ML11263A031

E-mail with Attachment from T. Ghosh, NRR, to L. Perkins, NRR Et Al, on Salem/Hope Creek Chapter 5

ML11263A031
Person / Time
Site:	Salem, Hope Creek
Issue date:	09/23/2010
From:	Tina Ghosh NRC/RES/DSA
To:	Leslie Perkins License Renewal Projects Branch 2
References
FOIA/PA-2011-0113
Download: ML11263A031 (25)

Text

Perkins, Leslie From:

Ghosh, Tina Sent:

Thursday, September 23, 2010 12:09 PM To:

Perkins, Leslie Cc:

Harrison, Donnie; Gallucci, Ray; Pham, Bo

Subject:

RE: Salem/Hope Creek Chapter 5 Attachments:

DSEIS Salem-Hope Creek Chapter 5 Perkins draft.doc Hi Leslie, Changes look fine for the most part, but please correct the following:

p p. 5-7, second line of text, "determining form" should be "determined from" Sp. 5-7, 4 th-5th line, replace "higher truncation of 5 x 10-11 per a year used and" with "using a higher truncation of 5 x 1011 per year,"

p. 5-7, 5 th line of text, "inact" should be "intact"

p. 5-7, about middle of text paragraph, "Latter" and "Evaluations" should not be capitalized

p. 5-7, please insert the following sentence before the last sentence in the last paragraph (which begins with "The breakdown of CDF..."): "PSEG did not explicitly include the contribution from external events within the HCGS risk estimates; however, it did account for the potential risk reduction benefits associated with external events by multiplying the estimated benefits for internal events by a factor of 6.3."

IMPORTANT: pp. 5-8, Table 5-3 has typos, the numbers don't line up with the entries correctly, and the exponent formatting is lost. Please cut and paste Table 5-5 from what we provided you directly into your document.

IMPORTANT: p. 5-8, near end, "33 percent" should be "3 percent" (big difference - an order of magnitude!)

p. 5-9, under section 5.3.3, replace "PSA" with "PRA"

p. 5-10, please replace 1st full paragraph with the following: "PSEG removed two candidate SAMAs from further consideration for each site because they are not applicable at SGS or HCGS due to design differences, have already been implemented at SGS or HCGS, or were estimated to have implementation costs that would exceed the dollar value associated with completely eliminating all severe accident risk at SGS or HCGS. A detailed cost-benefit analysis was performed for the 25 and 21 remaining SAMAs for SGS and HCGS, respectively, as well as four additional SAMAs that were analyzed for SGS in response to an NRC staff request for additional information."

p. 5-10, please add the following at the end of the first sentence in section 5.3.4, ", as well as four additional SAMAs that were added for SGS in response to an NRC staff request for additional information."

p. 5-10, 2 nd full paragraph in section 5.3.4, replace the first sentence with the following: "PSEG estimated the costs of implementing the candidate SAMAs through the development of site-specific cost estimates."

p. 5-12, in the only full paragraph, replace "1.64" with "2.5"

p. 5-15, near the end, don't delete the phrase ", or a subset of" and don't insert "identified" I did not track these because it got too messy (Ray - FYI attached is Leslie's original).

Let me know if you have questions.

Best, Tina From: Perkins, Leslie Sent: Thursday, September 23, 2010 10:38 AM 1d

To: Perkins, Leslie; Ghosh, Tina; Harrison, Donnie

Subject:

RE: Salem/Hope Creek Chapter 5

Tina, I am following up to see when I will receive feedback on chapter 5. Is it possible to get feedback today? If not, please let me know when you think you will get feedback to me regarding chapter 5 for Salem/Hope Creek.

Thanks, Leslie From: Perkins, Leslie Sent: Wednesday, September 22, 2010 8:56 AM To: Perkins, Leslie; Ghosh, Tina; Harrison, Donnie

Subject:

RE: Salem/Hope Creek Chapter 5

Tina, I had added a table in chapter 5 and forgot renumber the rest of the tables in the document. Please review the revised chapter 5 attached.

Thanks, Leslie From: Perkins, Leslie Sent: Tuesday, September 21, 2010 5:30 PM To: Ghosh, Tina; Harrison, Donnie

Subject:

Salem/Hope Creek Chapter 5

Tina, As discussed early, I have shortened the chapter 5.input that you provided be consistent with the format used in other recently published SEISs. Attached is the revised chapter 5 for your review. Please review and make changes as necessary. Please contact if you have any questions.

Thanks, Leslie Perkins Project Manager Division of License Renewal Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 301-415-2375 2

1 5.0 ENVIRONMENTAL IMPACTS OF POSTULATED ACCIDENTS 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 This chapter describes the environmental impacts from postulated accidents that might occur during the period of extended operation, The term "accident" as defined in the Generic Environmental Impact Statement (GELS) (NRC, 1996) refers to any unintentional event outside the normal plant operational envelope that results in a release or the potential for release of radioactive materials into the environment_ Two classes of postulated accidents are evaluated under the National Environmental Policy Act (NEPA) in the license renewal review: design-basis accidents (DBAs) and severe accidents, In the GElIS, thle_._

NRC staff categorized accidents as "design basis" when the plant was designed specifically to accommodate these accidents or as "severe" for those accidents involving multiple failures of equipment or function whose likelihood is generally lower than design-basis accidents but where consequences may be higher" (NRC, 1996), These issues are_

evaluated in Chapter 5 of GElS, "Environmental Impacts of Postulated Accidents."

Table 5-1. Issues Related to Postulated Accidents. Two issues related to postulated accidents are evaluated under the National Environmental Policy Act (NEPA) in the license renewal review, DBAs, and severe accidents.

Issue GElS Section Category DBAs 5.3.2; 5.5.1 1

Severe accidents 5.3.3; 5.3.3.2; 5.3.3.3; 5.3.3.4; 2

5.3.3.5; 5.4; 5.5.2 I

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S-[ Deletedl. 'I 17 18 19 5.1 pDESIGN-BASIS ACCIDENTS 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 As described in 10 CFR 50.34(b), in order to receive NRC approval for an operating license, an applicant for an initial operating license must submit a final safety analysis report (FSAR) as part of its applicatiorn_ The FSAR presents the design criteria and design information for Deleted" the proposed reactor and comprehensive data on the proposed site The FSAR also discusses various hypothetical accident situations and the safety features that are provided to prevent and mitigate accidents, The NRC staff reviews the application to determine j

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whether or not the plant design meets the NRC's regulations and requirements and includes, in part, the nuclear plant design and its anticipated response to an accident.

The environmental impacts of postulated accidents were evaluated for the license renewal period in Chapter 5 of the GELS.,...cti0. 5.1 states:

All plants have had a previous evaluation of the environmental impacts of design-basis accident-In addition, the-licensee will be required to maintain______

acceptable design and performance criteria throughout the renewal period......

Therefore, the calculated releases from design-basis accidents would not be expected to change, Since the consequences of these events are evaluated Deleted:

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Deleted: October 2010.5-1 Draft NUREG-1437, Supplement 45

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12 3

4 5

6 78 9

10 11 12 for the hypothetical maximally exposed individual at the time of licensing, changes in the plant environment will not affect these evaluations, Therefore, the staff concludes that the environmental impacts of design-basis accidents are of small significance for all plants, Because-the environmental.....

impacts of design basis accidents are of small significance and because additional measures to reduce such impacts would be costly, the staff concludes that no mitigation measures beyond those implemented during the current term license would be warranted. T'his is a-Category 1 issue..........

No new and significant information related to DBAs was identified during the review of PSEG's environmental report (ER), site audit, scoping process, or evaluation of other available informatiorý_ Therefore, there are no impacts related to these issues beyond those discussed in the GELS.

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J 13 5.2 SEVERE ACCIDENTS 14 Severe nuclear accidents are those that are more severe than DBAs because they could 15 result in substantial damage to the reactor core, whether or not there are serious offsite 16 consequences, In the GELS, the staff assessed the impacts of severe-accidents duringthe 17 license renewal period, using the results of existing analyses and information from various 18 sites to predict the environmental impacts of severe accidents for plants during the renewal 19 period.

20 Severe accidents initiated by external phenomena such as tornadoes, floods, earthquakes, 21 fires, and sabotage have not traditionally been discussed in quantitative terms in the final 22 environmental impact statements and were not specifically considered for the_ Salem 23 Generating Station, Units 1 and 2 (SGS) and Hope Creek Generating Station (HCGS)-sites-24 in the GElS (NRC, 1996) The GEISJ however, did evaluate existing impact assessments---

25 performed by the NRC staff and by the industry at 44 nuclear plants in the United States and 26 segregated all sites into six general categories and then estimated that the risk 27 consequences calculated in existing analyses bound the risks for all other plants within each 28 category, The GElS further concluded that the risk from -beyond design-basis earthquakes _

29 at existing nuclear power plants is designated as SMALL, The Commission believes that 30 NEPA does not require the NRC to consider the environmental consequences of 31 hypothetical terrorist attacks on NRC-licensed facilities. However, the NRC staffs GElS for 32 license renewal contains a discretionary analysis of terrorist acts in connection with license 33 renewal. The conclusion in the GElS is that the core damage and radiological release from 34 such acts would be no worse than the damage and release to be expected from internally 35 initiated events.

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- " -I Deleted: Cooper Hope Creek Generationing Station L(HCGS) andJ Formatted: Not Highlight Deleted: Nuclear Station, Unit 1 (CNS-1) site Deleted:.

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36 37 38 39 In the GELS, the NRC staff concludes that the risk from sabotage and beyond design-basis earthquakes at existing nuclear power plants is designated as SMALL, and additionally, that the risks from other external events are adequately addressed by a generic consideration of internally initiated severe accidents (NRC, 1996).

1 Based on information in the GElS, the staff found that:

2 The generic analysis... applies to all plants and that the probability-weighted 3

consequences of atmospheric releases, fallout onto open bodies of water, 4

releases to ground water, and societal and economic impacts of severe 5

accidents are of small significance for all plants,_ However not all plants 6

have performed a site-specific analysis of measures that could mitigate 7

severe accidentk _ Consequently_, severe accidents are a Category 2 issue -

8 for plants that have not performed a site-specific consideration of severe 9

accident mitigation and submitted that analysis for Commission review ---

10 The staff identified no new and significant information related to postulated accidents during 11 the review of PSEG's environmental report, the site audit, the scoping process, or evaluation 12 of other available informatior. Therefore, there are no impacts related to postulated......

13 accidents beyond those discussed in the GEISV In accordance with 10 CFR 14 51.53(c)(3)(ii)(L), however, the NRC staff has reviewed severe accident mitigation 15 alternatives (SAMAs) forSGSand HGCS, Review results are discussed in Section 5.3.

16 5.3 SEVERE ACCIDENT MITIGATION ALTERNATIVES 17 Regulations under Section 51.53(c)(3)(ii)(L) of 10 CFR requires an ER to contain an analysis 18 of alternatives to mitigate severe accidents if the staff has not previously evaluated SAMAs 19 for the applicant's plant in an environmental impact statement (EIS), related supplement, or 20 in an environmental assessment. The Commission's reconsideration of the issue of severe, 21 accident mitigation for license renewal is based on the Commission's NEPA regulations that 22 require a consideration of mitigation alternatives in its environmental impact statements 23 (EISs) and supplements to EISs, as well as a previous court decision that required a review 24 of sever mitigation alternatives at the operating license stage.

25 5.3.1 Introduction 26 This section presents a summary of the SAMA evaluation for SGS and HCGS conducted by 27 PSEG and the NRC staffs reviews of those evaluations. The NRC staff performed its 28 review with contract assistance from Pacific Northwest National Laboratorý r,'he_-NRC 29 staffs reviews~are available in full in AppendiemsF and G; the SAMA evaluationsare 30 available in full inpSEG'ERs..

31 The SAMA evaluations forGS and j-CGS weraeconducted with a four-step-app[oach In _

32 the first step,,SEG quantified the level of risk associated with potential reactor accidents 33 using the plant specific probabilistic risk assessment (PRA) and other risk models, 34 In the second step, PSEG examined-the major risk contributors and identified possible ways 35 (SAMAs) of reducing that rislk Common ways of reducing risk are changes to com!ponents,,

36 systems, procedures, and training,,5PSEG identified-__7 potential SAMAs for.SGS, and 23 for_-

37 HCGS4 JPSEG performed an initial screening to determine if any SAM As could be eliminated-38 because they are not applicable to-GS_or HCGS due to design-differences, or have ---

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Deleted: for CNS-1 conducted by NPPD Energy Company, LLC (NPPD) and the staffs review of that evaluation.

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Deleted: Subsequent to the ER, NPPD discovered a problem with the process they used to numerically average the site-specific meteorological data. NPPD performed a sensitivity analysis of the population dose risk and offsite economic cost risk using corrected meteorological data. and found that the population dose and offsite economic cost values for each of the release categories would be slightly less than reported in the ER, and that the conclusions of the SAMA remain valid (NPPD, 2009b).

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'It Deleted: NPPD's Deleted: ER and the supplement submitted in December 2009 Deleted: CNS-1HGCS

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1 estimated implementation costs that would exceed the dollar-value associated with 2

completely eliminating all severe accident risk atSGS and HOGS,,Four,,SAMAs were 3

eliminated based on this screening, jeaving 25 ar1_

for further evaluatiorgfor SGS and 4

HCGS, respectively.

5 In the third step, PSEG estimated the benefits and the costs associated with each of the 6

SAMAsk Estimates were made of how much each SAMA could reduce risk Those 7

estimates were developed in terms of dollars in accordance with NRC guidance for 8

performing regulatory analyses (NRC, 1997,.T.he cost of implementing the proposed 9

SAMAs was also estimated.

10 Finally, in the fourth step, the costs and benefits of each of the remaining SAMAs were 11 compared to determine whether the SAMA was cost beneficial, meaning the benefits of the 12 SAMA were greater than the cost (a positive cost benefit_ _PSEG, concluded in its ERs that 13 several of the SAMAs evaluated are potentially cost-beneficial (PSEG 2009a, PSEG 2009b).,

14 The potentially cost-beneficial SAMAs do not relate to adequately managing the effects of 15 aging during the period of extended operatiorV Therefore, they need not be i mp!emented as 16 part of license renewal pursuant to 10 CFR Part 54, PSEG'sSAMA analysis and the NRC 17 staffs review are discussed in more detail below.

18 5.3.2 Estimate of Risk SDeleted CNS-1 Deleted:.

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" Deleted: N S Deleted: NPPD Deleted: (NPPD, 2008).

Deleted: (e.g., none of the potentially cost-0 I beneficial SAMAs would reduce the frequency or risk

I associated with aging-related failures)

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I 19 20 21 22 23 24 25 26 27

,SEG submitted an assessment of SAMAs forSGS and HCGS as part of the ERs (PSEG 2009a, PSEG 2009b. _For each_, two distinct analyses are combined to form the basis for the risk estimates used in the SAMA analysis: (1) the plant-specific Level-1 and Level-2 PSA models, which are updated versions of the IPEs (PSEG 1993, PSEG 1994, PSEG 1995); (2) a supplemental analysis of offsite consequences and economic impacts (essentially a Level-3 PSA model) developed specifically for the SAMA analysis. The most recent plant-specific Level-1 and Level 2 PSA models consisted of the following Internal Events PSAs: (1) for SGS, Salem PRA, Revision 4.1, September 2008, model of record (MOR); (2) for HCGS, the HC108B update. Neither of these includes external events,-------------------.

Deleted: NPPDs 1

4Deleted: NPPD Deleted: CNS-1 as part Deleted:.

Deleted: This assessment was based on the most recent CNS-1 probabilistic safety assessment (PSA) available at that time, a plant-specific offsite consequence analysis performed using the MELCOR Accident Consequence Code System 2 (MACCS2) computer program, and insights from the CNS-1 Individual Plant Examination (IPE) (NPPD, 1993) and Individual Plant Examination of External Events (IPEEE) (NPPD, 1996). As mentioned above, NPPD discovered an error in the method used to average their wind data. NPPD performed a sensitivity analysis using the corrected meteorological data.

Their analysis found that the error was conservative relative to the average population dose and economic cost, and that no SAMAs were inappropriately excluded from consideration in the LRA as a result of the error in wind direction. NPPD submitted their analysis and changes to the LRA in a letter dated December 7, 2009 (NPPD. 2009b).¶

1 The SGS CDF s approximately4.8 x 10 per year for internal events as determined from 2

quanti-fication of the Level-l-PRA model at a truncation of 1 x 10.11 per year. When 3

determined from the sum of the containment event tree (CET) sequences, or Level 2 PSA 4

model, the release frequency (from all release categories, which consist of intact 5

containment, late release, and early release) is approximately 5.0 x 10-5 per year, also at a 6

truncation of 1 x 1011 peryear. The latter value was used as the baseline CDF in the SAMA 7

evaluations (PSEG 2009a).- The CDF is base-d-roi-the Fik-s-s&ssrfie-nt for interhrllinitiatd -

8 events, which includes internal flooding. PSEG did not explicitly include the contribution 9

from external events within the SGS risk estimates; however, it did account for the potential 10 risk reduction benefits associated with external events by multiplying the estimated benefits 11 for internal events by a factor of 2i 4rhe breakdown of CDF by initiating event provided6in.

12 Table 5-1 13 Table 5-1., Salem Nuclear Station Core Damage Frequency for Internal Events Event CDF

% Contribution to (per year)

CDF Formatted: Font: Not Bold DeleteO: a Formatted: Font: Not Bold l(

ueetea:

Formatted: Font: Not Bold Deleted: The CNS-1 core damage frequency (CDF) is approximately 9.3 x 10n per year for internal events as determined from the quantification of the Level 1 PSA model. When determined from the sum of t Deleted: Transients Deleted: 3.0 x 10-6 Deleted: 2

• jFormatted: Superscript

/ Deleted Loss of DC Power Deleted: 2.1 Deleted: Loss of Coolant Accidents (LOCA)

[Deleted: 1.4 F-Deleted: 15 I

,4oss 9f Control AreaVentilatioln oqss ofOffsite Power (LOOP)

!-qss of Service water -

t.1 x 10-6....

~.6§2 1Q§.

1

.-1* Deleted: Loss of Feedwater irnternal Floods

.,.§ 5 Q-§ 9

,Transients -

,Steam Generator Tupe Rupture (SGTR).....

,Loss of Component Cooling Water (CCW)..

,.nticipated Transient Without Scram (A_'SI

,oss of 125 V DC Bus A Others (less than 1 percent) 3

,4.9 2ý1(-,

_2-.Y_ 3ý j 0 ---

Ox9

ý q7-,

.7._4 x 10-7..

9 x

.I.0-7 1

1.8x10-6,------------------------------

-( Deleted: 1.0 Deleted: 11 Deleted: Loss of Offsite Power (Deleted:

6.5....0.10-7

.fI

'{Deleted:... 7 Deleted: Loss of Service Water eleted: 6.0....7.

10-7 rTi]

t Deleted: 7 Deleted: Loss of AC Buses Deleted: 2.6....

.0 x10-7 Deleted: 3 Deleted: Internal Flood

• ,{ Deleted: 2.6 Total CDF (Internal Events) 4.8,x _0-*5 100 14 15 16 17 18 19 20 21 22 As shown in this table, events initiated byjosses of control area ventilation, offsite power, or service water are the dominant contributors to the MDr. PSEG identified that Station

\\

Blackout (SBO) contributes to 8 x 10,per year_(PSEG_2010_a_)._

,PSEG estimated the dose to the population within 50 miles (80 kin) of theSGS site to be approximately 0,78 person-sievert (p_erson-Sv__,7 person-rem)_pe_ r year The breakdown of the total population dose by containment release mode is summarized in Table 5-Containmentb~ypass events (.sUh_as.SGTR-initiated large early release freque ncy (LERF).

accidents) and late containment failures without feedwater dominate the population dose risk at SGS, eleted: 3 it Deleted: Interfacing System LOCA Deleted: 5.

"Formattedteri Deleted: 8 Deleted: 9.3...x 10-6 Deleted: transients...osses of control ar 8

Formatted: Superscript I Deleted: NPPD... SEG estimated the d

Table 5-2,Breakdown of Population Dose by Containment Release Mode For SGS Population Dose

% Contribution=

~(Person-RemI Per Year)

Containment Release Mode(ProRe 1 PrYa)

%Ctrbtn--

,ontainment over-pressure (late) 4 2.9 5-..

Steam Generator Rupturs 31.9 41 Containment Isolation Failure.2.3

.3_

jnact Containm ent,..

...-.-.-.. - <1..

Snterface systemLOCA.....

6 _

-1 Catastrophic Islaotion Failue 0.4

<1 Basemat melt-through (late) negligible negligilbe.

Total.............--

78.2--.-.-.-.-..

1_00 1One person-rem = 0.01 person-Sv - -.................................

• A" Formatted Fr -in-11 2

3 4

5 6

7 8

9 10 11 12 13 14

,The HCGS CDF is apjroximately5. 1x1_0_ per year as determined from quantification of the Level 1 PRA at a truncation of I x 1012 per year. When determing form the sum of the containment event free (CET) sequences, or Level 2 PRA model, higher truncation of 5 x 10 11per a year used and the resulting release frequency (from all release categories, which consist of inact containment, late release, and early release) is approximately 4.4 x 10.6 per year. The Latter value was used as the baseline CDF in the SAMA Evalutions (PSEG 2009b). Although this is about 16% less that the internal events CDF of 5.1 x 10°' per year obtained from the Level-1 model, the NRC staff considers that its use will have a negligible impact on the results of the SAMA evaluation because the external event multiplier and uncertainty multiplier used in the SAMA analysis have a much greater impact on the SAMA evaluation results than the small difference arising from the model quantification approach.

The breakdown of CDF by initiating event is provided in Table 5-3.

Table 5-,3, Hope Creek Nuclear Station Core Damage Freauencv for Internal Events

__ FomatDeletd:

4.

/ Deleted: Early Containment Failure

/'*(Deleted: 78 Formatted Table (Deleted: 1.67 (Formatted

.131 (Deleted: Intermediate

. Deleted: 0.47 Formatted Deleted: 22 Deleted: Late... nact Containment Failure Deleted: <0.1 Formatted 4'

'\\ \\Deleted: Intact Containment Deleted: Negligible Formatted

... 17 0'

Deleted: Negligible¶

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Formatted Deleted: 2.14

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Formatted

.. f[91 Formatted

[01

\\j Deleted: ¶ Formatted

.2 1 Deet: d....22 Deleted: Loss of Control Area Ventilatioln Deleted: 1.8-....3-x-10-6 (Deleted: 37 Deleted: Loss of Offsite Power (LOOP)-

Deleted:,

[Deleted: 17

, Deleted: Loss of Service water Deleted: 6.6.... 7 x 10-6

... [241 J,

/" Deleted: 14 Deleted: Internat Floods

-,,fDeleted:

4.5...2.~10-6 TT5 Deleted: 9 Deleted: Transients Deleted: 4.o....8 x10-6 F.

261

- Deleted: S G

/ Deleted: Steam Generator Tupe Rupture(STR CDF

% Contribution to Initiating Eventyear)

CDF

,Loss of Offsite Power 9._3 x 1 W

.18

,Loss of Service Water_(S W) 8.1'x10-,-.-.-15

,.Vlanual Shutdown..7.7.

0-v J 5 juqrine Trip with Bypass 2.x10--

,Small Loss of Coolant Accident (LOCA)-Water_(Below.

. *.8x 1 Top of Active Fuel)

Small LOCA-Steam (Above Top of Active Fuel32 Deleted: 2.7....3 x 10-6

-- ff[271--- ýfl

.. 2.3 o,

-4

-- Deleted: 6 I

toss of Condenser Vacuum tire Protection System Rupture Outside Control Room

,solation LOCA in Emergency Core Cooling System (ECCS) Discharge Paths

,Main Steam Isolation Valve (MMSIV)_Closure Internal Flood Outside Lower Relay Room Loss of Feedwater Loss of Safety Auxilaries Cooling System

,Reactor Auxiliaries Cooling Systemn (RACS) Common...

Header Unisolable Rupture Unisolable SWA Pipe Rupture in RACS Room Unisolable SW B Pipe Rupture in RACS Room Others (less than 1% each) 1_,ýj 7

1.1x10-7 2

1.1 x 10-7 2

9.7 x 1 -8 2

8.8 x 10-8 2

7.9 x 10-8 2

7.6 x 10-8 1

5.7 x 10-8 1

5.7 x 10-8 1

4.1x10-6 Deleted: Loss of Component Cooling Water (CC\\AO Total CDF (Internal Events) 5.t1 x 10q-6 100 1

As shown in this table, events initiated by los4pf offsite power, 3loss of service water and 2

other transients (manual shutdown and turbine trip with bypass) are the dominant 3

contributors to the CDF. Aticipated transient without s*ram (ATWS) sequence.s a c.unt for 4

33 percent of the CDF, station blackout accounts for 12 percent of the CDF (PSEG 2010b).

5 PSEG estimated the dose to the population within 50 miles (80 km) of theHCGS site to be 6

approximately 023_person-sievert (person-Sv)_(.2.9 person-rem) per year*_ Te breakdown 7

of the total population dose by containment release mode is summarized in Table 5-A.4 8

Releases from the containment within the early time frame (0 to less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following 9

event initiaton) and intermediate time frame (4 to less that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following event initiation) 10 dominate hhe population dose risk atHCGS.

11 Table 5-4 Breakdown of Population Dose by Containment Release Mode For HCGS Deleted: 1.0.0 10-6.1..

[281 "1" Deleted: 2 Deleted: Anticipated Transient Without Scram

~'1(ATWS)9 D leted: 7.4

,1Deleted: 2 SDeleted: Lossof 125 V DC Bus A

>1 Deleted: Others (less than 1 percent)3

]

Formatted: Tab stops: 3.04", Right

-.1 Deleted: Loss of Safety Deleted: 11l

/ 4

'Delleted: 6.9 10-7¶

'//1.8.10-6 Formatted: Left Deleted: 4.8....1 x 10-5 8.

29

/Deleted: es...

of control area ventilatioq-r--

/ 7 Deleted: SGS...CGS site to be approx' Deleted: Containment over-pressure (late)

Formatted Table Deleted: 42.9 Formatted: Centered

/ Deleted: 55 Deleted: Steam Generator Rupturs Deleted: 31.9 Deleted: 41 Deleted: Containment Isolation Failure

/

Deleted: 2.3 Formatted: Centered Deleted: 3 Formatted: Centered eDeleted: 0.2 Deleed

<1]

Deleted: Interface system LOCAJI Catastrophic Islaotion Failue¶ Basemat melt-through (late)

Deleted: 0.6¶S

' {Formatted: Centered SDeleted: 41

",'), negligible Deleted: <1ij negligilbe Containment Release Mode Population Dose

% Contribution (Person-Rem1 Per Year)

,Early Releases (.< 4h.rs).

jnterrnediate Releasesj4 to< 24hrs-Jate ReleasesA *_24hbr!s)_

Inact Containment v11.9 5 2. _

_ _ __,43...

4J

<0.1

<0. 1.

negligible,

Total p22.9 100 1One person-rem = 0.01 person-Sv Deleted: 78.2 - -I Deleted:

78.2 jJ Formatted: centered 1 IThe NRC staff has reviewed PSEG's data and evaluation methods and concludes that the Deleted: ¶ 2

quality of the risk analyses is adequate to support an assessment of the risk reduction.

Deleted: NPPDs 3

potential for candidate SAMAS,_ Accordingly, the staff based its assessment of offsite risk on.

Deleted:.

4 the OD~

s and offsite doses reported byPSEG..

Deleted: NPPD in their December 2009 letter

[Deleted: (NPPO, 2009b) 5 5.3.3 Potential Plant Improvements 6

Once the dominant contributors to plant risk were identified,,PSEG searched for ways to_ - -

Deleted: NPPD 7

reduce that risk. In identifying and evaluatingpotential SAMAs, _pSEG considered insights -

Deleted:.

8 from the plant-specific PSA, and SAMA analyses performed for other operating plants that Deleted: NPPD 9

have submitted license renewal applications PASEG identified,27and 23 potential risk-4_.

D 10 reducing improvements (SAMAs) to plant components, systems, procedures and training for Ded 11 SGS and HCGS, respectively.

Deleted: NPPD Deleted: 244 12 PSEGpemoved all but 25 and 21of the SAMAs from further consideration for SGS and Deleted: NPPD 13 HCGS because they are not applicable at SGS or HCGS, due to design differences, have.

Deleted: 80 14 already been implemented atSGS or HCGS, or are similar in nature and could be combined (Deleted: CNS-1 15 with another SAMA candidate_ Adetailed cost-benefit analysis was performed for each of.

Deleted: CNS-1 16 the remaining SAMAs.

Deleted:.

I 17 18 19 20 The staff concludes thatPSEG used a systematic and comprehensive process for identifying potential plant improvements forSGS and HCGS, and that the set of potential plant improvements identified byPSEG is reasonably cormprehensive and. therefore...

acceptable.

- Deleted: NPPD

-C et:

CNS-1 Deleted: NPPD 21 5.3.4 Evaluation of Risk Reduction and Costs of Improvements 22 JPSEGevaluated the risk-reduction potential of the remaining 25,4anrdl2_1 SAMAs for SGS Deleted: NPPD 23 and HOGS, respectively,_ _T~h~ern~Majority of the SAMA evaluations were performed in a

,Deleted:

29 24 bounding fashion in that the SAMA was assumed to completely eliminate the risk associated -

Deleted: 2380 25 with the proposed enhancement.

Deleted:.

26 PSEG estimated the costs of implementing the candidate SAMAs through the application of Deleted: NPPD 27 engineering judgment and use of other licensee's estimates for similar improvements, The..... -(

Deleted:.

28 cost estimates conservatively did not include the cost of replacement power during extended 29 outages required to implement the modifications, nor did they include rontingency cost for........

- Deleted: maintenance and surveillance costs of the 30 unforeseen difficulties, installed equipment 31 The staff reviewedPSEG's bases for calculating the risk-reduction for the various plant Deleted: NPP's 32 33 improvements and concludes that the rationale and assumptions for estimating risk reduction are reasonable and generally conservative (i.e., the estimated risk reduction is

1 higher than what would actually be realized)' Accordingly, the staff based its estimates of

].

- Deleted:.

2 averted risk for the various SAMAs onSE(G's riskl reduction estimates.-

Deleted: NPPD's 3

The staff reviewed the bases for the applicant's cost estimates, For certain impropvements

-",Deleted:. and advanced light-water reactors.

4 the staff also compared the cost estimates to estimates developed elsewhere for similar Deleted: nd 5

improvements, including estimates developed as part of other licensee's analyses of SAMAs Deleted: NPPD 6

for operating reactors. The staff found the cost estimates to be reasonable,_and generally _

,' 9Deleted: NPPD 7

consistent with estimates provided in support of other plants' analyses.

i Deleted:.

/

Deleted:.

8 The staff concludes that the risk reduction and the cost estimates provided by,PSEG are _ _

,, Deleted:.

9 sufficient and appropriate for use in the SAMA evaluation.

"'," Deleted: NPPD

]

]

-J

-J 4

Deleted: both IN Deleted-: both 10 5.3.5 Cost-Benefit Comparison 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 Deleted: NPPD The cost-benefit analysis performed bypýSEG was based primarily on NUREG/BR-0184 Deleted: 8 (NRC, 1997) and was executed consistent with this guidance, NUREG/BR-0058 has Deleted: NPPD recently been revised to reflect the agency's revised policy on discount rate-s, Revision 4 of Deleted: eight seveteen NUREG/BR-0058 states that two sets of estimates should be developed - one at 3 percent Dle and the other at 7 percent (NRC, 2004..

PSE_ providedboth~setsofestimates forSGS and De' l HCGS (PSEG 2009a,2009b.......................................

.,.. Deleted: for SGS.

S Deleted: 25 j

f fthe renactor'core isolatic For SGS, PSEG identifiedpleven potentially cost-beneficial SAMAs in the baseline analysis f'

DtDevelop proct lures to allow bypass on cooline (ROClC contained in the EP, Thepotentially cost-beneficial SAMAs are:

0 0

0 0

0 0

SAMA 1,-.Enhance proqed.u res -andlprovide additional equipment to respond to loss of control area ventilation.

J SAMA2 -_,l-p_-cornfigure Salem 3_to provide a moe expedient backupp to, AC power source for Salem 1 and 2..

SAMA,4_-_,lnstall fuel oit transfer pump on "C" emergency diesel generator (EDG) and provide procedural guidance for using "C" EDG to power selected "A" and "B" loads.

SAMA

-,6_-ýEnnce flood_detection for 84' auxiliary building and enchance procedural guidance for responding to service water flooding SAMA,9_-_,ConnectHope Creek cooling tower basin to Salem service water system as alternate service water supply,----------------

SAMA 10 -Provide procerdural guidance for faster cooldown on Loss of reactor coolant pump (RCP) Seal i/ I turbine exhaust pressure trip, extending the time available for RCIC operation Deleted: 30 Deleted: Revise SDeleted: procedures to allow manual

/

alignment of the fire water system to the

/

residual heat removal (RHR) heat exchangers, providing improved ability to coot the RHR heat exchangers in a loss of service water (SW)

Deleted: 33

-4 Deleted: Provide for the ability to establish an emergency connection

- Deleted: of existing or new water sources to feedwater and condensate systems, increasing availability of feedwater

- D-eleted: 40 Deleted: Revise procedures to provide additional space cooling to the emergency diesel generator (EDG) room via the use of portable equipment, increasing availability of the EDG.

Deleted: 453 Deleted: Provide an alternate means of supplying the instrument air header, increasing availability of Deleted: Instrument air 8Deleted:

6 Deleted: Revise procedures to allow thf 12

.-t Deleted: 78 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 SAMA 11,Modify p!ant procedures to make use of other Unit's_ P!DP for RCP seal.

0 SAMA,2 -,Im.poye flood barriers outside 220/440VAC sWitchgear............... -

rooms.

0 SAMA 14 -. Expand ATNýT nitigation system actuation circuitry ------------

(AMSAC) function to include backup breaker trip on RPS failure.

SAMA 17 -,Enhance pro9cedures and provide additional quipment to respond to loss of EDG control room ventilation.-

SAMA 24 -,Provide proceduralguidance to cross-tie Salem 1 and 2 service water systems.

PSEG performed additional analyses to evaluate the impact of parameter choices and uncertainties on the results of the SAMA assessment (PSEG, 200,9_4)_ If the benefits are -

increased by an additional factor of.1.64 to account for uncertainties five additional SAMA candidates were determined to be potentially cost-beneficial. The ER also showed that the sentivity case SAMA (SAMA 5A) was potentially cost-benificial:

Deleted: Improve training on alternate injection via the fire water system, increasing the availability of alternate injection Deleted: 79 Deleted: Revise procedures to allow use of the residual heat removal service water (RHRSW) system without a service water booster pump, increasing availability of the RHRSW system.

Deleted: Revise procedures to allow the ability to cross-connect the circulating water pumps and the service water going to the turbine equipment cooling (TEC) heat exchangers, which allow continued use of the power conversion system after service water is lost Deleted: Revise procedures to allow the ability to cross-connect the circulating water pumps and the service water going to the turbine equipment cooling (TEC) heat exchangers, which allow continued use of the power conversion system after service water is lost Deleted: Revise procedures It cross-connect the circulating w service water going to the turbi (TEC) heat exchangers, which the power conversion system a lost 0

0 0

0 Deleted: VI uW" NPPD SAMA 3 - Install limited emergency diesel generator (EDG) cross-tie

'I',

Deleted: (NPPD capability between Salem 1 and 2.,,

',,\\

Formatted: Not Highlight 10Deleted:86 SAMA5--Jnstall portable diesel genetrators to charge station battery Formatted: Highlight and circulating water batteries and replace PDP with air-cooled pump.

Deleted Deleted: 3 SAMASA-_jnstal! portable diesel generators to charge station battery.

Deleed: thre and circulating water batteries.

Fomttd Highlight

\\* ormatd: Unnumbered List SAMA 7,-.,nstall "B" Train auxiliary feedwater storage tank (AFWST)

Formatted: Font color: Auto makeup including alternative water source.

Deleted: 14 o allow the ability to ater pumps and the ne equipment cooling allow continued use of ifter service water is SER 1st LINE enerator to supply DC ng a station blackout lable for AC power o allow use of 33 SAMA 8 - Install high pressure pump powered with portable diesel generator and long-term suction source to supply the AFW Header.

SAMA 27 - In addition to the equipment installed for SAMA 5, install permanently piped seismically qualified connections to alternative AFW water sources.

iiDeleted: Provide a portable g power to individual panels duri

'I*

(SBO), Increasing the time ava Lo recovery

,0 Deleted: 64

[Deleted: Revise procedures t "I Formatted: Indent: Left: 0.56", Hanging: 0.82" J Deleted: 75 I

I Deleted: Implement Generation Risk Asses(f341

,PSEG indicated that all 17_potential!y cost-beneficial SAMAs will be considered for -

- -I Deleted: T..51 implementation through the established Salem Plant Health Committee process5- - - - - - - - - - -

" Deleted: detailed engineering project cost-b[

'" Formatted: Not Highlight

1 2

3 4

5 ForHCGS, PSEG identifiednine potentially cost-beneficial SAMAs in the baseline analysis Deleted: SGS contained in the ER. The potentially cost-beneficial SAMAs are:

Detd: eleven Deleted: Enhance procedures and provide additional equipment to respond to loss of control area ventilation I

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 0

0 SAMA 1 -,Remove automatic depressurizationsyst~er (ADS)_inhibit.

from non-AtWS emergency operating procedures.

SAMA3_--Jnstall backup air compressor to supply air-operated valves..

SAMA 4 -,Provide procedural guidance to cross-tie residual heat removal (RHR) trains.

SAMAII

_0-,Pjrgvide proceduralrguidance to use B.5.b low pressure pump for non-security events______

SAMA,17 -,Replace a supply fan witha different design in service water pum p room,_

SAMA,18 -- Replacea return fan with a different design In service water pump room.

SAMA,30-,Provide proceduralrguidance for partial transfer function of control functions from the control room to the remote shutdown panel. -

SAMA 35,-,Relocate, minimize, and/or eliminate electrica l _heaters__in electrical access room.

SAMA 39,Provide procedural gdane to _bypass reactor core isolation cooling turbine exhaust pressure trip.

! DeltedDeleted:

Re-configure Deleted: Aovs l Deleted: Salem 3 to provide a moe expedient b bacukup to AC power source for Salem 1 and 2.¶

• Deleted: Install fuel oit transfer pump on "C" emergency diesel generator (EDG) and provide procedural guidance for using "C" EDG to power selected "A" and "B" loads.

Dele, Enhance Deleted: flood detection for 84' auxiliary

'\\

building and enchance procedural guidance for I

responding to service water flooding F\\rDeleted: 9 Deleted: Connect Hope Creek cooling tower basin to Salem service water system as alternate service Deleted: water supply.

1 V Deleted: 10

'Deleted: Provide procerdural guidance for faster cooldown on Loss of RCP Seal

]1 Deleted: 11 3

Deleted: Modify plant procedures to make use of other Unit's PDP for RCP seal.

'j Deleted: 12 Deleted: Improve flood barriers outside S220/44VAC switchgear rooms.

J Deleted: 14 Deleted: Expand AMSAC function to include backup breaker trip on RPS failure.

l ' Deleted: RCIC

,PSEG performed additional analyses to evaluate the irmpa.ct of parameter choices and uncertainties on the results of the SAMA assessment CPSEG, 2009b,._ Iff the benefits are increased by an additional factor of 2.84,to account for uncertainties,,our additional SAMA candidates were determined to be potentially cost-beneficial:

• \\1 SAMA 8 - Convert selected fire protection piping from wet to dry pipe system.

SAMA 32 - Install additional physical barriers to limit dispersion of fuel oil from DG rooms.

SAMA 7 - Provide procedural guidance for loss of all 1E 120V AC poweý.------------------------------------------

" Deleted:.. SAMA 17 - Enhance procedures IW, and provide additional quipment to respond to I loss of EDG control room ventilation.¶

,,w

• . SAMA 24-Provide procedural guidance to 11W cross-tie Salem 1 and 2 service water systems.¶ 0

%11' Formatted: Not Highlight

( Deleted: a

( Deleted: 1.64

( Deleted: three

( Deleted: a

1 2

3 4

5 6

0 SAMA 37 - Reinforce 1 E 120V AC distribution panels.

7 8

9 10 11 12 13 14 15 16 17 18 19 PSEG jindicated thatalL 13potentially cost-beneificial SAMAs will be considered for Deleted:

to.

SAMAr generators to charge implementation through the established-HCGS Plant Health" Committ-ee-process,..

water batteries and re

-~pump.¶

,ased on the staffs review. the NRC staff concludes that, with the exception of the 4

chSAMA sAt-Instatt charge station batter~

p oten i ally cost-b-en e ficial- -

SAMAs-di scu ssed-above, t he costs of the SA M As evalu ated-SAMA 7 - install would be higher than the associated benefits.

including alternative S

• . SAMA 8 - Instaltl with portable diesel g 5.3.6 Conclusions source to supply the,

-. SAMA 27-tn add for SAMA 5, install pe The staff reviewed PSEG's analysis and concluded that the methods used and the

,,,IM qualified connections implementation of those methods were sound. The treatment of SAMA benefits and costs sources support the general conclusion that the SAMA evaluations performed by1PSEG are I,

Formatted: Not Higi reasonable and sufficient for the license renewal submittal.

II'I. Deleted: detailed II Formatted: Not Higi Based on its review of the SAMA analysis, the staff concurs with,PSEG_ s identification of II Deleted: engineerin areas in which risk can be further reduced at both SGS and HCGS in a cost-beneficial I t i have been initiated fo areas 1

beneficial SAMAs (Pt manner through the implementation of all identifiedpotentially cost-beneficial SAMA~-s --

Given the potential for cost-beneficial risk reduction, the staff considers that further I I I I Formatted: Not Higl consideration of these SAMAs byPSEG is warrantedL However, none of the potentially__...

Formatted: Not Higi cost-beneficial SAMAs relate to adequately managing the effects of aging during the period I I Formatted: Not Higi of extended operation forSGS or HCGS Therefore, they need not be implemented as part Deleted: NRC staff r of the license renewal pursuant to 10 CFR Part 54.

analysis as submitted I

II error was conservativ population dose and i

no SAMAs were inap I

consideration in the L ti 1 3 B III Formatted: Normal

- Install portable diesel station battery and circulating eplace PDP with air-cooled portable diesel generators to and circulating water batteries.¶ B" Train AFWST makeup water source.¶ igh pressure pump powered enerator and long-term suction AFW Header.$

ition to the equipment installed rmanently piped seismically to alternative AFW water hlight Wlight g project cost-benefit analyses r the 1316 potentially cost-SEG 2009ba) hlight hlight hlight eviewed NPPD's PSEG re-1 by NPPD and.agrees that the

'a relative to the average offsite economic cost and that propriately excluded from RA as a result of the error.¶(

II M

Deleted: and the supplemental information I"

provided by NPPD Deleted: NPPD's I

Deleted: NPPD IIDeleted: NPPDs Deleted: or a subset of Deleted:

Deleted:.

Deleted: NPPD Deleted:.

I( Deleted: fpr I

Deleted:. (e.g., none of the potentially cost-beneficial SAMAs would reduce the frequency or risk associated with aging-related failures).

', I Formatted: Font: Arial Bold, 12 pt, Bold, All 1 Lcaps I

1

5.4 REFERENCES

Deleted: <#>es%

2

,Public Service Electric and Gas Company (PSEG)_. 1993. Letter from Stanley LaBruna, -

Electric Power Research Institute (EPRI). 1989.

SProbabilistic Seismic Hazard Evaluations at 3

PSEG, to NRC Document Control Desk.

Subject:

"Generic Letter 88-20; Individual Plant Nuclear Plant Sites in the Central and Eastern 4

Examination (IPE) Report, Salem Generating Station, Unit Nos. 1 and 2, Docket Nos. 50-United States; Resolution of the Charleston 5

272 and 50-311," Hancocks Bridge, New Jersey. July 30, 1993. Accessible at Earthquake Issues." EPRI NP-6395-D, EPRI Project P101-53.. Palo Alto. CA. April 1989.¶ 6

ML080100047.

Electric Power Research Institute (EPRI). 1991.

"A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Implementation Guide NP-6041, Revision 1. Palo Alto, CA.

7 Public Service Electric and Gas Company (PSEG). 1994. "Hope Creek Generating Station.

August 1991.¶ 8

Individual Plant Examination." April 1994. Accessible at ML080160331.

Electric Power Research Institute (EPRI). 1993.

"Fire Induced Vulnerability Evaluation (FIVE)

Methodology." TR-100370, Revision 1, Palo Alto, CA. September 19, 1993.¶ 9

Public Service Electric and Gas Company (PSEG). 1995. Letter from E. Simpson, PSEG, Nuclear Energy Institute (NEI). 2005. "Severe 10 to NRC Document Control Desk.

Subject:

"Response to Generic Letter 88-20 Individual Accident Mitigation Alternative (SAMA) Analysis Guidance Document", NEI 05-01, Rev. A.

11 Plant Examination for Severe Accident Vulnerabilities - 1 OCFR50.54 (f) Request for Washington. D.C. November 2005.1 12 Additional Information Salem Generating Station, Unit Nos. 1 and 2 Facility Operating Formatted: No bullets or numbering 13 License Nos. DRR-70 and DPR-75 Docket Nos. 50-272 and 50-311," Hancocks Bridge, New 14 Jersey. August 01, 1995. Accessible at ML080100021.

15 Public Service Electric and Gas Company (PSEG). 1996. Letter from E. C. Simpson, 16 PSEG, to NRC Document Control Desk.

Subject:

"Response to Generic Letter No. 88-20, 17 Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident 18 Vulnerabilities, Salem Generating Station Units Nos. 1 and 2, Facility Operating License 19 Nos. DPR-70 and DPR-75, Docket Nos. 50-272 and 50-311," Hancocks Bridge, New Jersey.

20 January 29, 1996. Accessible at ML080100023.

21 Public Service Electric and Gas Company. (PSEG). 1997. "Hope Creek Generating Station 22 Individual Plant Examination of External Events (IPEEE) for Severe Accident 23 Vulnerabilities." July 1997. Accessible at ML080160320.

24 PSEG Nuclear, LLC (PSEG). 2009a. Salem Nuclear Generating Station -- License 25 Renewal Application, Appendix E: Applicant's Environmental Report; Operating License 26 Renewal Stage. Hancocks Bridge, New Jersey. August 18, 2009. Accessible at 27 ML092400532.

28 29 PSEG Nuclear, LLC (PSEG). 2009b. Hope Creek Generating Station - License Renewal 30 Application, Applicant's Environmental Report, Operating License Renewal Stage, August 31 2009. Accessible at ML092430484.

1 PSEG Nuclear, LLC (PSEG). 2010a. Letter from Paul. J. Davison, PSEG, to NRC 2

Document Control Desk.

Subject:

"Response to NRC Request for Additional Information 3

dated April 12, 2010, related to the Severe Accident Mitigation Alternatives (SAMA) review 4

of the Salem Nuclear Generating Station, Units 1 and 2," Hancocks Bridge, New Jersey.

5 May 24, 2010. Accessible at ML101520326.

6 PSEG Nuclear, LLC (PSEG). 2010b. Letter from Paul J. Davison, PSEG, to NRC 7

Document Control Desk.

Subject:

"Response to NRC Request for Additional Information 8

dated April 20, 2010, related to the Severe Accident Mitigation Alternatives (SAMA) review 9

associated with the Hope Creek Generating Station License Renewal Application,"

10 Hancocks Bridge, New Jersey. June 1, 2010. Accessible at ML101550149.

11 PSEG Nuclear, LLC (PSEG). 2010c. Letter from Christine T. Neely, PSEG, to NRC 12 Document Control Desk.

Subject:

"Supplement to RAI responses submitted in PSEG Letter 13 LR-N10-0164 dated May 24, 2010, related to the Severe Accident Mitigation Alternatives 14 (SAMA) review of the Salem Nuclear Generating Station, Units 1 and 2," Hancocks Bridge, 15 New Jersey. August 18, 2010. Accessible at ML102320211.

16 PSEG Nuclear, LLC (PSEG). 2010d. Letter from Christine T. Neely, PSEG, to NRC 17 Document Control Desk.

Subject:

"Supplement to RAI responses submitted in PSEG Letter 18 LR-N10-0181 dated June 1, 2010, related to the Severe Accident Mitigation Alternatives 19 (SAMA) review of the Hope Creek Generating Station," Hancocks Bridge, New Jersey.

20 August 18, 2010. Accessible at ML102320212.

21 U.S. Nuclear Regulatory Commission (NRC). 1989. Fire Risk Scoping Study. NUREG/CR-22 5088. January 1989. Washington, D.C.

23 U.S. Nuclear Regulatory Commission (NRC). 1991b. "Procedural and Submittal Guidance 24 for the Individual Plant Examination of External Events (IPEEE) for Severe Accident 25 Vulnerabilities." NUREG-1407. Washington, D.C. June 1991.

26 U.S. Nuclear Regulatory Commission (NRC). 1994. Revised Livermore Seismic Hazard 27 Estimates for Sixty-Nine Nuclear Plant Sites East of the Rocky Mountains. NUREG-1488, 28 April 1994. Washington, D.C.

29 U.S. Nuclear Regulatory Commission (NRC). 1997. Regulatory Analysis Technical 30 Evaluation Handbook. NUREG/BR 0184, Washington, D.C. January 1997.

31 32 U.S. Nuclear Regulatory Commission (NRC). 1998. Code Manual for MACCS2.

NUREG/CR 6613, Washington, D.C. May 1998.

1 U.S. Nuclear Regulatory Commission (NRC). 1999a. Letter from Patrick D. Milano, U.S.

2 NRC to Harold W. Keiser, PSEG.

Subject:

Generic Letter 88-20, Supplement 4, "Individual 3

Plant Examination for External Events for Severe Accident Vulnerabilities," Salem Nuclear 4

Generating Station, Unit Nos. 1 and 2 (TAC Nos. M83669 and M83670). May 21, 1999.

5 U.S. Nuclear Regulatory Commission (NRC). 1999b. Letter from Richard B. Ennis, U.S.

6 NRC, to Harold W. Keiser, PSEG.

Subject:

"Review of Individual Plant Examination of 7

External Events (IPEEE) Submittal for Hope Creek Generating Station (TAC No. M83630)".

8 April 26, 1999.

9 U.S. Nuclear Regulatory Commission (NRC). 2001. "Review of Columbia Generating 10 Station Individual Plant Examination of External Events Submittal (TAC No. M83695)."

11 Washington, D.C. February 26, 2001. (ADAMS Accession No. ML010570035) 12 U.S. Nuclear Regulatory Commission (NRC). 2004. Regulatory Analysis Guidelines of the 13 U.S. Nuclear Regulatory Commission. NUREG/BR-0058, Revision 4, Washington, D.C.

14 September 2004.

15 U.S. Nuclear Regulatory Commission (NRC). 2010a. Letter from Charles Eccleston, U.S.

16 NRC, to Thomas Joyce, PSEG.

Subject:

Request for Additional Information, Regarding 17 Severe Accident Mitigation Alternatives for the Salem Nuclear Generating Station, Units 1 18 and 2. April 12, 2010. Accessible at ML100910252.

19 U.S. Nuclear Regulatory Commission (NRC). 2010b. Letter from Charles Eccleston, U.S.

20 NRC, to Thomas Joyce, PSEG.

Subject:

Revised Request for Additional Information 21 Regarding Severe Accident Mitigation Alternatives for Hope Creek Generating Station. May 22 20, 2010. Accessible at ML101310058.

23 U.S. Nuclear Regulatory Commission (NRC). 2010c.

Subject:

Summary of Telephone 24 Conference Held on July 29, 2010 between the U.S. Nuclear Regulatory Commission and 25 PSEG Nuclear LLC, Concerning Follow-up Questions Pertaining to the Salem Nuclear 26 Generating Station, Units 1 and 2, and Hope Creek Generating Stations License Renewal 27 Environmental Review. August 13, 3010. Accessible at ML102220012.

Deleted: 10 CFR Part 50. Code of Federal Regulations, Title 10, Energy, Part 50, 'Domestic Licensing of Production and Utilization Facilities." 7 10 CFR Part 51. Code of Federal Regulations, Title 10, Energy, Part 51," Environmental Protection Regulations for Domestic Licensing and Related S,---Regulatory Functions."¶ 10 CFR Part 54. Code of Federal Regulations, Title 10, Energy, Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." I 10 CFR Part 100. Code of Federal Regulations, Title 10, Energy, Part 100, "Reactor Site Criteria." 7 Nebraska Public Power District (NPPD). 1993.

"Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54(f), IPE Report Submittal, Cooper Nuclear Station, Docket No. 50-298, DPR-46," March 1993.

ADAMS Accession Nos. ML073600192 (Volume 1) and ML073600193 (Volume 2).7 Nebraska Public Power District (NPPD). 1996.

"Individual Plant Examination for External Events Report - 10 CFR 50.54(f), Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46,"

October 1996. ADAMS Accession No. ML073580487.1 Nebraska Public Power District (NPPD).2008.

Cooper Nuclear Station - License Renewal Application, Appendix E: Applicant's Environmental Report, Operating License Renewal Stage.

Columbus, Nebraska, September 24, 2008 ADAMS Accession Nos.ML083030246 (main report) and ML083030252 (attachments).¶ Nebraska Public Power District (NPPD).2009. Letter from Stewart B. Minahan, NPPD to NRC Document Control Desk.

Subject:

Response to Request for Additional Information for License Renewal Application - Severe Accident Mitigation Alternatives, Cooper Nuclear Station, Docket No. 50-298, DPR-

46. July 1, 2009. ADAMS Accession No. ML0918803193.,

Nebraska Public Power District (NPPD). 2009b.

Letter from Stewart B. Minahan, NPPD to NRC Document Control Desk.

Subject:

SAMA Meteorological Anomaly Related to the Cooper Nuclear Station License Renewal Application, Cooper Nuclear Station, Docket No. 50-298, DPR-

46. December 7, 2009. ADAMS Accession No. ML0934909971 U.S. Nuclear Regulatory Commission (NRC). 1996.

Generic Environmental Impact Statement for License Renewal of Nuclear Plants. NUREG-1 437. Vols. 1 and 2, Washington, D.C. ADAMS Accession No. ML0617706057 U.S. Nuclear Regulatory Commission (NRC). 1997.

Regulatory Analysis Technical Evaluation Handbook.

NUREG/BR-0184, Washington, D.C.7

.U.S. Nuclear Regulatory Commission (NRC). 2004.

Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission. NUREG/BR-0058, Rev. 4, Washington. D.C.

-Section Break (Next Page)

1 THIS PAGE IS INTENTIONALLY LEFT BLANK

Page 5: [1] Deleted Author The CNS-1 core damage frequency (CDF) is approximately 9.3 x 10-6 per year for internal events as determined from the quantification of the Level 1 PSA model. When determined from the sum of the containment event tree sequences, or Level 2 PSA model, the release frequency is approximately 1.2 x 105 per year. The latter value was used as the baseline CDF in the SAMA evaluations. The CDF value is based on the risk assessment for internally-initiated events. NPPD did not include the contributions from external events within the CNS-1 risk estimates; however, it did account for the potential risk reduction benefits associated with external events by increasing the estimated benefits for internal events by a factor of 3. The breakdown of CDF by initiating event is provided in Table 5-3.

Page 5: [1] Deleted Author The CNS-1 core damage frequency (CDF) is approximately 9.3 x 106 per year for internal events as determined from the quantification of the Level 1 PSA model. When determined from the sum of the containment event tree sequences, or Level 2 PSA model, the release frequency is approximately 1.2 x 10-5 per year. The latter value was used as the baseline CDF in the SAMA evaluations. The CDF value is based on the risk assessment for internally-initiated events. NPPD did not include the contributions from external events within the CNS-1 risk estimates; however, it did account for the potential risk reduction benefits associated with external events by increasing the estimated benefits for internal events by a factor of 3. The breakdown of CDF by initiating event is provided in Table 5-3.

Page 5: [1] Deleted Author The CNS-1 core damage frequency (CDF) is approximately 9.3 x 10-6 per year for internal events as determined from the quantification of the Level 1 PSA model. When determined from the sum of the containment event tree sequences, or Level 2 PSA model, the release frequency is approximately 1.2 x 10.5 per year. The latter value was used as the baseline CDF in the SAMA evaluations. The CDF value is based on the risk assessment for internally-initiated events. NPPD did not include the contributions from external events within the CNS-1 risk estimates; however, it did account for the potential risk reduction benefits associated with external events by increasing the estimated benefits for internal events by a factor of 3. The breakdown of CDF by initiating event is provided in Table 5-3.

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Revise procedures to allow the ability to cross-connect the circulating water pumps and the service water going to the turbine equipment cooling (TEC) heat exchangers, which allow continued use of the power conversion system after service water is lost.

Page 10: [33] Deleted Author Revise procedures to allow use of a fire pumper truck to pressurize the fire water system, increasing availability of the fire water system.

Page 10: [34] Deleted Author Implement Generation Risk Assessment (trip and shutdown risk modeling) into plant activities, decreasing the probability of trips/shutdown.

Page 10: [35] Deleted Author NPPD Page 10: [36] Deleted Author detailed engineering project cost-benefit analyses have been initiated for the 1176 potentially cost-beneficial SAMAs (NPPD PSEG 2009a2008, 2009 ).