L-2015-046, Response to Request for Additional Information Regarding License Amendment Request No. 229, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance
| ML15069A153 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/20/2015 |
| From: | Kiley M Florida Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-2015-046, TAC MF3931, TAC MF3932 | |
| Download: ML15069A153 (17) | |
Text
0 L-2015-046 10 CFR 50.90 February 20, 2015 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Re: Turkey Point Nuclear Plant, Units 3 and 4 Docket Nos. 50-250 and 50-251 Response to Request for Additional Information Regarding License Amendment Request No. 229, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program"
References:
- 1. Florida Power & Light Company letter L-2014-033, License Amendment Request No.
229, Application for Technical Specification Change Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program," April 9, 2014 [ML14105A042]
- 2. NRC letter "Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Request for Additional Information on License Amendment Request to Revise Technical Specifications to Implement TSTF-425, Revision 3, 'Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specifications Task Force (RITSTF) Initiative 5B' (TAC Nos. MF3931 and MF3932)," August 7, 2014 [ML14212A713]
- 3. Florida Power & Light Company letter L-2014-266 "Response to NRC Technical Specifications Branch Request for Additional Information Regarding License Amendment Request No. LAR-229, 'Application for Technical Specification Change Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program'," August 29, 2014
[ML14252A228]
- 4. NRC letter "Request for Additional Information Re. LAR 229 for Turkey Point 3 & 4 (TACs MF3931 & MF3932)," January 22, 2015 [ML15023A080]
In Reference 1 and supplemented by Reference 3, Florida Power & Light Company (FPL) submitted a request for an amendment to the Technical Specifications (TS) for Turkey Point Units 3 and 4. The proposed amendment would modify the TS by relocating specific surveillance frequencies to a licensee-controlled program with implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specification Initiative 5b, Risk-informed /VJ6 Florida Power & Light Company 9760 SW 3 4 4 th St., Florida City, FL 33035
Florida Power & Light Company L-2015-046 Page 2 of 2 Method for Control of Surveillance Frequencies." The changes are consistent with U.S. Nuclear Regulatory Commission (NRC)-approved TS Task Force Standard TS change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-Informed TS Task Force] Initiative 5b," Revision 3.
In Reference 3, the NRC staff requested additional information in order to complete its review of the requested amendment. The enclosure to this letter provides FPL's response to the request for additional information (RAI).
This response to the RAI does not alter the conclusion in Reference 1 that the proposed changes do not involve a significant hazards consideration.
This RAI response contains no new regulatory commitments and does not modify any existing commitments.
Should you have any questions regarding this submittal, please contact Mr. Mitch Guth, Licensing Manager, at 305-246-6698.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on February ,2015 Sincerely, Michael Kiley Site Vice President Turkey Point Nuclear Plant Enclosure cc: NRC Regional Administrator, Region II NRC Senior Resident Inspector NRC Project Manager Ms. Cindy Becker, Florida Department of Health
Florida Power & Light Company L-2015-046 Enclosure ENCLOSURE Response to Request for Additional Information Regarding License Amendment Request No. 229, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program" APLA RAI-1 APLA RAI-2 APLA RAI-3 APLA RAI-4
Florida Power & Light Company L-2015-046 Enclosure APLA RAI-1 Nuclear Energy Institute (NEI) 04-10, Revision 1 (ADAMS Accession No. ML071360456),
Section 4,0, Step 8, states:
The risk impact of a proposed [Surveillance Test Interval (STI)] adjustment shall be calculated as a change of the test-limited risk (see Regulatory Guide 1.177, Section 2.3.3). Since the test-limited risk is associated with failures occurring between tests, the failure rate that shall be used in calculating the risk impact of a proposed STI adjustment is the time-related failure rate associated with failures occurring while the component is in standby between tests (i.e., risk associated with the longer time to detect standby-stress failures).
Describe how the Turkey Point Surveillance Frequency Control Program will address the standby time-related contribution for extended surveillances.
Response
The standby time-related contribution evaluation will be performed in accordancewith NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b Risk-Informed Method for Control of Surveillance Frequencies,"Revision 1. Any changes to the frequencies listed in the Surveillance FrequencyControl Program(SFCP)will comply with the following guidance from NEI 04-10, Revision 1:
In general, the failure probability values of components used in probabilisticrisk assessments (PRAs) consist of a time-related contribution(i.e. the standby time-related failure rate) and a cyclic demand-relatedcontribution (i.e. the demand stress failure probability). The risk impact of a proposed STI adjustment shall be calculatedas a change of the test limited risk (see Regulatory Guide 1.177, Section 2.3.3). Since the test-limited risk is associatedwith failures occurringbetween tests, the failure rate that shall be used in calculating the risk impact of a proposed STI adjustmentis the time-related failure rate associatedwith failures occurring while the component is in standby between tests (i.e., risk associatedwith the longertime to detect standby-stress failures).
Therefore, caution should be taken in dividing the failure probabilityinto time-related and cyclic demand-relatedcontributionsbecause the test-limited risk can be underestimated when only part of the failure rate is considered as being time-related while this may not be the case. Thus, if a breakdown of the failure probabilityis considered,it shall be justified through data and/or engineeringanalyses. When the breakdown between time-related and demand-relatedcontributionsis unknown, all failures shall be assumed to be time-related to obtain the maximum test-limited risk contribution.
Page 1 of 14
Florida Power & Light Company L-2015-046 Enclosure APLA RAI-2 NEI 04-10, Revision 1, Section 4.0, Step 10, provides guidance on the initial assessment of Internal Events, External Events, and Shutdown Events. Describe how shutdown events will be assessed as part of the Turkey Point Surveillance Frequency Control Program.
Response
The shutdown risk evaluation will be performed in accordance with NEI 04-10, Revision 1, which permits quantitativeor qualitative assessment of shutdown risk impacts. Fleet procedures will be written consistent with Nuclear Energy Institute (NEI) industry guidance document, NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b Risk-Informed Method for Control of Surveillance Frequencies,"Revision 1, for performing the shutdown risk assessment.
Documentation of the assessment will include the following:
- Identification of applicable MODES of Operation that were used.
- If shutdown risk can be quantified,then core damage frequency (CDF)and large early release frequency (LERF) will be calculatedfor shutdown risk and included in the cumulative risk of all changes assessed. Turkey Point (PTN) does not currently have a RG 1.200 shutdown model. As such the shutdown risk assessments will be based on the PTN shutdown safety program developed in support of NUMARC 91-06, an application-specificshutdown analysis, a bounding sensitivity analysis, or other acceptable method described in NEI 04-10, Revision 1.
- Justificationfor a qualitativeanalysis (if quantitative was not used).
" Shutdown risk will be included in the comparison to applicablethresholds.
APLA RAI-3 The enclosure to Attachment 2 of the LAR provides the peer review Facts and Observations (F&Os). Address the impact of the following F&Os on this application, clarifying the disposition of the F&O, as necessary:
DA-D5-01, DA-D6-01, DA-D6-02, IE-C14-01, IE-C14-02, IE-C14-03, IE-C14-04, IE-C14-05, IE-C14-06, and IE-C14-07.
Response
See table below.
Page 2 of 14
Florida Power & Light Company L-2015-046 Enclosure FPR For several CCF groups, a "global Two alternatives. The compiexe Lt,,r moaeiing upoate aouressing mis 2013 common cause event" (as described at missing CCF terms could issue is planned for the next internal events the end of Section 4.2 of PTN-BFJR- be added to the CAFTA model update. In fact, there is a working model 2008-012, Rev. 0) is used. While this is fault trees and CCF with the CCF update already implemented. If a a reasonable simplification, the global basic events calculated 5b application is started prior to the internal common cause event needs to account for the new terms. A events model update, the working model with the for the common cause combinations that simpler alternative is to CCF update will be used to perform a sensitivity are not included explicitly. However, for revise the calculation of analysis for the application.
several 6-component groups (AFW the a6 term to include AOVs FTO, AFW CVs FTO, AFW MOVs the missing a5 value.
FTO), the 5-of-6 term was not included Thus, a6' = a5 + a6. This and the 6-of-6 term was not adjusted. A overestimates the a5 similar issue appears to be present for contribution, since it is SG SVs FTO (4-component group), applied to the case where only the 4-of-4 term is included where all 6 components (the 2-of-4 and 3-of-4 terms are missing fail, but this should be a and the 4-of-4 term was not adjusted). small and conservative approximation. (Similar correction for the 4-component group, a4'=
a2 + a3 + a4).
Page 3 of 14
Florida Power & Light Company L-2015-046 Enclosure DA-D36-01 FPR The CCF notebook did not include a Review plant-specific During the review of plant records for the data 2013 review of plant failure data for common component failure events analysis used in the current model-of--record, no cause events. from the most recent common cause failure events were found. This data update to identify fact will be added to the CCF notebook.
any common cause failures. If CCFs are identified, verify that the CCF is modeled for the specific component and failure mode. If this data indicates a significantly larger fraction of failures are CCFs than the generic CCF parameters would predict, plant-specific CCF parameters should be calculated. If the data is limited (one or two failures in a specific component group), this would not be sufficient evidence to
- ustify plant-specific CCF
ýparameters.
Page 4 of 14
Florida Power & Light Company L-2015-046 Enclosure FPR Section 3.0 of the CCF Notebook rroviae a oasis Tor L.ors are inciuoeo Tor me components in me 2013 includes the assumption that CCFs are excluding CCFs from initiating event fault trees. For example, in the not included in fault tree initiating events system initiating events CCW system where 2/3 pumps are normally with year-long mission times due to and include CCFs where running, there are AND gates with a single FTR excessive conservatism in applying CCF a basis for exclusion event of one of the normally running pumps with factors that are developed for 24-hr cannot be established. an 8760-hour mission time and CCF events for mission time. However, this is not For example, include the other 2 running pumps with mission times sufficient basis for excluding CCFs for CCF in system initiating equal to the MTR of the pumps. There is not a fault tree IE models. event models only for CCF for all 3 pumps with a mission time of 8760 active components that hours, nor should there be; all 3 pumps are not are in the same normally running at the same time, and certainly configuration (i.e., not for 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />.
between normally This F&O will therefore not be a factor in 5b Dperating pumps in the applications.
same system but not between operating and standby pumps in the same system).
Page 5 of 14
Florida Power & Light Company L-2015-046 Enclosure rrir RCP TBHX rupture probability - The IE Assess the tube rupture mesomuion o[ mis r-,u is piannea Tor me nexi 2013 frequency for tube rupture is based on a original source data and internal events model update. In fact, there is a Reference 5 value of 3.48E-08/hr (peer whether it is applicable Norking model with the RCP TBHX rupture review did not verify this reference) for to each thermal barrier probability change already implemented. If a 5b "HX Tube External Leak Large >50 cooler/RCP. Revise application is started prior to the internal events gpm". This hourly frequency is initiator %ZZISLTBCCW model update, the working model with this change multiplied by 8760hr/yr for an annual IE and document any Nill be used to perform a sensitivity analysis for frequency of 3.05E-04/yr. Depending on changes or basis the application.
the application of the data, this IE accordingly.
frequency could be applied at each RCP, thus event tree top event "RCP TBHX Tubes Intact?" would be multiplied by a factor of 3. Applicability of the TBHX data to one or all RCPs should be examined/documented for impact on the total %ZZISLTBCCW initiator/results.
Page 6 of 14
Florida Power & Light Company L-2015-046 Enclosure manuai operator action is creoixea Tor r-vaiuaie ano aocumeni i ne Tact inai me pressure increase in me I.A.VV 2013 local manual closure of MOV-*-626 whether the operator system due to the TBHX tube rupture would be (should it fail to close) and/or to local action should be credited mitigated by the CCW surge tank expansion closure of manual valve *-736. Operator and remove credit for the volume and the relief valve RV-3/4-707 opening success ensures that the CCW piping action if it cannot be at 50 psig are obviously the reason some credit is remains intact. Although the HEP for the justified given to closing a valve to isolate the leak. The local action is 0.5, the time window basis time available for performing the isolation will should document to ensure that the depend upon the size of the rupture as well as operator has sufficient time to perform other factors.
these actions before the CCW piping This will be addressed in the next internal events boundary fails. model update. If a 5b application is started prior to the internal events model update, a sensitivity analysis will be performed by setting the HEP to 1.0 for the application.
Page 7 of 14
Florida Power & Light Company L-2015-046 Enclosure 2013 RCP Thermal Barrier CCW Supply the TBCCW supply examined in the next internal events model Penetration #3 - This penetration is not penetration for possible update; however, the risk impact if these evaluated for potential ISLOCA ISLOCA initiating events. penetrations are included in the ISLOCA model contribution. This penetration is Should also assess the will likely be minimal. The CCF of the two check protected by two normally open, active impact on CCW return valves to close is 5.2E-06. The frequency of a check valves (717 and 721ANB/C) inside line from RCP motor thermal barrier tube breach is likely less than 1 E-containment and two normally open cooling and lifting of RV- 03 per year, bringing us to a frequency of (1 E-03 MOVs (716AIB) outside 729 ifV-712A fails open. per year)*(5.2E-06) = 5.2E-09 per year. If no containment. The associated piping Ensure that these credit is assumed for the closing of the MOVs inside containment appears to be penetrations are also 716A1B, the ISLOCA will, at worst, fail the unit's designed for full RCS identified in Table 1, list CCW pumps. For a LOCA at PTN, all four HHSI pressure. However, given a thermal of penetrations. pumps start on the SI signal and inject to the barrier tube breach, the active check stricken unit, and the opposite-unit HHSI pumps valves could fail to close (w/CCF). The will not be affected by the loss of CCW.
active failure of the outboard MOVs Even if a CCDP of 1.0 is assumed, a delta CDF of (also w/CCF) may be highly unreliable 5.2E-09 per year will not be significant for any 5b due to low differential pressure design application.
capability and lack of relevant closure signals, and there might not be sufficient time for manual action. Failure of this penetration should be assessed for possible contribution to the TBCCW ISLOCA event frequency and sequences.
Page 8 of 14
Florida Power & Li, ]ht Company L-2015-046 Enclosure ISOAassmn fPntain1 Evaluate and document Resolution of this F&O is planned for the next 2013 (RH sutin SD lne)didno cosidr he RHR small ISLOCA internal events model update. In fact, there ise that the common suction piping beyond sequences taking no Norking model with this modeling change alreac the RHR pumps could be affected by the credit for associated Unit implemented. If a 5b application is started priol over-pressurization event. This would HHSI pumps and RWST. the internal events model update, the working impact the function of the high head SI model with this change will be used to perform a pumps and the RWST (and sensitivity analysis for the application.
Containment Spray pumps, which are not important in ISLOCA scenarios). As a result, the current RHR small ISLOCA event sequences apply too much credit for the associated Unit's RWST and HHSI pumps.
Page 9 of 14
Florida Power & Light Company L-2015-046 Enclosure reneirations oOIowiOu: kiIrici cola ieg rteview inese I ne reievanT Taiiure moae nere is a cnecK vaive 2013 injection) - These penetrations are penetrations and provide transferring open against the pressure that is qualitatively screened from further further basis for holding it closed - difficult to conceive of the detailed evaluation on the basis that screening. motive force causing such a failure. While
.... "the combination of three check perhaps not quite as secure of an isolation as valves is equivalent to three three, locked-closed isolation valves, the series of locked/closed isolation valves", for 3 closed check valves is considered to be meeting NUREG/CR-5928 criterion (c), adequate for screening out these penetrations.
systems isolated by redundant normally closed and locked manual valves that are independently verified to be closed and locked before plant startup". This comment is also applicable to Penetration 18. Additional basis is needed to support this equivalency assertion for screening these penetrations.
IE-C14-06 FPR Suggestion. Consider updating the This is a suggestion only. At the next opportunity, 2013 The PTN ISLOCA analysis is based on ISLOCA evaluation to the ISLOCA analysis will be updated to the latest early NUREG information and industry current industry practice guidance, but a sensitivity analysis for 5b is not practice, which continue to provide a and reference material. It necessary.
reasonable source of inputs/practice for is noted that there are consideration in ISLOCA modeling. In limitations in the WCAP-general however, the evaluation might 17154, Revision 1 benefit from aspects of the latest methodology and its industry ISLOCA best complete adoption is not practice/methodology presented in recommended.
WCAP-17154, Rev.1.
Page 10 of 14
Florida Power & Light Company L-2015-046 Enclosure IE--C14-07 FPR Suggestion. Consider updating the This iss a suggestion only, dealing with 2013 Table 1 "Potential ISLOCA Flow Paths" - ISLOCA report to dIocumentation only. A sensitivity analysis for ,5b Consider adding more detail in the ISL improve the details in is not necessary.
Screening Results column. For example, Table 1, primarily the Penetrations 13 and 14 (Letdown and column information Charging) may not cleanly screen. Both under"ISL Screening systems interface with low pressure Results" systems (letdown-purification piping and charging-pump suction). Typically there are redundant isolation means to isolate
- thus IE frequency should be low.
However, this cannot be concluded from the table details. Also, Penetration 3, "RCP CCW Supply" indicates that this penetration was screened based on "not connected to the RCS". However, this penetration provides the CCW supply to RCP thermal barrier cooling and should be assessed (refer to F&O IE-C14-2).
Page 11 of 14
Florida Power & Light Company L-2015-046 Enclosure APLA RAI-4 The NRC staff notes that the response to PRA Request for Additional Information (RAI) 22.01 (ADAMS Accession No. ML14113A176) associated with the licensee's request to adopt National Fire Protection Association (NFPA) Standard 805 (NFPA 805) provides the results of the Turkey Point Gap Analysis to the American Society of Mechanical Engineers (ASME) /
American Nuclear Society (ANS) Probabilistic Risk Assessment (PRA) Standard (ASME/ANS RA-Sa-2009) as endorsed by Regulatory Guide (RG) 1.200, Revision 2 (ADAMS Accession No. ML090410014).
Confirm whether the results of the assessment for F&Os LE-F1 -01 and LE-G5-01, as discussed in PRA RAI 22.01, are applicable to the LAR to relocate specific surveillance frequency requirements to a licensee controlled program. If the results are not applicable, address the impact of the F&Os for this application.
Response
See table below.
Page 12 of 14
Florida Power & Light Company L-2015-046 Enclosure LE-Fl-01 Endstate frequency totals are PDS relative Perform summary calculation This finding only addresses the given in Table 5 of the Level 2 contribution to to quantify PDS relative categorization of LERF results. This notebook, PTN-BJFR-99-010, LERF is not contribution to LERF. will be done in the next model Rev. 1, and results by release provided as update, but will have no effect on 5b category are given in Table 6. specified in the SR. applications.
However, results using the Plant Damage State definitions of Section 4.2 are not provided.
CC II is not met because relative contribution to LERF by PDS is not shown, although information is available to provide such data.
Page 13 of 14
Florida Power & Light Company L-2015-046 Enclosure limitations of severe accident intent of providing possible limitations of the uncertainty discussion of LERF understanding and modeling. a discussion LERF analysis based on, for results. This will be done in the next This includes such matters as regarding example, limitations on the model update, but will have no effect the impact of uncertainty limitations on the state of severe accident on 5b applications.
regarding thermally induced understanding of understanding and level 2 SGTR on quantification, the severe accident PRA analysis. Briefly describe uncertainty of ISLOCA break phenomenology, how key uncertainties in the size and location on timing and and how the Level LERF quantification could source term, and the 2 modeling impact risk-informed changes assignment of CET to uncertainties could to the licensing basis under endstates. Conservative impact LERF RG 1.174, for example.
treatment of some phenomena quantification and can affect LERF quantification, potential risk-which in turn impacts LERF informed and delta LERF results when applications.
applying RG 1.174 guidelines in risk-informed changes to the licensing basis, for example.
Page 14 of 14