ML11263A031
| 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:
pp. 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. 5-10, 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 This chapter describes the environmental impacts from postulated accidents that might 3 occur during the period of extended operation, The term "accident" as defined in the I -" Deleted:.
4 Generic Environmental Impact Statement (GELS) (NRC, 1996) refers to any unintentional 5 event outside the normal plant operational envelope that results in a release or the potential 6 for release of radioactive materials into the environment_ Two classes of postulated J .. Deleted:.
7 accidents are evaluated under the National Environmental Policy Act (NEPA) in the license 8 renewal review: design-basis accidents (DBAs) and severe accidents, In the GElIS, thle_._ J.1
. Deleted: I 9 NRC staff categorized accidents as "design basis" when the plant was designed specifically 10 to accommodate these accidents or as "severe" for those accidents involving multiple 11 failures of equipment or function whose likelihood is generally lower than design-basis -- Deleted:.
12 accidents but where consequences may be higher" (NRC, 1996), These issues are_
13 evaluated in Chapter 5 of GElS, "Environmental Impacts of Postulated Accidents."
14 Table 5-1. Issues Related to Postulated Accidents. Two issues relatedto postulated 15 accidents are evaluated under the National EnvironmentalPolicy Act (NEPA) in the license 16 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 17 18 S-[ Deletedl. 'I 19 5.1 pDESIGN-BASIS ACCIDENTS 20 As described in 10 CFR 50.34(b), in order to receive NRC approval for an operating license, 21 an applicant for an initial operating license must submit a final safety analysis report (FSAR) 22 as part of its applicatiorn_ The FSAR presents the design criteria and design information for _ ..... Deleted" 23 the proposed reactor and comprehensive data on the proposed site The FSAR also 24 discusses various hypothetical accident situations and the safety features that are provided 25 to prevent and mitigate accidents, The NRC staff reviews the application to determine j -- Deleted:
26 whether or not the plant design meets the NRC's regulations and requirements and 27 includes, in part, the nuclear plant design and its anticipated response to an accident.
28 The environmental impacts of postulated accidents were evaluated for the license renewal 29 period in Chapter 5 of the GELS.,...cti0 . 5.1 states: .......... ............ ..-....... .-... Deleted:
30 All plants have had a previous evaluation of the environmental impacts of Deleted:.
31 design-basis accident- In addition, the-licensee will be required to maintain______ Deleted:.
32 acceptable design and performance criteria throughout the renewal period ......
fDeleted:.
33 Therefore, the calculated releases from design-basis accidents would not be 34 expected to change, Since the consequences of these events are evaluated Deleted: October 2010 .5-1 Draft NUREG-
-" ,' 1437, Supplement 45
-'I
1 for the hypothetical maximally exposed individual at the time of licensing, 2 changes in the plant environment will not affect these evaluations, .- --'tDeleted:.
3 Therefore, the staff concludes that the environmental impacts of design-basis Deleted:.
4 accidents are of small significance for all plants, Because-the environmental ..... J 5 impacts of design basis accidents are of small significance and because 6 additional measures to reduce such impacts would be costly, the staff 7 concludes that no mitigation measures beyond those implemented during the 8 current term license would be warranted. T'his is a-Category 1 issue.......... Deleted:.
9 No new and significant information related to DBAs was identified during the review of 10 PSEG's environmental report (ER), site audit, scoping process, or evaluation of other 11 available informatiorý_ Therefore, there are no impacts related to these issues beyond those . Deleted:.
12 discussed in the GELS.
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 Deleted:.
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 -" -I Deleted: Cooper Hope Creek Generationing Station 23 Generating Station, Units 1 and 2 (SGS) and Hope Creek Generating Station (HCGS)-sites- L(HCGS) andJ 24 in the GElS (NRC, 1996) The GEISJ however, did evaluate existing impact assessments--- Formatted: Not Highlight 25 performed by the NRC staff and by the industry at 44 nuclear plants in the United States and . Deleted: Nuclear Station, Unit 1 (CNS-1) site 26 segregated all sites into six general categories and then estimated that the risk Deleted:.
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 _ - Deleted:
-. { Deleted:.
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.
36 In the GELS, the NRC staff concludes that the risk from sabotage and beyond design-basis 37 earthquakes at existing nuclear power plants is designated as SMALL, and additionally, that 38 the risks from other external events are adequately addressed by a generic consideration of 39 internally initiated severe accidents (NRC, 1996).
1 Based on information in the GElS, the staff found that:
Deleted:.
Deleted:.
2 The generic analysis... applies to all plants and that the probability-weighted Deleted:.
3 consequences of atmospheric releases, fallout onto open bodies of water, I, 4 releases to ground water, and societal and economic impacts of severe Deleted:.
I, 5 accidents are of small significance for all plants,_ However not all plants _, Deleted:.
6 have performed a site-specific analysis of measures that could mitigate / Ii Deleted: HGCS 7
8 severe accidentk _Consequently_, severe accidents are a Category 2 issue -
for plants that have not performed a site-specific consideration of severe I, 'I
'I Deleted: ope Creek Generating Station Deleted: Salem Generating StationGS 9 accident mitigation and submitted that analysis for Commission review --- 'I I,
Deleted:, Units land 2.
I I'
Deleted:.
10 The staff identified no new and significant information related to postulated accidents during Ill 11 the review of PSEG's environmental report, the site audit, the scoping process, or evaluation Deleted: for CNS-1 conducted by NPPD Energy Company, LLC (NPPD) and the staffs I
12 of other available informatior. Therefore, there are no impacts related to postulated...... review of that evaluation.
13 accidents beyond those discussed in the GEISV In accordance with 10 CFR Deleted:.
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. 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 I
sensitivity analysis of the population dose risk 16 5.3 SEVERE ACCIDENT MITIGATION ALTERNATIVES I I I and offsite economic cost risk using corrected meteorological data. and found that the population dose and offsite economic cost 17 Regulations under Section 51.53(c)(3)(ii)(L) of 10 CFR requires an ER to contain an analysis values for each of the release categories would 18 of alternatives to mitigate severe accidents if the staff has not previously evaluated SAMAs be slightly less than reported in the ER, and that the conclusions of the SAMA remain valid 19 for the applicant's plant in an environmental impact statement (EIS), related supplement, or 'if (NPPD, 2009b).
20 in an environmental assessment. The Commission's reconsideration of the issue of severe, I Deleted: is 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 PDeleted: x 23 (EISs) and supplements to EISs, as well as a previous court decision that required a review II Deleted: is 24 of sever mitigation alternatives at the operating license stage. 'It Deleted: NPPD's Deleted: ER and the supplement submitted in December 2009 25 5.3.1 Introduction Deleted: CNS-1HGCS
, 4 Deleted: SGS 26 This section presents a summary of the SAMA evaluation for SGS and HCGS conducted by Deleted: as 27 PSEG and the NRC staffs reviews of those evaluations. The NRC staff performed its Deleted:
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 Deleted: NPPD 30 available in full inpSEG'ERs.. , .
M'Deleted:
Deleted: NPPD 31 The SAMA evaluations forGS and j-CGS weraeconducted with a four-step-app[oach In _ 1,/ Deleted:.
32 the first step, ,SEG quantified the level of risk associated with potential reactor accidents / ', Deleted:.
33 using the plant specific probabilistic risk assessment (PRA) and other risk models, Deleted: NPPD
,* Deleted: 33 34 In the second step, PSEG examined-the major risk contributors and identified possible ways fDeleted: 9 35 (SAMAs) of reducing that rislk Common ways of reducing risk are changes to com!ponents,,
CNS-l 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- fDeleted, 38 because they are not applicable to-GS_or HCGS due to design-differences, or have --- SDeleted: NPPD
- Deleted: CNS-l
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 SDeleted CNS-1 3 eliminated based on this screening, jeaving 25 ar1_ for further evaluatiorgfor SGS and Deleted:.
4 HCGS, respectively. ' (Deleted: 24
,\ \ Formatted: Not Highlight 5 In the third step, PSEG estimated the benefits and the costs associated with each of the \\\'* Deleted:
- No ,
6 SAMAsk Estimates were made of how much each SAMA could reduce risk Those M j Formatted: Not Highlight 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 Deleted: m 9 SAMAs was also estimated. Deleted: leaving all 33 Formatted: Not Highlight 10 Finally, in the fourth step, the costs and benefits of each of the remaining SAMAs were 'j Formatted: Not Highlight 11 compared to determine whether the SAMA was cost beneficial, meaning the benefits of the Deleted: NPPD 12 SAMA were greater than the cost (a positive cost benefit_ _PSEG, concluded in its ERs that Deleted:.
13 several of the SAMAs evaluated are potentially cost-beneficial (PSEG 2009a, PSEG 2009b).,
\, '
- Deleted:.
"\
Deleted:.
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 imp!emented as " Deleted: N 16 part of license renewal pursuant to 10 CFR Part 54, PSEG'sSAMA analysis and the NRC S Deleted: NPPD Deleted: (NPPD, 2008).
17 staffs review are discussed in more detail below.
'> Deleted: (e.g., none of the potentially cost-0 I beneficial SAMAs would reduce the frequency or risk 18 5.3.2 Estimate of Risk I associated with aging-related failures)
Dlted:. I 19 20
,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 Deleted: NPPDs 4Deleted: NPPD 1
21 the risk estimates used in the SAMA analysis: (1) the plant-specific Level-1 and Level-2 PSA Deleted: CNS-1 as part 22 models, which are updated versions of the IPEs (PSEG 1993, PSEG 1994, PSEG 1995); (2) 23 a supplemental analysis of offsite consequences and economic impacts (essentially a Level- Deleted:.
24 3 PSA model) developed specifically for the SAMA analysis. The most recent plant-specific 25 Level-1 and Level 2 PSA models consisted of the following Internal Events PSAs: (1) for 26 SGS, Salem PRA, Revision 4.1, September 2008, model of record (MOR); (2) for HCGS, 27 the HC108B update. Neither of these includes external events,-------------------. 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).¶
Formatted: Font: Not Bold 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 . ..
DeleteO: a 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 Formatted: Font: Not Bold 5 containment, late release, and early release) is approximately 5.0 x 10-5 per year, also at a ueetea:
l(
6 truncation of 1 x 1011 peryear. The latter value was used as the baseline CDF in the SAMA Formatted: Font: Not Bold 7 evaluations (PSEG 2009a).- The CDF is base-d-roi-the Fik-s-s&ssrfie-nt for interhrllinitiatd -
Deleted: The CNS-1 core damage frequency 8 events, which includes internal flooding. PSEG did not explicitly include the contribution (CDF) is approximately 9.3 x 10n per year for 9 from external events within the SGS risk estimates; however, it did account for the potential internal events as determined from the 10 risk reduction benefits associated with external events by multiplying the estimated benefits quantification of the Level 1 PSA model. When determined from the sum of t 11 for internal events by a factor of 2i 4 rhe breakdown of CDF by initiating event provided6in.
Deleted: Transients 12 Table 5-1 Deleted: 3.0 x 10-6 Deleted: 2 13 Table 5-1., Salem Nuclear Station Core Damage Frequency for Internal Events I *jFormatted: Superscript Event (per CDF year) % Contribution CDF to / Deleted Loss of DC Power Deleted: 2.1
,4oss 9f Control AreaVentilatioln oqss ofOffsite Power (LOOP) _ t.1 x 10-6. . . . Deleted: Loss of Coolant Accidents (LOCA)
[Deleted: 1.4
!-qss of Service water - ~.6§2 1Q§. -
.... 1 .... .. F-Deleted: 15
.-1* Deleted: Loss of Feedwater irnternal Floods .,.§ 5 Q-§ 9 -( Deleted: 1.0
,Transients - ,4.9 2ý1(-, _ Deleted: 11 Deleted: Loss of Offsite Power
,Steam Generator Tupe Rupture (SGTR) ..... _2-.Y_ 3ýj 0--- , (Deleted: 6.5....0.10-7 .fI
,Loss of Component Cooling Water (CCW) . . ý Ox9 q7-,
'{Deleted: ...7 Deleted: Loss of Service Water
,.nticipated Transient Without Scram (A_'SI .7._4x 10-7 ..
eleted: 6.0....7. 10-7 rTi]
,oss of 125 V DC Bus A 9x 1
.. .I.0-7
............ t Deleted: 7 3 Deleted: Loss of AC Buses Others (less than 1 percent) ..............
1.8x10-6,------------------------------
Deleted: 2.6.....0x10-7 Total CDF (Internal Events) 4.8,x _0-*5 100 Deleted: 3 Deleted: Internal Flood 14 As shown in this table, events initiated byjosses of control area ventilation, offsite power, or *,{ Deleted: 2.6 15 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_)._ eleted: 3 16 it Deleted: Interfacing System LOCA Deleted: 5.
17 ,PSEG estimated the dose to the population within 50 miles (80 kin) of theSGS site to be 18 approximately 0,78 person-sievert (p_erson-Sv__,7 person-rem)_pe_r year The breakdown of "Formattedteri 19 the total population dose by containment release mode is summarized in Table 5-20 Containmentb~ypass events (.sUh_as.SGTR-initiated large early release freque ncy (LERF). Deleted: 8 21 accidents) and late containment failures without feedwater dominate the population dose Deleted: 9.3...x 10-6 22 risk at SGS, Deleted: transients...osses of control ar 8
' Formatted: Superscript I Deleted: NPPD ... SEG estimated the d
- A" Formatted Fr -in-11
__FomatDeletd:
4.
Table 5-2,Breakdown of Population Dose by Containment Release Mode For SGS Containment Release Mode(ProRe ~(Person-RemI Population 1 Dose PrYa)
Per Year) % Contribution=
%Ctrbtn-- / Deleted: Early Containment Failure
-- /'*(Deleted: 78
,ontainment over-pressure (late) 4 2.9 ... 5-..
. .... Formatted Table (Deleted: 1.67 Steam Generator Rupturs 31.9 41 (Formatted .131 Containment Isolation Failure .2.3 .3_ (Deleted: Intermediate jnact Containm ent,.. . . .. -, .................
. Deleted: 0.47
-- ...-.-.-.. - <1 .. .
Formatted Snterface systemLOCA ..... 6 _ <
-.- -1
- Deleted: 22 Catastrophic Islaotion Failue 0.4 <1 Deleted: Late... nact Containment Failure Deleted: <0.1 Basemat melt-through (late) negligible negligilbe. ", Formatted 4' '\ \Deleted: Intact Containment Total .............-- - - - - -. 78.2-- .-.-.-.-.. -- 1_00
.*. Deleted: Negligible 1
One person-rem = 0.01 person-Sv - -................................. . .................... Formatted ... 17 0' Deleted: Negligible¶ 2 ,The HCGS CDF is apjroximately5. 1x1_0_ per year as determined from quantification of the \' Formatted 3 Level 1 PRA at a truncation of I x 1012 per year. When determing form the sum of the Deleted: 2.14 4 containment event free (CET) sequences, or Level 2 PRA model, higher truncation of 5 x 5 10 11per a year used and the resulting release frequency (from all release categories, which \\ Formatted Formatted
.. f[91
[01 6 consist of inact containment, late release, and early release) is approximately 4.4 x 10.6 per 7 year. The Latter value was used as the baseline CDF in the SAMA Evalutions (PSEG \j Deleted: ¶ 8 2009b). Although this is about 16% less that the internal events CDF of 5.1 x 10°' per year Formatted :.21 9 obtained from the Level-1 model, the NRC staff considers that its use will have a negligible Deet: d....22 10 impact on the results of the SAMA evaluation because the external event multiplier and Deleted: Loss of Control Area Ventilatioln 11 uncertainty multiplier used in the SAMA analysis have a much greater impact on the SAMA Deleted: 1.8-....3-x-10-6 12 evaluation results than the small difference arising from the model quantification approach. (Deleted: 37 13 The breakdown of CDF by initiating event is provided in Table 5-3.
Deleted: Loss of Offsite Power (LOOP)-
14 Table 5-,3, Hope Creek Nuclear Station Core Damage Freauencv for Internal Events Deleted:,
[Deleted: 17 CDF % Contribution to Initiating Eventyear) CDF , Deleted: Loss of Service water Deleted: 6.6....7 x 10-6 ... [241
,Lossof Offsite Power 9._3x1 W .18 J, /" Deleted: 14 Deleted: Internat Floods
,Loss of Service Water_(S W) 8.1'x10-,-.-.-15
-,,fDeleted: 4.5 ...2.~10-6 TT5
,.Vlanual Shutdown . .7.7. 0-v _ J5 Deleted: 9 juqrine Trip with Bypass __ 2.x10-- .. Deleted: Transients Deleted: 4.o....8 x10-6 F. 261
,Small Loss of Coolant Accident (LOCA)-Water_(Below . . *.8x 1 - Deleted: S G Top of Active Fuel) / Deleted: Steam Generator Tupe Rupture(STR Small LOCA-Steam (Above Top of Active Fuel32 .. 2.3 o, -4 Deleted: 2.7....3 x 10-6 -- ff[271--- ýfl
-- Deleted: 6 I
Deleted: Loss of Component Cooling Water toss of Condenser Vacuum 1_,ýj 7 ..... .. ... (CC\AO Deleted: 1.0.0
.. 10-6.1.. [281 tire Protection System Rupture Outside Control Room "1"Deleted: 2
,solation LOCA in Emergency Core Cooling System 1.1x10-7 2 Deleted: Anticipated Transient Without Scram (ECCS) Discharge Paths ~'1(ATWS)9 1.1 x 10-7 2 D leted: 7.4
,Main Steam Isolation Valve (MMSIV)_Closure 9.7 x 1 -8 2 ,1Deleted: 2 Internal Flood Outside Lower Relay Room SDeleted: Lossof 125 V DC Bus A 8.8 x 10-8 2 3 Loss of Feedwater >1 Deleted: Others (less than 1 percent) ]
7.9 x 10-8 2 Formatted: Tab stops: 3.04", Right Loss of Safety Auxilaries Cooling System 7.6 x 10-8 1 - .1 Deleted: Loss of Safety
,Reactor Auxiliaries Cooling Systemn (RACS) Common ...
Header Unisolable Rupture 1 Deleted: 11l 5.7 x 10-8 / 4 Unisolable SWA Pipe Rupture in RACS Room 1 'Delleted: 6.9 10-7¶
'//1.8.10-6 5.7 x 10-8
' . Formatted: Left Unisolable SW B Pipe Rupture in RACS Room 4.1x10-6 Others (less than 1% each) Deleted: 4.8....1 x 10-58. 29
/Deleted: es. ..of control area ventilatioq-r--
Total CDF (Internal Events) 5.t1 x 10q- 6 100 / 7 Deleted: SGS.. .CGS site to be approx' Deleted: Containment over-pressure (late) 1 As shown in this table, events initiated by los4pf offsite power, 3loss of service water and Formatted Table 2 other transients (manual shutdown and turbine trip with bypass) are the dominant Deleted: 42.9 3 contributors to the CDF. Aticipated transient without s*ram (ATWS) sequence.s a c.unt for
' Formatted: Centered 4 33 percent of the CDF, station blackout accounts for 12 percent of the CDF (PSEG 2010b).
/ Deleted: 55 Deleted: Steam Generator Rupturs 5 PSEG estimated the dose to the population within 50 miles (80 km) of theHCGS site to be Deleted: 31.9 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 Deleted: 41 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 Deleted: Containment Isolation Failure 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) / Deleted: 2.3 10 dominate hhe population dose risk atHCGS.
Formatted: Centered Deleted: 3 11 Table 5-4 Breakdown of Population Dose by Containment Release Mode For HCGS Formatted: Centered eDeleted: 0.2 Deleed <1]
Containment Release Mode Population 1 Dose % Contribution (Person-Rem Per Year) 4J Deleted: Interface system LOCAJI
,Early Releases (.<4h.rs) . v11.9 Catastrophic Islaotion Failue¶
....52. _
Basemat melt-through (late) jnterrnediate Releasesj4 to< 24hrs- -
_ _ __,43 . . . Deleted: 0.6¶S Jate ReleasesA*_24hbr!s)_ <0.1 ' {Formatted: Centered Inact Containment <0. 1. negligible, SDeleted: 41
",'), negligible Deleted: <1ij negligilbe
Total p22.9 100 Deleted: 78.2 - -I Deleted:78.2 jJ 1
Formatted: centered One person-rem = 0.01 person-Sv 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 intheir 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
- Deleted: NPPD 17 The staff concludes thatPSEG used a systematic and comprehensive process for -C et: CNS-1 18 identifying potential plant improvements forSGS and HCGS, and that the set of potential 19 plant improvements identified byPSEG is reasonably cormprehensive and. therefore... ..... Deleted: NPPD 20 acceptable.
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 improvements and concludes that the rationale and assumptions for estimating risk 33 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:. -J
/ Deleted:. -J 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 4
Deleted: both
- IN Deleted-: both 10 5.3.5 Cost-Benefit Comparison Deleted: NPPD 11 The cost-benefit analysis performed bypýSEG was based primarily on NUREG/BR-0184 ,, ,,, Deleted: 8 12 (NRC, 1997) and was executed consistent with this guidance, NUREG/BR-0058 has Deleted: NPPD 13 recently been revised to reflect the agency's revised policy on discount rate-s, Revision 4 of Deleted: eight seveteen 14 NUREG/BR-0058 states that two sets of estimates should be developed - one at 3 percentand , Dle 15 and the other at 7 percent (NRC, 2004.. PSE_ providedboth~setsofestimates forSGS De'l 16 HCGS (PSEG 2009a,2009b ....................................... ., .. Deleted: for SGS.
S Deleted: 25 lures to allow bypass 17 For SGS, PSEG identifiedpleven potentially cost-beneficial SAMAs .in..the ....baseline
. ... . . analysis
. .. . . j f' fDtDevelop proct on cooline (ROClC fthe renactor'core isolatic 18 contained in the EP, Thepotentially cost-beneficial SAMAs are: i/ I turbine exhaust pressure trip, extending the time available for RCIC operation Deleted: 30 Deleted: Revise 19 0 SAMA 1,-.Enhance proqed.u res -andlprovide additional equipment to J 20 respond to loss of control area ventilation. 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 21 0 SAMA2 -_,l-p_-cornfigure Salem 3_to provide a moe expedient backupp to, exchangers in a loss of service water (SW) 22 AC power source for Salem 1 and 2..
Deleted: 33
-4Deleted: Provide for the ability to establish an 23 0 SAMA,4_-_,lnstall fuel oit transfer pump on "C" emergency diesel :* emergency connection 24 generator (EDG) and provide procedural guidance for using "C" EDG to - Deleted: of existing or new water sources to 25 power selected "A" and "B" loads. systems, increasing feedwater and condensate availability of feedwater
- D-eleted: 40 26 0 SAMA -,6_-ýEnnce flood_detection for 84' auxiliary building and _.
27 enchance procedural guidance for responding to service water flooding 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 28 0 SAMA,9_-_,ConnectHope Creek cooling tower basin to Salem service EDG.
29 water system as alternate service water supply,---------------- Deleted: 453 Deleted: Provide an alternate means of 30 0 SAMA 10 -Provide procerdural guidance for faster cooldown on Loss supplying the instrument air header, increasing 31 of reactor coolant pump (RCP) Seal availability of Deleted: Instrument air 68Deleted:
Deleted: Revise procedures to allow thf 12
.-t Deleted: 78 1
- SAMA 11 ,Modify p!ant procedures to make use of other Unit's_ P!DP Deleted: Improve training on alternate injection via 2 for RCP seal. the fire water system, increasing the availability of alternate injection Deleted: 79 3 0 SAMA,2 -,Im.poye flood barriers outside 220/440VAC sWitchgear ............... -
4 rooms. 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.
5 0 SAMA 14 -. Expand ATNýT nitigation system actuation circuitry ------------
(AMSAC) function to include backup breaker trip on RPS failure. Deleted: Revise procedures to allow the ability to 6 cross-connect the circulating water pumps and the service water going to the turbine equipment cooling (TEC) heat exchangers, which allow continued use of 7
- SAMA 17 -,Enhance pro9cedures and provide additional quipment to the power conversion system after service water is 8 respond to loss of EDG control room ventilation.- lost Deleted: Revise procedures to allow the ability to cross-connect the circulating water pumps and the 9
- SAMA 24 -,Provide proceduralguidance to cross-tie Salem 1 and 2 service water going to the turbine equipment cooling 10 service water systems. (TEC) heat exchangers, which allow continued use of the power conversion system after service water is lost 11 PSEG performed additional analyses to evaluate the impact of parameter choices and Deleted: Revise procedures Ito allow the ability to 12 uncertainties on the results of the SAMA assessment (PSEG, 200,9_4)_ If the benefits are - cross-connect the circulating water pumps and the 13 increased by an additional factor of.1.64 to account for uncertainties five additional SAMA service water going to the turbine equipment cooling 14 candidates were determined to be potentially cost-beneficial. The ER also showed that the ,,, (TEC) heat exchangers, which allow continued use of the power conversion system aifter service water is 15 sentivity case SAMA (SAMA 5A) was potentially cost-benificial: lost Deleted: VI uW" NPPD 16 0 SAMA 3 - Install limited emergency diesel generator (EDG) cross-tie 'I',', ,\ Deleted: (NPPD 17 capability between Salem 1 and 2.,, . .. .. .. .. .. .. .. .. . . . . . . . ... Formatted: Not Highlight 10Deleted:86 SAMA5--Jnstall portable diesel genetrators to charge station battery Formatted: Highlight 18 0 19 and circulating water batteries and replace PDP with air-cooled pump. Deleted Deleted: 3 20 0 SAMASA-_jnstal! portable diesel generators to charge station battery . Deleed: thre 21 and circulating water batteries. Fomttd Highlight
\*ormatd: Unnumbered List SER 1st LINE 22 0 SAMA 7,-.,nstall "B" Train auxiliary feedwater storage tank (AFWST) Formatted: Font color: Auto 23 makeup including alternative water source. Deleted: 14 iiDeleted: Provide a portable g enerator to supply DC power to individual panels duri ng a station blackout
'I* (SBO), Increasing the time ava lable for AC power 24 SAMA 8 - Install high pressure pump powered with portable diesel Lo recovery 25 generator and long-term suction source to supply the AFW Header.
,0 Deleted: 64
[Deleted: Revise procedures t o allow use of 33 26 SAMA 27 - In addition to the equipment installed for SAMA 5, install "I Formatted: Indent: Left: 0.56", Hanging: 0.82" 27 permanently piped seismically qualified connections to alternative AFW 28 water sources. J Deleted: 75 I I Deleted: Implement Generation Risk Asses(f341 29 ,PSEG indicated that all 17_potential!y cost-beneficial SAMAs will be considered for - - -I Deleted: T..51 30 implementation through the established Salem Plant Health Committee process5 - - - - - - - - - - - " Deleted: detailed engineering project cost-b[
'" Formatted: Not Highlight
1 ForHCGS, PSEG identifiednine potentially cost-beneficial SAMAs in the baseline analysis Deleted: SGS 2 contained in the ER. The potentially cost-beneficial SAMAs are: - Detd: eleven 3
4 SAMA 1 -,Remove automatic depressurizationsyst~er from non-AtWS emergency operating procedures.
(ADS)_inhibit .
Deleted: Enhance procedures and provide additional equipment to respond to loss of control area ventilation I 5 SAMA3_--Jnstall backup air compressor to supply air-operated valves..
Re-configure
! DeltedDeleted:
Deleted: Aovs 6 SAMA 4 -,Provide procedural guidance to cross-tie residual heat l Deleted: Salem 3 to provide a moe expedient bbacukup to AC power source for Salem 1 and 2.¶ 7 removal (RHR) trains.
- Deleted: Install fuel oit transfer pump on "C" emergency diesel generator (EDG) and provide 8 SAMAII _0-,Pjrgvide proceduralrguidance to use B.5.b low pressure procedural guidance for using "C" EDG to 9 pump for non-security events______ power selected "A" and "B" loads.
Dele, Enhance 10 0 SAMA,17 -,Replace a supply fan witha different design in service water 11 pum p room ,_ .----------------...............
. ..................... ....... ......................... Deleted: flood detection for 84' auxiliary
'\ building and enchance procedural guidance for I responding to service water flooding 12 SAMA,18 -- Replacea return fan with a different design In service water F\rDeleted: 9 13 pump room.
0 Deleted: Connect Hope Creek cooling tower basin to Salem service water system as alternate service 14 SAMA,30-,Provide proceduralrguidance for partial transfer function of Deleted: water supply.
15 control functions from the control room to the remote shutdown panel. -
1 V Deleted: 10
'Deleted: Provide procerdural guidance for 16 SAMA 35,-,Relocate, minimize, and/or eliminate electrica l _heaters__in _--
faster cooldown on Loss of RCP Seal 17 electrical access room. ]1Deleted: 11 3 Deleted: Modify plant procedures to make use 18 SAMA 39,Provide procedural gdane to _bypass reactor core of other Unit's PDP for RCP seal.
19 isolation cooling turbine exhaust pressure trip. 'j Deleted: 12 Deleted: Improve flood barriers outside 20 ,PSEG performed additional analyses to evaluate the irmpa.ct of parameter choices and S220/44VAC switchgear rooms.
21 uncertainties on the results of the SAMA assessment CPSEG, 2009b,._ Iffthe benefits are J Deleted: 14 22 increased by an additional factor of 2.84,to account for uncertainties,,our additional SAMA *\1 23 candidates were determined to be potentially cost-beneficial: Deleted: Expand AMSAC function to include backup breaker trip on RPS failure.
l ' Deleted: RCIC 24 SAMA 8 - Convert selected fire protection piping from wet to dry pipe "IW,Deleted:..
and SAMA 17 - Enhance procedures provide additional quipment to respond to 25 system. I loss of EDG control room ventilation.¶
,,w * . SAMA 24- Provide procedural guidance to 11Wcross-tie Salem 1 and 2 service water systems.¶ 26 SAMA 32 - Install additional physical barriers to limit dispersion of fuel %11' Formatted: Not Highlight 27 oil from DG rooms. ( Deleted: a 28 29 0 SAMA 7 - Provide procedural guidance for loss of all 1E 120V AC poweý.------------------------------------------
(Deleted: 1.64 (Deleted: three
( Deleted: a
1 0 SAMA 37 - Reinforce 1 E 120V AC distribution panels.
SAMAr generators toto.
PSEG jindicated thatalL 13potentially cost-beneificial SAMAs will be considered for. charge - Install portable diesel 2 ..... ..... . . Deleted:
station battery and circulating 3 implementation through the established-HCGS Plant Health" Committ-ee-process,.. water batteries and reeplace PDP with air-cooled
,ased on the staffs review. the NRC staff concludes that, with the exception of the
-~pump.¶ 4 chSAMAsAt- Instattportable diesel generators to 4 charge station batter~ and circulating water batteries.¶ 5 poten ially cost-b-en eficial--SAMAs-di
--- scussed- above, t he costs of the SA MAs evalu ated- -'_
- SAMA 7 - install B" Train AFWST makeup 6 would be higher than the associated benefits. , including alternative water source.¶ S *. SAMA 8 - Instaltl igh pressure pump powered with portable diesel generator and long-term suction 7 5.3.6 Conclusions , source to supply the, AFW Header.$
-. SAMA 27-tn add ition to the equipment installed for SAMA 5, install pe rmanently piped seismically 8 The staff reviewed PSEG's analysis and concluded that the methods used and the ,,,IM qualified connections to alternative AFW water 9 implementation of those methods were sound. The treatment of SAMA benefits and costs sources 10 support the general conclusion that the SAMA evaluations performed by1PSEG are I, Formatted: Not Higihlight 11 reasonable and sufficient for the license renewal submittal. II'I. Deleted: detailed II Formatted: Not HigiWlight 12 Based on its review of the SAMA analysis, the staff concurs with,PSEG_s identification of II Deleted: engineerin g project cost-benefit analyses 13 areas inwhich risk can be further reduced at both SGS and HCGS in a cost-beneficial Iareas t i have been initiated for the 1316 potentially cost-1 beneficial SAMAs (Pt SEG 2009ba) 14 manner through the implementation of all identifiedpotentially cost-beneficial SAMA~-s-- ,--
15 Given the potential for cost-beneficial risk reduction, the staff considers that further III Formatted: Not Higlhlight 16 consideration of these SAMAs byPSEG is warrantedL However, none of the potentially___ ... Formatted: Not Higihlight 17 cost-beneficial SAMAs relate to adequately managing the effects of aging during the period II Formatted: Not Higihlight 18 of extended operation forSGS or HCGS Therefore, they need not be implemented as part ,, Deleted: NRC staff reviewed NPPD's PSEG re-19 of the license renewal pursuant to 10 CFR Part 54. analysis as submitted1by NPPD and.agrees that the III error was conservativ 'a relative to the average population dose and offsite economic cost and that i no SAMAs were inap propriately excluded from I consideration in the LRA as a result of the error.¶(
ti 1 B3 III Formatted: Normal 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 1
5.4 REFERENCES
__ Lcaps I Deleted: <#>es%
2 ,Public Service Electric and Gas Company
...... (PSEG)_.SProbabilistic 1993. Letter from Stanley LaBruna, - Electric Power Research Institute (EPRI). 1989.
Seismic Hazard Evaluations at 3 PSEG, to NRC Document Control Desk.
Subject:
"Generic Letter 88-20; Individual Plant Nuclear Plant Sites inthe 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 - 10CFR50.54 (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 U.S. Nuclear Regulatory Commission (NRC). 1998. Code Manual for MACCS2.
32 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 FederalRegulations, Title 10, Energy, Part 51," Environmental Protection Regulations for Domestic Licensing and Related S,---Regulatory Functions."¶ 10 CFR Part 54. Code of FederalRegulations, Title 10, Energy, Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." I 10 CFR Part 100. Code of FederalRegulations, 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.
CooperNuclear 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 NuclearPlants. NUREG-1 437. Vols. 1 and 2, Washington, D.C. ADAMS Accession No. ML0617706057 U.S. Nuclear Regulatory Commission (NRC). 1997.
RegulatoryAnalysis Technical Evaluation Handbook.
NUREG/BR-0184, Washington, D.C.7
.U.S. Nuclear Regulatory Commission (NRC). 2004.
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-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 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 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 ).