L-MT-17-022, Response to Second Round PRA Related RAIs for the License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval
| ML17089A321 | |
| Person / Time | |
|---|---|
| Site: | Monticello  (DPR-022) |
| Issue date: | 03/29/2017 |
| From: | Gardner P Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CAC MF7359, L-MT-17-022 | |
| Download: ML17089A321 (14) | |
Text
2807 West County Road 75 Monticello, MN 55362 Xcel Energy@
RES P 0 N S I B L E BY NATURE 800.895.4999 xcelenergy.com March 29, 2017 L-MT-17-022 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Response to Second Round PRA Related RAis for the License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type A Test Interval (CAC No. MF7359)
References:
- 1) NSPM (P. Gardner) to NRC, "License Amendment Request: Revise Technical Specification 5.5.11 to Provide a Permanent Extension of the Integrated Leakage Rate (Type A) Test Frequency from Ten to Fifteen Years," (L-MT-16-001), dated February 10,2016 (ADAMS Accession No. ML16047A272 and ML16047A273)
- 2) NRC (R. Kuntz) to NSPM (R. Loeffler), "Request for Additional Information RE: Monticello license amendment request for ILRT extension (CAC MF7359)," dated September 9, 2016 (ADAMS Accession No. ML16256A004)
- 3) NSPM (P. Gardner) to NRC, "Response to Request for Additional Information:
License Amendment Request for a Permanent Extension of the 1 0 CFR 50 Appendix J Containment Type A Test Interval (CAC No. MF7359)," (L-MT-16-044), dated October 10, 2016 (ADAMS Accession No. ML16284A015)
- 4) NRC (R. Kuntz) to NSPM (R. Loeffler), "Monticello ILRT extension amendment Request for additional information (CAC No. MF7359)," dated November 18, 2016 (ADAMS Accession No. ML16323A242)
- 5) NSPM (P. Gardner) to NRC, "Part 1 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type A Test Interval (CAC No. MF7359)"
(L-MT-16-062), dated December 16, 2016 (ADAMS Accession No. ML16355A183)
L-MT 022 Page 2 of 3
- 6) NSPM (P. Gardner) to NRC, "Part 2 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type A Test Interval (CAC No. MF7359)"
(L-MT-17-002), dated January 31,2017
- 7) NSPM (P. Gardner) to NRC, "Part 3 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 1 0 CFR 50 Appendix J Containment Type A Test Interval (CAC No. MF7359)"
(L-MT-17-007), dated February 7, 2017
- 8) NRC (R. Kuntz) to NSPM (R. Loeffler), "Request for Additional Information RE: Monticello Nuclear Generating Plant Integrated Leak Rate Test Interval Extension Request (CAC NO. MF7359)," dated March 27, 2017 On February 10, 2016, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, submitted a license amendment request (LAR) proposing a change the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). The proposed change is to permanently revise the frequency specified in Specification 5.5.11 "Primary Containment Leakage Rate Testing Program," to increase the containment integrated leakage rate test (ILRT) program Type A test interval from 10 years to 15 years (Reference 1).
On September 9, 2016, the U.S. Nuclear Regulatory Commission (NRC) requested additional information pertaining to the primary containment performance history and a clarification of ILRT test results (Reference 2). The responses to these requests for additional information (RAis) were provided on October 10, 2016, in Reference 3.
From October 13 through 14, 2016, the NRC conducted a regulatory audit to gain a better understanding of the containment accident pressure risk assessment in the MNGP LAR. On November 18, 2016, the NRC requested additional information pertaining to probabilistic risk assessment (PRA) related considerations (Reference 4). The responses to RAis 2 and 3 were provided by letter on December 16, 2016 (Reference 5). The responses to RAis 1.b, 1.c, 5.c, 5.d, and 5.e were provided by letter on January 31, 2017 (Reference 6). The responses to the remaining PRA related RAis were provided by letter on February 7, 2017 (Reference 7).
On March 27, 2017, the NRC requested further clarification of the PRA analyses and provided a second round of PRA related RAis (Reference 8). The response to this request for additional information is provided herein.
L-MT-17-022 Page 3 of 3 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
I declare under penalty of perjury, that the foregoing is true and correct.
Executed on March.li_, 2017.
Peter A Gardner Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota
Enclosure:
Response to Second Round PRA Related RAis for the LAR for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type A Test Interval cc:
Administrator, Region Ill, US NRC Project Manager, Monticello Nuclear Generating Plant, US NRC Resident Inspector, Monticello Nuclear Generating Plant, US NRC State of Minnesota
ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO SECOND ROUND PRA RELATED RAIS FOR THE LICENSE AMENDMENT REQUEST FOR A PERMANENT EXTENSION OF THE 10 CFR 50 APPENDIX J CONTAINMENT TYPE A TEST INTERVAL
( 1 0 pages follow)
L-MT-17-022 Enclosure Page 1 of 10 RAI1 RESPONSE TO SECOND ROUND PRA RELATED RAIS FOR THE LICENSE AMENDMENT REQUEST FOR A PERMANENT EXTENSION OF THE 10 CFR 50 APPENDIX J CONTAINMENT TYPE A TEST INTERVAL As provided in the response to request for additional information (RAI) 6.a, the change in population dose of 1.69 person-rem/year or 2% of the total population dose is larger than the acceptance criteria from Section 3.2.4.6 of the Safety Evaluation for the Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2, of 1.0 person-rem/year or 1% of the total population dose, whichever is less. restrictive. The table provided in response to RAI 6.a indicates that EPRI Class 7 and ?a (severe accident phenomena-induced containment failures) are significant contributors to the change in population dose results.
In addition, according to the response to RAI 6.a [in Reference 7 in the cover letter], the change in the conditional containment failure probability (CCFP) of 1.66% is above the acceptance criteria of 1.5% from Section 3.2.4.6 of the Safety Evaluation for EPRI TR-1 009325, Revision 2.
Given that the reported results are above the criteria for small change, the NRC staff requests that additional quantitative evaluation be performed to determine if the acceptance criteria can be met. Provide additional evaluation of these criteria, a description of the technical evaluation with justification for the approach, and the updated results.
RAI 1 Response To reduce calculated values for Lldose and LlCCFP, the Internal Events (IE) and Fire probabilistic risk assessment (PRA) containment accident pressure (CAP) fault tree models were refined to increase the realism of the calculated additional risk from CAP. The following response describes these refinements. For an explanation of the assignment of risk values to Class 7 and ?a, see the response to RAI 3i.
Internal Events (IE) PRA Model Changes Modular Accident Analysis Program (MAAP) cases were run to demonstrate the impact of a large loss of coolant accident (LOCA) with a loss of CAP on the net positive suction head (NPSH) available to the low pressure emergency core cooling systems (ECCS). These MAAP cases demonstrated greater than three hours are available to establish containment cooling. This prevents the loss of low pressure ECCS pumps from inadequate NPSH for the most limiting Large LOCA case. The conservative assumption that low pressure ECCS fails for all Large LOCAs when CAP is not available, made previously in the response to RAI 4b (in Reference 7 in the cover letter), was modified to reflect actual plant conditions; this was done by retaining and crediting adequate NPSH by establishing containment
L-MT-17-022 Enclosure Page 2 of 10 cooling so that a Large LOCA would not fail low pressure ECCS outright due to lack of NPSH. Re-quantifying the IE PRA CAP model with this more realistic assumption resulted in improvement to the change in core damage frequency (LlCDF), change in large early release frequency (LlLERF), LlCCFP, and Lldose metrics.
Fire PRA Model Changes It was determined the change in Fire CDF due to a loss of CAP was artificially high due to unrealistic conseNatisms in the Fire PRA model leading to unrealistic increases in dose and CCFP. These conseNatisms relate to the time available to establish torus cooling and to establish alternate injection sources that do not depend on the torus, given an initial injection success and using operator actions. In certain fire-induced accident sequences, the Fire PRA model did not include credit for specific operator actions relating to torus cooling and establishing alternate long-term injection when initial injection succeeds. The results indicate that adequate time is available to perform the actions required to establish torus cooling or establish alternate injection with pumps that do not depend on the torus for suction water.
Thus, the Fire PRA model was adjusted to apply the existing operator actions to scenarios where credit could be taken for establishing torus cooling and alternate injection. The impact on overall Fire CDF and LERF was not significant, but the impact to the importance of CAP credit was reduced because the reliability of establishing torus cooling and establishing alternate injection increased, which produced more realistic estimates of the LAR risk metrics LlCDF, LlLERF, LlCCFP, and Lldose.
Refined IE and Fire Risk Metrics The overall LlCDF due to loss of CAP attributed to the integrated leak rate test (ILRT) suNeillance frequency change was refined. The resulting decrease in total LlCDF reduces the impact to both population dose and CCFP by decreasing the total change in frequency for the application. The table on the following page shows the refined IE and Fire PRA results for the risk metrics.
L-MT-17-022 Enclosure Page 3 of 10 Class 2
3a 3b 4-6 7
7a 8
Total Dose (person-rem) 7.29E+03 2.34E+06 7.29E+04 7.29E+05 O.OOE+OO 1.88E+06 3.87E+05 1.53E+06 ILRT Dose Rate LlTotal From 3 Years Dose Rate From 10 Years
%L1Dose From 3 Years Rate From 10 Years LERF Total LERF From 3 Years LlLERF From 10 Years CDF From 3 Years LlCDF From 10 Years CCFP%
From 3 Years LlCCFP%
From 10 Years ILRT Extension Summary Base Case Extend to Extend to 3 in 10 Years 1 in 10 Years 1 in 15 Years CDFNear Person-CDFNear Person-CDFNear Person-RemNear RemNear RemNear 1.25E-05 9.14E-02 1.15E-05 8.42E-02 1.08E-05 7.90E-02 1.77E-06 4.13E+OO 1.77E-06 4.13E+OO 1.77E-06 4.13E+OO 3.31 E-07 2.41E-02 1.12E-06 8.17E-02 1.69E-06 1.23E-01 8.23E-08 6.00E-02 2.79E-07 2.03E-01 4.20E-07 3.06E-01 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 4.22E-05 7.93E+01 4.23E-05 7.95E+01 4.23E-05 7.96E+01 3.47E-07 1.34E-01 5.10E-07 1.97E-01 6.28E-07 2.43E-01 8.33E-07 1.28E+OO 8.33E-07 1.28E+OO 8.33E-07 1.28E+OO 5.81 E-05 8.50E+01 5.83E-05 8.55E+01 5.85E-05 8.58E+01 N/A 0.443 0.763 N/A N/A 0.320 N/A 0.521%
0.898%
N/A N/A 0.375%
7.00E-06 7.20E-06 7.34E-06 N/A 1.94E-07 3.34E-07 N/A N/A 1.39E-07 N/A 2.62E-07 4.52E-07 N/A N/A 1.90E-07 N/A 0.437%
0.749%
N/A N/A 0.312%
As shown in this table, the change in population dose rate is 0.763 person-rem/yr, and the percent change in dose rate is 0.898%. As a result of additional quantitative evaluation performed, the criteria of population dose being less than 1.0 person-rem/year or 1% of the total population dose, whichever is less restrictive, is met. As shown in the table, the
L-MT-17-022 Enclosure Page 4 of 10 LlCCFP is 0.749%. As a result of the additional quantitative evaluation performed, the criteria of LlCCFP being less than 1.5% is met.
With the refinements made as described in this RAI response, the LlCDF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 4.52E-7/year. As such, the estimated LlCDF is determined to be "very small" using the acceptance guidelines of Regulatory Guide (RG) 1.174 (Reference C.5).
When totaling the seismic analysis from RAI 1 a (in Reference 7 in the cover letter)
(assuming guaranteed crack growth), and the IE and fire CAP analysis from this RAI response, the most conservative estimate (the 3 in 10 to 1 in 15 case) of LlLERF comes from adding the seismic LlLERF with the combined IE and Fire LlLERF:
LlLERF =seismic LlLERF +combined IE and Fire LlLERF = 4.42E-7 + 3.34E-7 = 7.76E-7 Total LERF comes from adding the seismic baseline LERF, seismic LlLERF (assuming guaranteed crack growth), combined IE and Fire baseline LERF, and combined IE and Fire LlLERF:
Total LERF =seismic baseline LERF +seismic LlLERF +combined IE and Fire baseline LERF + combined IE and Fire LlLERF
= 1.25E-6 + 4.42E-7 + ?.OOE-6 + 3.34E-7 = 9.03E-6 Based on the results of this evaluation, it is reasonable to conclude that the change in CDF and LERF are "small" per the RG 1.17 4 risk acceptance guidelines (Reference C.5).
RAI2 The response to RAI 4.e [in Reference 7 in the cover letter] indicates that change in large early release frequency (ilLERF) from containment accident pressure (CAP) is estimated using the probabilistic risk assessment (PRA) models for internal and fire events. This is a change in methodology from that presented in the license amendment request (LAR) in which ilLERF was assumed to be equal to the change in core damage frequency (ilCDF) for CAP.
The RAI responses did not provide sufficient information on the evaluation of L1LERF due to CAP. Comparing the reported L1LERF of 4.65E-7/yr (which includes non-CAP related contributors) to the CAP-related L1CDF of 2.47E-6/yr provided in the response to RAI 6.a [in Reference 7 in the cover letter] indicates that over 81% of CAP related change in CDF is classified as non-LERF. Given that the methodology for this application assumes a large pathway to the environment exists through a non-detected large pre-existing leak in the containment:
L-MT 022 Enclosure Page 5 of 10
- i.
Explain how the PRA model distinguishes between LERF and non-LERF for loss of CAP; Response to RAI2i The PRA model distinguishes between LERF and non-LERF for loss of CAP via the success or failure of release mitigation. Given core damage due to loss of CAP attributed to a large pre-existing leak (i.e., the Class 3b failure) and failure of torus cooling (which precludes the need for CAP to ensure adequate ECCS NPSH), the PRA modeled that a large-early release would occur unless release mitigation succeeds.
The LERF PRA model includes credit for release mitigation via drywell spray; successful release mitigation via drywell spray would scrub the fission product release and reduce the dose such that it is no longer large in magnitude. Failure of release mitigation via drywell spray would result in a large-early release. The LERF PRA model includes logic to model the loss of systems and components that can provide drywell spray via sources other than the low pressure ECCS pumps that depend on the torus as a suction source. This includes the Residual Heat Removal Service Water (RHRSW)
System, which takes suction from the ultimate heat sink and does not depend on CAP to ensure adequate NPSH.
ii.
Explain and justify why there is a significant 11LERF reduction from the 11CDF; Response to RAI 2ii The results in the response to RAI 6a (in Reference 7 in the cover letter) indicate a significant portion of CAP risk was classified as non-LERF. As detailed in the response to RAI1 in this submittal, the IE and Fire PRA CAP fault tree models were refined to increase the realism of the calculated additional risk from CAP. This resulted in a significantly lower 11CDF and slightly lower 11LERF; therefore, the difference between LlLERF and LlCDF is smaller. As stated in the response to RAI 1, the 11CDF due to the ILRT frequency change was reduced by applying credit for torus cooling and injection using sources that do not depend on the torus as a suction source; these model refinements were made to portions of the model where credit was conservatively not applied in the model that supported the results in RAI 6a (in Reference 7 in the cover letter).
There is a LlLERF reduction from the LlCDF because the LERF PRA model includes credit for release mitigation via drywell spray, which scrubs the fission products using pumps that do not depend on the torus as a suction source; this includes RHRSW System, which takes suction from the ultimate heat sink, as described in the response to RAI 2i. The 11LERF reduction is justified based on the MNGP plant-specific LERF analysis of the release magnitude given an un-isolated containment with a large pathway to the environment but successful drywell spray operation.
L-MT-17-022 Enclosure Page 6 of 10 111.
Discuss the mechanisms and considerations credited for reducing LERF to non-LERF; Response to RAJ 2iii As described in the responses to RAis 2i and 2ii, the LERF PRA model includes credit for release mitigation via drywell spray, with flow provided by the RHRSW systems, which takes suction from the ultimate heat sink. Drywell spray scrubs the fission products, which reduces the source term. As stated in the LERF analysis (Reference C.4), drywell spray is capable of scrubbing fission products from the containment atmosphere, which is especially important when containment fails, and has a significant effect on preventing the volatile and non-volatile fission products from being released to the environment. Reference C.4 documents crediting scrubbing in the LERF analysis.
iv.
Provide technical justification for those credited for ilLERF reduction and their assigned likelihoods in the PRA model.
RAJ3 Response to RAJ 2iv The technical justification for those methods credited for ilLERF reduction and their assigned likelihoods in the PRA model is based on the MNGP plant-specific LERF PRA analysis. Reference C.4 credits scrubbing in the LERF analysis. The assigned likelihoods for the systems that support the drywell spray release mitigation function are from the MNGP IE and Fire LERF fault trees that are part of the peer reviewed PRA models. The MNGP IE and Fire LERF PRA models represent the as-built, as-operated plant, including credit for drywell spray for release mitigation. The LERF PRA model has been peer reviewed to ASME/ANS-Ra-Sa-2009 with clarification from RG 1.200 Revision 2, as discussed in the PRA quality statement portion of the LAR (Appendix A of Reference C.6), and includes failure modes of the systems that support the drywell spray function and estimates of probabilities in accordance with the standard.
The NRC staff's review found further explanation is necessary regarding results reported in the responses to the RAis. Address the following:
- i.
In the LAR, Class 7 does not show a dependency on the integrated leakage rate test (ILRT) frequency; however, in the response to RAI 6.a [in Reference 7 in the cover letter], Class 7 and 7a have a dependency on the ILRT frequency which results in significant contribution to the ILRT risk results. Explain why this dependency has been introduced.
L-MT-17-022 Enclosure Page 7 of 10 Response to RAI 3i Regarding the increase in Class 7 risk from extending the ILRT interval; in some cases the resulting release due to a pre-existing leak is scrubbed by drywell spray, and prevented from becoming a large early release. This portion of the change in risk was assigned to Class 7, as Class 3b is intended to reflect large and early releases due to the pre-existing leak. Class 7 was used for this portion of risk because the mitigated release is small and early, and Class 7 is used to represent releases that are not part of the other EPRI classes. Class 7 was subdivided into Class 7 and 7a, to show the small early risk associated with a mitigated release from a pre-existing leak, as further described in the response to RAI 3ii.
- 11.
Class 3a represents small releases. If Class 3a and Class 7a both represent small releases with dependencies on the ILRT frequency, explain the difference between these categories in the evaluation of the change in CDF over the ILRT frequency.
Response to RAI 3ii Class 3a and Class 7a both reflect small releases, but the difference between the two classes depends on the physical containment failure and success or failure of release mitigation. Class 3a is a small release reflective of a 1 Ola leak in containment, given core damage, with associated dose rate, as calculated per the EPRI methodology of scaling the Peach Bottom dose to determine the dose for Classes 1, 3a, and 3b.
Class 7a is a small release reflective of a 100La pre-existing leak in containment with release mitigation success following core damage. The dose for this case is determined using MNGP-specific analysis for small-early dose (Reference C.4).
As stated in the response to RAI 3i, initially all non-LERF LlCDF was assigned to Class 7 because Class 7 was used to capture all LlCDF that was not reflective of the other EPRI classes. In order to provide a more refined dose estimate, the Class 7 risk was redefined to not include the small-early release state, because the small-early release state is representative of a core damage event with a large pre-existing leak in containment (i.e., the Class 3b failure) with successful release mitigation. The frequency and dose associated with a Class 3b pre-existing leak with successful release mitigation (quantified as a small-early release in the Level 2 PRA model) was assigned to Class 7a. Once the release category frequencies were appropriately correlated with population doses, the change in population dose was estimated and compared against the criteria.
Although the EPRI guidance does not provide specific guidance for evaluating CAP scenarios, the EPRI guidance provides precedent for sub-dividing Class 7 based on release characterization. Section 5.2 of Revision 2-A of EPRI 1009325 (Reference C.2)
L-MT 022 Enclosure Page 8 of 10 divides Class 7 into multiple sub-categories based on release timing and whether or not the release is scrubbed. Therefore, the EPRI guidance provides precedent for dividing Class 7 accident sequences into sub-categories for the purpose of appropriately correlating release category frequencies with population doses.
iii.
EPRI TR 1009325, Revision 2, guidance stipulates that external events, including fire, should be included in the risk assessment. The February 7, 2017 letter indicates that Fire PRA model changes and enhancements were made, but did not provide the new values for fire CDF and LERF. Please provide these updated values for the Fire PRA baseline risk.
Response to RAI 3iii The base Fire CDF and LERF with CAP failure values are 5.01 E-5 and 6.02E-6, respectively. These values are results from the refined Fire PRA model with CAP failure, as detailed in the response to RAI 1.
L-MT 022 Enclosure Page 9 of 10 IV.
EPRI TR 1 009325, Revision 2, uses the Regulatory Guide 1.17 4 change in LERF (L1LERF) acceptance guidelines for the ILRT extension. The response to RAI 6.a [in Reference 7 in the cover letter] did not provide a clear presentation of the various contributors to the risk estimates to the Class 3b ilLERF. Summarize in a table format the numerical contributions to the Class 3b ilLERF from each hazard/contributor (e.g.,
internal events, seismic, fire, other external hazards, steel liner corrosion, and containment accident pressure).
Response to RAJ 3iv The table below presents contributions to Class 3b ilLERF from each hazard/contributor.
Class 3b LERF and illERF by Hazard/Contributor Hazard/Contributor Class 3b: Base Case Class 3b: Extend to Class 3b ~LERF 3 in 10 Years 1 in 15 Years Internal Events 4.12E-09 2.16E-08 1.74E-08 Internal Events CAP1 5.42E-09 2.76E-08 2.22E-08 Fire 5.92E-08 2.96E-07 2.37E-07 Fire CAP1 1.36E-08 7.12E-08 5.76E-08 Seismic2 1.1 OE-07 5.52E-07 4.42E-07 Seismic CAP3
£
£
£ Other External Events4
£
£ E
Steel Liner Corrosion 1.1 OE-11 5.63E-11 4.47E-11 Total 1.92E-07 9.69E-07 7.76E-07 Note 1: These values include the refinements detailed in the RAI 1 response.
Note 2: This seismic LlLERF from RAI 1 a (in Reference 7 in the cover letter) response conservatively assumes guaranteed crack growth.
Note 3: As discussed in the response to RAI 4b (in Reference 7 in the cover letter) for seismic risk, the LlCDF due to CAP can be considered insignificant because seismic risk is driven by key structure and correlated system failures in which containment credit for CAP has no impact. For the less significant portion of seismic risk where key structures and systems survive, torus cooling can be initiated to preclude the need for an intact containment. If torus cooling is not available with all power failed (thus failing the Residual Heat Removal and Core Spray Systems), the FLEX strategy can be used to provide reactor inventory makeup and decay heat removal.
Note 4: The LlLERF from other external events is negligible. As stated in response to RAI 1 b (in Reference 6 in the cover letter) (Reference C.3), the external event contribution from external floods, high winds and tornadoes, transportation and nearby facility accidents, or other external hazards, remains small and falls well within the bounding assessment for external events impact used in the LAR such that there is no impact on the ILRT extension application.
L-MT-17-022 Enclosure Page 10 of 10 REFERENCES C.1 Calculation PRA-CALC-16-006, "MAAP Analysis for ILRT Extension," Revision 1.
C.2 Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA. 1018243, October 2008.
C.3 "Part 2 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type A Test Interval (CAC No. MF7359),"
January 31, 2017, ML17032A038.
C.4 Monticello Nuclear Generating Plant, Application for Renewed Operating License, Appendix E - Environmental Report, 2005.
C.5 An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, May 2011.
C.6 Calculation 54005-CALC-01, "Monticello Nuclear Generating Station: Evaluation of Risk Significance of Permanent ILRT Extension," February 2016, ML16047A273. of L-MT-16-001, "License Amendment Request: Revise Technical Specification 5.5.11 to Provide a Permanent Extension of the Integrated Leakage Rate (Type A) Test Frequency from Ten to Fifteen Years," ML16047A272.