Text

Response to Request to Additional Information Regarding RBS License Amendment Request to Extend Type a and Type C Test Frequencies
	ML16215A194
Person / Time
Site:	River Bend
Issue date:	07/27/2016
From:	Chase M; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RBG-47695
	Download: ML16215A194 (36)
	v • d • e

Entergy Operations, Inc.

River Bend Station

--- Enter~

5485 U.S. Highway 61N St. Francisville, LA 70775 Tel 225-381-3612 Marvin L Chase Director, Regulatory & Performance Improvement RBG-47695 July 27, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Response to NRC Request for Additional Information - RBS License Amendment Request to Extend Type A and Type C Test Frequencies River Bend Station, Unit 1 Docket No. 50-458 License No. NPF-47 Reference 1) Entergy Letter; License Amendment Request for change to Technical Specification 5.5.13 to be extended to 15 years, Drywall Bypass Test Frequency to 15 Years and Type C Test Frequency to 75 Months (RBG-47620) dated October 29, 2015 r

2) NRC Email; River Bend Station, Unit 1, Request for Additional Information - RBS License Amendment Request to Extend Type A and Type C Test Frequencies (NEI 94-01, Rev. 3-A) - TAC No.

MF7037, dated March 21, 2016

3) Entergy Letter; Response to NRC Request for Additional Information - RBS License Amendment Request to Extend Type A and Type C Test Frequencies (RBG-47675) dated April 19, 2016

4) NRC Email; Request for Additional Information - RBS License Amendment Request to Extend Type A and Type C Test Frequencies (NEI 94-01, Rev. 3-A) and Drywall Bypass Test frequency- TAC No.

MF7037, dated May 20, 2016

Dear Sir or Madam:

In Reference 1, Entergy Operations, Inc. (Entergy) submitted a request for an amendment to the Technical Specifications (TS) for River Bend Station (RBS), Unit 1. The proposed amendment modifies the existing requirements related to containment leak rate testing.

In Reference 2, the NRC staff requested additional information (RAI) in support of this request. This information was submitted in Reference 3.

RBG-47695 Page 2 of 3 In Reference 4, the NRG staff requested additional information (RAI) in support of this request. Attachment 1 provides responses to the RAI.

This letter does not contain commitments.

If you have any questions or require additio'nal information, please contact B. Burmeister at (225) 381-4148.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July,27, 2016.

Sincerely,

~~~

MLC/KYH/bmb

Attachment:

Response to Request for Information cc: Regional Administrator U. S. Nuclear Regulatory Commission, Region IV 1600 East Lamar Blvd.

Arlington, TX 76011-4511 NRG Senior Resident Inspector

. P. 0. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Attn: Mr. Stephen S. Koenick

MS 8 B1A One White Flint North 11555 Rockville Pike Rockville, MD 20852

RBG-47695 Page 3 of 3 Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312 Public Utility Commission of Texas Attn: PUC Filing Clerk 1701 N. Congress Avenue P. 0. Box 13326 Austin, TX 78711-3326 RBf 1-16-0078 LAR 2014-04

Attachment 1 RBG-47695 Response to Request for Information

RBG-47695 Attachment 1 Page 1of32 By application dated October 29, 2015 (Agencywide Documents Access and Managef'T!ent System (ADAMS) Accession No. ML15307A293), Entergy Operations, Inc. (Entergy, the licensee), submitted a License Amendment Request (LAR) for River Bend Station, Unit 1 (RBS). The LAR would revise Technical Specification (TS) 5.5.13, "Primary Containment Leakage Rate Testing Program," to incorporate Nuclear Energy Institute (NEI) Topical Report 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"' which would allow for the extension of the Type A Test (Integrated Leak Rate Test, or ILRT) and Type C Test (Local Leak Rate Test) frequencies from 1O to 15 years and ao to 75 months, respectively. Surveillance Requirement (SR) 3.6.5.1.3, would al~o be revised to extend the maximum interval for performing the Drywell Bypass Test (DWBT) from 1Oto 15 years in order to remain consistent with the proposed extended Type A Test frequency provided for in NEI 94-01 Revision 3-A. ,

The U.S. Nuclear Regulatory Commission (NRG) staff has determined that additional

information is required in order to complete its review of the LAR. This set of questions relates to the first and second proposed changes to extend the Type A ILRT and the DWBT as they are supported by risk information. The specific questions relate to the NRG Safety Evaluation Limitations and Conditions for EPRI Report No. 1009325, Revision 2 1 and for Regulatory Guide 1.1742

APLA RAl-1

1. The LAR, Attachment 3, Section 5.7, provides the evaluation of contributors from hazard groups other than the internally initiated events modeled in the Probabilistic Risk Assessment (PRA). Table 5.7-1 shows the evaluation of an "external events multiplier." As shown in Table 5.7-2, the Large Early Release Frequency (LERF) increase due to external events is derived from the LERF increase due to internal events times the external events multiplier.

a. For seismic events, the seismic risk analysis should consider the River Bend Mark Ill containment performance during a seismic event with potentially pre-existing flaw,s. A pre-existing flaw classified in Class_3a may grow due to the seismic event and may not remain a Class_3a flaw type for some seismic initiators. The external events multiplier method assumes that the initiating event has no impact on the flaw size, whereas a flaw may have growth

.potential due to seismic initiating event stresses prior to core damage occurring.

1 U.S. Nuclear Regulatory Commission, Final Safety Evaluation for Nuclear Energy Institute (NEI) 11 Topical Report (TR) 94-01, Revision 2, lndustry Guideline for Implementing Performance-Based 11 Option of 10 CFR Part 50, Appendix 3 and Electric Power: Research Institute (EPRI) Report No.

11 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (TAC No. MC9663), Accession Number MLOB1140105, June 25, 2008.

2 U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis", Regulatory Guide 1.174, Revision 2, May 2011.

RBG-47695 Attachment 1 Page 2 of 32 Provide an updated seismic risk contribution due to the ILRT frequency extension, accounting for the River Bend Mark Ill containment performance with potentially pre-existing flaws, given a seismic initiating event, and describe your method. Include in the discussion your technical justification for the method and results. Alternatively, periorm an appropriate sensitivity study to determine the risk significance of Class_3a flaws for the application due to seismic events.

b. The LAR not~s that the RBS seismic GDF used for the ILRT extension is one order of magnitude smaller than an NRC-e~timated seismic GDF; Determine whether, when using the NRG estimated seismic GDF, the risk acceptance criteria for the ILRT frequency extension can be met, otherwise provide the technical justification for reducing the seismic risk one order of magnitude.

RESPONSE

1a. While only a fraction of Class_3a flaws would grow to a Class_3b flaw due to a seismic initiator, a sensitivity study has been performed by conservatively assuming that all of the Class 3a seismic contribution also goes to LERF (i.e., is equivalent to Class 3b). To do this, the first step is to exclude the seismic contribution from the external events multiplier fromTable 5.7-1 of the LAR such that the seismic impact can be accounted for separately. The revised Table 5.7-1 is shown below.

)

Revised Table 5.7-1 Other Hazard Group Contributor Summary OTHER HAZARD INITIATOR GROUP CDF (1NR)

Seismic [8] N/A Internal Fire [9] 2.25E-05 Internal Flood [1 O] 4.97E-06 High Winds [11] 1.81E;;.()7 External Floods [9] Screened Transportation and Nearby Facility Accidents [9] Screened Total (for initiators with CDF available) 2. 77E-05/vr Internal Events CDF 2.60E-06 External Events Multiplier (Excluding Seismic) 10.64 For this bounding sensitivity case, the seismic impacts are calculated separately assuming the Class 3a contribution also goes to LERF (i.e., is equivalent to Class 3b). This change requires that the base calculations be re-performed rather than using a straight multiplier approach since the increase in the Class 3b frequency will also influence the calculated change in person-rem and the change in the

RBG-47695 Attachment 1 Page 3 of 32 conditional containment failure probability. Based on this revised conservative

assumption, Table 5.7-2, Table 5.7-3, and Table 5.7-4 from the LAA can be updated accordingly. The revised tables are shown below.

Revised Table 5.7-2 RBS 3b {LERF) as; a Function of ILRT/DWBT Frequency for Internal and External Events (Including Age Adjusted Steel Corrosion Likelihood)

38 38 38 LERF FREQUENCY FREQUENCY FREQUENCY INCREASE<1>

(3-PER-10 (1-PER-10 (1-PER-15 VEAR YEAR YEAR ILRT/DWBT) ILRT/DWBT) ILRT/DW8T)

Internal Events 6.03E-09 2.04E-08 3.11E-08 2.51E-08 Contribution Other Hazard Group Contribution (Internal 6.41E*08 2.16E*07 3.31E-07 2.67E-07 Events CDF x 1o.64)

Seismic Contribution 2.85E-08 9.53E*08 1.44E-07 1.15E-07 Combined 9.87E*08 3.32E-07 5.05E*7 ) 4.07E-07 1

( l Associated with the change from the baseline 3-per-1 O year frequency to the proposed 1-per-15 year frequency.

RBG-47695 Page 4 of 32 Revised Table 5.7-3 Comparison to Acceptance Criteria Including Other Hazard 1

Groups Contribution for RBS - \_

Contributor ALE RF APerson-rem/yr ACCFP RBS Internal Events 2.51 E-8/yr 7.22E-03/yr (0.72%) 1.15%

,r RBS Other *Hazard 2.67E-7/yr 7.68E-02/yr (0.72%) 1.15%

Groups RBS Seismic 1.15E-7/yr 1.91 E-02/yr (1.96%) 4.79%

RBS Total 4.07E-7/yr 1.03E-01/yr (0.81 %) 1.43%

Acceptance Criteria* <1.0E-6/yr <1.0 person-rem/yr or S1.5%

<1.0%

\.

-- In this bounding sensitivity case, all of the acceptance criteria are met and the bounding 4.07E-07/yr increase in LEAF due to the combined hazard events from extending the RBS ILRT/DWBT frequency from 3-per-1 O years to 1-per-15 years still falls within Region II between 1E-7 to 1E-6 per reactor year ("Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when the calculated increase in LEAF due to the proposed plant change is in the "Small Change" range, the risk assessment must also reasonably show that the total LEAF is less than 1E-5/yr. Similar bounding assumptions regarding the external event contributions that were made above are used for the total LEAF estimate.

Revised Table 5.7-4 Impact of 15-yr ILRT Extension on LERF (3b) for RBS Internal Events LERF 2.48E-08/yr Other Hazard Group LERF 2.63E-07/yr (Internal Events LERF x 10.64)

\ ...__

Seismic LERF 2.38E-08/yr Internal Events LERF due to* 3.11 E-08/yr ILRT (at 15 years) ')

Other Hazard group LERF due 3.31 E-07/yr to ILRT (at 15 years)

RBG-47695 Page 5 of 32 Revised Table 5.7-4 Impact of 15-yr ILRT Extension on LERF (3b) for RBS Seismic LERF due to ILRT (at 1.44E-07/yr 15 years)

Total 8.17E-07/yr As can be seen for this bounding sensitivity case, the estimated upper bound LEAF for RBS is estimated as 8.17E-07/yr, which is still less than the RG 1.174 requirement to demonstrate that the total LEAF due to internal and external events is less than 1.0E-5/yr.

In summary, the results of the bounding sensitivity case that conservatively assumes that all of the Class 3a seismic contribution also goes to LEAF indicated that the acceptance criteria would all still be met. This is a very conservative and bounding assumption as only a fraction of Class_3a flaws would grow to a Class_3b flaw due to.a seismic initiator, dependent on both pre-existing flaiJv size*

and magnitude of the seismic initiator.

1b. The NRG estimate of Seismic Core Damage Frequency (SCDF) of 2.5E-5/yr for River Bend from the Safety/Risk Assessment (SRA) [1] was based on the 2008 U.S. Geological Survey (USGS) seismic hazard curves and a very conservative estimate of the plant-level seismic capacity. The NRG used information from the

\ River Bend IPEEE submittal [9] to derive the plant-level fragility used to calculate the SCDF. River Bend was identified as a reduced scope IPEEE plant in

accordance with NUREG-1407 [2]. Thus, a reduced scope seismic margins assessment (SMA) was performed for \

the IPEEE. For plants that performed a reduced scope SMA for the IPEEE, the NRG used the plant Safe Shutdown Earthquake (SSE) as the-High Confidence of Low Probability of Failure (HCLPF) plant-level seismic capacity value since a HCLPF was not ceported. As such, NRG utilized a plant-level seismic capacity of just 0.1 g in the SRA for River

~ .

Bend.

To provide a better estimate of seismic risk at River Bend, Entergy assembled a Seismic Review Team (SRT) tasked with developing an SCDF estimate that more closely reflects the robustness of River Bend [8]. The SRT exqmined the assumption used by the NRG that the plant HCLPF is equal to the SS,E. The SRT calculated revised fragility values by two independent methods. It then selected the more conservative of the two results for use in re-assessing the SCDF. The SRT concluded that a plant-level HCLPF of 0.3g was more appropriate for estimating the seismic risk for River Bend. The SRT reproduced the NRC's reported SCDF results for the PGA, 1O Hz, 5 Hz, and 1 Hz 2008 USGS hazard curves. Then the SRT re-performed the calculations using a plant-level HCLPF of 0.3g instead of 0.1 g and estimated a revised SCDF of 2.5E-6/yr (one order of magnitude below the NRG SCDF weakest link value of 2.5E-5/yr). Using the same methods but with the 201 O EPRI hazard curves for River Bend [3], the SRT re-performed the analysis and estimated a revised SCDF of just 8.3E-7/yr [8].

HBG-47695 Page 6 of 32 To investigate the importance of the plant-level HCLPF assumption, the SCDF estimates for the 1 Hz, 5 Hz, 1O Hz, and PGA 2008 USGS hazard curves and corresponding weakest link estimates were re-performed over a range of HCLPF values. The results from the NRC analysis at 0.1 g are provided first as a reference to demonstrate that the revised analysis can reproduce the NRC methods. Then the results from the updated analysis over a range of HCLPF values are reported in the table which follows. As can be seen, even a modest increase in the assumed plant-level HCLPF value (to just 0.2g) results in an estimated SCDF value of about 5.0E-6/yr, and the revised estimated value with a pla_nt HCLPF at 0.3g is about an order of magnitude l~_ss than the NRC estimate.

2008 USGS Hazard Curve SCDF Estimates (/yr) Over a Range of HCLPF Values HCLPF-> 0.1g {NRC) 0.1g 0.15g 0.2g 0.25g 0.3g A 1 Hz 1.5E-5 1.5E-05 2.0E-06 5.2E-07 2.3E-07 1.4E-07 5 Hz 5.9E-6 6.0E-06 -2.1E-06 1.1 E-06 6.7E-07 4.7E-07 10 Hz 9.SE-6 9.9E-06 4.4E-06 2.4E-06. 1.5E-06 1.1E-::06 PGA 1.6E-5 1.6E-05 8.1E-06 4.9E-06 3.1E-06 2.1E-06 Weakest 2.5E-5 2.5E-05 8.9E-06 5.0E-06 3.2E-06 2.2E-06 Link A Estimated HCLPF value established applicable to RBS.

In 2014, Entergy submitted a response ~o the NRC request for information pursuant to 10 CFR 50.54(f) regarding recommendation 2.1 of the near term task force review of insights from the Fukushima Dai-ichi accident [4]. The results of the seismic screening evaluation were successful and no further seismic evaluations need to be performed for River Bend. In that assessment, a site-specific control point hazard curve for a broad range of spectral accelerations was computed given the site-specific bedrock hazard curve and site-specific estimates of soil and soft-rock response and associated uncertainties. When these more recent 1 Hz, 5 Hz, 1O Hz, and PGA Hazard curves are utilized, updated SCDF estimates as a function of assumed plant-level HCLPF are provided below. As can be seen in this case, even a modest increase in the assumed plant-level HCLPF value (to just 0.2g) results in an estimated SCDF value of less than 2.SE-6/yr, and the revised estimated value with a plant HCLPF at 0.3g*is close to the SRT estimate [8] of 8.3E-7/yr using the 2010 EPRI curves.

_p j

RBG-47695 Page 7 of 32 2014 Hazard Curve SCDF Estimates (/yr) Over a Range of HCLPF Values HCLPF-> 0.1g 0.15g 0.2g 0.25g 0.3g A 1 Hz 6.2E-06 1.8E-06 ' 8.0E-07 4.3E-07 2.7E-07 5 Hz 4.7E-06 1.7E-06 7.SE-07 4.1E-07 2.SE-07 10 Hz 4.7E-06 1.8E-06 8.SE-07 4.8E-07 3.0E-07 PGA 1.0E-05 4.4E-06 2.3E-06 1.3E-06 8.3E-07 Weakest 1.1E..;05 4.4E-06 2.3E-06 1.3E-06 8.6E-07 Link A Estimated HCLPF value established applicable to RBS.

Additionally, note that none of these values account for the risk mitigation capabilities of RBS "FLEX" equipment, implemented in response to NRG Order EA-.12-049. The seismic contribution to both GDF and LEAF is reduced when these risk mitigation capabilities are considered.

Based on these assessments, accounting for the most recent information and

. additional plant capabilities that now exist to respond to a seismic event, an upper bound estimate for SCDF at River Bend of 2.5E-06/yr (consistent with the LAA) is reasonable, and the actual value is expected to be a much lower value of 8.6E-07/year when the weakest link calculation is performed for a plant level HCLPF of 0.3g with the latest hazard curves. When this value is used for the seismic GDF, then even the bounding sensitivity for Class 3a above does not significantly challenge the acceptance criteria as shown in the revised Table 5.7-3 below.

Revised Table 5.7-3 Comparison to Acceptance Criteria Including Other Hazard Groups Contribution for RBS (Updated Hazard Curve and 0.3g Plant Level HCLPF)

) Contributor LiLERF Li Person-rem/yr LiCCFP RBS Internal Events 2.51 E-8/yr 7.22E-03/yr (0.72%) 1.15%

RBS Other Hazard 2.67E-7/yr 7.68E-02/yr {0.72%) 1.15%

Groups

RBG-47695 Page 8 of 32 RBS Seismic 3.96E-8/yr 6.57E-03/yr (1.96%) 4.79%

(Updated)

I RBS Total 3.31E-7/yr 9.05E-02/yr (0.75%) 1.25%

Acceptance Criteria <1.0E-6/yr <1.0 person-rem/yr or S1.5%

<1.0%

APLA RAl-2

2. The LAR Table 5.3-2 contains the following note:

"(3) The DWBT leakage cases of 1Ox and 1OOx with unit coolers unavailable are assumed to lead to an increased frequency of Class 7 (non-LEAF)."

Provide justification for not increasing the Class 7 frequency of LERF also, or update the Class 7 frequency and the risk results for the application.

RESPONSE

Note (3) from Table 5.3-2 refers to the assumption described in Section 5.1 of the LAR for~

Class 7 sequences. That is, for the core damage scenarios that previously resulted in an.

intact containment, it was assumed that these DWBT leakage rates could lead to containment failure if unit coolers are unavailable. The assignment to non-LERF was based on the containment capacity where an extended time would be available before failure would occur if the unit coolers were unavailable. With drywell bypass per Technical Specifications, operation of both containment unit coolers is capable of preventing containment failure in transient scenarios in the RBS PRA.

A sensitivity analysis has been performed to demonstrate that the assumption of excluding Class 7 DWBT cases from Class 3b LERF cases does not affect the conclusions of the analysis. In the LAR, the Class 3b contribution excluded the assumed "Class 7 (Non-LERF)" frequency increase as indicated by the "Class 7DWBT" designator below.

Class_3b = 0.0023 *[GDF - (Class 2 +Class 7 LERF +Class 8 +Class 7 0 wsr)]

= 0.0023 * [2.60E (6.64E-1.1 + 5.35E-09 + 1.93E-08 + 1.24E-09)]

=5.91 E-09/yr To address the impact of this assumption, the Class 3b (and Class 3a) contributions can be revised to not exclude the contribution from "Class 7DWBT' as indicated in the Class_3b example below.

Class_3b = 0.0023 * [GDF - {Class 2 + Class 7 LERF + Class 8)]

= 0.0023 * [2.60E (6.64E-11 + 5.35E-09 + 1.93E-08)]

=5.91 E-09/yr

RBG-47695 Page 9 of 32 As can be seen, this has a negligible impact on the overall results (i.e., it is below the resolution of the significant digits displayed), and therefore has a very negligible impact on the results.

APLA RAl-3

3. In the LAR Figure 4.1-1 the highest leakage from the drywell boundary is assumed

to be 1OODWLti, consistent with the limitations and conditions noted in the NRG safety evaluation report dated June 25, 2008 (ADAMS Accession No. ML081140105) for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2.

This represents~an increase from 35DWLb, the highest drywell leakage assumed for the one-time DWBT extension license amendment request dated February 16, 2004. The NRG staff safety evaluation for the one-time drywell bypass test extension (ADAMS Accession number ML043200567) stated that for events in which containment unit coolers operate, drywell leakage was assumed to have no impact on the containment's existing leakage category, since the containment coolers would condense any steam that bypasses the suppression pool. Address how the increase in drywell leakage rate from 35DWLb to 1OODWLb impacts the assumed containment leakage categories and include updated risk assessment if necessary.

RESPONSE

Note that the overall approach for the ILRT/DWBT extension uses a bounding approach which includes several conservatisms .. The DWLti of 800 scfm was chosen as a reference leakage value for consistency with the prior DWBT extension requests for River Bend.

This was conservatively chosen even though none of the DWBTs performed at River Bend have ever exceeded this value. It should also be noted that this reference value is far greater than the allowable containment leakage rate, La of 138,434 seem (i.e., < 5 scfm) as defined for the River Bend ILRT acceptance criteria [5] for which the accepted multipliers of 10 (for Class 3a) and 100 (for Class 3b) were derived. Additionally, per TS SR 3.6.5.1.3 and USAR Section 6.2, the current acceptable A/kv2 design drywell bypass leakage area is 0.81 ft2 at 3 psid, which corresponds to a flow rate of approximately 32500 scfm (based on 1 ft2 equating to 4011 O scfm in the 2007 DWBT submittal for River Bend

[6]).

As noted above, the allowable drywell bypass leakage has considerable margin compared to the allowable containment leakage. The frequencies used for the small and large leakages were also separately derived in Section 4.6.1 of the LAR compared to the frequencies utilized for the containment leakage multipliers for Classes 3a and 3b.

Therefore, in retrospect, it is inappropriate to use the same multipliers on the drywell leakage rate as is done on the containment leakage rate. Figure 4.6-1 from the LAR is reproduced below.

RBG-47695 Page 10 of 32 10.00 - , - - - - - - - - - - - - - - - - - - - - - - - - - - ,

8.00 - i - - - - - - - - - - - - - - - - - - - - - - 1 6.00 + - - - - - - - - - - - - - - - - - - - - - - !

. 4.00 +----------------------*

2.00 +------",,,IP~!'ll!v*,._,v _ _ _ _ _ _ _ _ _ _ _ _ _- i 0.00 **

A6

6 ***

..,_....,,*.,"'"""~*----,...--~"""--"f,~----"l"-r-~---.-----1 0 5 10 15 20 25 30 Figure 4.6-1 Mark Ill DWBT Results Compared to 800 SCFM In the LAR, the two events above the line were conservatively applied to the i1 Ox category even though none of the values exceeded a factor of 4. If 4x for that category is used as a more reasonable estimate for the multiplier to apply, then correspondingly a value of 40x is more reasonably applied as an upper bound estimate to the large leakage category represented by the Jeffrey's non-informative prior likelihood value.

Consistent with the LAR, when the three revised data points (i.e., > 1Lb, 4Lb, and 40Lb) are plotted on a curve, the trend still appears reasonable as shown in revised Figure 4.6-

2. These values are therefore used to provide a more reasonable representation for the base case assessment to represent the OW bypass leakage behavior. Increases to these values are assumed to occur for the different test intervals consistent with the ILRT methodology. The refinement to the drywell leakage values does not change the results of the frequency analysis, but provides additional justification that the larger drywell leakage category is less than the allowable design leakage (i.e., the upper bound 40x DWLb value is less than allowable design leakage). Therefore, a more reasonable upper bound.of 40x DWLb is very close to the 35 DWLbused in the prior assessment. As such, the analysis is consistent with the prior DWBT analysis and . associated SER (ADAMS Accession number ML043200567) which stated that for events in which containment unit coolers operate, drywell leakage can be similarly assumed to have no impact on the containment's existing leakage category, since the containment coolers would condense any steam that bypasses the suppression pool.

RBG-47695 Attachment 1 Page 11of32 0.30 ~---------------------

Q.25 --l--*"'.11.-----------------------*-*-*---*------

n20 + - 1 r - - - - - - - - - - - - - - - - - - - - - - -

0.15 *----**-*------*--------

0:10.-+_ ____,,._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~--

0~00 +,_-*--~-~---..----r-~--,------;.----,----,

0 5 10 15 :25. 3o .35 Revised Figure 4.6-2 Estimated Mark Ill DWBT Leakage Probability Compared to 800 SCFM APLA RAl-4

4. The LAR Section 5. 7 of Attachment 3 to the tAR discusses external events designated as "other" such as external floods and transportation and nearby facility accidents. The LAH references the results from the Individual Plant Examination of External Events (IPEEE) which concluded that no undue risks are present that might contribute to CDF with predicted frequency in excess of 1E-6/yr. However, since the I PEEE is outdated, assess these external events for the current plant, and discuss your assessment for the ILRT extension application.

RESPONSE

In 2014, Entergy submitted a flood hazard re-evaluation report in response to the NRC request for information pursuant to 10 CFR 50.54(f} regarding recommendation 2.1 of the near term task force review of, insights from the Fukushima Dai-ichi accident [7]. In 2015, NRC provided an assessment of the flood hazard reevaluation report (FHRR) submitted

/ . by Entergy as well as supple!'Tlental information resulting from requests for additional information and audits [12]. Part of the NRC's assessment indicates the following for River Bend:

The NRC staff has concluded that the licensee's reevaluated flood hazards information, as summarized in the. Enclosure, is suitable for the assessment of mitigating strategies developed in response to Order EA 049 (i.e., defines the mitigating strategies flood hazard information described in guidance documents currently being finalized by the industry and NRC staff), for River Bend. Further, .the staff has concluded that the licensee's reevaluated flood hazard information is a suitable input for other

RBG-47695 Page 12 of 32 assessments associated with Near Term Task Force Recommendation 2.1 "Flooding". The NRC staff plans to issue a staff assessment documenting the basis for these conclusions at a later time.

In addition, NEI 12-06 "Diverse and Flexible Coping Strategies (FLEX)

Implementation Guide" is currently being revised. This revision will include a methodology to perform a Mitigating Strategies Assessment (MSA) with respect to the reevaluated flood hazards. Once this methodology is endorsed by the NRC, flood event duration parameters and applicable flood associated effects should be considered as part of the River Bend MSA. The NRC staff will evaluate the flood event duration parameters (including warning time and period of inundation) and flood-related associated effects developed by the licensee during the NRC staff's review of the MSA.

The River Bend MSA will follow the guidance in Appendix G of NEI 12-06, Revision 2 [13]

which was issued in December 2015 to ensure that appropriate mitigating strategies exist to deal with the new flood hazard information. These mitigating strategies combined with the very unlikely nature of these types of events helps to ensure that the risk from external flooding impacts at River Bend remains low.

As recommended by Generic Letter 88-20, Supplement 4, the IPEEE employed a methodology for analyzing other external events at River Bend Station which was a screening approach. The first step in the screening approach was to determine, if the criteria of the 1975 Standard Review Plan (SAP) were met. The RBS IPEEE screened external events due to transportation accidents as well as due to accidents at nearby industrial facilities.

RBS design processes continue to assure that there is no adverse impact on the original design basis regarding transportation accidents or accidents at nearby industrial facilities.

The only major change to areas near the plant since the IPEEE was the opening of Louisiana State Highway 1O south of the plant, leading to the John James Audubon Bridge across the Mississippi River, and turning U.S. Highway 61 from a two-lane to a four-lane highway. State Highway 1O runs roughly one mile south of the plant, leading to the John James Audubon Bridge which opened in May 2011. This is roughly the same distance as U'.S. Highway 61 is to the northeast of the plant. The distance to both roads exceeds the acceptance criteria on distance of about 1700 feet per Regulatory Guide 1.91, Rev.1, Figure 1 or about 1500 feet per Regulatory Guide 1.91, Rev.a, Figure 2 for Tornado Region I which is applicable to RBS per Regulatory Guide 1.76.

SAR Figure 2.2-1 shows industrial firms and major transportation routes within 5 miles of River Bend Station. The IPEEE documented that the nearest railroad to the plant at the time of plant licensing was the Illinois Central Gulf line, which passed through the plant site. The spur of this line passing through the RBS site is no longer in service; the tracks crossing Parish Road 965 south and west of the plant no longer exist. Consistent with this, there are no railroad crossings on the Highway 1O approach to the Audubon Bridge or on Parish Road 964. The Illinois Central Gulf spur to the Hood Container* (formerly Crown Zellerbach at the time. of the IPEEE) Mill at the end of Parish Road 964 is currently decommissioned. The Kansas City, Southern Louisiana, and Arkansas Railway line

RBG-47695

) Attachment 1 Page 13 of 32 discussed in the IPEEE and shown in SAR Figure 2.2-1 has had its Highway 61 crossing deactivated. Thus, the closest active railway to the plant is the spur to the Big Cajun power plant, described in the IPEEE. This is greater than 3 miles from River Bend Station, across the Mississippi River; this exceeds the roughly 2500 foot distance criteria for boxcars of Figure 1 of Regulatory Guide 1.91, Rev.1, Figure 1 as well as the 3000 foot boxcar criteria of Figure 2 of Regulatory Guide 1.91, Rev.a, for Tornado Region I.

No new industrial facilities have been built in the vicinity of RBS since the IPEEE. FAA Aeronautical Charts for the Baton Rouge area were reviewed, which indicated no new pathways proximate to the River Bend site. SAR Figure 3.5~6 shows FAA aircraft pathways near River Bend Station.

The IPEEE section on Transportation and Nearby Facility Accidents was reviewed by the West Feliciana Parish Emergency Operations Coordinator and by the RBS Emergency Planning department. No additional hazards to the plant beyond those addressed in the IPEEE were identified.

As documented in letter RBG-47618 to the NRC dated September 29, 2015, RBS has completed the required actions and is in full compliance with NRC Order EA-12-049 for Mitigation Strategies for Beyond Design Basis External Events. Implementation of these "FLEX" actions increases mitigation capabilities to restore or maintain core cooling, containment, and spent fuel pool cooling capabilities in the event of a beyond-design-basis external event, thus will significantly reduce the risk associated with such events.

In summary, the contribution to external events risk from external floods and transportation and nearby facility accidents is still judged to be 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.

APLA RAl-5

5. If the evaluations or updates in RAls 1 through 4 resulted in changes to the LAR results, provide the updated cumulative risk results for the application.

RESPONSE

The results of the bounding sensitivity case in response to APLA RAl-1 a and APLA RAl-1b indicate that the acceptance criteria are still met. The responses to the other RAls were shown to have very negligible impact or not require any changes to the assumptions in the LAR.

APLA RAl-6

6. Appendix A to Attachment 3 to the LAR discusses the peer review of the internal events PRA.

a. Confirm that the 2011 peer review was a full scope peer review.

b. The LAR, Section A.2.4, states that the peer review team generated 59 findings. However, only 29 findings are provided in the LAR. Please provide

RBG-47695 Page 14 of 32 J

the remaining findings (or observations) other than suggestions, following the same tabular format as in Appendix A to Attachment 3 to the LAR.

RESPONSE ,

a. The 2011 BWROG peer review of the RBS PRA was a full scope peer review of the RBS internal events PRA, incluping internal flooding. ,

b. There were a total of 59 Findings identified during the 2011 RBS PRA peer review.

Consistent with what was observed in other License Amendment Requests for extending Integrated Leak Rate Test (ILRT) frequency to 15 years on a permanent basis, Table A.2-1 of Appendix A to Attachment 3 of the RBS submittql documented the 29 open findings, the status of the resolution for .each finding, and the potential impact of each finding on this application.

Table RAl-6.1 below provides this information for the 30 Findings which River Bend has resolved and which are considered closed. Note ~hat the RBS LERF model is a NUREG/CR-6595 model, which is defined as Category I per the PRA Standard; thus, since the RBS PRA was assessed against Category II of the standard, seven of these Findings document that the RBS PRA uses a NUREG/CR-6595 LERF model. 1

RBG-47695 Page 15 of 32 Table RAl-6.1 Summary of Industry Peer Review Findings for the RBS Internal Events PRA Model Update (Closed Findings from 2011 PRA Peer Review)

Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT 6.1.1 SY-All INCLUDE in the system model those Based on a sampling review of system In the RPS model for mechanical failure to Revise the RPS fault The RBS model does include Common Cause Failure of the failures of the equipment and components notebooks (PRA-RB-01-002Sll) and scram, the only failures that are considered are tree model to consider control rods to insert, with a probability of 2.SE-07. This (Met) that would affect system operability (as the CAFTA PRA model, it was those that affect the SDV valves. In reality other the complete spectrum plant specific CCF calculation is considered more identified in the system success criteria), confirmed that the system models failure modes (including mechanicalbinding of of possible failure applicable to RBS than the generic NUREG/CR-5500 value.

except when excluded using the criteria in include failures of equipment and the control rods themselves) maybe more mechanisms for the Accounting for the 2.lE-06 probability of mechanical SY-A15. This equipment includes both components that would affect system likely. NUREG/CR-5500 Volume Ill, for RPS and the control - common cause failure of the Reactor Protection System active components (e.g., pumps, valves, and operability. The equipment included example, estimates control rod binding for a rods. from NUREG/CR-5500 has only a miniscule impacton air compressors) and passive components both active and passive components. BWR at 2.lE-6. Inclusion of this additional calculated core damage frequency. Using the cutsets pre-(e.g., piping, heat exchangers, and tanks) failure mode would increase the computed generated for MSPI purposes with a E-13 truncation limit, required for system operation. However, in the RPS model for failure probability of the RPS by a significant the probability of corresponding basic event C71-CRD-CF-mechanical failure to scram, the only amount. CTROD was adjusted from 2.SE-07 to 2.lE-06. The failures that are considered are those resulting CDP increased from 2.642E-06 to 2.688E-06, a that affect the SDV valves. In reality, " 1.7% increase. Thus, this finding questioning the modeling other failure modes (including of common cause mechanical failure probability in the RBS mechanical binding of the control rods ~ PRA does not impact the ability of the RBS Rev.5 PRA themselves) may be more likely. model to be used.

NUREG/CR-5500 Volume Ill, for example, estimates control rod binding Note the applicable Supporting Requirement from the for a BWR at 2.lE-6. Inclusion of this Standard was judged to be Met.

additional failure mode would increase the computed failure This closed Finding does not impact the RBS ILRT probability of the RPS by a significant extension request.

amount.

6.1.2 IFQU-AlO For each flood scenario, REVIEW the LERF This is a finding because the technical While the LERF model is used to quantify LERF Review the LERF At the time of the RBS Rev.5 PRA peer review, the Internal analysis to confirm applicability of the requirements were not met. impacts due to flooding, there is no discussion model to ensure that Flooding PRA remained based in the previous Rev. 4 PRA.

(Not Met) LERF sequences. in PRA-RB-01-006 that the non-flood LERF no new flood-related RBS has subsequently (2012) re-performed the internal If appropriate LERF sequences do not exist, model was reviewed to determine if any LERF scenarios are flooding quantification using Revision 5 of the RBS PRA.

MODIFY the LERF analysis as necessary to changes were necessary to consider unique created. If new At that time, the RBS LERF model was reviewed for account for any unique flood-induced -flooding impacts. c scenarios are potential impacts due to Internal Flooding. It was scenarios or phenomena in accordance with necessary, then update determined and documented that no LERF model changes the applicable requirements described in 2- It is possible that new LERF scenarios would be the LERF model. In were required for the Internal Flooding PRA.

2.8. necessary (e.g., for non-recoverable SBO). any case, document Therefore, it is necessary that the LERF model the review. This finding is closed and has no impact upon the RBS be reviewed to ensure that no changes are ILRT Extension Request.

necessary and to document that review.

6.1.3 IFQU-A6 For all human failure events in the internal This is a finding because the technical PRA-RB-01-006 Appendix A documents a Review the in-control In-control room actions were assumed to have the same flood scenarios, INCLUDE the following requirements for in-control room review of existing HFEs in the internal events room operator actions HFE probabilities for flooding as for Internal Events based (Not Met) scenario-specific impacts on PSFs for operator actions have not been model to determine if modifications are needed to assess the flooding on operator interviews conducted specifically in support of control room and ex-control room actions performed. to reflect flooding conditions. For actions impacts pertaining to the Internal Flood PRA. This is documented in the as appropriate to the HRA methodology outside the control room, affected events are set workload, stress, and flooding quantification calculation and in the HRA being used: to true (failed) which would be conservative. impacts on indication. calculation. Flooding specific operator actions credited in

RBG-47695 Page 16 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT (a) additional workload and stress (above However in-control room actions are assumed the PRA have been explicitly addressed in the RBS HRA that for similar sequences not caused by to be unaffected and the evaluation does not calculation based upon operator interviews. Only limited internal floods) consider additional workload or stress, or the credit is taken in the RBS Internal Flood PRA for operator (b) cue availability potential for control room indications to be actions in response to sump level indications and the (c) effect of flood on mitigation, required impacted by the flood. IFPRA currently does not credit any sump pumps, one of response, timing, and recovery activities the conservative simplifications in the model. Thus, this (e.g., accessibility restrictions, possibility item is considered to be fully addressed in the RBS PRA of physical harm) and to have no impact on the Internal Flooding PRA (d) flooding-specific job aids and training results; (e.g., procedures, training exercises)

This finding does not impact the ILRT Extension Request.

6.1.4 IFEV-B2 IFEV-AS: DETERMINE the flood initiating This is a finding since the Flooding initiating event frequencies are Document the basis This Finding has been addressed through improved .

(Met) event frequency for each flood scenario requirements of this SR are not met. documented in the individual flood zones for the frequencies documentation as part of the 2012 update to the Internal group by using the applicable contained within PRA-RB-01-004 revision 0. that were applied and Flooding PRA. The flooding quantification calculation IFEV-A5 requirements in 2-2.1. However, adjustments to initiator frequencies for exclusions of addresses that operator actions to isolate any failures are are made based on judgment with only limited certain break sizes. not accounted for in initiating event frequencies. The (Not Met) IFEV-B2: DOCUMENT the process used to discussion of the basis. Also, scenarios that Update initiator Internal Flooding Analysis document, Section 3.2, identify applicable flood-induced include failure of operator isolation as part of frequencies_ if addresses adjustments in failure frequency for low risk initiating events. For example, this the initiator frequency should be explicitly- necessary. CNS fiberglass piping, and Section 3.1.6 documents the documentation typically includes addressed. review of RBS internal flooding operating experience that (a) flood frequencies, component does not call into question use of EPRI pipe failure unreliabilities/unavailabilities, and HEPs frequencies.

used in the analysis (i.e., the data values unique to the flooding analysis)

Thus, this finding is Closed and does not impact the ILRT (b) calculations or other analyses used to Extension Request.

support or refine the flooding evaluation (c) screening criteria used in the analysis 6.1.5 HR-D5 ASSESS the joint probability of those This is a finding because the technical Dependent pre-initiator dependencies have Reevaluate the HEP The statement of the peer review report that certain pre-HFEs identified as having some degree of requirements of the SR are not met been assessed. However, several independent estimation initiator human error event assessments are non-(Not Met) dependency (i.e., having some common pre-initiator human error event assessments are methodology such conservative is an incorrect statement. The ASEP elements in their causes, such as performed non-conservative and judged to be evaluated that stated guidelines methodology requires separate inputs on whether or not by the same crew in the same time-frame). too low by RBS analysts. are being followed. there is a daily or per shift parameter check separately for determination of the Basic Human Error Probability For example, in worksheet hfe_a.xls, the (BHEP) and in a separate calculation for the duration that independent event B21-XHE-MC-V658A twice such an error might be in effect.

credits the 'Status Check Each Shift or Day'.

Once in the ASEP screening questions leading Additionally, the events such as B21-XHE-MC-V658A to ASEP case VII and a second time in the which have low calculated probabilities are not directly

'Adjustment for Average Unavailability'. This used in the PRA. Rather, these are events which are double counts potential recovery actions developed as part of the calculation of common leading to a very low estimate of the miscalibration included in the PRA model. It was independent event at -1.3 E-:07. confirmed that, with a single exception which is being fixed, all these common miscalibration events specified a EN-NE-G-013 (HRA) specifies a minimum floor probability value of l.OE-06 in the RBS PRA database, individual HEP of lE-5 and a combined (joint) consistent with methodology for combinations of events.

HEP of lE-6. Therefore, an independent value For example, event B21-XHE-MC-V658A would be an assessed at 1.3 E-07 deviates from the Entere:v input into calculation of event B21-XHWE-MC-N058 for

RBG-47695 Page 17 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT Guidance Document. miscalibration of ATWS RPT sensors B21-PTN058A, B, E, and F; however, despite the individual miscalihration event Also, the NUREG on good HRA practices is having a calculated probability of l.3E-07, the common NUREG-1792. This NUREG states the miscalibration event has an applied probability of l.OE-06 following: in the PRA database. The peer review finding had overlooked this adjustment of the values actually used in The total combined probability of all the HFEs the model.

in the same accident sequence/ cut set should not be less than a justified value. It is suggested Thus, the subject Finding had mischaracterized this issue that the value not be below ~0.00001 since it is for the River Bend PRA and has been closed. There is no typically hard to defend that other dependent impact on the ILRT Extension Request.

failure modes that are not usually treated (e.g.,

random events such as even a heart attack) cannot occur. Depending on the independent HFE values, the combined probability may need to be higher.

6.1.6 IE-C2 When using plant-specific data, USE the This is considered a finding since the PRA-RB-01-002S06, Section 5 Generic data was Document the process Section 5.0 of the RBS PRA Data calculation includes most recent applicable data to quantify the SR required justification of the data updated with plant trip data from January 1, used and the justification for use of data from January, 2004, for (Not Met) initiating event frequencies. JUSTIFY excluded. This justification is not 1987 to May 31, 2009 for all transient events justification for initiating events with a relatively high (>0.5/year) excluded data that is not considered to be provided. except for T3A (which used January 1, 2004 to screening /.grouping frequency. This applies only to the IE-T3A reactor scram/

either recent or applicable (e.g., provide May 31, 2009). However, justification for actual plant trip data. turbine trip initiator. A review of data showed only evidence via design or operational change excluding data is not provided as required by Also, correct the negligible impact of neglecting the older data; the IE-T3A that the data are no longer applicable). the SR. assumption number frequency changed only from 1.39 per calendar year to 1.32 from #3 to #1. per calendar year when considering all data since 1987 or In addition, the description in Section 5.1 the smaller 2004-2009 interval; this wollld have a miniscule alludes to assumption #3 as the basis for this impact upon the final value of 0.846 per reactor critical year exclusion. This assumption instead defines the calculated using the Bayesian update process. This only T3A plant initiator. The appropriate impacts IE-T3A, which is the highest frequency IE and is by assumption is #1 of Section 2.0. its nature the initiating event with the lowest CCDP.

LOSP initiating event frequencies as Thus, this finding has been closed and has negligible documented in PRA-RB-01-002809, revision 1, impact upon the ILRT Extension Request.

section 4.2, encompasses genel'.ic data from 1999 to 2008. It is documented. that River Bend has not had any LOSP events; 6.1.7 LE-Bl Findings 19 through 25 refer to Supporting These are findings because use of RBS PSA LERF assessment does not identify the Include the LERF RBS PRA RS Peer Review findings 19 through 25 document LE-B2 Requirements of the Standard which NUREG/CR-6595 LERF contributors credible LERF contributors identified in ASME contributors as listed Findings against Cat. II of the RBS LERF model, consistent document that a NUREG/CR-6595 LERF and containment and Table 2-2.8-3 to support Capability Category II. in ASME Table 2-2.8-3 with BWROG Peer Review practice. The RBS LERF model model meets Capability Category I of the phenomenological analysis is not In addition, applicable generic or plant-specific to support Capability is a Cat. I NUREG/CR-6595 simplified LERF model that PRA Standard. adequate to meet the Capability analyses based on LERF contributors identified Category II. meets Cat. I of the PRA Standard but is not intended to Category II requirements of SRs. in ASME Table 2-2.8-3 are not used for most meet Cat. II.

The River Bend LERF model is a containment challenges. Use applicable Mark NUREG/CR-6595 LERF model intended to III generic or best These Peer Review Findings confirm that the RBS model meet Capability Category I. estimate plant-specific meets Capability Category I for the LE Supporting analyses for all Requirements of the Standard.

containment

RBG-47695 Page 18 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT challenges listed in Thus, these findings do not impact the ILRT Extension ASME Table 2-2.8-3. Request.

6.1.8 LE-Cl Plant specific accident sequence challenges are Develop plant specific LE-C4 not treated on a plant specific basis and the accident sequences on definition of radionuclide releases is not a plant specific basis developed using plant specific analysis. to adequately address Used NUREG/CR-6595 generic approach for SRs LE-Bl and LE-B2 the containment and phenomenological (as described in analysis. Entergy guidance document EN-NE-G-011, Steps 5.2 and 5.3).

Provide a realistic estimation of the severe accident sequence progression (take credit for mitigating actions such as fission product scrubbing as accounted for in MAAP analyses).

6.1.9 LE-C2 Operator actions are not explicitly evaluated to Incorporate operator LE-C3 assess the procedural directions and the time actions into the available post core damage. :No substantial Containment Event credit has been given for repair. Tree (CET) top event fault trees in a realistic manner.

Credit repair as deemed appropriate to implement primary and alternate mitigating actions post core-damage.

6.1.10 LE-C5 PRA-RB-01-002512, Rev 1- Used NUREG/CR- Use engineering 6595 generic approach for the containment and guides EN-NE-G-003 phenomenological analysis. and EN-NE-G-011, Sections 5.2, 5.4, and Attachment 6.2 to generate plant specific Level 2 success criteria.

6.1.11 LE-ClO No credit is taken for equipment survivability Perform review of LE-Cll or operator actions in adverse environment in significant accident LE-C12 PRA-RB-01-002512. progression sequences LE-C9 resulting in large early

RBG-47695 Page 19 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT release to determine if engineering analysis could support continued equipment operation or operator actions that could

reduce LERF.

6.1.12 LE-C13 PRA-RB-01-002Sl2, Rev 1 - Containment bypass Containment bypass was treated in a conservative manner (vessel events due to vessel rupture and ATWS sequences going to core rupture and ATWS damage results in containment bypass). events should be treated realistically.

No credit for fission product scrubbing was modeled per NUREG/CR-6595 (although basic event 12-ABSCRUB was modeled to account for RPV venting through the MSIVs with direct release to the condenser, the probability for this event was set to 1.0) 6.1.13 LE-E2 PRA-RB-01-002Sl2, Rev 1- Generic approach to Use realistic LE-E3 the containment and phenomenological analysis parameter estimates based on NUREG/CR-6595 is used. throughout the LERF accident progression analysis.

Include LERF contributors identified from the results of the accident progression analysis by performing appropriate source term analyses using MAAP and other available sources.

6.1.14 IE-A3 REVIEW the plant-specific initiating event PRA-RB-01-002S06, Section 4.2 and This is considered a finding since a re-screening Re-screen plant Refer to the response to Finding 6.1.6 (SR IE-C2) which experience of all initiators to ensure that Appendix A lists the plant-specific would result in a change in the initiating event specific events and establishes that justification was provided for exclusion of (Met) the list of challenges accounts for plant events. frequency, unless exclusion is justified/ provide a technical older IE-T3A events, and that there was negligible impact experience. See also IE-A7. documented. basis (e.g., from that exclusion. The 1/16/87 trip is one of those older The IE notebook provides a list of implementation of a IE-T3A events for which exclusion has been justified.

plant trips identified by LERs between Reactor /Turbine trip initiators that occurred scram reduction 1987 and 2009. Each event is either prior to the last 5 years were discarded. The program), rather than Note the applicable Supporting Requirement from the screened, with justification, or events should be retained unless it can be an age basis, for Standard was judged to be Met.

categorized into the appropriate IE. shown that the plant design/ operation has exclusion of events.

However, many events that would changed. For example, the frequent use of a 25 - This Finding has been closed and has negligible impact have been categorized as reactor trip I 40% scram in lieu of a controlled shutdown upon the ILRT Extension Request.

turbine trip were screened based solely would be a valid reason for exclusion. A reactor

RBG-47695 Page 20 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Commenf Possible Resolution Impact on ILRT on age. trip due to FW reg. valve failure (1/16/87) should not be eliminated if that same failure PRA-RB-01-002S09, revision 1, section would cause the same event today.

4.2.1.1, documents that no LOOP events have occurred at the River Bend site.

6.1.15 IE-A7 In the identification of initiating events, This is a finding since screening out The IE notebook provides a list of plant trips Re-screen events to This concerns identification of 3 shutdowns as Manual INCORPORATE events that resulted in a scram because identified by LERs between 1987 and 2009. which section b Shutdowns vice Scrams.

(Met) (a) events that have occurred at conditions they occurred as part of a manual Each event is either screened, with justification, applies.

other than at-power operation (i.e., during shutdown sequence is not in or categorized into the appropriate IE. Several 9/1/08 was the shutdown for Hurricane Gustav. This was low-power or shutdown conditions), and accordfil!ce with section b of this of the events that screen in were at low power a controlled plant shutdown. The plant was not in danger for which it is determined that the event could also occur during at-power operation element. ___, operations. At least 3 events (dated 5/23/07 3/21/08, and 9/1/08) were identified as scrams of automatic scram at the time but was doing a manual shutdown due to grid stability issues. The plant was shut (b) events resulting in an unplanned but were screened out as 'manual shutdowns.' down prior to what would have been a partial loss of controlled shutdown that includes a scram offsite power event a couple of hours after the plant scram prior to reaching low-power conditions, from low power.

unless it is determined that an event is not applicable to at-power operation. 5/23/07 was a shutdown for Recirculation Flow Control Valve repairs. This was a controlled plant shutdown and the order was given to scram from low power (-30%)

instead of continuing to drive rods in.

3/21/08: repairs for a stem-disk separation on a feedwater heater valve. Inserted manual scram once downpowered to 15%.

Thus, these were all controlled evolutions and*i.1.o scram was required as part of the plant shutdown process.

Thus, none of these three events qualify as unplanned scrams prior to reaching low power conditions, thus, per SR IE-A7, can be excluded from consideration.

Note the applicable Supporting Requirement from the Standard was judged to be Met.

Further note these events would be considered IE-T3A events, which are low CCDP events with corresponding low contribution to plant risk (only 3.7% of plant risk per the Rev. 5 Summary calculation) despite having the highest initiating event frequency of 0.846/year of all initiating events.

Thus, it is concluded this Finding does not impact the ILRT Extension Request.

6.1.16 IE-Cl CALCULATE the initiating event frequency PRA-RB-01-002S06, Section 4, IE The plant-specific reactor/turbine trip rate This indicates that the Since the lognormal distribution is based on plant specific

RBG-47695 Page 21of32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT accounting for relevant generic and plant- frequencies were calculated using (T3A) is significantly higher than the generic, PRA should use the data and that for generic data overlap, it is consistent with (Met) specific data unless it is justified that there generic and plant-specific data except with the plant-specific mean well above the plant-specific values Entergy and industry practice to use Bayesian updating for are adequate plant-specific data to those initiators that are rare (LOCAs) generic and updated 95th. without Bayesian deriving the IE-T3A frequency for use in the RBS PRA.

characterize the parameter value and its in which generic data alone is used updating (reference IE Specifically, the 5% value for the plant specific distribution uncertainty. (See also IE-Cl3 for and the special initiators for which a notebook, Appendix of 0.637 /rx-year is less than that for the generic (0.725) requirements for rare and extremely rare fault tree is developed. B, page 53) even though the plant specific mean of 1.46 is greater than events.) Generic data for some special initiators that of the generic. Entergy practice would be to use only was screened based on non- the plant specific data if the 5%-tile of the higher applicability to River Bend. No IE probability event is higher than the 95%-tile of the lower frequencies based solely on plant event, a criteria which this data did not meet, thus Bayesian specific data were observed. updating is considered valid. The primary reason why LOSP initiating events are analyzed in Bayesian updating is done in the first place is because it is PRA-RB-01-002S09 revision 1. Removal difficult to obtain a high degree of statistical confidence of events that were determined to not with plant-specific data only. Again, the underlying be applicable to the River Bend Site is assumption in Bayesian analysis is that the generic data is discussed in section 4.2. Events that representative of the plant-specific data. Thus, Bayesian occurred in shutdown conditions are updating of data where there is an overlap of the shown in Table 4. It was documented distributions should be used in order to obtain a more that RBS has not experienced a LOSP, accurate estimate of actual future performance.

therefore, plant data specificity is not an issue. NUREG/CR-6823 as well as standard textbook on the However, the plant-specific subject (Martz, Harry F and Waller, Ray A, "Bayesian reactor/turbine trip rate (T3A) is Reliability Analysis") do not provide any specific criteria significantly higher than the generic, based on distribution overlap or lack thereof for when with the plant-specific mean well Bayesian update results would not be legitimate. The RBS above the generic and updated 95th. IE-T3A frequency meets the requirements of Entergy Data Analysis guide EN-NE-G-007 to "Check that.the mean value is reasonable when compared to the generic value."

Even if this Finding had a correct basis, its impact on risk would be small. The site specific raw data resulted in a 1.45/year frequency vice a 0.846 value used in the RBS PRA. Based on a CCDP of 1.15E-07, this would result in a slight increase (2.6%) in base CDF from 2.604E-06/year to 2.673E-06/year.

Note the applicable Supporting Requirement from the Standard was judged to be Met.

This closed Finding is concluded to have no impact on the ILRT Extension Request.

6.1.17 IE-C12 COMPARE results and EXPLAIN PRA-RB-01-002S06, Comparison of IE Table 10 of the IE notebook provided the Revise the method to Initiating Event frequencies for ISLOCA and BOC are differences in the initiating event analysis results and explanation of differences required comparison for most initiators. LOOP, determine the IE developed in the documents for those specific events.

(Met) with generic data sources to provide a is contained in Section 5.4 and breaks outside containment, and ISLOCAs are frequency for loss of reasonableness check of the results. presented in Table 10. not discussed in that table. From table 10, the vital de busses to use a When reviewed, the 3 events which are the basis of the frequency of loss of vital DC buses is fault tree model. NUREG/CR-5750 Loss of DC initiating event frequency are

RBG-47695 Page 22 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding* Peer Review Comment Possible Resolution Impact on ILRT Reasons for the differences are approximately a factor of 10 below the generic either applicable to other higher CCDP transients or simply provided and appear reasonable. value. The frequency of these 2 initiating events not a transient initiator. Comparison to other plants shows LOOP, breaks outside containment, was based on screened generic data, indicating that the RBS value is comparable to that of other similar and ISLOCAs are not discussed in that .that 90% of the generic events were considered design plants.

table. From table 10, the frequency of not applicable to RB.

loss of vital DC buses is approximately Since Loss of DC events have a low CCDP of a factor of 10 below the generic value. This is a finding because a possible technical approximately E-06, and a small generic frequency of The frequency of these 2 initiating error may e.xist. The screened IE frequency for 1.17E-03, highly conservative use of the generic frequency events was based on screened generic loss of vital DC bus is almost a factor of 10 would only result in a near-negligible CDF contribution of data. lower than the generic value (reference IE approximately lE-09.

notebook, table 10), meaning that almost 90% of No comparison of LOSP initiating the generic events were deemed not applicable Note the applicable Supporting Requirement from the event analysis was noted. to RB. Three other PRAs were examined; each, Standard was judged to be Met.

through the use of a fault tree, calculated the frequency for loss of a vital DC bus at This Finding has been closed and is concluded to have no approximately a factor of 10 larger than that for impact on the ILRT Extension Request.

RB. This great of a difference seems improbable, and should be re-visited.

6.1.18 IFSN-B2 IFSN-B2: IFSN-B2: Appendix B, plant walkdown, provides the SSC Update This finding has been addressed through Revision 1 of the (Not Met) DOCUMENT the process used to identify The documentation provides some of identification and spatial locations. The documentation to RBS Internal Flooding Analysis and Internal Flooding applicable flood scenarios. For example, this the information suggested. However, walkdown sheets for two flood areas (AB141- clearly identify the Quantification, calculations. All PRA components subject IFSN-AS documentation typically includes as noted in other SRs, improvements 1/2/3 I 4 and AB70-3) were inspected. In both SSC's included in the to internal flooding are accounted for in the Internal (Not Met) (a) propagation pathways between flood need to be made to fully meet this SR. cases inspected, SSC spatial location and internal events Flooding Analysis; Appendix E of that document cross-areas and assumptions, calculations, or other vulnerability to spray was identified. analysis. Verify that references equipment modeled in the PRA to Internal bases for eliminating or justifying IFSN-AS: all internal event SSCs Flood zones. Explicit identification of components propagation pathways This is a finding because the IF The list of SSCs for AB141-l/2/3/4 included susceptible to failure assumed damaged by flooding (e.g., spray or (b) accident mitigating features and barriers documentation does not clearly many items that were not described in the by flooding are submergence) was included in Appendix Dfor each credited in the analysis, the extent to which identify SSCs included in the internal associated accident sequence description accounted for in the individual flooding scenario).

they were credited, and associated events analysis. Therefore, the located in section 4.2.1.5. The list of SSCs for Internal Flooding justification accuracy of the scenarios cannot be AB70-3 included RCIC components that were analysis. The results of the updated IFPRA have been used in the (c) assumptions or calculations used in the determined. not included on the RCIC system simplified ILRT Extension Request report.

determination of the impacts of diagram.

submergence, spray, temperature, or other This finding has been closed and has no impact on the flood-induced effects on equipment There appears to be no documented linkage ILRT Extension Request.

operability (d) screening criteria used in the between SSCs identified for the IF analysis and analysis those identified for analysis in the internal (e) flooding scenarios considered, screened, events PRA. Furthermore, there is no explicit and retained identification of PRA SSCs that would be

(£) description of how the internal event affected due to flooding.

analysis models were modified to model these remaining internal flood scenarios (g) calculations or other analyses used to support or refine the. flooding evaluation (h) any walkdowns performed in support of the identification or screening of flood scenarios

RBG-47695 Page 23 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT IFSN-A5:

For each flood area not screened out using the requirements under other Internal Flood Supporting requirements (e.g. IFSO-A3 and IFSN-A12), IDENTIFY the SSCs located in each defined flood area and along flood propagation paths that are modeled in the internal events PRA model as being required to respond to an initiating event or whose failure would challenge normal plant operation, and are susceptible to flood. For each identified SSC, IDENTIFY, for the purpose of determining its susceptibly per IFSN-A6, its spatial location in the area and any flooding mitigative features (e.g.,

shielding, flood, or spray capability ratings).

6.1.19 HR-E3 TALK THROUGH (i.e., review in detail) This is a finding. While the PRA did Appendix C to PRA-RB-Ol-002S03 documents Review the Findings 5-5 and 5-6 addresses the modeling of terminate-with plant operations and training personnel perform talk-throughs, the information operator input for the HRA. Per discussion information and-prevent EOP actions.

(Met) the procedures and sequence of events to communicated in the talk throughs with the RBS HRA analyst, interviews involved documented on the confirm that interpretation of the procedures (as-operated plant) were not discussions of procedural application. talk throughs to Resolution of this finding is primarily documentation in is consistent with plant observations and effectively modeled in the operator However, operator action B21-XHE-FO-INHIB ensure the developed nature as consideration of operator tools is implicit in training procedures. response analysis. as presented (page 127) coupled the inhibit ADS* HEP's model the as- operator interviews in support of HRA analyses and in action with the HPCS terminate and prevent operated facility. other interactions with Operations staff that results in action. In the development of the action in inputs to the plant PRA. Continued interaction with spreadsheet HFE_CP.xls, the action was Operations staff, including regular simulator observations modeled only as the inhibit ADS function. by the site PRA engineer, has not identified any deviations When the use of operational "hardcards" (as between the "Hard Cards" of procedure OSP-0053 documented on the operator interview sheet) (Emergency and Transient Response Support Procedure) was asked, the more experienced PRA staff did and detailed Abnormal Operating Procedure (AOP) not indicate awareness of this application. instructions. This is as anticipated, as OSP-0053 is intended HPCS terminate and prevent is a specific to provide the same important guidance as in AOP's and "hardcard" shown in OSP-0053 as Attachment 5. per EOP's but in a more streamlined fashion. Per While an extensive review could not be OSP-0053, "Hard Cards" were developed to reduce the conducted, this singular thread could indicate a probability of operator error in carrying out these actions weakness in confirming (understanding) actual and are considered an expedited "short form" for response plant operational aspects and procedural usage. to transients.

Note OSP-0053 Hard Cards were mentioned only once during operator interview related to Initiating Events or HRA or for interviews conducted during the Internal Flooding Analysis. Hard Cards also were not a topic brought up during Expert Panel meeting or in various discussions that have occurred with Operations staff regarding PRA models over the past several years. Thus, it is concluded that the reliance on procedures (and plant

RBG-47695 Page 24 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT focus on Procedural Adherence) provides the basis for confirming adequate consideration of plant operational aspects and procedural usage with regard to the HRA and other aspects of PRA modeling. This is consistent with recent observations of simulator scenarios by PRA staff. It is thus concluded that the basis for modeling of Operator actions in HRA analyses is robust and proper.

Use of OSP-0053 Hard Cards is sometimes noted as part of simulator observations performed by the plant PRA engineer. The simulator observations along with communications with Operations on risk assessment

issues ensures that the plant PRA for River Bend reflects actual plant operational aspects and procedural usage.

Thus, this finding has been closed and does not impact the ILRT Extension Request.

6.1.20 IFEV-A7 IFEV-A7: INCLUDE consideration of IFEV-A7: PRA-RB-01-004 revision 0 Scope of Analysis Consider maintenance Maintenance induced flooding is now addressed in section (Not Met) human-induced floods during maintenance The requirement is to consider human discussion in Section 2 notes that "In this induced flood events 3.1.6 of the revised Internal Flooding Analysis calculation.

through application of generic data. induced floods during maintenance. analysis, all causes of flooding were considered and provide Columbia and Clinton Internal Flooding PRA's were IFEV-B2 The scope of analysis excluded this except plant-specific maintenance activities". documentation of this consulted in developing Section 3.1.6.

(Not Met) IFEV-B2: DOCUMENT the process used to consideration. Hence this is a finding analysis.

Identify applicable flood-induced initiating However, maintenance-induced floods are not Thus, this finding has been closed and does not impact the events. For example, this documentation IFEV-B2: considered. ILRT Extension Request.

typically includes The documentation provides some of (a) flood frequencies, component the information suggested. However, unreliabilities I unavailabilities, and HEPs as noted in other sRs; improvements used in the analysis (i.e., the data values need to be made to fully meet this SR.

unique to the flooding analysis)

(b) calculations or other analyses used to support or refine the flooding evaluation (c) screening criteria used in the analysis 6.1.21 IFEVA6 IFEV-A6: IFEV-A6: PRA-RB-01-004 revision 0 Section 3.2 provides a Provide a more As part of revising the Internal Flooding Analysis (Cat.I) This is considered a finding as the discussion of 'focused' problematic issues complete and subsequent to the peer review, additional discussion of Cat. I: In determining the flood initiating requirement requires the gathering of applicable to the flooding analysis. Judgment I systematic review of River Bend internal flooding operating IFEV-B2 event frequencies for flood scenario plant specific information related to assessment has been provided for not being demonstration of experience was provided in Section 3.1.6, along with (not met) groups, USE ONE OF THE FOLLOWING: plant design, operating practices applicable. However, the discussion appears to plant-specific reference to additional RBS flood history reviews (a) generic operating experience (maintenance induced flood be limited, indicating a lack of depth. applicability. conducted for the 2009 study and additional (b) pipe, component, and tank rupture potentials), etc., which was only Operating practices (digging in the yard, documentation for the 2012 revision. The RBS specific failure rates from generic data sources identified in a limited manner. clearance program discussion, etc.), or a internal flooding operating experience does not call into (c) A COMBINATION OF (A) OR (B) detailed screening of past plant issues related to question the use of EPRI TR-1013141 pipe failure ABOVE with engineering judgment IFEV-B2: flooding potentials was not identified. Generic frequencies except for fiberglass CNS piping, for which The documentation provides some of equipment rupture and leak frequencies used RBS applied a plant specific failure rate based on plant Cat. II: GATHER PLANT-SPECIFIC the information suggested. However, are presented in Tables 3.2.1.1and3.2.1.2. As experience; note this CNS piping is a very minor INFORMATION ON PLANT DESIGN, as noted in other SRs, improvements an all-inclusive gathering of plant specific data contributor to flooding risk.

RBG-41695 Page 25 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT OPERATING PRACTICES, AND need to be made to fully meet this SR. is not apparent, this SR is viewed to only satisfy CONDITIONS THAT MAY IMPACT Category I. Thus, this finding has been closed and does not impact the FLOOD LIKELIHOOD (I.E., MATERIAL ILRT Extension Request.

CONDITION OF FLUID SYSTEMS, This is considered a finding as the requirement EXPERIENCE WITH WATER HAMMER, requires the gathering of plant specific AND MAINTENANCE-INDUCED information related to plant design, operating FLOODS). practices (maintenance induced flood In determining the flood-initiating event potentials), etc., which was only identified in a frequencies for flood scenario groups, USE limited manner.

A COMBINATION OF THE FOLLOWING:

(a) generic and PLANT-SPECIFIC operating experience (b) pipe, component, and tank rupture failure rates from generic data sources AND PLANT-SPECIFIC experience (c) engineering judgment FOR CONSIDERATION OF THE PLANT-SPECIFIC INFORMATION COLLECTED IFEV-B2: (documentation)

See Finding 6.1.20. (IFEV-A7) 6.1.22 LE-A2 IDENTIFY the accident sequence This is a finding since the evaluation PRA-RB-01-002S12, Rev 1-Accident sequence Employ the LERF This finding concerns the LERF model. Review of (Met) characteristics that lead to the physical for significant accident progression characteristics that influence LERF are definition provided in sequences was conducted, and several sequences were characteristics identified in LE-Al. Examples sequences resulting in a large early identified in Section 4.1, RBS Simplified PRA-RB-01-002S12 to identified which could potentially contribute to LERF due include release was not conservative. Containment Event Tree. A cursory review of determine the to timing of declaration of General Emergency with respect (a) type of initiator Level 1 event trees in PRA-RB-01-002S01 reveals appropriate accident to any subsequent release. A sensitivity study was (1) transients can result in high RCS pressure that these physical characteristics were progression conducted with the LERF model which concluded this (2) LOCAs usually result in lower RCS appropriately considered for each core damage sequences. would result in a LERF increase of no more than 5.7E-09, pressure end state. resulting in a total LERF of 3.04E-08/reactor year. This (3) ISLOCAs, SGTRs can result in small increase in the base PRA LERF value would not containment bypass. However, the definition of LERF listed in PRA- impact the conclusions or acceptability of the ILRT (b) status of electric power: loss of electric RB-01-002S12 is releases before the effective Extension Request since those sequences which are already power can result in loss of ECC injection implementation of off-site emergency response characterized as LERF in the base model are excluded from (c) status of containment safety systems such and protective actions. The document the ILRT extension delta- risk assessment.

as sprays, fan coolers, igniters, or venting continues to assess timing from the initiating systems: operability of containment safety event to off-site emergency response, but did Additional discussions with Operations and Emergency systems determines status of containment not account for when the sites emergency plan Planning personnel result in a conclusion that plant heat removal References [2-14] and [2-15] would implement the actions following an Emergency Implementing Procedures (EIP's) and EOP's provide example lists of typical event. The actual emergency classification aid will result in an early declaration of General Emergency for characteristics. EIP-2-001 may not implement off-site those scenarios, thus those scenarios would not contribute emergency response until late in the toLERF..

progression. For example a loss of decay heat removal is only a site area emergency. Therefore Note the applicable Supporting Requirement from the the early dismissal of longer term sequences in Standard was judged to be Met.

the event tree as applied is not appropriate for the LERF definition and is not a conservative As there will be NTTF-related and FLEX-related changes to

RBG-47695 Page 26 of 32 Disposition SR and and Finding Assessment SR description. Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT evaluation. EOP's and EIP's implemented, this Finding will be revisited as part of the next PRA Revision to confirm that no changes to the LERF model are required and provide better documentation on the interface between Emergency Declarations and LERF analysis. Other changes, such as credit for improved RCIC room heatup calculations and FLEX plant modifications, are expected to further mitigate LERF as part of that PRA update.

This finding thus does not impact the ILRT Extension Request.

6.1.23 IFSO-A3 IFSO-A3: IFSN-A12 thru A16, IFSO-A3: No explicit documentation of flood area Document screening While there is not a comprehensive list of areas screened, (Not Met) SCREEN OUT flood areas with none of the There is not a documented listing of screening is provided. However, it appears performed at each step the requirements are fulfilled. The Internal Flooding potential sources of flooding listed in IFSO- which sources were screened from a review of Section 4.2, table 4.1.1.1, and of the analysis. Analysis has subsequently been revised. Areas without (also IFSN- Al and IFSO-A2. the walkdown sheets in appendix B that areas flooding sources have been noted on Table 4.1.1.1. Flood A12 IFSN-B2: which contain no flood sources were excluded zones without walkdown notes and which are screened thru-A16) IFSN-A12 thru -A16: The documentation provides some of from the analysis. from further consideration are documented in Section 4.1.1 (Met) [flood source screening] the information suggested. However, and in the introduction to the walkdown notes. Flood as noted in other SRs, improvements This same issue also applies to flood source zones have not been screened based on operator mitigation IFSN-B2 IFSN-B2: (documentation) need to be made to fully meet this SR. screening. actions but a conservative approach is used where HRA's (Not Met) Refer to Finding 6.1.18 above. are developed when operator action is credited. Detailed scenario reviews are used to determine if flood sources are insufficient to propagate damage. Generally no credit has been taken for drains and sump pumps other than crediting Control Building drains subject to Preventive Maintenance tasks, and that has been addressed via a detailed scenario discussion rather than a screening.

This finding is concluded to identify weaknesses in documentation which have been addressed in the revision to the Internal Flooding Analysis.

This closed finding has no impact on the ILRT Extension Request.

6.1.24 IFSN-A2 For each defined flood area and each flood This is a finding since identification of Plant design features that affect flood Document alarms, Alarms associated with flooding events and plant locations source, IDENTIFY plant design features that alarms that are generated for each propagation are identified in the walkdown particularly alarms for alarms (i.e., Main Control Room vs. Auxiliary Control (Met) have the ability to terminate or contain the flood scenario is needed to support sheets shown in Appendix B. The effect of these that are used to cue Room) have been explicitly identified, including in flood propagation. proper evaluation of human failure features on flood scenarios appears to be operator response, individual building discussions in Scenario development INCLUDE the presence of events need to support internal addressed properly in the propagation analysis that are expected to section of the Internal Flooding Analysis. Specific alarm (a) flood alarms flooding quantification. shown in Section 4.1.2 of Calculation PRA-RB- occur for each flood. response procedures that would be used are included in (b) flood dikes, curbs, sumps (i.e., physical 01-004 Rev. 0 and the scenario development~ the spreadsheets of the RBS Human Reliability Analysis structures that allow for the accumulation shown in Section 4.2. However, explicit calculation.

and retention of water) identification of alarms that would be generated Note the applicable Supporting Requirement from the (c) drains (i.e., physical structures that can as a result of the flooding does not appear to Standard was judged to be Met.

function as drains) have been completed.

RBG-47695 Page 27 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT (d) sump pumps, spray shields, water-tight This closed Finding does not impact the ILRT Extension doors Request.

(e) blowout panels or dampers with automatic or manual operation capability 6.1.25 IFSN-A3 For each defined flood area and each flood IFSN-A3: Identification of actions to isolate or terminate Identify the actions Actions on flood termination actions have been added in (Not Met) source, IDENTIFY those automatic or No specific actions or responses that floods is required by the SR. Identification of that can terminate the scenario descriptions (4.2) and system information (4.1.4) operator responses that have the ability to will terminate any flood were these actions is needed to support evaluation of flood, include the sections of the Internal Flooding Analysis, and in the HRA IFSN-B2 terminate or contain the flood propagation. identified. General actions such as the feasibility the events when performing the procedural guidance calculation, where spreadsheet calculation sheets for (Not Met) "isolate the SSW system" are identified, HRA; hence this is a finding. for doing so and the flooding are now explicitly included in the calculation.

IFSN-B2: (documentation) however, no indication of whether or equipment needed. Increased discussion of procedural instructions has been Refer to Finding 6.1.18 above. not the actions are procedurally provided. Quantification includes scenarios where directed is provided. isolation is successful as well as where isolation fails.

The quantification methodology will This Finding is Closed and does not impact the ILRT apply a generic isolation operator Extension Request.

failure probability to many different scenarios without considering the cues or timing that will initiate the actions.

The calculated HEPs for different scenarios could be different for different scenarios. Furthermore, only scenarios where isolation failure occurs are evaluated in the quantification.

IFSN-B2:

The documentation provides some of the information suggested. However, as noted in other SRs, improvements need to be made to fully meet this SR.

6.1.26 IFSN-A4 ESTIJ\tlATE the capacity of the drains and Identification of these parameters is No estimation of the amount of water retained Evaluate and Information on sump pump flow rates and sump capacities the amount of water retained by sumps, required by the SR; hence this is a by sumps, berms, or curbs was identified in document the capacity has been added to building discussions (e.g., see discussion berms, dikes, and curbs. ACCOUNT for finding. Furthermore, these factors can Calculation PRA-RB-01-004 Rev. 0. No of flood mitigation of AB70-7 /8 sumps and pumps about six pages into these factors in estimating flood volumes impact the accident progression by calculations for estimating the capacity of drains features and revise Section 4.2.1). Condensate pit volume had been and SSC impacts from flooding. changing timing of cues and, therefore, was identified. However, drain capacity is flooding scenarios to determined and considered in the Rev.0 analysis.

operator mitigation response. These assumed for s9me scenarios. account for these Discussion on drain systems were added in Section 4.1.4 of effects can be significant for small and effects. the Internal Flooding Analysis; however, drains and sump moderate flow rate scenarios. pumps are generally not credited in the IFPRA, thus impact on timing of cues and operator mitigation actions for scenarios without detailed HRA's (where this is considered) is negligible. Effects of curb heights have been explicitly considered in the Internal Flooding Analysis. A specific determination of fuel in the fuel oil area beneath the east end of the DG rooms was performed for Rev.1 of the Internal Flooding Analysis, which reduced the assumed

RBG-47695 Page 28 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT time to fill that volume (however, that time is not actually credited in the Internal Flooding PRA analyses).

This Finding is Closed and does not impact the ILRT Extension Request.

6.1.27 IFSN-A6 IFSN-A6: IFSN-A6: Consideration of these effects is required by the Perform.and This was considered a documentation issue, as noted in the (Not Met) For the SSCs identified in IFSN-A5, Section 3.2 of Calculation PRA-RB SR; hence this is a finding. These effects can be document review comment. Revision 1 of the Internal Flooding IDENTIFY the susceptibility of each SSC in a 004 Rev. 0 states that the effects of significant contributors to overall risk because consideration of HELB Analysis included in Appendix D listings of the

( IFSN-A7) flood area to flood-induced failure steam from pipe breaks are limited to they are difficult for operators to diagnose and induced floods. components and/ or zones which are subject to local (Met) mechanisms. the room in which the break occurs. mitigate. Document how damage mechanisms (e.g., spray, HELB effects, etc.) and INCLUDE FAILURE BY SUBMERGENCE This assumption is not supported. potential jet subject to flood propagation effects (e.g., submergence).

AND SPRAY in the identification process. Effects of a high-energy line break impingement could Due to the general small volumes of important RBS EITHER (HELB) can propagate across large affect equipment locations, generally all equipment in a Flood Zone was (a) ASSESS qualitatively the impact of flood- areas of the plant and effect equipment availability and flood conservatively assumed subject to local damage induced mechanisms that are not formally through temperature or humidity. scenario progression. mechanisms, so that there is no difference in the addressed (e.g., using the mechanisms listed Furthermore, HELBs can actuate fire components assumed damaged due to HELB effects versus under Capability Category ill of this protection systems thereby spray effects in the same flood zone.

requirement), by using conservative exacerbating the flood. Operator assumptions; OR actions to mitigate a HELB-induced Environmental effects, including those due to HELB, have (b) NOTE that these mechanisms are not flood can be significantly impaired been thoroughly considered, as documented for the case of included in the scope of the evaluation.

compared to a flood from a low- internal flooding in Att.2 to letter RBG-46944 dated August temperature fluid. Pipe whip and jet 11, 2009, and as discussed in the Internal Flooding Analysis IFSN-A7: impingement were not addressed. document. Steam propagation effects are accounted for In applying SR IFSN-A6 to determine Consideration of these effects is through the RBS EQ program, as discussed in that letter.

susceptibility of SSCs to flood-induced required by the clarification in Reg Steam propagation for steam and feedwater lines were also failure mechanisms, TAKE CREDIT for the Guide 1.200. considered in development of damaged equipment sets for operability of SSCs identified in IFSN-A5 the Turbine Building.

with respect to internal flood impacts only if IFSN-A7:

supported by an appropriate combination of In general, all equipment in an affected This Finding is closed and does not impact the ILRT (a) test or operational data (b) engineering flood area is assumed failed. One Extension Request.

analysis (c) expert judgment exception is failure due to temperature or humidity effects noted in SR IFSN-A6. Another exception is failure of air-operated or hydraulic-operated valves.

Table 3.1.4.1 notes that these valves are assumed to remain functional for spray and steam environments.

However, this assumption is not supported.

6.1.28 IFSN-A7 In applying SR IFSN-A6 to determine In general, all equipment in an affected This is a finding since the SR requires that Provide a basis for air For the revised Internal Flooding Analysis, all active valves susceptibility of SSCs to flood-induced flood area is assumed failed. One availability of equipment after a flood must be and hydraulic valve are assumed to fail due to spray or steam unless otherwise (Met) failure mechanisms, TAKE CREDIT for the exception is failure due to temperature supported by data, analysis, or judgment. The operation for post- specifically justified in the scenario development.

operability of SSCs identified in IFSN-A5 or humidity effects noted in SR IFSN- ability of air or hydraulic valves to operate in a flood conditions or with respect to internal flood impacts only if A6. Another exception is failure of air- spray or steam environment is not supported. reevaluate scenarios Note the applicable Supporting Requirement from the

RBG-47695 Page 29 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT supported by an appropriate combination of operated or hydraulic-operated valves. Failure of such equipment could impact considering such Standard was judged to be Met.

(a) test or operational data Table 3.1.4.1 notes that these valves are accident progression and risk. equipment failed.

(b) engineering analysis assumed to remain functional for This Finding is Closed and does not impact the ILRT (c) expert judgment spray and steam environments. Extension Request.

However, this assumption is not supported.

6.1.29 IFSN-A8 IDENTIFY interarea propagation through Inter-area propagation via drain line It appears that propagation for pathways Confirm that there are It has been confirmed that there are no piping penetrations the normal flow path from one area to back flow is evaluated and described identified was properly considered. However, it no pipe penetrations which would serve as flood propagation pathways exist in (Met) another via drain lines; and areas connected in Section 4.1.2 of Calculation PRA-RB- is unusual to see no pipe penetrations below the between rooms in the the lower elevations of the auxiliary building. It has also via backflow through drain lines involving 01-004 Rev. 0. Propagation between level of the HVAC ducts in the lower levels of lower elevations of the been confirmed that HVAC ducts do not contain water failed check valves, pipe and cable rooms on the 70' aux building is the aux building. Also, chilled water is provided aux building. If some sources.

penetrations (including cable trays), doors, included for flow through the HVAC to cooling coils and failure of these cooling coils are identified, then stairwells, hatchways, and HVAC ducts. ducts. No pipe or cable penetrations could allow propagation through the ducting to evaluate their ability Note the applicable Supporting Requirement from the INCLUDE potential for structural failure are identified for walls betWeen these other equipment could occur and these to withstand floods Standard was judged to be Met..

(e.g., of doors or walls) due to flooding rooms. However, pipe penetrations scenarios are addressed. Consideration of these considered in the loads. below the level of the HVAC ducts situations is required by the SR; hence this is a PRA. Confirm that no This Finding is closed and does not impact the ILRT

.. would be expected. Although these finding. water sources are Extension Request.

penetrations would be sealed for a located inside HVAC design-basis flood, their ability to ducts. If some are, withstand a higher water level should include new scenarios be evaluated. No water-filled fluid to evaluate these systems were identified as being sources.

located inside HVAC ducts.

Therefore, failure of systems inside HVAC ducts is not evaluated.

The documentation for two flood areas (AB141-l/2/3/4 and AB70-3) was inspected. For these two areas, propagation of flooding to other areas was addressed: (a) backflow through drains, (b) pipe and cable penetrations (note -- no cable trays pathways were identified), (c) hatchways, and (d)

HVAC ducting. Propagation through a failed door was addressed in area AB70-3. Barrier unavailability was not addressed for either area.

6.1.30 IFSN-A3 IFSN-A3: IFSN-A3: Model and include in The generic operator actions to isolate within 60or120 (Not Met) For each defined flood area and each flood No specific actions or responses that This is a finding since following any break, loss the quantification minutes (FL-HEEHFL120MIN and FL-HEEHFL60MIN) are source, IDENTIFY those automatic or will terminate any flood were of equipment failed due to immediate impacts scenarios where included in the documentation but have not been used in IFQU-A5 operator responses that have the ability to identified. General actions such as along with loss of systems due to flood isolation isolation is successful the Update to the Internal Flooding Analysis. HRA's (Not Met) terminate or contain the flood propagation. "isolate the SSW system" are identified, should be considered as a minimum. Omission but equipment is specific to the system and building for the flood event are however, no indication of whether or of these scenarios couid significantly failed due to the developed and documented. The scenario development

RBG-47695 Page 30 of 32 Disposition SR and and Finding Assessment SR description Basis for Peer Review Finding Peer Review Comment Possible Resolution Impact on ILRT IFQU-AS: not the actions are procedurally underestimate risk. Also, application of a single immediate effects of process credits operator success to isolate a failure only for If additional human failure events are directed is provided. HEP to multiple scenarios may not be the break along with extremely long times (e.g., six hours). Thus, isolation required to support quantification of flood appropriate, since the timing and other loss of systems that failure is assumed in the quantification process except scenarios, PERFORM The quantification methodology will conditions may differ. may occur due to when a successful operator mitigating action is applied.

any human reliability analysis in accordance apply a generic isolation operator flood isolation. Note in many cases the successful operator action does not with the applicable requirements described failure probability to many different Ensure that the eliminate Core Damage but mitigates (greatly reduces) the in2-2.5. scenarios without considering the cues scenario-specific CCDP I CLERP for the scenario. More aggressive or timing that will initiate the actions. operator actions are application of operator recovery actions would reduce the The calculated HEPs for different appropriate. summed CDF /LERF for flooding, as discussed in the scenarios could be different for Internal Flooding Quantification document.

different scenarios. Furthermore, only scenarios where isolation failure The Internal Flooding Analysis and its quantification occurs are evaluated in the considered immediate effects of pipe failures, including quantification. specifying the loss of systems due to the flood initiator.

This is part of the quantification process. Timing and cues IFQU-AS: for operator response are documented in the spreadsheets Additional human error events were for the HRA calculation. For simplicity due to the number added to model termination of various of RBS Internal Flooding scenarios, conservative bounding flood scenarios. These are noted in the calculations are performed where HRA operator actions flood quantification notebook (PRA- are credited and applied for multiple scenarios.

RB-01-006, Appendix A) and Differences in results with and without successful isolation documented in. the HRA notebook are considered in the quantification process.

(PRA-RB-01-002503. The flooding HFEs are defined and quantified in a This finding is closed and does not impact the TLRT manner similar to the other HFEs in Extension Request.

the non-flooding model. However, it appears that a single HFE is applied to multiple flooding scenarios. Given that the time available to terminate various floods may vary, it is not clear if the HEPs used are appropriate for all of the scenarios in which termination is credited.

RBG-47695 Page 31of32 References

[1] USN RC, Safety/Risk Assessment Results for Generic Issue (GI) 199, Implications of Updated Probabilistic Hazard Estimates in Central and Eastern United States on Existing Plants, September 2, 201 O (ADAMS Accession Number ML100270582).

[2] USN RC, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, NUREG-1407, June 1991.

[3] EPRI, Updated Seismic Hazard Results for the Arkansas, Fitzpatrick, Grand Gulf, Indian Point, Pilgrim, River Bend, Vermont Yankee, and Waterford Nuclear Sites.

August 201 O (Not Published, Proprietary).

[4] Entergy, Entergy Operations Inc. Seismic Hazard and Screening Report (CEUS Sites), Response to NRG Request for Information Pursuant to 10 CFR 50.54(f)

Regarding Recommendation 2. 1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, River Bend Station - Unit 1, March 26, 2014 (ADAMS Accession Number ML14091A426).

[5] Entergy, River Bend Station Primary Containment Leakage Rate testing (Appendix J) Program, SEP-APJ-004, Revision 4, June 2015.

[6] Entergy, License Amendment Request, LAR 2007-010, One-time Extension of the Integrated Leak Rate Test Interval River Bend Station, Unit 1, August 17, 2007 (ADAMS Accession Number ML072340544). *

[7] Entergy, River Bend, Unit 1, Response to Request for Information Regarding Recommendation 2. 1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 22, 2014 (ADAMS Accession Number ML14073A647).

[8] Entergy Engineering Report, Re-Assessment of River Bend Seismic Core Damage Frequency, RBS-SA-11-00001, Revision O, June 2011.

[9] Entergy Engineering Report, Individual Plant Examination for External Events (IPEEE), SEA-95-001, June 1995.

[1 O] Entergy Calculation, RBS Internal Flooding Analysis - Quantification and Results, PRA-RB-01-006, Revision 2, March 2012.

[11] Entergy Calculation, River Bend Nuclear Generating Station Probabilistic Safety Assessment High Winds Analysis, PRA-RB-09-001, Revision O, August 2004.

[12] USNRC, River Bend Station, Unit 1 - Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10 CFR 50.54(f) Information Request- Flood-Causing Mechanism Reevaluation (TAC No. MF3675), September 4, 2015, (ADAMS Accession Number ML15212A727).

RBG-47695 Page 32 of 32

[13] Nuclear Energy Institute, Diverse and Flexible Coping Strategies (Flex)

Implementation Guide, NEI 12-06, Revision 2, December 2015.

[,14] Entergy Engineering Report, RBS-NE-14-00002, RBS Rev.5 PRA Technical

) Adequacy, September 2014.

v t e Letter Sequence Request
CAC:MF7037, (Approved, Closed)
Initiation Request, Request, Request Acceptance Administration Questions Answers 19 April 2016 Results Approval Other:
MONTHYEAR2016-10-27 [Table View]

RBG-47695, Response to Request to Additional Information Regarding RBS License Amendment Request to Extend Type a and Type C Test Frequencies

Text

SUBJECT:

Dear Sir or Madam:

Attachment:

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

Navigation menu

RBG-47695, Response to Request to Additional Information Regarding RBS License Amendment Request to Extend Type a and Type C Test Frequencies

Text

SUBJECT:

Dear Sir or Madam:

Attachment:

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

Navigation menu

Search