View source for NL-05-107, Proposed Change to Indian Point 2 Technical Specifications Regarding LBLOCA Analysis Methodology

{{Adams
| number = ML052770536
| issue date = 09/26/2005
| title = Proposed Change to Indian Point 2 Technical Specifications Regarding LBLOCA Analysis Methodology
| author name = Dacimo F
| author affiliation = Entergy Nuclear Northeast
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000247
| license number = DPR-026
| contact person = 
| case reference number = NL-05-107
| document type = Letter type:NL, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Amendment
| page count = 18
}}

=Text=
{{#Wiki_filter:Entergy Nuclear Northeast Indian Point Energy Center Entergy                                                                        450 Broadway, GSB PO. Box 249 Buchanan, NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration September 26, 2005 Re:    Indian Point Unit 2 Docket 50-247 NL-05-107 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

==SUBJECT:==
Proposed Change to Indian Point 2 Technical Specifications Regarding LBLOCA Analysis Methodology

==REFERENCE:==
: 1. Entergy letter (NL-05-058) to NRC; "Reanalysis of Large Break Loss of Coolant Accident Using ASTRUM", dated April 22, 2005.

==Dear Sir:==

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc, (Entergy) hereby requests an amendment to the Operating License for Indian Point Nuclear Generating Unit 2 (IP2) to adopt the use of ASTRUM for the licensing basis analysis of the Large Break Loss of Coolant Accident (LBLOCA), as stated in Reference 1.
Entergy has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1) using the criteria of 10 CFR 50.92 (c) and Entergy has determined that this proposed change involves no significant hazards considerations, as described in Attachment 1. The proposed changes to the Technical Specifications are shown in Attachment 2.
A copy of this application and the associated attachments are being submitted to the designated New York State official.
Entergy requests approval of the proposed amendment by June 2006, to be implemented within 60 days. There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Kevin Kingsley at 914-734-6695.
I declare under penalty of perjury that the foregoing is true and correct. Executed on            4!/   '
Fred R. Dacimo Site Vice President Indian Point Energy Center A.~

NL-05-107 Docket No. 50-247 Page 2 of 2 Attachments:
: 1. Analysis of Proposed Technical Specification Change
: 2. Proposed Technical Specification Change (markup) cc:    Mr. John P. Boska, Senior Project Manager, NRC NRR Mr. Samuel J. Collins, Regional Administrator, NRC Region 1 NRC Resident Inspector's Office, Indian Point Unit 2 Mr. Peter R. Smith, President, NYSERDA Mr. Paul Eddy, New York State Dept. of Public Service

ATTACHMENT 1 TO NL-05-107 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING USE OF ASTRUM FOR LBLOCA ANALYSIS ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247

NL-05-1 07 Docket No. 50-247 Attachment 1 Page 1 of 11

==1.0     DESCRIPTION==

This letter requests an amendment to Operating License DPR-26, Docket No. 50-247 for Indian Point Nuclear Generating Unit No. 2 (IP2).
The proposed amendment will revise the analysis method used for the Large Break Loss of Coolant Accident (LBLOCA) by incorporating the use of a new approach (ASTRUM) for the treatment of parameter uncertainties. The new approach is described in Westinghouse Topical Report WCAP-1 6009-P-A, approved by NRC in Reference 1.
Changes to the Technical Specifications to reflect the proposed use of ASTRUM in LBLOCA analyses consist of revisions to the list of references provided in Technical Specification Section 5.6.5, Core Operating Limits Report.

==2.0     PROPOSED CHANGE==
S Technical Specification Section 5.6.5 (Core Operating Limits Report); three references in line item b.6 are replaced by a new reference, WCAP-16009-P-A. Refer to Attachment 2 for markup page.

==3.0     BACKGROUND==

The current methodology used for analysis of LBLOCA at IP2 is based on Westinghouse Topical Report WCAP-12945 and the plant-specific application of the methodology to IP2 as approved by NRC in 1997 (Reference 2). Since that time, the analyses have been updated to account for various plant change evaluations and model correction items in accordance with 10 CFR 50.46(a)(3). The most recent analysis update was performed as part of a stretch power uprate project in 2004 (Reference 3).
Entergy committed, in Reference 4, to perform a reanalysis of LBLOCA using the ASTRUM methodology. NRC subsequently approved (Reference 5) Westinghouse Topical Report WCAP-16009-P, which describes the ASTRUM methodology and Entergy performed the LBLOCA reanalysis as committed. Entergy reported the results of that analysis in Reference 6 and stated the intent to submit a license amendment request to formally adopt the ASTRUM methodology.

NL-05-1 07 Docket No. 50-247 Attachment I Page 2 of 11

==4.0     TECHNICAL ANALYSIS==

The original application of the Westinghouse Best Estimate Methodology to Indian Point Unit 2 Nuclear Plant, approved by the NRC in 1997, employed the NRC approved 1996 Evaluation Model (Reference 2). Westinghouse recently underwent a program to revise the statistical app roach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95 percentile. This method is still based on the Code Qualification Document (CQD) methodology (Reference 7) and follows the steps in the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case. The ASTRUM evaluation model is documented in WCAP-16009-P-A (Reference 1).
This section summarizes the application of the Westinghouse ASTRUM Best Estimate Loss of Coolant Accident (BELOCA) evaluation model to the Indian Point Unit 2 Nuclear Plant for analysis of the large break LOCAs (LBLOCA). The analysis was performed in compliance with all the NRC conditions and limitations as identified in WCAP-16009-P-A.
The current WCOBRA/TRAC model for Indian Point Unit 2 is based on the methodology of WCAP-1 2945 and uses the current uprated power level of 3216 MWth. Use of the best estimate LBLOCA methodology for Indian Point 2 was initially approved by NRC in 1997 (Reference 2). The WCOBRAITRAC noding that was developed at that time remains unchanged for the best estimate LBLOCA ASTRUM analysis. The ASTRUM best estimate LBLOCA analysis was performed for a full core of upgraded fuel. Table 1 lists the major plant parameter assumptions used in the analysis for Indian Point Unit 2. The axial power distribution envelope assumption is shown in Figure 1. Table 2 summarizes the results of the ASTRUM best estimate LBLOCA analysis, as previously reported in Reference 6. Table 3 contains a sequence of events for the limiting PCT transient. Based on these results, Indian Point Unit 2 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.
The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 95th percentile of the Peak Clad Temperature (PCT), Local Maximum Oxidation (LMO), and Core Wide Oxidation (CWO) with 95% confidence level. These parameters are needed to satisfy the 10 CFR 50.46 criteria with regard to PCT, LMO, and CWO. From these 124 calculations, Run 76 proved to be the limiting PCT transient and the limiting IMO transient, and Run 11 the limiting CWO transient.
The scatter plot presented on Figure 2 shows the effect of the effective break area on the final PCT. The effective break area is calculated by multiplying the discharge coefficient (CD) with the sample value of the break area, normalized to the cold-leg cross sectional area. Figure 2 is provided because the break area is a significant contributor to the variation in PCT.
Figures 3 and 4 are presented to show the limiting cladding transient for each criterion. Figure 3 shows the predicted clad temperature transient at the PCT and LMO limiting elevation for Run 76 and Figure 4 presents the PCT trace for the CWO limiting transient from Run 11.

NL-05-107 Docket No. 50-247 Attachment I Page 3 of 11 Table 1: Major Plant Parameter Assumptions Used in the Best Estimate Large Break LOCA Analysis for Indian Point Unit 2 Parameter                                          Value                   Documentation Plant Physical Description
* SG Tube Plugging                         l 10%   **                                   UFSAR 14.3 Plant Initial Operating Conditions
* Reactor Power                             5102% of 3216 MWt                           UFSAR 14.3
* Peaking Factors                           Fas 2.5                                     UFSAR 14.3
_ __ __ __ __ _ _ _ _ _ _ _ _ _ _ _  FAH~s 1.7_                                    _ _   _  _  _
* Axial Power Distribution                  See Figure 1                                UFSAR 14.3 Fluid Conditions
* TAVG                                      549 - 3.3 0F 5 TAVO < 572 + 3.3 &deg;F ()*      UFSAR 14.3
* Pressurizer Pressure                      2250 - 25 psia s PRCS s 2250 + 25 psia (2)  UFSAR 14.3
* Reactor Coolant Flow                       2 80700 gpm/loop                           UFSAR 14.3
* Accumulator Temperature                    80 1Fs TAccs 130 OF                        UFSAR 14.3
* Accumulator Pressure                      612.7 psia s PACC S 699.7 psia **           UFSAR 14.3
* Accumulator Water Volume                  723 ft3 s VAcc s 875 ft3                    UFSAR 14.3 Accident Boundary Conditions
* Single Failure Assumptions                 Loss of one ECCS train                     UFSAR 14.3
* Safety Injection Flow                      Minimum                                    UFSAR 14.3
* Safety Injection Temperature               35 0F STs1 s 110 OF                        UFSAR 14.3
* Safety Injection Initiation                s 38 seconds (with offsite power)              AR Delay Time                                 s 45 seconds (without offsite power)       UFS     14.3
* Containment Pressure                       Bounded (minimum)                           UFSAR 14.3 (1) Include -3 (bias)*
(2) Include -3, +12 (bias)*
* Bias sign convention: u+emeans indicated value is higher than actual and "-. means indicated value is lower than actual.
    ** The current version of the UFSAR does not contain this value, or range. However, this value or range will be updated in a subsequent UFSAR revision.

NL-05-107 Docket No. 50-247 Attachment 1 Page 4 of 11 Table 2: Indian Point Unit 2 Best Estimate Large Break LOCA Results 10 CFR 50.46 Requirement                Value             Criteria 95/95 PCT (0F)                          1962        < 2200 95/95 LMO 1(%)                           2.39       < 17 95/95 CWO (%)                            0.35       <1 Core remains        Core remains C    abGe     ercoolable                            coolable Core remains cool Long Term Cooling Core     term cool in long term in longremains 1 Local Maximum Oxidation 2 Core Wide Oxidation Table 3: Indian Point Unit 2 Best Estimate Large Break LOCA Sequence of Events for Limiting PCT Transient Event                                   Time (sec)
Start of Transient                           0.0 Safety Injection Signal                      6.0 Accumulator Injection Begins                10.0 End of Blowdown                             28.0 Accumulator Empty                           39.0 Bottom of Core Recovery                     40.0 Safety Injection Begins                      51.0 PCT Occurs                                 123.0 PCT Elevation Quench                       330.0 End of Transient                           500.0

NL-05-107 Docket No. 50-247 Attachment I Page 5 of 11 Figure 1: Indian Point Unit 2 Best Estimate Large Break LOCA Analysis Axial Power Shape Operating Space Envelope PBOT: integrated power fraction in the lower third of the core PMID: integrated power fraction in the middle third of the core

NL-05-107 Docket No. 50-247 Attachment I Page 6 of 11 Figure 2: Indian Point Unit 2 Best Estimate Large Break LOCA Analysis HOTSPOT PCT vs. Effective Break Area Scatter Plot 2000
                                                                  .
* aU          .
1800  -
mE.
N    a
                                                                               ; U 1600-
* 0
* affi               .1.           .
Q    mu ONU
         "  1400  -
* E*-u C-,)
I-Cl-                       *       *              *       . U C)                                                  -
                                                   *_
1200  -
                                                 -;
U M
a
* U a
1000  -
M
                       *;.g 800
_0 600  -I II I I      I  I I      I  I    I  I    I     I  I   I I  I I    I  I    I 0                      i               15                       2.5          3 CdsAbreak/ACL

NL-05-1 07 Docket No. 50-247 Attachment 1 Page 7 of 11 Figure 3: Indian Point Unit 2 Best Estimate Large Break LOCA Analysis HOTSPOT Clad Temperature Transient at the Limiting Elevation for the PCT and LMO Limiting Case Cladding  Temperature  at Limiting PCT and LMO Elevation 2000 1500-D 1000 so-0

NL-05-1 07 Docket No. 50-247 Attachment 1 Page 8 of 11 Figure 4: Indian Point Unit 2 Best Estimate Large Break LOCA Analysis WCOBRAITRAC PCT Transient for the CWO Limiting Case Peck Cladding   Temperature 2000 1500 -
31000 E
500 0
Time After Break (s)

NL-05-1 07 Docket No. 50-247 Attachment 1 Page 9 of 11

==5.0 REGULATORY ANALYSIS==

5.1     No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) has evaluated the safety significance of the proposed changes regarding use of the ASTRUM methodology in the analysis of the Large Break Loss of Coolant Accident (LBLOCA) for Indian Point 2 (IP2) according to the criteria of 10 CFR 50.92, "Issuance of Amendment". Entergy has determined that the subject changes do not involve a Significant Hazards Consideration as discussed below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:        No The proposed change modifies the analysis methodology used to account for the variation in parameters that are used for the safety analysis of the LBLOCA. This proposed change has no effect on the design or operation of plant equipment.
Use of the new methodology will revise the results of the current analysis, but there will be no change in initiating events for this accident scenario or the ability of the plant equipment or plant operators to respond.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:        No The proposed change does not involve modifications to existing plant equipment or the installation of any new equipment. The proposed change only affects the analysis methodology that is used to evaluate the response of existing plant equipment to the LBLOCA scenario. Plant operating and emergency procedures that are in place for the LBLOCA scenario are also not being changed by this proposed amendment. This proposed change does not create new failure modes or malfunctions of plant equipment nor is there a new credible failure mechanism.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response:        No

NL-05-1 07 Docket No. 50-247 Attachment I Page 10 of 11 The proposed license amendment revises the analysis methodology which is used to assess the impact of the LBLOCA scenario with respect to established acceptance criteria. Margins of safety for LBLOCA include quantitative limits for fuel performance established in 10 CFR 50.46. These acceptance criteria and the associated margins of safety are not being changed. The evaluation of the LBLOCA scenario, using the proposed new methodology must still meet the existing established acceptance criteria.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Entergy Nuclear Operations, Inc. concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92 (c), and, accordingly, a finding of uno significant hazards consideration' is justified.
5.2     Applicable Requlatorv Requirements / Criteria The applicable regulatory requirement for this license amendment request is 10 CFR 50.46, which includes requirements and acceptance criteria pertaining to the evaluation of emergency core cooling system (ECCS) performance.
This regulation includes the requirement that "... uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria ... there is a high level of probability that the criteria would not be exceeded."
The proposed license amendment requests approval to use the ASTRUM methodology (WCAP-16009) for the treatment of uncertainties in the inputs used for the LBLOCA analysis. There is no change being proposed to the analysis acceptance criteria specified in the regulations. NRC has reviewed WCAP-16009 and found it acceptable for referencing in licensing applications for Westinghouse and Combustion Engineering designed pressurized water reactors. WCAP-16009 is applicable to Indian Point 2 and the plant-specific application of the ASTRUM methodology to the IP2 LBLOCA analysis has been performed in accordance with the conditions and limitations of the topical report and the associated NRC Safety Evaluation.
5.3      Environmental Considerations The proposed change to the IP2 Technical Specifications do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

NL-05-1 07 Docket No. 50-247 Attachment 1 Page 11 of 11 6.0     PRECEDENCE NRC has reviewed and accepted the Westinghouse topical report (WCAP-16009) which describes the ASTRUM methodology and NRC is in the process of reviewing a plant-specific application (Docket 50-244; April 29, 2005) which includes use of ASTRUM.

==7.0     REFERENCES==
: 1.      Nissley, M. E., et.al, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty (ASTRUM)," WCAP-16009-P-A, January 2005.
: 2.      NRC letter to Consolidated Edison; Issuance of Amendment [188] for Indian Point Nuclear Generating Unit No. 2", dated March 31, 1997.
: 3.      NRC letter to Entergy; "Issuance of Amendment [241] Re: 3.26 Percent Power Uprate (TAC MC1865)0, dated October 27, 2004.
: 4.      Entergy letter (NL-04-081) to NRC; "Proposed Schedule for Reanalysis of Large Break Loss of Coolant Accident", July 2, 2004.
: 5.      NRC letter to Westinghouse; "Final Safety Evaluation for WCAP-1 6009-P, Revision 0, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty (ASTRUM)," November 5, 2004.
: 6.      Entergy letter (NL-05-058) to NRC; "Reanalysis of Large Break Loss of Coolant Accident Using ASTRUM", April 22, 2005.
: 7.      Bajorek, S. M., et. al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis,' WCAP-1 2945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary).

ATTACHMENT 2 TO NL-05-107 MARKUP OF TECHNICAL SPECIFICATION PAGES FOR PROPOSED CHANGES REGARDING USE OF ASTRUM FOR LBLOCA ANALYSIS AFFECTED PAGE 5.6-3 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247

Reporting Requirements 5.6 5.6 Reporting Requirements NO CHANGES THIS PAGE - INFORMATION ONLY 5.6.3        Radioactive Effluent Release ReDort
                                                     -NOTE -
A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.
5.6.4        Not Used 5.6.5        CORE OPERATING LIMITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1. Technical Specification 2.1, Safety Limits (SL);
: 2. Technical Specification 3.1.1, SHUTDOWN MARGIN (SDM);
: 3. Technical Specification 3.1.3, Moderator Temperature Coefficient (MTC);
: 4. Technical Specification 3.1.5, Shutdown Bank Insertion Limits;
: 5. Technical Specification 3.1.6, Control Bank Insertion Limits;
: 6. Technical Specification 3.2.1, Heat Flux Hot Channel Factor (FO(Z));
: 7. Technical Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor, INDIAN POINT 2                              5.6- 2                         Amendment No. 241

INSERT REFERENCE 6 FOR SECTION 5.6.5.b:
WCAP-16009-P-A, uRealistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," M. E. Nissley, et al., January 2005.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
: 8. Technical Specification 3.2.3, Axial Flux Difference (AFD);
: 9. Technical Specification 3.3.1, Reactor Protection System Instrumentation;
: 10. Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and
: 11. Technical Specification 3.9.1, Boron Concentration.
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
: 1. WCAP-9272-P-A,         "Westinghouse       Reload     Safety    Evaluation Methodology," July 1985;
: 2. WCAP-8385, nPower Distribution Control and Load Following Procedures - Topical Report", September 1974;
: 3. T.M. Anderson to K. Kniel (NRC) January 31, 1980 - 

==Attachment:==

Operation and Safety Analysis Aspects of an Improved Load Follow Package;
: 4. NUREG-0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981, including Branch Technical Position CPB 4.3-1,Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981;
: 5. WCAP-1 1397-P-A, "Revised Thermal Design Procedure', April 1989; Replace with                 6. Y         P12945-P, "Code Qualification Document for Best Es mte new Ref 6,                        /LOCA Analysis", June 1993, as supplemented up to June 13, 1996 as see insert        at        /       follows:
next page
                    \    l          Westinghouse letter (N. J. Liparulo) to USNRC, "Re-Analysis Work Plans Using Final Best Estimate Methodology", NSD-NRC-96-4746,
                          \         June 13, 1996, and
* USNRC letter (J. Harold) to Consolidated Edison Company (S.
Quinn), "Issuance of Amendment [188] for Indian Point Nuclea
                                 \ Generating Unit No. 2 (TAC No. M96370)", March 1997.
: 7. WCAP-8745-P-A, Design Bases for the Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Functions", September 1986; INDIAN POINT 2                                5.6- 3                           Amendment No. 241}}