ML052030315

Supplement to Request for License Amendment Errata and Addenda for NEDC-33066P, Rev 2, Arts/Mella Implementation

ML052030315
Person / Time
Site:	Hope Creek
Issue date:	07/08/2005
From:	Barnes G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H04-01, LR-NO5-0234, NEDC-33066O, Rev 2
Download: ML052030315 (8)

Text

I PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 JUL 0 8 2005 PSEG J NTuclear LLC LR-N05-0234 LCR H04-01 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT ERRATA AND ADDENDA FOR NEDC-33066P REVISION 2 ARTS/MELLLA IMPLEMENTATION HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354

Reference:

LR-N05-0032, "Supplement to Request for License Amendment; ARTS/MELLLA Implementation," dated February 18, 2005.

By the referenced letter, PSEG Nuclear LLC (PSEG) submitted NEDC-33066P, Revision 2, "Hope Creek Generating Station, APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," February 2005. Attachment 1 to this letter contains Errata and Addenda No. 1 to NEDC-33066P, Revision 2. Attachment 2 contains Errata and Addenda No. 2. The Errata and Addenda provide replacement pages containing the results of a more limiting evaluation case for the Anticipated Transient Without Scram (ATWS) event for the GE14 equilibrium core. Replacement pages for NEDO-33066, Revision 2 are also provided.

While NEDC-33066P, Rev. 2 contains proprietary information, there is no proprietary content on the replacement pages.

PSEG has determined that the attached information does not alter the conclusions reached in the 10CFR50.92 no significant hazards analysis previously submitted.

Ifyou have any questions or require additional information, please contact Mr. Paul Duke at (856) 339-1466.

PFI 95-2168 REV. 7199

~1

• Document Control Desk LR-N05-0234 JUL O8 2005 I declare under penalty of perjury that the foregoing is true and correct.

Executed on 7871o*

(date) George P. Barnes Site Vice President Hope Creek Generating Station Attachments (2)

C Mr. S. Collins, Administrator- Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. S. Bailey, Project Manager - Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625

1

• Attachment 1 LR-N05-0234 LCR H04-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT ERRATA AND ADDENDA FOR NEDC-33066P REVISION 2 ARTSIMELLLA IMPLEMENTATION Errata and Addenda Number 1 to NEDC-33066P, Revision 2 APRM/RBM/Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)

Errata and Addenda Sheet GE Energy Nuclear 3901 Castle Hayne Rd Wilmington, NC 28401 Applicable to: Hope Creek Generating Station E&A Number I Publication No. NEDC-33066P Revision 2 Date: March 31, 2005

Title:

APRM/RBM/Technical Specifications /Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)

Issue Date: February 2005 Note: Correct all copies of the applicable publication as specified below.

-Itemi -(Section, Page Coections And Additions

.Pragr aph,'Line) I (Supplied s Replacement Pages)

The limiting ATWS analysis (PRFO at EOC) was re-evaluated to account for an Increased boron transport delay time. Therefore, the following two changes were made to MSAR Table 10-2: the peak suppression pool temperature and peak containment pressure values were revised and a note was added explaining the change.

1. Table 10-2 Revised the Peak Suppression Pool Temp (OF) and Peak Containment Pg 10-4 Pressure (psig) for the GE14 PRFO @ EOC to 176 IF and 8.0 psig.

2. Table 10-2 Added note (4) explaining the different boron transport delay times used Pg 10-4 for some of the GE14 cases.

Note: There is no proprietary content on this page.

The Non-Proprietary version is also attached.

r NEDO-33066 Revision 2 With Errata-Addenda No. 1 Table' 10-1 Initial Conditions for ATWS Analyses C,,,,,,,;,, .t, !,- ,,,~

.,,a, Mra eters ' i ' ' ' ' ' I Dome Pressure (psia) 1020 Core Flow (Mlb/hr / % of RCF) 76.6 / 76.6 Core Thermal Power (100% of CLTP, MWt) 3339 Steam/Feed Flow (Mlb/hr / % of rated) 14.38 / 100 Feedwater Temperature ( 0F) 422.6 Sodium Pentaborate Solution Concentration in the 13.6 SLCS Storage Tank (% by weight)

Nominal Boron 10 Enrichment (%) 19.8 SLCS Injection Location Core Spray Nozzles Number of SLCS Pumps Operating 2 SLCS Injection Rate per pump (gpm) 41.2 Initial Suppression Pool Temperature (IF) 95 Initial Suppression Pool Mass (Mlbm) 7.32 Service Water Temperature ( 0F) 95 High Dome Pressure ATWS-RPT Setpoint (psig) 1101 Number of SRVs OOS I Table 10-2 Results for Limiting ATWS Events GE14 Equilibrium / Cycle 13 Mixed Core Event J Exposure ^ ;Peak Vessel Peak Cladding -- Pea Suppression Pea Contairmentl

___',_',_- :w- .i Pressure'(Psig) Ternperature (oJ)(I) Pool'Temp (oF)(3 ) pressure P (Psig)

MSIVC( 4 ) BOC 1333 / 1303 (2) /1147 173 / 169 7.5 / 7.0 MSIVC(4 ) EOC 1334 / 1312 1420/ 1360 174/172 7.8 /7.4 PRFO( 4 ) BOC 1343 / 1312 (2) /1436 173 / 169 7.6/7.0 PRFO EOC 1340/ 1322 1589/ 1588 176/ 172 8.0/ 7.3 (1) The fuel clad oxidation is insignificant and is less than 17%. The values reported for the Cycle 13 Mixed Core are the limiting results of GE14 and SVEA-96+ fuel.

(2) The PCT values for CLTP/BOC conditions are not bounding and not calculated.

(3) The peak suppression pool temperatures were also validated to meet the ATWS acceptance criteria through depressurization of the reactor. The depressurization evaluation was conservatively performed at 3952 MWt.

(4) In the full core GE14 analysis, a shorter boron transportation delay time of 86 seconds was used in the evaluation of these non-limiting cases. All other cases are evaluated based on an updated boron transportation delay time of 104.4 seconds. The non-limiting cases would be similarly impacted by this revised input and remain bounded by the historically limiting PRFO-EOC case.

10-4

r LR-N05-0234 LCR H04-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT ERRATA AND ADDENDA FOR NEDC-33066P REVISION 2 ARTS/MELLLA IMPLEMENTATION Errata and Addenda Number 2 to NEDC-33066P, Revision 2 APRM/RBM/Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)

Errata and Addenda Sheet GE Energy Nuclear 3901 Castle Hayne Rd Wilmington, NC 28401 Applicable to: Hope Creek Generating Station E&A Number 2 Publication No. NEDC-33066P Revision 2 Date: May 5, 2005

Title:

APRM/RBM/Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)

Issue Date: February 2005 Note: Correct all copies of the applicable publication as specified below.

I:

.tm' :-(Secton, Page

!Paragaph LieSupplied

' -i  :.; Corrections AndAdditio:ns-As Replacement Pages)

The Errata and Addenda Number 1 changes to Table 10-2 missed instances of the same numbers in the text in Section 10.3, Pg 10-2.

1. I Section 10.3 Revised the Peak Suppression Pool Temp (IF) and Peak Containment Pg 10-2 l Pressure (psig) for the limiting case to 176 IF and 8.0 psig.

Note: There is no proprietary content on this page.

The Non-Proprietary version is also attached.

NEDO-33066 Revision 2 With Errata-Addenda No. 2 10.2 Input Assumptions The following initial conditions and assumptions were used in'the analysis:

Analytical Assumptions Bases/Justifications The reactor is operating at 3339 MWt (100% Consistency with HCGS current licensing of CLTP). basis.

Initial core flow is 76.6% of RCF Lowest core flow at rated power range to maximize the initial void fraction in the coolant, and thus more severe pressurization transient consequences.

Both Beginning-of-Cycle (BOC) and EOC Consistency with generic ATWS nuclear dynamic parameters were used in the evaluation bases.

calculations.

Sodium Pentaborate Solution Concentration is Minimum solution concentration to meet 13.6% by weight. the ATWS requirements.

One SRVOOS, specified as the valve with the Consistency with the Technical lowest setpoint. Specifications.

SRV setpoints adjusted to be consistent with Consistency with the Technical the 3% setpoint tolerance relaxation. Specifications.

The initial operating conditions are presented in Table 10-1.

10.3 Analyses Results Table 10-2 presents the results for the MSIVC and PRFO events. The limiting ATWS event for HCGS is the PRFO. The peak vessel bottom pressure for this event is 1343 psig at BOC, which is below the ATWS vessel overpressure protection criterion of 1500 psig.

The highest calculated peak suppression pool temperature is 1760 F at EOC, which is below the ATWS limit of 2010 F. The highest calculated peak containment pressure is 8.0 psig at EOC, which is below the ATWS limit of 62 psig. Thus, the containment criteria for ATWS are met.

Coolable core geometry is ensured by meeting the 22000 F PCT and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46. The highest calculated PCT is 15890 F, which is significantly less than the ATWS limit. The fuel cladding oxidation is insignificant and less than the 17% local limit.

The maximum SLCS pump discharge pressure and timing depend primarily on the SRV setpoints. The maximum SLCS pump discharge pressure during the limiting ATWS event is approximately 1258 psig. This value is based on a peak reactor vessel upper plenum pressure of 10-2