Level 3.7.15 Westinghouse STS 3.7.15-1 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX 13 13 1 3CTS Spent 13.7 PLANT SYSTEMS 3.7.15 Fuel Storage Pool Water Level LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water level not within limit. A.1 --------------NOTE-------------- LCO 3.0.3 is not applicable. -------------------------------------
Suspend movement of irradiated fuel assemblies in the fuel storage pool. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.
[ 7 days OR In accordance
with the Surveillance
Frequency Control Program
] APPLICABILIT Y ACTION 13 13 13 SR 4.9.11 1 2 13.9.11 1 Spent spent spent 1spent Spent 1 1spent INSERT 1 AND A.2 Restore the spent fuel pool water level to within limit.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Fuel Storage Pool Water Level 3.7.15 Westinghouse STS 3.7.15-1 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX 13 13 1 3CTS Spent 13.7 PLANT SYSTEMS
3.7.15 Fuel Storage Pool Water Level
LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
A. Fuel storage pool water level not within limit.
A.1 --------------NOTE--------------
LCO 3.0.3 is not applicable.
Suspend movement of irradiated fuel assemblies in
the fuel storage pool.
Immediately
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.
[ 7 days OR In accordance
with the Surveillance
Frequency Control Program
] APPLICABILIT Y ACTION 13 13 13 SR 4.9.11 1 2 13.9.11 1 Spent spent spent 1spent Spent 1 1spent INSERT 1 AND A.2 Restore the spent fuel pool water level to within limit.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> JUSTIFICATION FOR DEVIATIONS ITS 3.7.13, SPENT FUEL STORAGE POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 1 of 1 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 Ai r 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.
Fuel Storage Pool Water Level B 3.7.15 Westinghouse STS B 3.7.15-2 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1 13 1 13 Spent 2BASES APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does
not apply.
When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the
fuel storage pool water level is lower than the required level, the 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.
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 assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.
15.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event 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.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.
]
During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.
4 5 13 1spent spent spent spent 1 1 1 1 INSERT 2 With the 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.
Fuel Storage Pool Water Level B 3.7.15 Westinghouse STS B 3.7.15-2 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1 13 1 13 Spent 2BASES APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does
not apply.
When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the
fuel storage pool water level is lower than the required level, the 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.
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 assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.
15.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event 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.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.
]
During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.
4 5 13 1spent spent spent spent 1 1 1 1 INSERT 2 With the 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.
JUSTIFICATION FOR DEVIATIONS ITS 3.7.13 BASES, FUEL STORAGE POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 1 of 1 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.
LCO Applicability B 3.0 Westinghouse STS B 3.0-4 Rev. 4.0 1Revision XXX Sequoyah Unit 1 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.
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 of this is in LCO 3.7.
15 , "Fuel Storage Pool Water Level." LCO 3.7.
15 has an Applicability of "During movement of irradiated fuel assemblies in
the fuel storage pool." Therefore, this LCO can be applicable in any or all 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 placing the unit in a shutdown condition. The Required Action of LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the 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.
13 13 13 Spent 1spent spent LCO Applicability B 3.0 Westinghouse STS B 3.0-4 Rev. 4.0 1Revision XXX Sequoyah Unit 2 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.
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 of this is in LCO 3.7.
15 , "Fuel Storage Pool Water Level." LCO 3.7.
15 has an Applicability of "During movement of irradiated fuel assemblies in
the fuel storage pool." Therefore, this LCO can be applicable in any or all 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 placing the unit in a shutdown condition. The Required Action of LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the 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.
13 13 13 Spent 1spent spent Licensee Response/NRC Response/NRC Question Closure Id419 NRC Question Number KAB066 Select Application Licensee Response Attachment 1 Attachment 1 supplement for KAB066 01_12_2015 -Copy.pdf (2MB)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 SQN's 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 markup s 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)Page 1of 2 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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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 Date/Time 1/14/2015 6:25 AMClosure Statement Question Closure Date Notification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder Grover Khadijah HemphillAndrew Hon Lynn MynattRay Schiele Added By Scott Bowman Date Added 1/14/2015 5:24 AMDate Modified Modified By Page 2of 2 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=41 9
Page 1 of 22 REFUELING OPERATIONS 3/4.9.11 SPENT FUEL PIT WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY
- Whenever irradiated fuel assemblies are in the spent fuel pit.
ACTION: 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 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. SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the sp ent 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. SEQUOYAH - UNIT 1 3/4 9-11 ITS 3.7.13ITS LCO 3.7.13 Applicabilit y During movement of irradiated fuel assemblies in the spent fuel pool.
ACTION A ACTION A Note in accordance with the Surveillance Frequency Control Program SR 3.7.13.1 A01 LA01 L01 L02 POOL pool A01 A01 L01 immediatelyirradiated A02 L02 L01 Page 12 of 22 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL-SPENT FUEL PI T LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY
- Whenever irradiated fuel assemblies are in the spent fuel pit.
ACTION: 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 leve l 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 />. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 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.
SEQUOYAH - UNIT 2 3/4 9-13 ITS 3.7.13ITS LCO 3.7.13 Applicabilit y During movement of irradiated fuel assemblies in the spent fuel pool.
ACTION A ACTION A Note in accordance with the Surveillance Frequency Control Program SR 3.7.13.1 A01 LA01 L01 L02 POOL pool A01 A01 L01immediately irradiated A02 L02 L01 DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 2 of 3 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 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.
DISCUSSION OF CHANGES ITS 3.7.13, SPENT FUEL POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 3 of 3 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.
Fuel Storage Pool Water Level 3.7.15 Westinghouse STS 3.7.15-1 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX 13 13 1 3CTS Spent 13.7 PLANT SYSTEMS 3.7.15 Fuel Storage Pool Water Level LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water level not within limit. A.1 --------------NOTE-------------- LCO 3.0.3 is not applicable. -------------------------------------
Suspend movement of irradiated fuel assemblies in the fuel storage pool. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.
[ 7 days OR In accordance
with the Surveillance
Frequency Control Program
] APPLICABILIT Y ACTION 13 13 13 SR 4.9.11 1 2 13.9.11 1 Spent spent spent 1spent Spent 1 1spent INSERT 1 AND A.2 Res tore spent fuel p ool water level to within limit.
4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> s Fuel Storage Pool Water Level 3.7.15 Westinghouse STS 3.7.15-1 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX 13 13 1 3CTS Spent 13.7 PLANT SYSTEMS
3.7.15 Fuel Storage Pool Water Level
LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
A. Fuel storage pool water level not within limit.
A.1 --------------NOTE--------------
LCO 3.0.3 is not applicable.
Suspend movement of irradiated fuel assemblies in
the fuel storage pool.
Immediately
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.
[ 7 days OR In accordance
with the Surveillance
Frequency Control Program
] APPLICABILIT Y ACTION 13 13 13 SR 4.9.11 1 2 13.9.11 1 Spent spent spent 1spent Spent 1 1spent INSERT 1 AND A.2 Restore spent fuel pool water level to within limit.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> JUSTIFICATION FOR DEVIATIONS ITS 3.7.13, SPENT FUEL STORAGE POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 1 of 1 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 Ai r 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.
Fuel Storage Pool Water Level B 3.7.15 Westinghouse STS B 3.7.15-2 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1 13 1 13 Spent 2 BASES APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists. ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the
fuel storage pool water level is lower than the required level, the 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. 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 assemblies is not sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.7.
15.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event 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.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.
] During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.
4 5 13 1spent spent spent spent 1 1 1 1 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 Com pletion 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.
Fuel Storage Pool Water Level B 3.7.15 Westinghouse STS B 3.7.15-2 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1 13 1 13 Spent 2 BASES APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists. ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the
fuel storage pool water level is lower than the required level, the 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. 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 assemblies is not sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.7.
15.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event 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.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.
] During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.
4 5 13 1spent spent spent spent 1 1 1 1 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 Com pletion 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.
JUSTIFICATION FOR DEVIATIONS ITS 3.7.13 BASES, FUEL STORAGE POOL WATER LEVEL Sequoyah Unit 1 and Unit 2 Page 1 of 1 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.
LCO Applicability B 3.0 Westinghouse STS B 3.0-4 Rev. 4.0 1Revision XXX Sequoyah Unit 1 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.
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 of this is in LCO 3.7.
15 , "Fuel Storage Pool Water Level." LCO 3.7.
15 has an Applicability of "During movement of irradiated fuel assemblies in
the fuel storage pool." Therefore, this LCO can be applicable in any or all 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 placing the unit in a shutdown condition. The Required Action of LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the 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.
13 13 13 Spent 1spent spent LCO Applicability B 3.0 Westinghouse STS B 3.0-4 Rev. 4.0 1Revision XXX Sequoyah Unit 2 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.
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 of this is in LCO 3.7.
15 , "Fuel Storage Pool Water Level." LCO 3.7.
15 has an Applicability of "During movement of irradiated fuel assemblies in
the fuel storage pool." Therefore, this LCO can be applicable in any or all 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 placing the unit in a shutdown condition. The Required Action of LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the 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.
13 13 13 Spent 1spent spent Licensee Response/NRC Response/NRC Question Closure Id437NRC Question Number KAB066Select Application NRC Question ClosureAttachment 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 Date 4/23/2015Notification Mark BlumbergScott BowmanMargaret ChernoffMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 4/23/2015 11:26 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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ITS NRC Questions Id 191NRC Question Number KAB067Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s) NRC Reviewer Supervisor Roger PedersonTechnical Branch POC Mark Blumberg Conf Call Requested NNRC Question RAI ARCB/SCVB2-2 (in response to KAB
-044)Calculation LTR-CRA 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 convec tion 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 volu me 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 Page 1of 2 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=191 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 9/30/2014Added By Khadijah HemphillDate Modified Modified By Date Added 9/30/2014 4:02 PMNotification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder GroverKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottPage 2of 2 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=191 Licensee Response/NRC Response/NRC Question Closure Id 405NRC Question Number KAB067Select Application Licensee ResponseAttachment 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.At tachment 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 th at 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 th e water pool is assumed to be released linearly from the pool to the environment within two hours.No credit is taken for mi xing in the containment volume.Response Date/Time 12/18/2014 4:00 PM Closure Statement Question Closure Date Notification Mark BlumbergScott BowmanKristy BucholtzMargaret ChernoffMichelle ConnerRobert ElliottKhadijah HemphillAndrew Hon Lynn MynattRay SchieleAdded By Michelle Conner Date Added 12/18/2014 3:04 PMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id438NRC Question Number KAB067Select Application NRC Question ClosureAttachment 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 Date 4/23/2015Notification Mark BlumbergScott BowmanMargaret ChernoffMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 4/23/2015 11:27 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 04/27/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=438 ITS NRC Questions Id192 NRC Question Number KAB068 Category TechnicalITS Section ASTITS Number DOC Number JFD Number JFD Bases Number Page Number(s)
NRC Reviewer Supervisor Roger PedersonTechnical Branch POC Mark BlumbergConf Call Requested N NRC Question RAI ARCB2-3 (in response to KAB
-044)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 prov ided is given below:Changes are made (additi ons, 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 c onflicts with the staff's assessment of Sequoyah's 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 staff's concern, or replace the text proposed to be removed.
Attach File 1 Attach File 2 Issue Date 9/30/2014Page 1of 2 Sequoyah ITS Conversion Databas e 3/16/201 5htt p s://members.excelservices.com/rai/index.
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Added By Khadijah HemphillDate Modified Modified By Date Added 9/30/2014 4:03 PMNotification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder Grover Khadijah HemphillAndrew Hon Lynn MynattRay SchieleRoger ScottPage 2of 2 Sequoyah ITS Conversion Databas e 3/16/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id380 NRC Question Number KAB068Select Application Licensee ResponseAttachment 1 Attachment 1 for RAI KAB068.pdf (31KB)Attachment 2 Response Statement In response to KAB068, the ITS 3.9.4 Bases Applicable Safety Analyses Section, on pages 124 and 135 of
, Volume 14, w ill 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 te xt will be retained.See Attachment 1 for the draft revised ITS 3.9.4 Bases.Response Date/Time 11/24/2014 4:55 AMClosure Statement Question Closure Date Notification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder GroverKhadijah HemphillAndrew HonLynn Mynatt Ray SchieleAdded By Scott Bowman Date Added 11/24/2014 3:55 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 3/16/201 5htt p s://members.excelservices.com/rai/index.
p h p?re q uestT yp e=areaItemPrint&itemId=380 Containment Penetrations B 3.9.4 Westinghouse STS B 3.9.4-2 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 2BASES
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 su bsystem, 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, 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.
]
The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent. 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 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 accident is a postulated event that involves damage to irradiated fuel (Ref. 2). Fuel handling accidents, analyzed in Reference 3, include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The requirements of LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum 2 1 1INSERT 1 (either open or closed) 2 2 Containment Penetrations B 3.9.4 Westinghouse STS B 3.9.4-2 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 2BASES
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 su bsystem, 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, 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.
]
The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent. 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 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 accident is a postulated event that involves damage to irradiated fuel (Ref. 2). Fuel handling accidents, analyzed in Reference 3, include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The requirements of LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum 2 1 1INSERT 1 (either open or closed) 2 2 Licensee Response/NRC Response/NRC Question Closure Id406NRC Question Number KAB068Select Application NRC Question ClosureAttachment 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 Date 12/19/2014Notification Mark BlumbergScott BowmanMargaret ChernoffMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 12/19/2014 7:21 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 3/16/201 5htt p s://members.excelservices.com/rai/index.
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ITS NRC Questions Id 194NRC Question Number KAB070Category TechnicalITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s) NRC Reviewer Supervisor Roger PedersonTechnical Branch POC Mark Blumberg Conf Call Requested NNRC Question RAI ARCB2-5/SCVB2-5 (in response to KAB
-044)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 [contr ol room envelope] integrity should be performed while the CRE, its a ssociated ventilation systems, and the ventilation systems located in, traversing, or serving areas adjacent to the CRE are function ing in a manner that reflects CRE inleakage when these ventilation syst ems are operati ng in response to a particular challenge. and, In addition to the ab ove, CRE testing should be performed when changes are made to the structures, systems, components, and procedures that could im pact 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 limiti ng condition. Test ing should be Page 1of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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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 HemphillDate Modified Modified By Date Added 9/30/2014 4:06 PMNotification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder GroverKhadijah HemphillAndrew Hon Lynn MynattRay SchieleRoger ScottPage 2of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id 416NRC Question Number KAB070 Select Application Licensee Response Attachment 1 KAB070 Attachment 1.pdf (188KB)Attachment 2 Response Statement An inoperable Auxiliary Building Gas Treatment System (ABGTS) or inoperable ABGTS actuation in strumentation 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 ac tuation instrumentation are not required by Technical Specificati ons 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 a nd 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, "Param eters Used in Fuel Handling Accident Analyses,"51 cfm unfilt ered 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 C onsequences 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, determi ning 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 tra cer gas test was 22 +/-9 cfm for Page 1of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Train A and 15 +/-7 for Train B; i.
e., a 39% margin and a 50% margin for Train A and Train B, respectivel y, 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 freque ncies 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 bounda ry 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 ) are not in th e auxiliary building secondary containment enclosure and therefor e, are not affected by ABGTS operation.As such, an inoperabl e ABGTS or inoperable ABGTS actuation instrumentation will not affect operation of the CREVS or alter the MCR unfiltered inleak age assumptions in any SQN DBA analysis, including the FHA.Response Date/Time 1/4/2015 8:20 PMClosure Statement Question Closure Date Notification Mark BlumbergScott BowmanKristy BucholtzMargaret ChernoffMichelle ConnerRobert ElliottMatthew HammKhadijah HemphillAndrew HonLynn MynattRay Schiele Added By Michelle ConnerDate Added 1/4/2015 7:20 PMDate Modified Modified By Page 2of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id442NRC Question Number KAB070Select Application NRC Question ClosureAttachment 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 Date 4/29/2015Notification Mark BlumbergMargaret ChernoffMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 4/29/2015 10:06 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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ITS NRC Questions Id 195NRC Question Number KAB071Category Technical ITS Section AST ITS Number DOC Number JFD Number JFD Bases Number Page Number(s) NRC Reviewer Supervisor Roger PedersenTechnical Branch POC Mark Blumberg Conf Call Requested NNRC Question RAI ARCB2-7 (in response to KAB
-044)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 219, Revision 0, assumed that all releases from a FHA within containment would be through the containment equipment door after 30 seconds.Cal culation 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 af ter 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 c ontainment 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 revi sed X/Q value.Attach File 1 Attach File 2 Page 1of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Issue Date 9/30/2014Added By Khadijah HemphillDate Modified Modified By Date Added 9/30/2014 4:07 PMNotification Mark BlumbergScott BowmanKristy BucholtzMichelle ConnerRavinder GroverKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottPage 2of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id 417NRC Question Number KAB071 Select Application Licensee ResponseAttachment 1 Attachment 2 Response Statement TVA provided a revised limiting atmospheric dispersion factor, /Q value, 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," da ted 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 provid es 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 poin t source release locations, the AB vent stack is the limi ting location for th e calculation of MCR doses based on its proximity to the MCR air intake locations.Therefore, any release fr om containment would be bounded by the release from the AB vent stack.The revised /Q value of 2.56E-3 sec/m 3 is the limiting atmospheric dispersion valu e for calculating the radiological consequences to personnel in the MCR following a design basis FHA at Sequoyah Nuclear Plan t, Units 1 and 2.Response Date/Time 1/4/2015 8:30 PM Closure Statement Question Closure Date Notification Mark BlumbergScott BowmanKristy BucholtzMargaret ChernoffMichelle ConnerRobert ElliottPage 1of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Khadijah HemphillAndrew HonLynn MynattRay SchieleAdded By Michelle Conner Date Added 1/4/2015 7:31 PMDate Modified Modified By Page 2of 2 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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Licensee Response/NRC Response/NRC Question Closure Id447NRC Question Number KAB071Select Application NRC Question ClosureAttachment 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 Date 5/6/2015Notification Mark BlumbergMargaret ChernoffMichelle ConnerKhadijah HemphillAndrew HonLynn MynattRay SchieleRoger ScottAdded By Khadijah Hemphill Date Added 5/6/2015 4:32 PMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 05/07/201 5htt p s://members.excelservices.com/rai/index.
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ITS NRC Questions Id 6NRC Question Number KNH-001 Category TechnicalITS Section TSTF-425 - PRAITS Number DOC Number JFD Number JFD Bases Number Page Number(s) NRC Reviewer Supervisor Rob ElliottTechnical Branch POC Jonathan EvansConf Call Requested NNRC 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 HemphillDate Modified Modified By Date Added 2/4/2014 4:04 PMNotification Michelle Conner Khadijah HemphillRay SchieleGerald WaigPage 1of 1 Sequoyah ITS Conversion Databas e 8/21/201 4htt p s://members.excelservices.com/rai/index.
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6 Licensee Response/NRC Response/NRC Question Closure Id 5NRC Question Number KNH-001 Select Application Licensee ResponseAttachment 1 Cover Letter from Westinghouse LTR-RAM-II-11-010.pdf (94KB)Attachment 2 Appendix A NUC-SQN-MEB-MDN-000-2010-0200 REV 1.pdf (485KB)Response Statement 0 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 Date/Time 2/5/2014 12:30 PMClosure Statement Question Closure Date Notification Scott BowmanMichelle ConnerRobert ElliottKhadijah HemphillLynn MynattLisa RegnerRay SchieleRoger ScottGerald Waig Added By Scott BowmanDate Added 2/5/2014 11:24 AMDate Modified Modified By Page 1of 1 Sequoyah ITS Conversion Databas e 8/21/201 4htt p s://members.excelservices.com/rai/index.
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5 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 Nucl ear 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 Nu clear 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
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PLANT PROBABILISTIC RISK ASSESSMENT -
SUMMARY
NOTEBOOK Appendix A - Resolution of F&Os Finding Level F&Os F&ONumberF&ODetails1 4MDN 000000 2010 0203doesnotdocumentanassessmentoftheimpactoffloodingeventsonexistingHFEscarriedoverfromtheinternaleventsscenariousedtorepresentthefloodingevent.(ThisF&OoriginatedfromSRIFQU A6)AssociatedSR(s)IFQU A6BasisforSignificanceAlthoughthisisadocumentationissue,itisimportanttoanunderstandingoftheresultsandtoshowthatthetechnicalelementissatisfied.PossibleResolutionDocumentaprocessforassessmentoftheimpactofthefloodscenariosonexistingHFEsfromtheinternaleventsPRAsequencesusedtorepresentthefloodscenarios.EPRI1019194Section7.3describesthetypesofHFEadjustmentstobeconsidered.ResponseToaddresshumanactionsandtheirmodificationduetofloodingeventsSection9.3wasaddedtothedocument.Section9.3addressesthechangestothehumanactionsinthemodelbyaccountingfor:1.HumanactionsthatareinfluencedbyHRAactions,theseareeventsthatoccurwithinanhouroffloodinitiation.2.Humanactionsthatarefaileddueflooding.
Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-2
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PLANT PROBABILISTIC RISK ASSESSMENT -
SUMMARY
NOTEBOOK F&ONumberF&ODetails1 7DependencyanalysiswasperformedforthepostinitiatorHEPsusingtheEPRIHRACalculator.However,severalissueswereidentifiedincluding:1) Useofthesamecuefortwoactionscanresultinconservativedependencyvalues.Forexample,theuseofthesamecueforactionsHARR1andAFWOP3resultedincompletedependencybetweentheactions.However,reviewofthecuesindicatedthatthecueforAFWOP3shouldbedifferentthanthatforHARR1.2) Inconsistententryofthetiminginformationcreatesresultsthatmayappearinvalid.Forexample,thetimingentriesforactionsHARR2andAFWOP3makeitappearthatcoredamageasaresultoffailureofHARR2wouldoccurbeforethecueforAFWOP3isreceived.DiscussionrevealedthattheTswforHARR2isbasedonthetimeatwhichtheRWSTwouldemptyratherthancoredamageasstatedintheHRACalculator.3) InclusionofscreeningHFEsinthedependencyanalysiscanresultinerrors.ThescreeningHEPsdonothaveinformationthatisnecessaryforthedependencyanalysis(e.g.,timinginputs).Thiscanresultinthewrongeventbeingtreatedastheindependenteventinthecombination.Forexample,reviewofdependencycombination41showsthatthedependencyanalysistreatsHACD1asthefirstorindependentHFEinthecombinationandAFWOP5asfollowingHACD1.ThisresultsinajointHEPof1.0basedoncompletedependency.However,thedescriptionofHFEHACD1,"PerformcooldownwithmainfeedwaterfollowingAFWfailure,"indicatesthatAFWOP5shouldbethefirstevent.ThiswouldresultinajointHEPof2.9E 03.4) ThedependencylevelofthecognitiverecoverieswerenotenteredintheHRACalculatordatabaseforthepostinitiators.Thisrequiresmanualentrybytheanalystanddoesnotdefaulttotherecommendeddependencelevel.FailuretoenterthisinformationmayunderestimateoroverestimatetheHEPdependingontheapplicabledependencelevel.SomeoftheseitemswerecorrectedduringthereviewbuttheyaredocumentedinanF&Oduetotheneedtoevaluatetheextentofthecondition.(ThisF&OoriginatedfromSRHR G7)AssociatedSR(s)HR G7BasisforSignificanceIncorrectassignmentofcues,timing,andresourceentriescanresultinincorrectdependencyanalysisresults.PossibleResolution Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-3
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PLANT PROBABILISTIC RISK ASSESSMENT -
SUMMARY
NOTEBOOK
- 1) Reviewthecues,timingassumptions,andresourcerequirementsforsignificantHEPstoensurethatthefactorsarecorrectlyassessedinthedependencyanalysis.2) ForcombinationswherethetimingindicatesTswisreachedforthefirstactionbeforethecueforthesecondisreceived,documentthebasisforacceptabilityofthedependentcombination.3) EnsureallinformationaffectingthedependencyanalysisisenteredintotheHRACalculatorforthescreeningHFEstoensuretheyaretreatedcorrectlyinthedependencyanalysis.4) EnsurethatthedependencelevelbetweencognitiveactionsandapplicablerecoveriesissetintheHRACalculatordatabase.Response1. CueforAFWOP3hasbeenupdatedtocorrectcue.Reviewhasbeenperformedforallremainingactionstodetermineifanyadditionalcuesneedtobeupdated.ThisreviewverifiedtheaccuracyofHRAcuesandupdatedsixoftheidentifiedcues.2. TheendpointforTswisanirreversibledamagestate.ForHARR2,thisirreversibledamagestateisthelossofallECCSpumpswhentheRWSTisdepletedandautoswaphasfailed.Thisisthecorrectirreversibledamagestateastheoperatordoesnothaveuntilcoredamagetoperformthatactionifthepumpsfailwhentheirsuctionsourcerunsdry.Thedependencyanalysiswasreviewedforoverlappingtimeframes.3. ScreeningvalueHEPswereremovedfromthedatabaseiftherevaluesweresetto1.0.TheHEPsthatwereoriginallyinthemodelwerenolongerrequiredandweredeletedfromthefaulttree.4. ThishasbeencorrectedforalloftheactionsintheSQNHRA.
Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-4
Subject:
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SUMMARY
NOTEBOOK F&ONumberF&ODetails1 8MDN 000000 2010 0203Section9.5onlyaddressesquantificationandresultsforCDF.ThereisnodiscussionofLERFforthefloodingscenariosordocumentationindicatingthatthefloodscenarioswerereviewedtodetermineiftheywouldhaveanimpactontheLevel2CETs.ThelinkedfaulttreemodelshouldhavethecapabilitytoproduceLERFresults,butthishadnotbeendoneatthetimeofthereview.Inaddition,therewasnodiscussionintheLevel2Notebook(MDN000 000 2010 0206)thatindicatestheresultsincludetheinternalfloodscenarios.(ThisF&OoriginatedfromSRIFQU A10)AssociatedSR(s)IFQU A10BasisforSignificanceNoLERFresultsforinternalfloodingscenarioswasprovidedforreview.PossibleResolution1) DocumentareviewofthetopeventsintheLevel2modeltoconfirmthattherearenouniquefloodingimpactsthataffecttheCETs.2) DocumenttheLERFresultsfortheinternalfloodscenariossimilartotheresultsforotherinitiatorsinSection11ofMDN 000 000 2010 0206.ThiscanbedoneinthefloodnotebookortheLevel2notebook,but,ifdoneintheLevel2notebookthisshouldbereferencedintheInternalFloodingAnalysisnotebook.ResponseTheinternalfloodingcalculationwasrevisedtoaddSection10(ResultsAnalysisforLargeEarlyReleaseFrequency).Section10.1addressestheeighteenquestionsconcerningLERFandtheirimpact.Section10.3and10.4addresstheLERFresultsduetofloodingToaddresstheadditionalinformationthefollowingAppendiceswereaddedtothemodel:AppendixQSignificantCutsetReviewforLargeEarlyReleaseAppendixRNonSignificantCutsetReviewforLargeEarlyReleaseAppendixSImportanceReportsforLargeEarlyRelease Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-5
Subject:
PLANT PROBABILISTIC RISK ASSESSMENT -
SUMMARY
NOTEBOOK F&ONumberF&ODetails1 10MDN 000000 2010 0206Section5.6notesthatcreditwastakenforscrubbingofreleasesfromarupturedSG.However,thetechnicaljustificationforthiscreditneedstobestrengthened.ThecurrentbasiscomparesthezeropowercollapsedleveltothetopoftheSGtubes.However,ES 3.1,Post SGTRCooldownUsingBackfillallowsthelevelintherupturedSGtobebetween20%narrowrangeand75%narrowrangeduringthecooldown(Step7).TheexpectedlevelsduringSGTRrecoveryshouldbeusedtojustifythescrubbingcredit.ItalsoappearsthattheanalysisimplicitlyassumesthatifFWwillbeappliedtotherupturedSGifFWisavailable.NoconsiderationofoperatorfailuretoprovideFWflowtotherupturedgeneratorisincludedintheanalysis.(ThisF&OoriginatedfromSRLE C4)AssociatedSR(s)LE C4LE C13LE E3BasisforSignificanceThetechnicalbasisofthecreditforscrubbingofSGTRreleasesdoesnotconsiderthelevelsallowedintheEOPs.PossibleResolutionRevisethejustificationinMDN 000000 2010 0206Section5.6toincludeconsiderationoftheSGlevelsmaintainedduringrecoveryusingtheapplicableEOPs.ResponseThedocumentationhasbeenupdatedtoincludeadiscussionofthewaterlevelsabovethesteamgeneratortubesduringtuberupturerecoveryactions.Thesewaterlevels(between4.7and9.8feet)shouldbesufficienttotakecreditforfissionproductscrubbing.Thisanalysisassumesthattheoperatorissuccessfulinprovidingfeedwaterflowtotherupturedsteamgenerator.
Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-6
Subject:
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SUMMARY
NOTEBOOK F&ONumberF&ODetails1 11ThetotalLERFiscomparedwithotherWestinghouse4 loopplantsandwithotherIceCondenserplants.However,thereisnocomparisonatthelevelofsignificantcontributorsorplantdamagestates.Withoutthecontributorinformation,itisnotreallypossibletodeterminehowsimilartheLERFresultsaretootherplants.(ThisF&OoriginatedfromSRLEF2)AssociatedSR(s)LE F2BasisforSignificanceThereisnoreviewofthecontributorstoLERFwiththeresultsforsimilarplantstoensurethatplant specificmodelingchoiceshavenotskewedtheresults.PossibleResolutionDocumentacomparisonoftheLERFresultstoplantsofsimilardesignatthesignificantcontributorandPDSlevels(similartoTables11 1and11 3).ResponseThedocumentationhasbeenupdatedtoincludecomparisonsbyinitiatingeventforseveralotherPWRsinTable11 7.
Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-7
Subject:
PLANT PROBABILISTIC RISK ASSESSMENT -
SUMMARY
NOTEBOOK F&ONumberF&ODetails1 14Demanddataisobtaineddirectlyfromtheplantprocesscomputerformostcomponents,asdescribedinSection7.3ofthedatanotebook(MDN 000 000 2010 0202).Thestatuschangeinformationfromthecomputerisfilteredandusedtodeterminethenumberofdemands.Theuseofautomaticdatacollection,however,meansthatstartandruneventsthatoccurinallmodesofoperationareincluded.Inaddition,postmaintenanceteststartsarealsoincludedinthedataset.Thisisidentifiedasasourceofuncertaintyinthesensitivitiesanduncertaintiesnotebook(MDN000 000 2010 0209)andaspecificsetofsensitivitystudieswereperformedthatassumedthatvariousnumbersofsuccessfulstartswereinvalid.TheresultsshowthattheimpactonCDFisrelativelysmall,unlessthenumberofsuccessfulstartsisoverestimatedbyalargeamount.However,thisSRisexplicitinitsrequirementtonotcountpostmaintenancetestevents.(ThisF&OoriginatedfromSRDA C6)AssociatedSR(s)DA C6BasisforSignificanceThisisconsideredtobeafindingsinceaspecifictechnicalrequirementoftheSRisnotmet.PossibleResolutionTocomplywiththisSR,thepostmaintenanceteststarts(followingacomponentfailure)shouldberemoved.Itwouldalsobemorecorrecttoalsoscreenoutcomponentdemandsthatoccurduringshutdownperiods(e.g.,byfilteringoutdatabasedonthedateoftheevent).ResponseTheworkordersforthecomponentsthatwerecreditedforsuccessinthedataanalysiswerereviewedtodiscoverthenumberofpostmaintenanceteststhatwereperformedonthecomponents.Table15wasaddedtodocumentthenumberofpostmaintenanceteststhatwereremovedfromtheanalysis.
Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-8
Subject:
PLANT PROBABILISTIC RISK ASSESSMENT -
SUMMARY
NOTEBOOK F&ONumberF&ODetails1 15Thesuperinitiator"generaltransient"mayoverlookcertaindifferencesamongitscontributors.Forexample,theimpactofspecificIEslikeLOSPandLossofDCthatmaypreventPORVoperationandchallengethePressurizerSafetiesdonotappeartobecaptured.Inaddition,failuretoprovideaseparateeventtreeforSBOmayoverestimatethesuccessofpowerrecoverybynotaddressingtheoperationofsystemssuchaschargingandAFWfollowingpowerrecovery.(ThisF&OoriginatedfromSRAS A10)AssociatedSR(s)AS A10AS B1SC B3BasisforSignificanceTheaccidentsequencesdonotcontainsufficientdetailtocaptureimportantsystemrequirementsandrequiredoperatorinteractionsforallinitiatingevents.PossibleResolution1) SubdividetheGeneralTransientseventtreetobetterrepresenttheuniquechallengespresentedbyspecificinitiatingevents(e.g.,TransientwithLossofPCS,TransientwithPCSAvailable,LOSP)ordocumenthowthosechallengesareaddressedinthetoplogicmodel.2) Modifytheexistingeventsequenceand/orlinkedfaulttreetoensurethatthechallengetothePressurizerSafetiesiscapturedforinitiatingeventsthatwouldpreventthePORVsfromopening.3) ExplicitlymodeltheSBOsequencestoensurethatthenecessarymitigatingsystemsareaddressedfollowingpowerrecovery.ResponseGTRANwasrestructuredtoaddressthiscomment.ThetreewasupdatedtoexplicitlyaskdemandforPORVsandSafeties Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-9
Subject:
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NOTEBOOK F&ONumberF&ODetails1 19ItwasnotedthatHFEHAPRZ(discussedinSection6.8andSection7.2)isnotcalculatedusingHRACalculator.ThiseventseemstohavebeencarriedoverfromtheWattsBaranalysisandistreatedasbasiceventU1_L2_NOTRCSDEPNOSBO.Inaddition,althoughSection6.8saysthattheNoRCSDepbranchissettoavalueof1forSBOcases,thevalueofbasiceventU1_L2_NOTRCSDEPSBOintheprovidedMASTERL2.CAFfaulttreewassetto0.9995.ThisalsoappearstobeacarryoverfromWattsBar.(ThisF&OoriginatedfromSRLE C7)AssociatedSR(s)LE C7BasisforSignificanceTheHFEsforintentionaldepressurizationneedstobeevaluatedtodeterminetheirapplicabilitytoSQN.PossibleResolution1) IncludeHAPRZintheHRAanalysisforSQNorjustifytheapplicabilityoftheWattsBarvalueandprovideanappropriatereferencetothesource.2) VerifythatthepropervalueofbasiceventU1_L2_NOTRCSDEPSBOisbeingusedinthequantification.ResponseThecurrentanalysishasbeenupdatedtochangethevalueoffailuretodepressurizetheRCSduringSBOscenariosto1.0(assumedfailure)inthemodel.ThebasiceventHAPRZ,whichrepresentsfailuretodepressurizefornon SBOscenarios,usesavalueof0.1forfailuretodepressurize,whichwastakenfromWCAP16341 P,revision0.Thelevel2eventtreesalsousesthecomplimenttothisactioncalledHAPRZ SUCwhichhasaprobabilityof0.9.
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NOTEBOOK F&ONumberF&ODetails2 1Section7.0oftheInitiatingEventsAnalysisobservesadecreasingtrendininitiatorfrequencyinthemorerecentgenericdatasources.However,thereisnocomparisonoftheSQNresultsagainstthegenericresultsnoranexplanationofanysignificantdifferences.(ThisF&OoriginatedfromSRIE C12)AssociatedSR(s)IE C12BasisforSignificanceThecurrentevaluationoftheinitiatorfrequencyresultsdoesnotcompareSQNresultstothegenericfrequencyresults.PossibleResolutionThesection7.0discussioncouldbeexpandedtoincludeacomparisonoftheSQNresultstothegenericresultsandanexplanationofanysignificantdifferences.ResponseAddedtexttoInitiatingeventsnotebookthatcomparesSequoyahinitiatorfrequenciestogenericindustrydata.
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NOTEBOOK F&ONumberF&ODetails2 3Section4.3.1oftheDataAnalysisnotebookdiscussesthebasiceventprobabilitymodelmethodology.GenericdatasourcesselectedforuseareapplicableforSQN.Forthosecomponentswhichhadafailureduringtheanalysistimeperiod(1/1/0311/30/09),thedistributionsareupdatedviatheBayesianupdateprogrambuiltintoCAFTAprogram.However,theintentofthissupportingrequirementistoassurerealisticparameterestimatesarecalculatedforSIGNIFICANTbasiceventsbasedonrelevantgenericandplant specificevidence,notjustthoseforwhichfailureshaveoccurred.Wherenofailureshaveoccurred,useofthegenericdatamaybeconservativesinceitincludesfailuresfrompotentiallylessreliablecomponentsacrosstheindustry.(ThisF&OoriginatedfromSRDA D1)AssociatedSR(s)DA D1DA D3BasisforSignificanceUsingpotentiallyconservativefailureratesforsignificantcomponentscanskewtheriskresults.Bothgenericandplantspecificexperienceshouldbeconsideredforthesignificantbasicevents.PossibleResolutionConsiderperformingaBayesianupdateforallsignificantbasiceventsnotjustthoseforwhichfailureshaveoccurred.ResponseSignificantcontributorsthatwerenotBayesianupdatedwereidentifiedas:BATFRBatteryFailstoOperateBUSFRBusFailstoOperateCBKFOCircuitBreakerFailstoOpenFNSFDStandbyfanfailstostartHXRPLHeatExchanger(RiverWater)PlugsorFoulsMOCXCMotorOperatedValveTransfersClosedPOEFRERCWpumpsfailtorunPSRFRRHRpumpsfailtorunSTRPLStrainersplug Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-12
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NOTEBOOK TSCPLTravelingwaterscreensplugXRFRTransformerfailstooperateTheseeventswereBayesianupdatedusingplantspecificdata.Thenotebookhasbeenupdatedtoreflecttheseadditionalupdates.
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NOTEBOOK F&ONumberF&ODetails2 4AppendixFoftheDataAnalysisnotebookprovidesgraphsthatshowthepriorandposteriordistributions.Table19listsgenericandBayesianupdatedmeanvalues,alongwitharatiooftheposteriortopriormeanvalue.However,therearenoconclusionsdrawnaboutwhetherornottheposteriordistributionsarereasonablegiventherelativeweightofevidenceprovidedbythepriorandtheplantspecificdata.(Note:thestatementthat"TherearenosignificantdifferencesbetweentheindustrydatafromNUREG/CR 6928andtheposteriordistributionsfortheSQNfailurerates"insection11.0isnotjudgedtobesufficient.Forexample,theratiooftheposteriortopriormeanfortheAHUFRtypecodeinTable19is10.6.FortypecodeLSTFR,theratiois4.3.Thesignificanceofthesedifferencesshouldbediscussed.)(ThisF&OoriginatedfromSRDA D4)AssociatedSR(s)DA D4DA E2BasisforSignificanceThereasonablenesscheckneedstoassesswhethertheBayesianupdatesyieldexpectedresultsgiventherelativeweightofevidenceprovidedbythepriorandtheplantspecificdata.PossibleResolutionDiscusstheobserveddifferencesinthepriorandposteriordistributionsanddrawconclusionsonthesignificanceassociatedwiththosedifferences.ResponseTheposteriordistributionswerevalidatedusingthefollowingprocess.UsingaMonteCarlosimulation,theposteriordistributionsweresamplestoseetheprobabilityofhavingarecurrenceinthenumberofeventsobservedinthedatawindowgiventhenumberofsuccessesinthedatawindow.Ifthemeanvaluewaswithin0.05to0.95theresultantdistributionwasusedwithinthemodel.AppendixFwasrewrittentoaddressthisanalysisaswellastopresenttheprior,posterior,andplantspecificdistributions.
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NOTEBOOK F&ONumberF&ODetails2 5ThemethodfromNUREG/CR 6823isusedtoBayesianupdateaJeffreysnoninformativepriordistributionwithplant specificexperience.However,thereisnocomparisonoftheposteriormeanstoplantspecificmeans.(SeethelastsentenceinNUREG/CR 6823,section6.7.1.2.)(ThisF&OoriginatedfromSRDA D4)AssociatedSR(s)DA D4DA E2BasisforSignificanceAreasonablenesscheckshouldbeperformedtoassuretheBayesian updatedmaintenanceunavailabilitesyieldexpectedresultswhencomparedtoplantspecificmeanvaluesgiventheamountofplant specificdata.PossibleResolutionComparetheBayesianupdatedmaintenanceunavailabilitestoplant specificmeanvalues,discusstheobserveddifferencesanddrawconclusionsonthesignificanceassociatedwiththosedifferences.ResponseThefundamentalassumptionusedintheBayesianupdateprocessdescribedintheDataAnalysisnotebookforunavailabilitycalculationsisthatthereisnopriorinformationfromwhichtoBayesianupdate.Therefore,themethodologyusedwastouseaJefferysnoninformativeprior(0.5)asthefoundationfortheupdateprocess.Alloftheavailabledatathatwasusedwasfromplantspecificdatacollection,thereforetheposteriormeanandplantspecificmeanaredirectlycorrelated.ThefollowingassumptionwasaddedtoSection3.0toaddressthenon informativeprior."Forunavailabilitycalculations,aJefferysnon informativepriorwasusedastherewasnoinformativepriorinformationavailable."
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NOTEBOOK F&ONumberF&ODetails2 8Theimportanceofcomponentsandbasiceventsareidentifiedinsections5.1and5.7oftheAccidentSequencenotebook,respectively.However,documentationthatdeterminedtheimportanceresultsmakelogicalsensecouldnotbeidentified.(ThisF&OoriginatedfromSRQU D7)AssociatedSR(s)QU D7BasisforSignificanceMultiplereviewsofthemodelsolutionresultsyieldedmodelchanges,asdocumentedinTable7.0 1andAppendixFoftheQuantificationnotebook.Importancemeasuresarecalculatedinsection5.7;however,theseneedtobeevaluatedinlightofthemodelsolutionresults.Inotherwords,dotheimportancemeasurereportsyieldtheexpectedresults?PossibleResolutionDocumentanevaluationoftheimportancemeasureresultsinlightoftheCDFresults.ResponseAreviewoftheimportanceofcomponentsandbasiceventshasbeenperformedtodeterminethattheymakelogicalsense.Thereviewshowsthattherisksignificantcomponentsareconsistentwiththemodelresultsandlimitations.Significantcontributorsincludebasiceventsassociatedwithdiesels,ERCW,ComponentCooling,RHR,AtmosphericReliefValves(ARVs)andAirCompressors.InSQN,failureoftheauxiliarycontrolairheadersimpactstheARVsthatareneededtocooldown/depressurizeinLOCAscenariossincethecondenserisunavailablefromaPhaseBisolation.Theemergencydiesel,ERCW,RCPbreakers,andRHRareimportantsincetheirfailureresultinscenariosinvolvingSBOandRCPsealLOCAs.
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NOTEBOOK F&ONumberF&ODetails3 1Section4.5,"Thecalculationaboveprovidesthatthecontainment'hole'sizemustliebetweena1inchequivalentpathanda4inchpath.Therefore,itisacceptabletousetheNRCvalueof2inches."Basedonthestatement,the1"equivalentholeshouldhavebeenconsidered.(ThisF&OoriginatedfromSRLE D7)AssociatedSR(s)LE D7BasisforSignificanceItisunknownwhattheapplicablebreaksizeisbetween1"and4",thereforetheconservativeapproachistouse1".PossibleResolutionPerformdetailedanalysistoensuretheuseofthe2"equivalentholeisallowableoruse1"andincludetheadditionalpenetrationsinthecontainmentisolationanalysis.ResponseSection4.4discussesthereasoningforconcludingthatthe2"holesizeisacceptableforuseintheSequoyahlevel2analysis.Thereferenceshowsthatthereleaseratecorrespondingtoa1771scfmratewouldberepresentedbyaventlinediametergreaterthan1"andslightlylessthan2".Becausethepointcorrespondingto1771scfmat19psig(whichishalfoftheassumedseverecontainmentchallengepressure)isonlyslightlybelowthe2"contourlineshowninReference33,andthereisconservatismbuiltintoboththeassumedcontainmentfailurepressureandtheassumedleakratesatthatpressure,itisjudgedappropriatetouse2"astheboundingvalueforalargeleakrate.
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NOTEBOOK F&ONumberF&ODetails3 7SeveralareaswereidentifiedthatneedadditionaldiscussionwithrespecttotheSuccessCriteriaAnalysis.Forexample:1) ThedifferencesbetweenplantresponsetoapipebreakSLOCAandaconsequentialPORVLOCAarenotfullydiscussed.Giventhedifferencesinbreaklocation,thereshouldbesomediscussionintheSuccessCriteriaNotebookofwhythepipe breakSLOCAanalysesboundtheconsequentialPORVLOCA.Inaddition,whilethereisadiscussionintheTHNotebookcomparingthevaluesofsomekeyparametersforthepipebreakSLOCAandtheconsequentialPORVLOCA,thisdoesnotfullyexploredifferencesinplantresponsethatmayaffectthesuccesscriteria.2) Thereneedstobemorediscussionofwhythe480gpmperpumpRCPSealleaksareincludedintheMediumLOCA(MLOCA)grouping.ItisstatedinSection4.4.10oftheTHNotebookthatthe480gpmsealLOCAmeetstheMLOCArequirementofnotrequiringAFWforaccidentmitigation,butthereisnodocumentationofsuccesscriteriaanalysesthatsupportthisstatement.3) ThebasisforassumingaSGTRflowof700gpminSection7.2.10oftheTHNotebookneedstobediscussedinmoredetailthansimplynotingthatnohistoricSGTRhasbeenofthemagnitudeofadoubleendedguillotineruptureofaSGtube.4) TheLOCAanalysisislimitedtotheupperandlowerendofthebreakrangeforeachclass.THanalysisatthemiddleofthebreakrangewithintheLarge,Medium,andSmallLOCAcategoriesmayprovideinsightsthathavenotbeenrevealedbytheupperandlowerendofthebreak.Forinstance,itisnotclearifsequenceMLOCA 011canbeasuccesspathforabreakinthe3to5inchrange.(ThisF&OoriginatedfromSRSC B3)AssociatedSR(s)SC B3BasisforSignificanceThereisalackofdiscussionregardinghowtheseitemsweretreatedinthesuccesscriteriaanalysis.PossibleResolutionExpandthediscussionofthenoteditemsintheSuccessCriteriadocumentation.Response1) ThesmallLOCAeventsassumethathebreakoccurslowwithinthephysicalstructureoftheRCS.ThesebreakswillalwayshaveahigherdeltaPvaluethan Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-18
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NOTEBOOK thoseofbreaksatthetopoftheRCS(PORVLOCA).Duetotheadditionalpressureandotherthermo hydrauliccharacteristicsthesuccesscriteriaisboundingfortheSLOCAcases.2) The480gpmsealLOCAisnowgroupedasaSLOCA.ThisrequirestheuseofAFWforsuccessfulaccidentmitigation.3) Thevalueof700gpmwasusedasanattempttoboundtheanalysis.Theselectionof700gpmwasdonetoassurethattheanalysiswasrealisticinnature,butconservativeaswell.4) TheMLOCAeventtreehasbeenrestructuredtorequiresuccessfulinjectionoftheCLAsthisistoassurethatanybreaksizewithintheMLOCArangecanbesuccessfullymitigatedafterfailureoftheCVCSsystemtoinject.
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NOTEBOOK F&ONumberF&ODetails3 9AllmitigationstrategiescreditedintheaccidentsequencemodelwhenthehighpressurerecirculationhasfailedarenotprescribedbythecorrespondingEOPs.Inotherwords,themitigationcreditintheeventtreemodelhasnobasis.Thisissuehasbeenself identifiedbytheSQNPRAstaffandacorrectiveactionreporthasbeenwrittenfortheEOPgrouptoresolvethisissue.AtthisstagethePRAgroup"firmly"believesthattheEOPwillbemodified,notthemodel.Thusitisatrackingissue.(ThisF&OoriginatedfromSRSCA3)AssociatedSR(s)SC A3BasisforSignificanceThereisaCRwrittenbytheTVA.PossibleResolutionTacktheCRandensurethattheproceduresaremodifiedorthatthemodelischangedtoreflecttheas operatedplant.ResponseEOPrevisionswereapprovedattheSQNPORCmeetingonMay6th2011.
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NOTEBOOK F&ONumberF&ODetails3 13Section4.4.2oftheTHNotebook(MDN 000 0002010 205)discussestheuseofMAAPforLLOCAinthecoldleg.TheconclusionisthatthelargeLOCA(LLOCA)limitationsarenotapplicabletobreaksizes<10inches.ThereferenceusedforthisisaMAAPtraininglecture.UseofMAAPtomodeltheinjectionphaseoftheLLOCAneedsadditionaljustificationwithreferencetotheapplicabletechnicaldocuments.(ThisF&OoriginatedfromSRSC B4)AssociatedSR(s)SC B4BasisforSignificanceMAAPisknowntohavedifficultymodelingtheinitialphaseoftheLLOCAevents.PossibleResolutionUseRELAPorotheralternativecodestoanalyzetheinitialphaseoftheLLOCAorprovideamorecomprehensivejustificationfortheuseofMAAPwhichincludesbenchmarkingagainstothercodes.ResponseThelimitationnotedforMAAPareforthelargerendoftheLLOCAspectrumperEPRITR 1020236.ThesuccesscriteriaforthelargeLOCAwasconsistentwithandlargelyderivedfromtheSQNdesignbasisanalysisandSAR.WhilethisdoesleadtoconservativeresultsintheLLOCAeventtree,theexpenditureofadditionalresourcesforthefurtherrefinementusingadditionalcodessuchasRELAPisnotwarranted,giventhatLLOCAeventsarenotrisksignificantintheSQNmodel.ThelowimportanceoftheLLOCAsequencesisconsistentwithotherPWRsintheindustry.TheMAAPanalysisfortheLLOCAeventswereusedmostlyasconfirmationoftheeventtreesbasedontheSQNSARandfortimingofHRAevents.SpecificallyfortheHRAevents,MAAPwasonlyusedtodeterminedepletionoftheRWSTandlongtermtimetocoredamagebasedonfailureofhotlegrecirculation.BothofthesecasesaresignificantlypasttheinitialstagesofaLLOCAwhereMAAPisnotedtolackthethermalhydraulicdetailrequiredtoevaluatetheinitialblowdown(EPRI TR1020236).
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NOTEBOOK F&ONumberF&ODetails3 14SeveraldocumentationissueswerenotedintheSuccessCriteriaandTHNotebooks.Specifically,1) Figures7 60and7 61oftheTHNotebook(MDN000 000 2010 205)needtobereplacedwithupdatedresults.2) ThediscussionofaccidentsequencenodeLPHinSection7.3.1oftheTHNotebook(MDN 000 000 2010 205)statesthat"ThetimeforswitchovertohotlegrecirculationisspecifiedintheEOPE 1as3hoursaftertheinitiationofalargeLOCA(Reference4,Step31c)."Intheparagraphimmediatelybelowthisstatement,thecalculationofthetimeavailableforrecoveryfromafailureofrecirculationusesaswitchovertimeof5hours.DiscussionwithTVApersonnelindicatedthatthe3hourvaluewascopiedfromtheWBNnotebook.TheactualtimespecifiedintheSQNproceduresis5hours.3) Table7 13oftheTHNotebook(MDN000 000 2010 205)doesnotincludesuccesspathISLM 014asshowninFigure6.4 10oftheAccidentSequenceNotebook(MDN 000 000 2010 0201).Inaddition,successpathISLM 017inTable7 13oftheTHNotebookisnotshowninFigure6.4 10oftheAccidentSequenceNotebook.4) Section4.4.11oftheTHNotebook(MDN 000 000 2010 205)discussestheclassificationofaStuckOpenPORVasasmallLOCA.Thebasisneedstobeprovided.(ThisF&OoriginatedfromSRSC C1)AssociatedSR(s)SC C1BasisforSignificanceThedocumentationneedstomatchthecurrentanalysis.PossibleResolution1) Replacefigures7 60and7 61withthecorrectfigures.2) RevisethetexttousethecorrectinformationforSQN.3) EnsurethesequencedesignationsinTable7 13oftheTHNotebook(MDN 000 000 2010205)matchthoseinFigure6.4 10oftheAccidentSequenceNotebook(MDN 000 000 2010 0201).4) JustifytheclassificationoftheStuckopenPORVasaSLOCA.Response1)Figures7 60and7 61wererevisedintheTHcalculationMDN000 000 2010 205.
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NOTEBOOK IntheoriginalMAAPruns,theSGARVswereopenedat30minutes,thisdroppedpressureintheRCS.OpeningoftheSGARVwasnotcreditedintheeventtreeforthesequencesevaluatedinfigures7 60and7 61.ThisisapplicabletotheWBNTHanalysisaswell.2)TheTHNotebookwasrevisedtobeconsistentwithEOIE 1step22.ThecorrecttimeofswitchingovertoHotLegRecirculationof5hourswasincludedinSection7.3.1oftheTHNotebook.3)Table7 13andFigures6.4 10wererevisedtobeconsistent.4)Additionalinformationwasincludedinsection4.4.11oftheTHnotebooktojustifytheclassificationofaStuckOpenPORV.Thisinformationincludesacomparisonofcoredamagetimingandmass/energyreleaseratesthroughaSOPORVandSLOCA.
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NOTEBOOK F&ONumberF&ODetails3 19Section7.2oftheHRANotebook(MDN 000 0002010 0204)doesnotexplicitlydiscusshowtherequiredandavailablemanpowerisaddressedintheanalysis.Manpowerrequirementsareincludedintheoperatorinterviewchecklistasitem37.However,itisnotclearhowthisinformationwasusedinthedevelopmentoftheHEPssincesomeinstanceswereobservedwheretheoperatorinterviewresponseswerenotusedintheHRAcalculator(seeHFEHARR1).(ThisF&OoriginatedfromSRHRH2)AssociatedSR(s)HR H2BasisforSignificanceSomeaccidentscenarioscanrequiremoremanpowerthanothersandthisisnotdiscussed.PossibleResolutionAddadiscussionofhowthemanpowerrequirementsareaccountedforintheHRA,especiallyforthoseHFEswhichrequirelocalactions.ResponseAdiscussionoftherequiredandavailablemanpowertoperformtheactionsandequipmentmanipulationswasdocumentedinsections7.1and7.2oftheHRAnotebook.Also,HARR1wasrevisedtomatchtheoperatorinterviewforthemanpowerrequirements.
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NOTEBOOK F&ONumberF&ODetails3 20SeveralissuesrelatedtotheTHanalysesusedtosupporttheHRAwereidentified.Specifically,1) SometimewindowsareburiedinMAAPoutputfileswhicharenotincludedintheTHNotebookandtaketimetoreview.Forexample,thetimewindowforAFWOP5isnoteasilyavailable.2) THNotebookMDN 000000 2010 205Section7.3.3discussestheactionsrequiredfollowingafailureofhighpressurerecirculation.Therequiredactionrelatedtofailureoftheautomaticrecirculationalignment(HARR1)hastwobigpieces.Thefirstistostopthepumptoavoidpumpdamage.Ifthepumpsaredamaged,highpressurerecirculationcan'tbesuccessful.ThetimewindowisshortforthisactionandisrelatedtoRWSTdepletion.Ifthepumpsarestoppedontimethenextactionistomanuallyestablishrecirculation.ThetimewindowforthatactionisbasedontheRCSinventorydepletionwhichis,relativelyspeaking,muchlonger.IfHPrecirculationisnotsuccessful,theRCSisdepressurizedtofacilitatelowpressurerecirculation(AFWOP3).Thesetwoactions(HPrecirculationandRCSdepressurizationandestablishLPinjection/recirculation)areforthesamemitigationfunction.Therefore,itisunclearwhytherearebigdifferencesbetweenthetimewindowsforthesetwoactions.Inaddition,theHRACalculatorinputfortheseactionsappearstobedifferentfromthedescriptionsinSection7.3.3ofMDN 000 000 2010205.3) TheuseofboundinganalysesfortheHFEsresultsinnonsequencespecifictiminginformationintheHRA.Forexample,HARR1isusedintheaccidentsequencesafterAFWSsuccessinSSBOandSSBIaccidentsequences.However,thetimingwindowofHARR1isbasedonthemediumLOCAanditisconservativeforthesesequences.(ThisF&OoriginatedfromSRHRI1)AssociatedSR(s)HR I1BasisforSignificanceThedocumentationoftheTHanalysesperformedtosupporttheHRAisdifficulttotraceandinsomecasescontainsconflictinginformation.PossibleResolution1) ReviseTable8 1intheTHNotebook(MDN 000 000 2010 205)toincludeadditionalinformationsuchasthetimetocoredamageandareferencetotheapplicableTHcases.
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NOTEBOOK
- 2) ReviewtheresultsofTHcasessupportingtheHRAtoensurereasonableconsistencyoftimewindowsfordifferentactionswiththesamepurpose.3) Refinethetiminganalysisasnecessarytoensuretheresultsarerealisticandrepresenttheaccidentsequence(s)inwhichtheactionsareused.Response1) THnotebookrevised-allHRAtiminginTable8.12) AllTHresultcaseswerereviewedtoensurethatthetimewindowsinusewereconsistentbetweendifferentactionswiththesamepurposes.3) AsstatedinthedetailsoftheF&O,theanalysisusedisconservative.Thetiminganalysisisforthemosttimelimitingbreakforwhichtheactionisapplied.Thisconservativetimingselectionaddressesallpotentialscenarios/breaksizesandwouldonlyreduceHEPandaddadditionalmargintotheanalysis.Thisisconsideredtobeappropriateduetotherangesofbreaksizesincludedinthebroadbandsofinitiatingeventgroupings.EvaluationoftherecoveryofadditionalmarginfromdevelopinglowerHEPindividualanalysesforeachapplicationofHARR1willbecompletedinfuturerevisionsoftheSQNPRAmodel.
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NOTEBOOK F&ONumberF&ODetails3 25Severaldocumentationissueswerenoted.Forexample:1) SequencesISLM 008andISLM 017weredeletedfromtheISLOCAeventtree.However,thereisnodiscussionofwhythiswasdone.2) Paragraphsinsection6.4.7needtoberevised.Specifically,thefirstsentenceinthefirstparagraphonpage62,startingwith"IfthetemperatureoftheRCSis557°Fanddropping,thesteamdumps,S/GPORVsandblowdownisolationvalvesareclosed."needstobefinished.Thereisthe"if"butno"then."ItisalsounclearhowthissentenceisrelatedtotheaccidentsequenceeventtreeorthefollowingstatementsintheparagraphrelatedtothePORVs.Thesecondparagraphonpage62hasgrammaticalerrors(e.g.,"-thepossibilityofhaveaRCPSealLOCA-").3) ThediscussionofmanualcontrolrodinsertionfollowingATWSinsection7.9needstoberevisedtoreflecttheintenttoremovecreditforthisactionfromthemodel.(ThisF&OoriginatedfromSRAS C1)AssociatedSR(s)AS C1BasisforSignificanceInconsistenciesinthedocumentationcanaffectmaintenanceandupdateofthemodel.PossibleResolution1) AddadiscussionexplainingwhysequencesISLM 008andISLM 017arenotusedorrenumberthesequencestoensuretherearenogapsinthenumbering.Also,ensureallrelateddocuments(e.g.,theSCandTHnotebooks)arerevisedforconsistency.2) Reviewsections6.4.7and7.9andrevise,asneeded,toensurethatthediscussionreflectstheaccidentsequencemodels.Response1) Thesequenceswerenotrenumberedfollowingthelatestupdatetotheeventtrees.Thenumberingschemewillbeupdatedinthenextrevisionofthenotebook.2) Thegrammaticallyerrorsnotedhavebeenupdatedandrevised.3) TheATWSdiscussionofMRIhasbeenupdatedtostatethatonlythemechanicalbindingofthecontrolrodsorthefailureoftheautomaticcontrolsystemare Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-27
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NOTEBOOK modeled.
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NOTEBOOK F&ONumberF&ODetails4 3Non waterfloodsourcesareexcludedonthebasisofAssumption11ofthenotebook.However,theStandardstates(inNote1forthisSR)thatnon watersourcesshouldbeconsidered,AmoredetailedbasisforexcludingthesesourcesshouldbedevelopedtomeettherequirementsofthisSR.(ThisF&OoriginatedfromSRIFSOA1)AssociatedSR(s)IFSO A1BasisforSignificanceThisisconsideredtobeafindingsincetherequirementsoftheSRhavenotbeenfullymet.PossibleResolutionUpdatetheanalysistoconsidernonwatersources,orbetterjustifywhythefloodingimpactsofthesenon watersourcesarenotsignificantandhencedonotrequireevaluation.ResponseAssumption11wasrewordedto:Allsourcesoffluidwithintheplantwereanalyzedforfloodingconsiderations.However,theglycolsystemistheonlysystemwhichcouldhaveanimpactonthefloodinganalysis.Allothersourcessucharesindidnothaveenoughvolumetocauseimpacttoplantoperation.Theglycolsystemalsohasaminimumvolume,butthelocationofthepiping,inthecontrolroddriverooms,causessystemtobeasourceofsprayinitiatingevents.
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NOTEBOOK F&ONumberF&ODetails4 7Nodiscussionofsourcesofuncertaintyassociatedwiththefloodinginitiatingeventsiscurrentlyprovidedinthefloodingnotebook(MDN 000 000 2010 0203).Itisnotedthatthenotebookincludesdocumentationofsourcesofuncertaintyforotherportionsofthefloodinganalysis.SourcesofmodeluncertaintyforinternalfloodingarealsodocumentedinMDN 000 000 2010 0209,UncertaintyandSensitivityAnalysis;however,againfloodinitiatoruncertaintiesarenotdiscussed.Ifnouncertaintiesareidentifiedforthefloodinitiatorfrequencyevaluation,thenthenotebookshouldstatethistobeconsistentwiththeapproachusedfortheIFPP,IPSO,andIFSNtasks.(ThisF&OoriginatedfromSRIFEV B3)AssociatedSR(s)IFEV B3BasisforSignificanceThisisconsideredtobeafindingsincetherequirementsofthisSRarenotmetPossibleResolutionProvideanassessmentofsourcesofmodelinguncertaintyforthefloodinitiatorfrequencydetermination.ResponseSection8.8wasaddedtotheInternalFloodingNotebookwiththefollowing:Theinternalfloodingfrequencycalculationhasseveraldifferentuncertaintiesassociatedwiththecalculation.Thecurrentmodelusesasummationofthreedifferentfrequencies,passivepipebreakfailures,humaninducedfloods,andmaintenanceinducedflooding.Eachofthesefloodingeventshasitsowninherentuncertainties.ForpassivepipebreakfailuresrateshavebeengivenanuncertaintyparameteraspresentedinSection8.5.TheimpactoftheseuncertaintiescanbetreatedbytheuseofarandomsamplingMonteCarloprocessasdiscussedinSection10.1.Humaninducedfloodingeventspresentanotherdifficultchallenge.TheuseoftheHRACalculatorprogramfromScientechcreatesanassumeduncertaintytermforanyHRAaction.Sincethehumaninducedfloodingeventsisacombinationofbothpreinitiatingeventandpostinitiatingevent,eachportionhasanindependentuncertaintyterm.TheHRACalculationprogramalsoarbitrarilyassignsanuncertaintytermtoHRAactionsbasedonthecalculatedprobabilities,seetheHRACalculationformoreinformationontheuncertaintyparameters(Reference68).Theotherfundamentalissuethatispresentedinhumaninducedfloodingeventsisthelocationofwork.Dependingonwheretheactualworkisbeingperformedinafloodarea, Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-30
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NOTEBOOK isolationcouldbeaconcernasthenextavailablevalvecouldbeinaninaccessiblearea.Additionally,therearenodetailedprocedurestoaddresshavingafloodoccurduringamaintenanceevent.Maintenanceinducedfloodingeventsalsopresentalevelofuncertainty.Thethreemaininputstothecalculationofthisfrequency,failurerateofanMOV,missiontime,andfrequencyoftheactivityallintroducesomelevelofuncertaintyintothecalculation.ThelargeinternalruptureofanMOVisassumedinNUREG/CR 6928tobeafactorof0.02lessthanthatofasmallinternalleakonanMOV(Reference104),astherehasbeennoactuallargeinternalruptureeventsintheindustry.Themissiontimeisalsoassumedbasedonasevendayrepairinterval,thisnumbercouldpotentiallybegreaterthanthatifthecomponentisnotcoveredbyanTechnicalSpecificationor,morelikely,lessthantheassumedsevendayrepairtime.Thefinalareaofuncertaintyisthefrequencyoftheactivity.MostoftheproceduresreviewedinAppendixJhavefrequenciesaswellasconditions.Theseconditionscouldcausetheactualmaintenanceactivitytooccurmoretimesthanthefrequencynotedintheprocedure.
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NOTEBOOK F&ONumberF&ODetails4 11WhilethePRAmodelconsidersthepossibilityoftwoPORVsbeingblockedatthesametime,theredoesnotappeartohavebeenaninvestigationofwhethercoincidentmaintenancecanoccurinthevariousSQNsystems(orifcoincidentinter systemmaintenancecanoccur).ThereforethisSRisnotmet.ItwasalsoobservedthatthePORVblockingbasiseventsnotedabovedidnotappeartobedocumentedineitherthedatanotebookortheappropriatesystemnotebook.(ThisF&OoriginatedfromSRDA C14)AssociatedSR(s)DA C14BasisforSignificanceThisisafindingsincethetechnicalrequirementsoftheSRarenotmet.PossibleResolutionAstudyshouldbeconductedtodetermineifcoincidentmaintenanceconditionscanoccur.Ifso,thesystemmodelsmayneedtobemodifiedandadditionalbasiceventstorepresentthecoincidentmaintenancestateswouldneedtobeadded.Ifitisdeterminedthatnocoincidentmaintenancecanoccur,thenthisshouldbedocumentedinthedatanotebookorwithinthesystemnotebooks.DocumentationforthecalculationofthetimethateitheroneorbothPORVscanbeblockedshouldalsobeaddedtoeitherthesystemnotebookorthedatanotebook.ResponseThefollowingwasaddedtothedataanalysisnotebooktoaddresscoincidentmaintenance:CoincidentmaintenanceisschedulingmaintenancewheremultipleSSCsareoutofserviceatthesametime.Specificallycomponentsonthesametrain,RHRtrainAandSItrainAforexample,beingoutofserviceformaintenanceatthesametime.TheOutageandSiteSchedulingDirectiveManual1.0(Reference28)dictatesthat:Twelve(12)weekschedulebyFEGgroupsensuresthatwithinatrainweek,notwo(2)accidentmitigatingdevicesareremovedfromserviceatthesametime[i.e.,"A"trainResidualHeatRemoval(RHR)isnotremovedfromserviceatthesametimeas"A"trainContainmentSpray.]ThisrequirementisfurtherdiscussedintheOutageandSiteSchedulingDirectiveManual4.7(Reference29)whichstatesthatanysystemsimportanttoPRAthatareunavailableatthesametimemustmeettherequiresoftheplantriskmatrix.NormallymaintenanceonanysystemsimportanttothePRAisnotscheduledatthesametime.Ifitistheseinstancesareextremelyrareandthecurrentmodeldoesnot Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-32
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NOTEBOOK excludecoincidentmaintenanceeventsfromappearinginasinglecutset.Thereforetheprobabilityofhavingcoincidentmaintenanceeventsisextremelyrareandaccountedforduringthenormalcutsetprocessing.
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NOTEBOOK F&ONumberF&ODetails5 2SomeHFEsaresettoavalueof0.0forquantification.Forexample,HACI1andHAAE1arerecoveryactionsforautomaticsignalsANDedwiththesignallogic.However,theHRAanalysissetstheHEPprobabilityto0.0basedonananalysisthattheoperatoractionisnotrequired.Thisscreeningapproach,combinedwiththemodelstructure,removestheautoactuationcontributiontomitigatingsystemfailureduringquantification.(ThisF&OoriginatedfromSRHR G1)AssociatedSR(s)HR G1BasisforSignificanceScreeningHFEsusingavalueof0.0removetheautoactuationhardwarefailurecontributionsinthequantificationresults.PossibleResolutionRevisethemodeltoremovenoncreditedoperatorrecoveryactionsfromthelinkedfaulttreeorsetallnoncreditedeventstoTRUEduringquantification.ResponseForthoseeventswhere0.0swereusedinthemodelthefaulttreewasupdatedtoremovetheeventssothattheconflictconcerninganANDgateandazeroeventwillnotlongerbeencounteredduringnormalquantification.
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NOTEBOOK F&ONumberF&ODetails6 2ThejustificationforexcludingplantdatapriortoJuly2002inthecalculationofplantspecificIEfrequenciesisnotdocumentedwellenoughtosupportIE C2.(ThisF&OoriginatedfromSRIE D2)AssociatedSR(s)IE C2IE D2BasisforSignificanceAjustificationwasprovidedduringthereview,butitisnotdocumentedinthenotebook.PossibleResolutionIncludeajustificationforexcludingdatapriortoJuly2002intheIEnotebook.ResponseAddeddiscussiontonotebookstatingthatdaterangewasadequatetogetagoodsampleofplantdatawithoutgoingtoofarbackandincludingeventsthatoccurredwhentheplantmayhavehaddifferentproceduresandoperatingpractices Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-35
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NOTEBOOK F&ONumberF&ODetails6 3ThealignmentflagsintheERCWsystemarenotfullyimplementedtorepresentthesystemalignmentwithintheInitiatingeventportionofthetree.Forexample,thegatesunderU0_AEX_G006shouldcontainflagstoindicatewhichpumpisrunningandwhichtwopumpsarenot,sothatthetwononrunningpumpswouldhaveconsiderationsforfailurestostart.(ThisF&OoriginatedfromSRIEA6)AssociatedSR(s)IE A6IE C10BasisforSignificanceThecurrentfaulttreeconfigurationdoesnotproperlyaccountforthesystemalignment.PossibleResolutionIncludethealignmentflagsintheindicatedandsimilargates.Reviewtheremainderofthetreetoensurethatthealignmentsareproperlyidentified.ResponseThecurrentflagalignmentforERCWhasbeenrevisedso,forthebaselinemodel,withoutsettingaspecificconfiguration,theflagfilesweresettotherespectivetimeineachconfigurationtothataprobabilityisnowusednotatrueorfalsevalue.
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NOTEBOOK F&ONumberF&ODetails6 5Thesupportsysteminitiatingeventtreesforthemostpartincludeprovisionsforcommoncausefailuresandroutinesystemalignments.TherearesomediscrepanciesinthemodelingofcommoncausefailuresintheERCWandCCSmodelsthatrequireattention,however.Forexample:1) Whileacommoncauseeventforall3ofthe1A,1B,andC Spumpsfailingtorunexists,therearenoteventsforthe1AandC Spumpsorthe1BandC Spumps.2) ThestructureoftheERCWtreeissuchthatpumpcommoncausefailurescouldresultinapumpfailingduetoanindependentfailureaswellasacommoncausefailureinasinglecutset.(SeegateU0_AEX_G001)3) ThecommoncauseinitiatingeventgroupU0_ERW08POEFRIisnotvalid,sinceitisentirelybasedon8760hourexposuretimeforallthecomponents.Thecommoncausefailurefrequenciesarethereforeoverestimated.TheCCStreeusesadifferentapproachthantheERCWtreeforcommoncauseinitiatingevents.AnalternateapproachisalsogiveninEPRIreports1013490and1016741.(ThisF&OoriginatedfromSRIEA6)AssociatedSR(s)IE A6IE C10SY B3BasisforSignificanceSupportsysteminitiatingeventfailuresareinconsistentlyappliedamongthesupportsystemmodels,andmaybegivingincorrectresults.PossibleResolutionReviewinitiatingeventcommoncauseeventsandselectaconsistentmodelingapproachamongthesupportsysteminitiatingeventmodels.ResponseWithrespecttothecommoncausefailureoftheCCSpumps:Thecommoncausefailureofthe1AandtheC Spumporthe1BandtheC SpumpwouldnotmeettherequirementstocauseaninitiatingeventfortheCCSsystem.OnlyfailureoftheAtrainwouldcausetheplanttohavetotripastheloadsonthecommontrainarenotrequiredforoperationatpower.Thereforeonlythecommoncausefailureofallthreepumpsismodeledinthefaulttree.
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NOTEBOOK WithrespecttothecommoncausefailureeventsfromtheERCWfaulttree:ThecommoncausefailureeventsintheERCWsystemwherecommoncausefailureandindependentfailuresshowupinthesamecutsetpresentaminimalandconservativeimpact.Withrespectthecommoncausecalculationofbasicevents:ThecommoncausefailureratesforERCWpumpsfailingtorunandCCSpumpsfailingtorunwererevisedbasedontheEPRIdocument1013490usingthediscussionpresentedonpage5 8.TheassumptionsandcalculationofthesebasiceventsisnotedinAppendixBofeachcalculation.
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NOTEBOOK F&ONumberF&ODetails6 6Section5oftheIEnotebookshowsaBayesianprocesswasusedtocombineplantspecificandgenericdata.However,LOCAfrequenciesfromNUREG 1829werealsoupdatedwithplantspecificdata.SincethefrequenciesinNUREG 1829werebasedonexpertjudgmentandnotactualindustrydata,anditisnotexpectedthataplantwouldexperiencesuchanevent,itdoesnotseemappropriatetousetheBayesianupdateprocessfortheseevents.TheupdatedidnotappeartosignificantlyaltertheIEfrequencies,however,sothereislittleimpactonCDF.(ThisF&OoriginatedfromSRIE C4)AssociatedSR(s)IE C4BasisforSignificanceFrequenciesinNUREG 1829werebasedonexpertjudgmentandnotactualindustrydata,anditisnotexpectedthataplantwouldexperiencesuchanevent,itdoesnotseemappropriatetousetheBayesianupdateprocessfortheseevents.PossibleResolutionUsethefrequenciesderivedfromNUREG 1829withoutBayesianupdatingwithplantdata.ResponseThefrequenciespresentedinNUREG1829representthebestestimatesavailableatthattime.Thereisnorestrictiononupdatinganexpertsolicitation,astheupdateprocesswillonlyserveastobetterestimatetheactualfailureratefortheinitialingevents.
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NOTEBOOK F&ONumberF&ODetails6 7Section6oftheInitiatingEventsAnalysis,theassociatedsystemnotebooks,andtheHRAnotebookdocumenttheuseofplantspecificinformationintheassessmentandquantificationofrecoveryactionswhereavailable,inamannerconsistentwiththeapplicableHRSRs.AnissuewasnotedwiththeERCWinitiatingeventtree.EventHAAEIE"OperatorFailstoStartERCWPump(InitiatingEvent)"hasbeensettozerobasedonananalysisthatfoundonepumpwassufficienttocoolplantloads,soifoneofthetworunningpumpstrips,operatoractionisnotrequiredtostartanotherpump.Operatoractiontostartastandbypumpwouldberequired,however,ifflowwastobelostfrombothrunningpumps.Thecurrentmodelessentiallyassumesasuccessfuloperatoractiontostartbothofthosepumps.(ThisF&OoriginatedfromSRIE C11)AssociatedSR(s)IE C11BasisforSignificanceTheoperatoractionHAAEIEhasinappropriatelybeenassumedtobe100%successful.PossibleResolutionRe evaluatethefailurerateforoperatoractionHAAEIE,giventherevisedrequirementsoftheERCWsystemwithregardstocausinganinitiatingevent.ResponseTheERCWinitiatingeventmodelhasbeenupdated.CalculationCN NUC SQN MEB MDQ000 067 2000 0095revisedtheexistingsuccesscriteriausedintheinitiatingeventmodel.Theresultsofthecalculationthataslongasthecontainmentsprayheatexchangerswerenotinservice,themaximumrequiredflowontheERCWsystemwouldberoughly9,000gallons.Thisiswithinthedesignflowrateof10,000gallonsperminutefromoneERCWpump.Duetothechangeinthesuccesscriteria,theinitiatingeventmodelwasupdatetorequiringthefailureoftworunningERCWpumpsaswellasfailureofbothstandbyERCWpumpstostart.TheHRAactionHAAEIEwasaddedtothemodelundertheappropriatefailuretostartgate,nolongerunderanANDgate.Additionally,thefaulttreelogicinquestionwasupdatesothatfailuretostarttakesintoaccountthefailureofoperationactionHAAEIE.
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NOTEBOOK F&ONumberF&ODetails6 10Tables42and43ofMDN 000 0002010 0209containalistofmodelingassumptionsandtheirimpactonthePRAmodel.However,themajorityofitemsinTable43haveanimpactof"Unknown."ClassificationofmodelimpactfortheseassumptionsisnecessarytomeetthisSR.(ThisF&OoriginatedfromSRQU E4)AssociatedSR(s)QU E4QU F4BasisforSignificanceTheSRrequiresidentificationoftheimpactofidentifiedassumptionsonthemodel.PossibleResolutionProvideanevaluationofanimpactoftheitemslistedas"Unknown"intable43.ResponseTheUncertaintyandSensitivityAnalysiscalculationhasbeenupdatedinthefollowingways:TextconcerningthediscussionofUnknownimpactsandperformingarespectiveuncertaintyanalysiswasremovedfromSection5.0.Table43wasupdatedtoremovethecolumn"ModelImpact"andthecolumn"Comments"wasupdatedto"ModelImpactsandComments"andexpanded.
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NOTEBOOK F&ONumberF&ODetails6 12Fromtheresultspresentedinsections5.2and5.7ofMDN 000000 2010 0208,itcanbeinferredthatthedefinitionofsignificantbasiceventandsignificantaccidentsequenceareconsistentwiththoselistedinPart2ofthestandard.Thisisnotexplicitlystatedinthedocumentation,however.Thedefinitionofsignificantcutsetisnotprovided,nordoesthe100cutsetlistprovidedinthedocumentationimplythatthepart2definitionwasused,asthe100cutsetsdonotrepresent95%oftherisk.(ThisF&OoriginatedfromSRQU F6)AssociatedSR(s)QU F6BasisforSignificanceDocumentationofthedefinitionof"significant"isrequiredbytheSR.PossibleResolutionProvideadefinitionofsignificantcutset,significantsequence,andsignificantbasiceventinthedocumentation.ResponseThedocumenteddefinitioninSection1 2.2oftheASME/ANScombinedstandardwasaddedtothequantificationcalculation.
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NOTEBOOK Suggestion Level F&Os F&ONumberF&ODetails1 1RG1.200Revision2documentsaqualifiedacceptanceofthisSR.TheNRCresolutionstatesthattomeetCapabilityCategoryII,theimpactsoffloodinducedmechanismsthatarenotformallyaddressed(e.g.,usingthemechanismslistedunderCapabilityCategoryIIIofthisrequirement)mustbequalitativelyassessedusingconservativeassumptions.(ThisF&OoriginatedfromSRIFSNA6)AssociatedSR(s)IFSN A6BasisforSignificanceThisisanenhancementrequiredtosatisfytheRG1.200qualification.PossibleResolutionDocumentaqualitativeassessmentoftheimpactsofjetimpingement,pipewhip,humidity,condensation,temperatureandotherfloodinducedmechanismsthatarenotexplicitlymodeled.ResponseTheanalysiswaschangedsothatallcomponentswithinafloodareaarefailedoninitiationregardlessoftheequipmentqualificationsorotherHELBmitigationfeatures.
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NOTEBOOK F&ONumberF&ODetails1 2Allcomponentswereassumedfailedifsubjectedtosubmergence.MDN 000000 2010 0203AppendixFdocumentswhethercomponentsweresprayvulnerable.Discussionwiththeresponsibleanalystrevealedthatfactorsconsideredinsprayvulnerabilitydeterminationsincludedshielding,sealing,andequipmentqualificationrecords.However,thisisnotdocumentedinthenotebook.(ThisF&OoriginatedfromSRIFSNA7)AssociatedSR(s)IFSN A7IFSN B2BasisforSignificanceThetechnicalrequirementismet,butthedocumentationcouldbeenhanced.PossibleResolutionDocumentthebasisfordeterminingwhetheracomponentis"sprayvulnerable"inMDN 000000 2010 0203.ResponseThefollowingtextwasaddedtoAppendixF:Duringthewalkdownssprayvulnerabilitywasdeterminedbyobservationsofthecomponents.ForMOVs,anobvioussealwithwaterproofinghadtobeobservedtodetermineifthecomponentwasvulnerabletospray.AfterthewalkdownswerecompleteacomparisontotheEQdatabasewasdonetoobserveifanycomponentswhichwereseenduringthewalkdownsandnotedasbeingvulnerabletosprayactualwereenvironmentallyqualified.ThosecomponentswereremovedfromthesprayanalysisaswellasAppendixGandAppendixH.TableswerealsoaddedtoAppendixFtoshowthosecomponentsthatarecurrentlylistedasenvironmentallyqualifiedintheMAXIMOdatabase.
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NOTEBOOK F&ONumberF&ODetails1 3ThewaterdepthrequiredtocausefailureofdoorsisdocumentedintheAppendixFwalkdownsheets,butthederivationofthisvalueisnotdocumented.(ThisF&OoriginatedfromSRIFSNA9)AssociatedSR(s)IFSN A9IFSN B2BasisforSignificanceDiscussionwiththeresponsibleanalystrevealedthatholdupofwaterbydoorswasnotcreditedinthefinalanalysis.Thereforethisisadocumentationissuethatdoesnotaffecttheresults.PossibleResolutionDocumentthebasisforthewaterdepthrequiredtocausedoorfailureusedinAppendixF.ResponseDoorfailureheightcalculationswereperformedpostwalkdown.Thefailureheightswerebrokenintotwodifferentsections.Ifthedoorwasobservedtobeafiredoor,thecalculationoffailureheightusedtheHELBanalysis,ifthedoorwaswiremeshoranonfiredoorthenthefailureheightwasassumedtobe0feet.Theheightofwaternecessarytofailadooriscalculatedbasedoniftheheightwillexceedtheactualheightofthedoor.Forcalculationswheretheheightofwaterislessthantheheightofthedoor,thefollowingequationcanbeused:Wherepisthefailurepressureandisthespecificweightofwater.Forthosecalculationswheretheheightofwaterwouldexceedtheactualheightofthedoor,adifferentequationmustbeused.Thedoorisnowconsideredtobeacompletelysubmergedsurface,sothefollowingequationcanbeused:Wherepisthefailurepressure,isthespecificweightofwater,andh dooristheactualheightofthedoor.ForthepurposesofAppendix F
,thefailurepressureofallfiredoorswastakenfromtheHELBanalysis,whichstatesthatapressureof1.5psidwillcausethedoorstofail.
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NOTEBOOK Usingtheequationsabove,thefailureheightofwateriscalculatedtobe6.92feet.
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NOTEBOOK F&ONumberF&ODetails1 5Documentationofthescenarioimpactsneedstobestrengthenedinsomeareas.Forexample:1) InMDN 000 000 20100203Section9.4.3.3forfloodarea734.0 A13itisstatedthat"thefloodfrequencyforthoseeventsthatimpactbothACAScompressorswillbe1/3oftheoriginalfrequency."Nobasisisprovidedforthisstatement.2) Section7.3makesthegeneralstatementthatfortheTurbineBuilding"floodoriginatedinanylevelwouldpropagatefreelytothebasementofthebuildingwithoutanyhindrance."ThissameassumptionwasappliedinpartitioningAuxiliaryBuildingarea930.0 A1,butthisisnotdocumented.(ThisF&OoriginatedfromSRIFSNA6)AssociatedSR(s)IFSN A6IFSN B2BasisforSignificanceThetechnicalprocessisacceptable,butthedocumentationcouldbestrengthened.PossibleResolutionDocumentassumptionspertinenttothefloodscenariossuchasthosenotedintheF&Odescription.ThiscouldbedoneintheNotebookortheFRANXdatabase.ResponseThetextquotedintheanalysishasbeenupdated.AfurtherwalkdownoftheplantwasdonetoreflectpipingthatcouldimpactboththeACASaircompressors.FurtherwalkdownsconductedinApril2011showedtherewasnopipingthatwaswithintwentyfeetofbothaircompressorsskidsontherefuelfloor.Thereforethesectionofthenotebookaddressingsuchissueswasremoved.Additionally,thebreakingapartof690intozoneshasbeenenhancedinthedocumentationandisdiscussedintheplantpartitioningsection.Thedocumentationofthefloodingscenariosandthefloodinganalysisnowincludesthepartitionsaspartoftheanalysis,notasachangeafterreviewofthecutsets.
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NOTEBOOK F&ONumberF&ODetails1 6MDN 000000 2010 0204Section7.3statesthatareasonablenesscheckwasperformedandgenerallydescribesthefactorsconsidered.However,itwouldbehelpfultoprovidetablesthatgroupedHEPsbytherelevantfactorstosupporttheconclusionsreached.(ThisF&OoriginatedfromSRHR G6)AssociatedSR(s)HR G6BasisforSignificanceThecheckwasperformed,buttheresultsarenotdocumentedinawaythatsupportsverificationoftheconclusions.PossibleResolutionIncludeatablethatgroupstheHFEsbythecomplexitytosupporttheconclusionthat"HFEsthataremorecomplexhavehigherfailureprobabilitiesthansimpleactions."Similarly,atablethatgroupstheHEPsbytheavailabletimewouldsupporttheconclusionthat"HFEswithshortertimewindowshavehigherfailureprobabilitiesduetofactorsincludinginsufficienttimetocreditreviewfromtheSTAandnegativeperformanceshapingfactors(e.g.,highstresslevels).ThesefactorsarenotincludedinthereferencedTable10 2.ResponseTableshavebeenaddedtoHRAnotebooksection7.3forthecomplexityandtimemargincomparisonscompletedfortheSQNRev5PRAandHRAmodelupdate.
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NOTEBOOK F&ONumberF&ODetails1 9TheCCSsystemsuccesscriteriaismodeledasdependentonthetemperatureoftheERCWsystem.MDN 000070 20100217AppendixDdocumentsthederivationoftheprobabilityforflageventFLG0070_ERCW_TEMP_GT_70(FLG_0024SUMMERinthelinkedfaulttree).However,thetemperaturedataispresentedinagraphicalformatratherthanatabularformat.Atabularformatwouldmakeiteasierforareviewertoperformavalidationofthedata.(ThisF&OoriginatedfromSRSY A10)AssociatedSR(s)SY A10BasisforSignificanceThisisadocumentationissuenotaffectingthetechnicalqualityofthemodel.PossibleResolutionProvidethedatausedtoderiveFLG0070_ERCW_TEMP_GT_70inatabularformat.ResponseThedatausedtocreatethegraphintheCCSnotebookis111,210cellslong.Thisdataisnotfeasibletobepresentedaspartofthesystemnotebook.Anelectroniccopywillbekeptwiththenotebookandwillbeavailabletoanypersonwishingtoreviewthedata.
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NOTEBOOK F&ONumberF&ODetails1 12SomesourcesofuncertaintywhicharecharacterizedashavinganimpactonLERFarenotanalyzedusingsensitivityanalysis(e.g.,coremeltarrestinvesselandmodelingoftheARFs).Inaddition,Section5.14.10statesthat"Althoughpost coredamagehumanactionssuchasintentionaldepressurizationoftheRCSaremodeledasrealisticallyaspossible,theuncertaintyrelatedtotheseactionsisaddressedinsensitivitystudies."Thenotebookreferencesthesensitivityanduncertaintyanalysisnotebookasthelocationforthesestudies.However,MDN 000 000 2010 0209performedtheuncertaintybysettingthevaluesofHEPsinthemodeltotheir5thand95thpercentilevalues.ThisdoesnotaddressuncertaintyinthevalueoftheHFEforintentionaldepressurizationoftheRCSwhichismodeledaspartoftheU1_L2_NOTRCSDEPNOSBObasicevent.(ThisF&OoriginatedfromSRLEF3)AssociatedSR(s)LE F3BasisforSignificanceThisisacompletenessissuewhichwouldenhancetheanalysis.PossibleResolutionAddsensitivityanalysestoaddresstheseadditionalsourcesofuncertaintyintheLERFresults.ResponseTheHFEforintentionaldepressurizationoftheRCSwaschangedfromU1_L2_NOTRCSDEPNOSBOtoHAPRZinthemodel,andwasincludedintheuncertaintystudyfortheHFEs.
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NOTEBOOK F&ONumberF&ODetails1 13Theoperatoractionsreflectedintheeventtrees,andthesequencespecifictiminganddependenciesthataretraceabletotheHRAfortheseactionsarediscussedintheHRAnotebook.TheoperatoractionsmodeledforeachsequencearelistedasaseparatesubsectioninMDN 000 000 2010 0201.However,itissuggestedthatasummarydiscussionofoperatoractionsaffectingtheaccidentsequences,includingadiscussionofthetopeventsimpacted,beincludedintheASnotebook.(ThisF&OoriginatedfromSRAS A4)AssociatedSR(s)AS A4AS C2HR G4BasisforSignificanceDocumentationenhancement.Thelinkscanbeidentifiedthroughcrossreferencetothesystem,successcriteria,andHRAnotebooks.PossibleResolutionProvideacrossreferenceforeachoperatoractiontotheaffectedeventtreetopeventintheAccidentSequenceNotebook.ResponseTheoperatoractionsarediscussedalongwitheachofthetopeventsforalleventtreesintheaccidentsequencenotebook.
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NOTEBOOK F&ONumberF&ODetails1 17CalculationType1isusedformissiontimeeventsandType2isusedforbasiceventswheretheprobabilityisbasedonperiodictests.TheCAFTAusersmanualstatesthat"-itisbettertousethemorepreciseformulasofcalculationtypes3,4,5or6.Thisisespeciallyimportantifyouareusinglargernumbers(e.g.,t>.05),orifyouwillbedoinguncertaintyanalysis."(seeCAFTAusersmanualTables6 2aand6 2bandthetextbelowTable6 2b.)(ThisF&OoriginatedfromSRQU E3)AssociatedSR(s)QU E3BasisforSignificanceThecurrentmethodyieldsavalidapproximationofthebasiceventprobabilities,butdoesnotrepresenttherecommendedpracticeforCAFTA.PossibleResolutionUsethemorepreciseformulasforthetimedependentbasiceventprobabilitycalculations.ResponseThecurrentmodeldoesnothaveanyeventswherethelambda*tvalueapproaches0.05.Themodelsusedwillhavetheircalculationtypesupdatedto3and5whereappropriatewhenanyupdateistobeperformed.Forthecurrentmodel,noeventsgeneratedarandomvaluegreaterthan1.0fortheuncertaintygraphs.
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NOTEBOOK F&ONumberF&ODetails1 18Thesystemnotebookstypicallystatethathighenergylinebreak(HELB)isconsideredintheInternalFloodingAnalysis(e.g.,ERCWNotebookSection3.4.7.2.6).However,assumption3.1oftheIFNotebookstatesthat"Additionalfailuremodes;jetimpingement,pipewhip,humidity,condensation,andtemperatureinducedfailuresareoutsidethescopeofthisanalysis."(ThisF&OoriginatedfromSRSY C2)AssociatedSR(s)SY C2BasisforSignificanceThetreatmentofHELBisnotclearlydocumentedinthesystemnotebooks.PossibleResolutionModifythestatementsinthesystemnotebookstoclearlystatethatHELBisnottreatedandtoprovideajustificationforthis.ResponseTheassumptionwasremovedfromtheIFnotebook.TheHELBeventsarenowaddressedintheIFanalysis.
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NOTEBOOK F&ONumberF&ODetails1 20TheSQNPRAconsideredEarlycontainmentfailureaswellasLatecontainmentfailureandbasematmeltthrough.Aftercontainmentfailure,thereisnoadditionalequipmentnorhumanactioncreditedtomitigatetheconsequences.ThereisalsonoevidencethatareviewwasperformedtodetermineifcreditingoperationofadditionalequipmentorhumanactionsaftercontainmentfailurewouldreduceLERF.(ThisF&OoriginatedfromSRLE C11)AssociatedSR(s)LE C11LE C12BasisforSignificanceRecommendationformeetingtherequirementsforCapabilityCategoryII/III.PossibleResolutionDocumentareviewoftheLERFresultstodetermineifcreditforequipmentoperationaftercontainmentfailureoradditionaloperatoractioncreditwouldbeeffectiveinreducingLERF.ResponseTherearenoadditionalactionsorequipmentcurrentlycreditedinthelevel2analysistomitigatetheconsequencesofareleaseaftercontainmentfailure.Thisresultsinsomewhatconservativeresults.AreviewhasbeenperformedtodetermineifcreditingadditionalequipmentorcreditingadditionalhumanactionscouldresultinaLERFreduction.AnactionidentifiedduringthisreviewinvolvescreditingmanuallyclosingtheRCPsealwaterreturnoutboardisolationvalvefollowingcoredamageintheeventthatitfailstocloseondemand.Asensitivitystudywasperformedtodeterminetheeffectofthisactionusingvariousassumedfailureprobabilities(seeSection12.7),althoughthefeasibilityofimplementingtheactionhasnotbeenstudiedindetail.
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NOTEBOOK F&ONumberF&ODetails1 21Athoroughlistofreferencesisdocumentedwitheachsystemnotebook.However,thereferencerevisionlevelisnotalwaysincluded(seetheDieselGeneratorandRPSnotebooks,forexample.)(ThisF&OoriginatedfromSRSY A2)AssociatedSR(s)SY A2SY C2BasisforSignificanceThesupportingrequirementismet,butthedocumentationcouldbeenhanced.PossibleResolutionProvidetheapplicablerevisionlevelforeachreferencetoimprovestraceabilityofthesourcedocuments.ResponseReferencelevelswereleftoffofthereferencesinallsystemnotebooksconsistentwiththeTVApracticesforPRAcalculations.
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NOTEBOOK F&ONumberF&ODetails2 2PerdiscussionwiththeSQNPRAdataanalyst,maintenanceisgenerallyperformedonatrainbasisratherthanacrossredundantcomponents.Whereredundantmaintenanceispermissible,e.g.,theERCWsystem,whichhas8pumps,thefaulttreeallowsforthegenerationofcutsetsthathavemultiplepumpsinmaintenance.However,itwouldbehelpfultodocumentaverificationthatsimultaneousunavailabilityofredundantequipmentisnothowworkisplanned.(ThisF&OoriginatedfromSRSY A20)AssociatedSR(s)SY A20SY C2BasisforSignificanceSincemaintenanceonredundantequipmentisnotmodeledasaplannedevent,documentationshouldbeprovidedorreferencedthatdescribeshowmaintenanceisplanned/coordinated.Thisassuresthatanymaintenancedependenciesarenotoverlooked.PossibleResolutionDocumentthemaintenanceapproachtakenonredundantequipmentandanyimpactonthePRAmodel.ResponseCoincidentunavailabilityisnowdiscussedinSection7.4.4ofthedataanalysisnotebook.
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NOTEBOOK F&ONumberF&ODetails2 6Nomodificationstoplantdesignoroperatingpracticeswereidentifiedthatleadtoaconditionwherepastdataarenolongerrepresentativeofcurrentperformance.Thuslimitingtheuseofolddatawasnotrequired.However,forcompleteness,itissuggestedthatthedataanalysisdocumenttheconsiderationofthissupportingrequirement.(ThisF&OoriginatedfromSRDA D8)AssociatedSR(s)DA D8DA E2BasisforSignificanceNodocumentationaddressingthissupportingrequirementwasidentified.PossibleResolutionDocumentaconsiderationofmodificationstoplantdesignoroperatingpracticesthatcouldleadtoaconditionwherepastdataarenolongerrepresentativeofcurrentperformanceintheDataAnalysisnotebook.ResponseSection7.2.1wasaddedtothedataanalysisnotebooktoaddressplantdesignchanges.
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NOTEBOOK F&ONumberF&ODetails2 7ThedataanalysisalignswellwiththePRAStandardrequirementsandisgenerallywelldocumented.Addinga'roadmap'tothePRAStandarddataSRsaswasdoneelsewhereinthePRAdocumentationwouldenhancetheperformanceofPRAapplications,upgrades,andpeerreview.(ThisF&OoriginatedfromSRDA E1)AssociatedSR(s)DA E1BasisforSignificanceAddinga'roadmap'tothePRAStandarddataSRswouldenhancetheperformanceofPRAapplications,upgrades,andpeerreview.PossibleResolutionIntheDataAnalysisnotebook,adda'roadmap'tothePRAStandarddataSRs.ResponseAppendixIwasaddedtothenotebooktoaddresstheASME/ANSstandardsections.
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NOTEBOOK F&ONumberF&ODetails3 4ThesuccesscriteriadescriptionneedstoincludetheboundaryconditionssuchasRCSpressure.Ingeneral,itisnotclearwhattheconditionisthatallowstheSIpumpstooperate.(ThisF&OoriginatedfromSRAS A3)AssociatedSR(s)AS A3AS B2SC A3BasisforSignificanceDocumentationofrequiredconditionspermittingsomeequipmenttooperateisnotprovided.PossibleResolutionProvidemoredetaileddiscussionofthesuccesscriteriaandmitigationsystemoperatingcharacteristics(e.g.pressure,flowrate)andhowtheconditionsareachieved.Forinstance,theSIpumpinjectionpressureandhowthepressureisachievedintheaccidentsequence(i.e.,byopeningpressurizerorSGPORVs)shouldbediscussed.ResponseAllboundaryconditionsarelistedineithertheparameterfileortheinputdeckselectroniccopiesoftheseareavailableonrequest.
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NOTEBOOK F&ONumberF&ODetails3 10TheEOPsassociatedwithaspecificaccidentsequencesuccesspatharenotidentifiedintheThermalHydraulicAnalysisortheSuccessCriteriaNotebook.ThesearealsonotexplicitlydiscussedintheAccidentSequenceNotebook.(ThisF&OoriginatedfromSRAS A5)AssociatedSR(s)AS A5BasisforSignificanceThereisnodiscussionrelatingtheEmergencyOperatingProcedure(EOP)withaccidentprogressioninASnotebook.PossibleResolutionProvidediscussionsrelatingEOPstotheaccidentsequenceandtopeventsordering.ResponseTheEOPstepsarenowincorporatedintotheaccidentsequencenotebook.
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NOTEBOOK AppendixAResolutionofF&OsFactsandObservationsSummary-SuggestionF&OsF&ONumberF&ODetails3 11Useofthedesignbasisforcertainsuccesscriteriamayresultinconservativemodeling.Forexample:1) Section4.4.3oftheTHNotebook(MDN 000 0002010 205)discussestheMAAP4.0.7limitationswhichpreventuseofMAAPfordeterminingthenumberofaccumulatorsrequiredforLargeLOCA(LLOCA)success.Theuseofthedesignbasisassumptionthat3of3intactloopaccumulatorsarerequiredislikelytobeconservative.2) Section4.4.7oftheTHNotebook(MDN 000 0002010 205)discussesthenumberoflinesneededfortheEmergencyCoreCoolingSystem(ECCS).BasedonMAAPlimitations,theconclusionisthatthecurrentanalysesonlysupportECCSflowthroughallintactlines.Thisconclusionislikelytobeconservativeforsomesequences.(ThisF&OoriginatedfromSRSC B1)AssociatedSR(s)SC B1BasisforSignificanceTheuseofdesignbasissuccesscriteriafortheaccumulatorsandfortherequirednumberofinjectionpathsmaybeconservative.PossibleResolutionPerformPRAspecificanalysisusinganalternativecodetodetermineifsuccesscanbeachievedwithfewerthan3accumulatorsorwithflowtofewerECCSinjectionpaths.ResponseMAAPcurrentlyistheconsensusmodelofchoiceforanalysissupportingthedevelopmentofthePRAmodel.TheuseofothercodesdoesnotfacilitatethedevelopmentofthePRAmodel.Thecurrentsuccesscriteriaof3of3CLAswillberetainedinthemodel.
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NOTEBOOK F&ONumberF&ODetails3 22Itisstatedthattheimpactsoftheinitiatingeventonmitigationsystemsarecapturedinthetopevents.However,thereisnodiscussionoftheseimpactsintheaccidentsequencenotebook.(ThisF&OoriginatedfromSRAS B1)AssociatedSR(s)AS B1BasisforSignificanceThereisnodocumentationfoundthatexplicitlydescribesthedependenciesbetweenthemitigationsystemsandtheinitiatingevents.PossibleResolutionDiscusstheimpactofinitiatingeventsonindividualmitigationsystemsundereachtopevent.Alternatively,provideaninitiatingeventtomitigatingsystemdependencymatrix.ResponseAninitiatingeventimpacttablewasaddedtotheSuccessCriteriaNotebook Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-62
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NOTEBOOK F&ONumberF&ODetails3 23Theimpactofthephenomenologicalconditionscreatedbytheaccidentprogressionisnotdiscussedintheaccidentsequencenotebook.(ThisF&OoriginatedfromSRAS B3)AssociatedSR(s)AS B3BasisforSignificanceThereisnodiscussionofphenomenologicalconditionsintheASnotebook.However,theenvironmentalconditionsaffectingequipmentoperationiscapturedinthesystemanalysisnotebooks.PossibleResolutionAddadiscussionofthephenomenologicalconditionscreatedbytheaccidentsequenceandtheirimpactonthecreditedmitigationequipment.ResponseThecurrentphenomenologicalconditions,initiatorimpact,arediscussedwithineachsystemnotebook.
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NOTEBOOK F&ONumberF&ODetails3 24Theintersystemdependenciesareembeddedintheaccidentsequences,butthereisnoexplicitdiscussionofthesedependencies.(ThisF&OoriginatedfromSRAS B5)AssociatedSR(s)AS B5BasisforSignificanceThereisnointersystemdependencydiscussionintheASnotebook.PossibleResolution1) Addanexplicitdiscussionoftheintersystemdependencytothediscussionofeachaccidentsequence,or2) IncludeasystemdependencymatrixintheASNotebooktoillustratethedependencies.ResponseAsystemdependencymatrixhasbeenincludedwithintheSCnotebook.
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NOTEBOOK F&ONumberF&ODetails4 1Section5.2oftheInternalFloodingnotebook(MDN 000 0002010 0203)considersfloodareasinthebuildingsofbothunits,andincludesallcommonbuildings.Atthebuildinglevel,thetextdiscusseswhetherthebuildingcontainssharedequipment;however,thetextandtablesdonotindicatewhichspecificfloodareascanimpactbothunits.Itwouldbehelpfultoenhancethedocumentationtoindicatewhichfloodareashavemultiunitimpacts.Similarly,thediscussionoffoodsourcesshouldattempttoidentifysourceswithmultiunitimpacts.(ThisF&OoriginatedfromSRIFPP A3)AssociatedSR(s)IFPP A3IFSO A2BasisforSignificanceThisisasuggestionsinceitpertainssolelytoenhancementofthedocumentationofthefloodareapartitioningandfloodsourceidentificationprocess.Thefloodanalysisitselfcorrectlyaddressesmultiunitimpacts.PossibleResolutionInclude(inthetextofsection5.2orwithinthetablesofincludedareas)indicationofwhatareashavemultiunitimpacts.Includesimilardocumentationinsection6.1forfloodsources.ResponseAllareascurrentlyanalyzedthatcontainERCW,CCS,HPFP,RCWoranyotherinfinitesourceofwaterareaddressedinSection5.2Thetablesprovidedlistallareasoftheplantincludingthosewherethereareandarenotmultiunitimpacts.NochangesweremadetotheinternalfloodingdocumentSection5.2or6.1 Calculation No. MDN-000-000-2010-0200 Rev: 001 Plant: SQN Page: A-65
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NOTEBOOK F&ONumberF&ODetails4 6TheSQNfloodinganalysishasaddressedsome,butnotalloftherequirementsforCategoryII/III.EPRIfloodingdatabasedongenericindustryexperienceisusedforfloodinitiatingeventsduetopiperuptures.Plant specificdatathatmightinfluencethepipefailuredata(e.g.,materialconditionofthefluidsystemsandwaterhammerexperience)arenotconsidered.However,areviewofplant specificmaintenanceinducedfloodingeventswasperformed(AppendixGofthefloodingnotebook)andwasconsideredinthecalculationofmaintenancefloodingfrequency.TofullymeetCategoryII/III,anassessmentshouldbemadeofplantmaterialconditionandwaterhammerexperience)andwhetherplantconditionswarrantanyadjustmentstothegenericfloodfrequenciesthatareused.(ThisF&OoriginatedfromSRIFEVA6)AssociatedSR(s)IFEV A6BasisforSignificanceThisisjudgedtobeasuggestion,sincetheresponsetothisF&Owillmostlikelyonlyimpactdocumentation.Also,theanalysismeetstheCategoryIrequirements,whichmaybesufficientformostapplicationsPossibleResolutionReviewplant specificexperiencepertainingtoplantmaterialconditionandwaterhammeranddocumenttheresultsofthereviewinthefloodingnotebook.ResponseAreviewofplantspecificfloodingeventswasperformedinAppendixG.Thisdatawasincorporatedintotheanalysisforinitiatingeventfrequencies.
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NOTEBOOK F&ONumberF&ODetails4 8Dependencybetweenpre initiatoreventswasdeterminedtonotbeapplicableduetothelargeamountoftimebetweentestandmaintenanceeventsbetweenvarioussystemtrainsandtheuseofdifferentcrewstoperformeachtrain'sactivities.Dependencebetweenpre initiatorsandpostinitiatorsisalsonotappropriate.Therationalefornotconsideringdependencyforpre initiatorsseemsappropriate.However,severalinconsistencieswerenotedintheHRAnotebookdocumentation(MDN 000 000 2010 0204)concerningthepreinitiatordependencytreatment.VariousHRAcalculatorentriesforthepre initiatorevents(inAppendixB)indicatethatdependencybetweeneventsistobeconsidered(seeforexample,eventSHEEMC_4).TheintroductorymaterialinAppendixFcontainssomestatementsindicatingpre initiatordependencywillbeconsidered,andotherstatementsexplainingwhydependencybetweentheseeventsisnotexpected.Theseinconsistenciesshouldbecorrected.(ThisF&OoriginatedfromSRHR D5)AssociatedSR(s)HR D5BasisforSignificanceThisisconsideredtobeasuggestionsinceitpertainstocorrectingdocumentationerrors.Theunderlyinganalysesthemselvesarecorrectandwillnotbeimpactedbytheseerrors.PossibleResolutionCorrectthedocumentationerrorsintheHRAnotebookasnoted.ResponseThedocumentationerrorsinAppendixBoftheHRAnotebookwerecorrected.
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NOTEBOOK F&ONumberF&ODetails4 9AppropriategenericdatasourcesappeartobeusedintheSNPRA,asdocumentedinthedataanalysisnotebook(MDN 000 000 20100202).ComponentfailureratesaretakenprimarilyfromNURG/CR 6928(withothersourcesusedincasesinwhichdataforspecificcomponenttypesarenotavailable).CommoncausedataisobtainedfromrecentNRC(INEL)andPWROGdatasources.OffsitepowerrecoverydataisobtainedfromNUREG/CR 6890.Componentrecoveryisnotused.Table2andAppendixAdescribetheboundariesassumedforeachmajorcomponenttype.TheSQNPRAmakesuseofgenericunavailabilitydatafromNUREG/CR 6928forcomponentsforwhichplantspecificdataisunavailable(asnotedinTable8).Itisassumed(seeAssumption1inthedatanotebook)thatallgenericdataisapplicabletoSQN;however,sincethisSRrequiresthattheconsistencyoftheSQNpracticesandphilosophiesbecheckedagainstthegenericdatasourceassumptions,additionaldocumentationneedstobeprovidedtobettermeettherequirementsofthisSR.Itisrecognizedthatassumption1islistedasanimportantuncertaintyandisdiscussedintheUncertaintiesnotebook(MDN000 000 2010 0209).However,sincetheuniqueattributesconcerningtheuseofgenericunavailabilitydataarenotdiscussed,addinganadditionalassumptionitemforthisissuemaybeappropriate.(ThisF&OoriginatedfromSRDA C1)AssociatedSR(s)DA C1DA E2BasisforSignificanceThisisconsideredtobeasuggestionasitpertainsprimarilytoadocumentationenhancement.Theuseofgenericunavailabilityestimatesforsomeplantcomponentsisprobablyacceptable;howeverdocumentationofthebasisforacceptingthisdataasappropriatetoSQNisrequired.PossibleResolutionEnhancethedocumentationinsection6.2tobetterdescribetheacceptabilityofthegenericestimatesforSQN.ConsiderationshouldbegiventospecificallyidentifyingthisgenericdatauseasanimportantassumptionintheUncertaintiesnotebook(MDN 000 000 2010 0209)aswell.Thatnotebookhasanoverallitemconcerningtheuseofgenericdata;however,aspecificitemfortheuseofgenericunavailabilitydatacouldalsobeadded.ResponseAdiscussionofthecomponentboundariesandmaintenancepracticeswasaddedtosection6.2.
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NOTEBOOK F&ONumberF&ODetails4 10Failuredatarecordsareobtainedfromtheplant'sCauseDeterminationandEvaluation(CDE)recordsthatarepreparedbysystemengineersinresponsetofailureevents.TheguidanceforCDEdevelopmentinplantprocedureSPP 6.6describesbasesforfailures,discussesdegradedconditions,andnotesthatTechnicalSpecificationfailuresoroperabilityissuesarenotautomaticallyMaintenanceRulefunctionalfailures(orPRAfailures).TheCDErecordsarealsothenreviewedbythePRAstafftodetermineifaPRAfailurehasoccurred.TheCDEsthatwereusedinthedataanalysisareincludedinAppendixDofthedatanotebook(MDN 000 000 2010 0202).BecausethereareseveralDA CSRsthatspecifyrequirementsforthedatacollectionandanalysisprocess,itissuggestedthatthedataanalysisdocumentationbeenhancedtospecificallynotetheserequirementsandhowtheyaremet,especiallysincetheotherplantproceduresdonotspecificallystatetheserequirements(sincetheproceduresareforsystemengineersandothernon PRApersonnel).(ThisF&OoriginatedfromSRDA C4)AssociatedSR(s)DA C4DA C5DA C11DA C12DA C13DA E2BasisforSignificanceThisisasuggestionsinceitpertainstoenhancingthedocumentationtoplaceallofthedataanalysisgroundruleswithinthedatanotebookforclarity.PossibleResolutionEnhancethedataanalysisnotebooktospecificallylistthedatacollectionrequirementsforDA C4,C5,C6,C11,C12,andC13.ResponseDA C4Functionalfailuresaredeterminedbasedonthesystemengineerandmaintenanceruleexpertpanel.ThesedeterminationsareoutlinedinSPP 6.6,andareonlymadebyaqualifiedindividual.
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NOTEBOOK DA C5Therewasanidentifiedeventwheremultiplerepeatfailuresoccurredwithinthesametime.EachoftheseeventswasassignedaspecificCDE,howeverasnotedinthedocumentationofCDE1615thethreeeventswereallassignedtoonefailureeventinthePRAmodel.DA C11UnavailabilityisdefinedinthemaintenanceruletechnicalinstructionTI 4.Thedefinitionpresentedstatesthatunavailabilityisonlycountedwhileatpower(mode1),additionallyinthedefinition,unavailabilityiscreditedwhenthecomponentwouldnotbeabletoperformitsdesignedfunction.DA C12ThedefinitionofthecomponentboundariesfortrackingunavailabilityaredocumentedinTI 4.Forfrontlinesystemsonlyfrontlineimpactsareassignedtothatsystem.IftheERCWheaderorothermulti systemimpactcomponentsareunavailablethentheunavailabilityistrackedatthatlevel.DA C13Forallsignificantunavailabilities,startandfinishtimesareaccuratelydocumentedinthemaintenancerulespreadsheets.