ML16007A180

Regulatory Audit Report Regarding License Amendment Request to Adopt Technical Specifications Task Force-490, Rev.0, and Implement Alternative Source Term

ML16007A180
Person / Time
Site:	Cook
Issue date:	01/20/2016
From:	Dietrich A Plant Licensing Branch III
To:	Gebbie J Indiana Michigan Power Co
Dietrich A
References
CAC MF5184, CAC MF5185
Download: ML16007A180 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 20, 2016 Mr. Joel P. Gebbie Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - REGULATORY AUDIT REPORT REGARDING LICENSE AMENDMENT REQUEST TO ADOPT TECHNICAL SPECIFICATIONS TASK FORCE-490, REV.O, AND IMPLEMENT ALTERNATIVE SOURCE TERM (CAC NOS. MF5184 AND MF5185)

Dear Mr. Gebbie:

By letter dated November 14, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14324A209), as supplemented by letter dated February 12, 2015 (ADAMS Accession No. ML15050A247), Indiana Michigan Power Company (l&M), submitted a license amendment request (LAR) for the Donald C. Cook Nuclear Plant, Units 1 and 2. The proposed amendment consists of adoption of Technical Specifications Task Force-490, Rev. 0, and implementation of a full-scope alternate source term (AST) radiological analysis methodology.

An audit was conducted from September 21, 2015, to September 24, 2015, at the offices of l&M in Buchanan, Michigan. The purpose of the audit was to review the thermal hydraulic (TH) analysis and parameter information sources for each accident considered in the AST analysis.

The NRC staff determined, based on the audit, that a request for additional information would be necessary regarding the use of plant operator simulator data for certain TH input parameters.

In addition, the NRG staff identified reference documents that would need to be made available for continued staff review in order to complete the evaluation for the LAR.

J. Gebbie The details of the results of the audit are set forth in the enclosed audit report. The NRC staff appreciates the resources that were made available by your staff during the audit. If you have any questions, please contact me at (301) 415-2846.

Sincerely,

_,/IJ /Jjfj, Allison W. Dietrich, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosure:

Audit Report cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AUDIT REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION LICENSE AMENDMENT REQUEST TO ADOPT TSTF-490. REVISION 0 AND IMPLEMENT FULL-SCOPE ALTERNATE SOURCE TERM DONALD C. COOK NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316

1.0 INTRODUCTION

By letter dated November 14, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14324A209), Indiana Michigan Power Company (l&M, the licensee), submitted a license amendment request (LAR) for the Donald C. Cook Nuclear Plant (CNP), Units 1 and 2. The proposed amendment consists of adoption of Technical Specifications Task Force-490, Rev. 0, and implementation of a full-scope alternate source term (AST) radiological analysis methodology. 2 of the LAR listed the various new AST input parameter values, including those based on reactor coolant system (RCS) performance for offsite and Control Room (CR) habitability doses, for each accident analysis. However, the LAR lacked sufficient information about the thermal hydraulic (TH) analyses that were applied as the basis for the values. This prompted a supplemental request for information from the NRC staff (ADAMS Accession No. ML14363A491). The licensee provided the requested supplemental information by letter dated February 12, 2015 (ADAMS Accession No. ML15050A247).

As stated in the February 12, 2015 letter, "The Current Licensing Basis (CLB) TH calculations were used to provide input to most of the new dose analyses, although some of those inputs are different from previous inputs that were derived from the same TH calculations." The supplemental letter also states that the majority of the input parameters originate from calculations performed for previous submittals, such as CNP Units 1 and 2 license amendment Nos. 271 and 252 for implementation of AST for CR habitability (ADAMS Accession No. ML022980619) and license amendment Nos. 256 and 239 to address Steam Generator Tube Rupture (SGTR) overfill (ADAMS Accession No. ML012690136). Other inputs were obtained from:

Projects implemented under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.59, such as the Unit 1 replacement steam generator (SG) modification;

• The CR habitability and offsite dose consequence analyses revised in 2011 and implemented under 10 CFR 50.59; Information obtained from actual plant post-trip data recorded by the CNP Units 1 and 2 Plant Process Computer; and

Information obtained from simulator data representing a Unit 1 SGTR transient including operator actions.

The actual source documents (e.g., volume data provided by Westinghouse Electric Company during a SG replacement project) for most of the accident analyses were not directly available to the NRC staff to verify the proper incorporation of the values as input parameters to the AST analysis. A request for additional information (RAI) was sent to l&M on July 14, 2015, requesting additional descriptions of the TH analysis and parameter information sources for each accident considered in the AST analysis (ADAMS Accession No. ML15195A698). The licensee provided the RAI response to the NRC in a letter dated August 24, 2015 (ADAMS Accession No. ML15238A726).

The August 24, 2015, letter provided additional information regarding the sources for the RCS input parameter values; however, it was not known whether the source documents had been reviewed by the NRC staff in prior LARs. Thus, the staff could not verify the authenticity of the RCS input parameter values to their source documentation. Nor could the staff verify whether the sources, other than those used for the previous LARs, had been reviewed and approved by the NRC, and were therefore part of the CLB. To verify the authenticity of the RCS input parameter values, the NRC staff held an on-site audit to review the supporting documentation and calculation files for the LAR.

2.0 AUDIT ACTIVITIES AND OBSERVATIONS The audit was conducted from September 21, 2015, to September 24, 2015, at the offices of l&M in Buchanan, Michigan. The purpose of the audit was as follows:

( 1) To review the documents listed as references in Enclosure 2 of the l&M letter dated August 24, 2015, in order to verify the sources of the TH input parameter values provided in Enclosure 12 of the LAR for the accidents specified by Regulatory Guide (RG) 1.183, which are:

SGTR; Main Steam Line Break (MSLB);

Locked Rotor Accident (LRA);

Control Rod Ejection (CRE); and Loss-Of-Coolant Accident (LOCA).

(2) To evaluate whether the TH input parameter values in the source documents identified in step ( 1) were properly set for each of the AST accidents.

(3) To identify the information that would need to be provided in order for the NRC staff to complete the accident analysis review of the AST LAR.

2. 1 Overview The audit began with a discussion of the timeline of the past analyses and related NRC reviews of the CR habitability and off-site dose consequences, in regards to the information that the licensee had provided to the staff for this LAR. Licensee responses to auditor inquiries indicated that the dose consequence analyses were updated under the 10 CFR 50.59 process

since the prior partial AST implementation from the years 2000 to 2002. The current analysis supporting the full implementation AST LAR was believed by the staff to be a further update.

However, it became clear from the initial audit discussions the recent analysis supporting the LAR was a complete re-baselining of the CR habitability and off-site dose consequences, and did not depend on the prior 10 CFR 50.59 updates. The auditors also determined that some key calculation files supporting the prior partial implementation of the AST were also the source of several key TH input parameter values for the full implementation of AST LAR.

Based on this revised understanding of the sources for the input parameter values for the accidents considered under this LAR for full implementation of AST, the NRC staff concentrated the audit on the analyses of accident events that would lead to the release of primary coolant, namely SGTR, MSLB, LRA, CRE, and LOCA.

The documents reviewed for the audit are presented in Table 1. The list of documents was provided in the August 24, 2015, l&M letter. The auditors first reviewed the documents to obtain an overall understanding of the material presented, and to gain comprehension on the analyses.

Table 1: Documents Reviewed Ref. Document No. Title Date I ML No.

No.a 1 AEP-NRC-2014-65 License Amendment Request To Adopt November 14, TSTF-490, Revision 0 2014 ML14324A209 7 NIA CNP Units 1 and 2 Technical N/A Requirements Manual (TRM), Revs. 37 (U1) and 36 (U2).

19 Calculation Red Wolf Cook Nuclear Plant AST Radiological July 2014 Associates (RWA)- Analysis Input Parameter Development 1313-001, Rev. 0 20 Calculation RWA-1313- Cook Nuclear Plant LOCA AST September 006, Rev. 1 Radiological Analysis 2014 22 Calculation RWA-1313- Cook Nuclear Plant Locked Rotor AST July 2014 009, Rev. 0 Radiological Analysis 23 Calculation RWA-1313- Cook Nuclear Plant Main Steam Line July 2014 010, Rev. 0 Break AST Radioloqical Analysis 24 Calculation RWA-1313- Cook Nuclear Plant Steam Generator July 2014 011, Rev. O Tube Rupture AST Radiological Analysis 25 Calculation RWA-1313- Cook Nuclear Plant Control Rod Ejection July 2014 012, Rev. 0 AST Radioloqical Analysis 33 Calculation TH-00-03, D.C. Cook Unit 2 Steam Generator Tube N/A Rev. 0 Rupture with Operator Actions 34 Calculation CN-CRA- D.C. Cook Units 1 & 2 Steam Releases N/A 99-047, Rev. 0 for Radiological Dose Calculation

Ref. Document No. Title Date I ML No.

No.a 35 Calculation CN-CRA- Donald C. Cook Steam Generator Tube N/A 99-55, Rev. 1 Rupture T&H Analysis for NUREG-1465 Dose Project- Revised 38 Engineering Change Unit 1 Cycle 25 Core Reload N/A EC-0000051727 39 Engineering Change Unit 2 Cycle 21 Core Reload N/A EC-0000052225 40 Engineering Evaluation Steam Generator Safety Valves NIA EE-2005-0139 42 Design Information Miscellaneous Input for Dose Reanalysis May 2014 Transmittal (DIT)-B- Effort (Contract# 01559762) 03594-00 43 Design Information Normal Operating Pressure (NOP) August 2013 Transmittal DIT-B- Normal Operating Temperature (NOT) 03526-02 Steam Line Break Mass & Energy Release Analysis 44 Design Information Reactor Coolant System Volumes November Transmittal DIT SGRP 1999 99035-00, Rev. 0 45 Design Information Unit 2 Steam Generator Design Moisture June 2000 Transmittal DIT-SGRP- Carryover 00064-00 46 AEP-13-63 American Electric Power Donald C. Cook August 2013 Units 1 and 2 Ultimate Heat Sink Program 49 Vendor Technical Section 8.2.1, "Babcock and Wilcox N/A Document (VTD)- Canada, Operating and Maintenance BAWl-0015 Manual for Unit 1 Replacement Steam Generators, PUB. #222-7803-0&M-1" 52 AEP-NRC-2015-19 Donald C. Cook Nuclear Plant Unit 1 and February 12, Unit 2 Supplemental Information for the 2015 License Amendment Request to Adopt ML15050A247 TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" and Implement Full-Scope Alternative Source Term 55 Letter from NRC to l&M Donald C. Cook Nuclear Plant, Units 1 October 24, AND 2 - Issuance of Amendments (TAC 2001 NOS. MB0739 AND MB0740), License ML012690136 Amendment Nos. 256 and 239 N/A NIA CNP Updated Final Safety Analysis N/A Report (UFSAR), Revision 25

Ref. Document No. Title Date I ML No.

No.a N/A AEP-NRC-2015-75 Donald C. Cook Nuclear Plant Unit 1 and August 24, Unit 2 Response to Second Request for 2015 Additional Information Regarding the ML15238A726 License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term a Reference numbers are from Enclosure 2, pages 19 through 22, of l&M letter dated August 24, 2015 2.2 Auditor Observations and Evaluation In reviewing the transient events given in RG 1.183, the NRC staff verified that the same RCS and SG parameter values were applied across the following four transient events: SGTR, MSLB, LRA, and CRE. The staff also reviewed the LOCA event input parameter values for this proposed full-implementation AST. The staff also reviewed information provided in UFSAR Chapter 14 on the same RG 1.183 events for consistency between the provided AST documentation and the CLB. The list of TH parameters that the staff reviewed for the four transient events is as follows:

RCS mass; SG secondary liquid mass;

• Intact SG steam release; Ruptured SG steam release; Pre-trip total steam flow rate through condenser;

• Primary-secondary leak rate;

• Ruptured tube break flow; Time to cool RCS to 212 degrees Fahrenheit; Intact SG moisture carryover fraction;

• Duration of intact SG tube uncovery after a reactor trip; and SG flashing fractions.

Each of the transient events has a corresponding overarching calculation file describing the licensee's analytical methodology, and providing references for the various input parameter values. The calculation files are References 20, 22, 23, 24, and 25 from Table 1. The calculation files all refer to RWA-1313-001 (Table 1, Ref. 19). This document provided each input parameter value along with a corresponding reference to the source document where the value was calculated or determined.

2.2.1 Input Parameters Common to Most Transients 2.2.1.1 Reactor Coolant System Mass The RCS mass TH parameter was changed from 499,325 pounds mass (lbm) to 466, 141.5 lbm for the AST LAR. The document RWA-1313-001 is referenced for the development of the new RCS mass TH parameter. The NRC staff reviewed the equations provided in the reference documents for calculating the new RCS mass for the SGTR analysis for Units 1 and 2. The staff also verified that the same value was applied in the MSLB, LRA, and CRE analyses.

2.2.1.2 Steam Generator Secondary Liquid Mass The SG secondary liquid mass TH parameter was changed from 91,000 lbm per SG to 97,515.7 lbm per SG for Hot Full Power, and 161,000 lbm per SG for Hot Zero Power. The NRC staff reviewed calculation CN-CRA-99-047 {Table 1, Ref. 34) and license amendment Nos. 271 and 252. The staff reviewed the equations and relationships provided in the reference documents for determining the SG secondary liquid mass. Additionally, the staff reviewed previously approved amendments to the CLB for methodology and adoption of an AST. The staff verified that the same value was applied for the SGTR, MSLB, LRA, and CRE analyses.

2.2.1.3 Intact Steam Generator Steam Release The intact SG steam release TH parameters were originally as follows:

from 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> - 565,000 lbm; from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> - 1,505,000 lbm; and from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - not modeled.

The intact SG steam release TH parameters for the proposed AST are as follows:

from Oto 30 minutes -198,515 lbm;

• from 30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> - 314,432 lbm;

• from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> - 1,367,475 lbm; and from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 1,347,000 lbm.

The NRC staff reviewed calculation CN-CRA-99-55, {Table 1, Ref. 35) and license amel")dment Nos. 271 and 252. The staff reviewed the calculations and explanations for determining the new intact SG steam release parameters for the proposed AST and verified that the same value was applied for the SGTR, MSLB, LRA, and CRE analyses.

2.2.1.4 Primary-Secondary Leak Rate The primary-secondary leak rate for the proposed AST is 0.25 gallons per minute {gpm) per SG compared to the CLB value of 0.1041667 gpm per SG. The NRC staff reviewed the LAR dated November 14, 2014, and verified that the value was applied consistently for the SGTR, MSLB, LRA, and CRE analyses.

2.2.1.5 Time to Cool Reactor Coolant System to 212 Degrees Fahrenheit The time to cool the RCS to 212 degrees Fahrenheit was changed from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the proposed AST. The NRC staff reviewed the Ultimate Heat Sink Program (Table 1, Ref. 46).

The staff reviewed the input parameters for the calculation, and verified that the output provided in the reference was consistently applied for the SGTR, MSLB, LRA, and CRE analyses.

2.2.1.6 Intact Steam Generator Moisture Carryover Fraction The intact SG moisture carryover fraction for the proposed AST is 0.2%, which had not been modeled previously. The NRC staff reviewed VTD-BAWl-0015, DIT-SGRP-00064-00, and

RWA-1313-011 (Table 1, Refs. 49, 45, and 24, respectively). These reference documents provided the explanations for the determined intact SG moisture carryover fraction for CNP Units 1 and 2. The staff found that the value was consistently applied for the SGTR, MSLB, LRA, and CRE analyses.

2.2.1.7 Duration of Intact Steam Generator Tube Uncovery The duration of intact SG tube uncovery after a reactor trip is 40 minutes for the proposed AST.

The NRC staff reviewed reference calculation RWA-1313-001 and DIT-B-03594-00 (Table 1, Ref. 42). These references provided an explanation for how 40 minutes was determined for the duration of intact SG tube uncovery after a reactor trip. The staff found that the value was consistently applied for the SGTR, MSLB, LRA, and CRE analyses.

2.2.1.8 Steam Generator Flashing Fractions The SG flashing fractions were reviewed by the staff through RWA-1313-011. This reference provided the calculation and explanation for how the flashing fractions for CNP were determined for the proposed AST. The NRC staff found that the value was consistently applied for the SGTR, MSLB, LRA, and CRE analyses. However, the NRC has yet to accept the method of applying transient data from the plant operator simulator for determining the flashing fractions versus time. Therefore, this will likely be the subject of an upcoming RAI.

2.2.2 Input Parameters Specific to Steam Generator Tube Rupture 2.2.2.1 Ruptured Steam Generator Steam Release For the SGTR analysis, the ruptured SG steam release TH parameter was changed from 73,000 lbm to 66, 171 lbm for the O-to-30 minute interval. The NRC staff reviewed calculation CN-CRA-99-55 and license amendment Nos. 271 and 252. The staff reviewed the calculations and NUREG-1465 analysis for determining the new ruptured SG steam release for the proposed AST.

2.2.2.2 Pre-trip Total Steam Flow Rate through Condenser The pre-trip total steam flow rate through the condenser was updated from 17,200,000 lbm per hour to 17, 153,800 lbm per hour for the proposed AST. The NRC staff reviewed the CNP UFSAR, Revision 25, and Engineering Evaluation EE-2005-0139, (Table 1, Ref. 40). The staff reviewed the referenced document for the relieving capacity for the valves in the steam lines for Units 1 and 2.

2.2.2.3 Ruptured Tube Break Flow The ruptured tube break flow for the proposed AST is 146,704 lbm compared to 162,000 lbm from the CNP CLB. The NRC staff reviewed calculation CN-CRA-99-55, license amendment Nos. 271 and 252, and the February 12, 2015, l&M letter. The staff reviewed the calculations and assumptions for the ruptured tube break flow from the provided references.

2.2.3 Input Parameters Specific to Loss-Of-Coolant Accident The NRC staff reviewed the LOCA input parameter values provided in Table 2 of Enclosure 12 of the LAR, and Table 2 of Enclosure 2 of the August 24, 2015 l&M letter, and traced them via the calculation file RWA-1313-006, (Table 1, Ref. 20) and associated references. The purpose of this part of the audit was to determine which values changed between the partial implementation of AST and the current LAR values. The staff reviewed the associated source reference documents as listed in Table 2, "LOCA Inputs and Assumptions," of Enclosure 2 of the August 24, 2015, letter. The staff found that the AST's LOCA input parameter values were derived from, and were similar to, prior calculation values with fractional changes (increase or decrease) for conservatism. However, there were some minor exceptions. For example, a higher value was assumed for the containment spray start time to support flexibility for future plant modifications. Another variation occurred for changes in values such as the time for specific decontamination factors to reach values of 50 and 200, which were based on the output of RADTRAD computer runs.

3.0 CONCLUSION

The NRC staff reviewed the documents listed in Table 1 to determine the connections from the principal calculation files for the current full-implementation AST LAR input parameter values, through the common parameter reference document, RWA-1313-001, to the source calculation files. The staff also found that two documents from the previously approved partial implementation amendment (Table 1, Refs 34 and 35) were relied upon for the full-implementation AST LAR. In addition, the staff identified the documents that should be docketed or otherwise made available for continued NRC staff review, in order for the staff to complete its evaluation of LAR. These documents are:

Calculation RWA-1313-001, Rev. 0, "Cook Nuclear Plant AST Radiological Analysis Input Parameter Development," July 2014; Calculation RWA-1313-011, Rev. 0, "Cook Nuclear Plant Steam Generator Tube Rupture AST Radiological Analysis," July 2014; Calculation RWA-1313-010, Rev. 0, "Cook Nuclear Plant Main Steam Line Break AST Radiological Analysis," July 2014; Calculation CN-CRA-99-047, Rev. 0, "D.C. Cook Units 1 & 2 Steam Releases for Radiological Dose Calculation," July 1999; Calculation CN-CRA-99-55, Rev. 1, "Donald C. Cook Steam Generator Tube Rupture T&H Analysis for NUREG-1465 Dose Project- Revised," September 1999; AEP-13-63, "American Electric Power Donald C. Cook Units 1 and 2 Ultimate Heat Sink Program," August 2013; Design Information Transmittal DIT-B-03594-00, "Normal Operating Pressure (NOP) I Normal Operating Temperature (NOT) Steam Line Break Mass & Energy Release Analysis," August 2013; and SAE-CRA-98-162 (Referenced in CN-CRA-99-055, Rev 1).

For the sources of the TH input parameter values themselves, the NRC staff understands the LOFTRAN calculations discussed in CNP's UFSAR Chapter 14 are not part of the supporting calculations for the current LAR under review by the staff. Concerning the reliance on certain

TH input parameter values derived from plant operator simulator data, the staff will likely issue an RAI regarding the validity of the use of such data in a safety analysis.

These conclusions and planned follow-up actions were discussed with the licensee at the audit closeout meeting on September 24, 2015.

4.0 LICENSEE PERSONNEL CONTACTED FOR THIS AUDIT Terry Curtiss Jason Wright Doug Badgero Gregory Hill Mickey Bellville Helen Kish Principal Contributors: Donald Palmrose, NRO Matthew Hardgrove, NRR Date: January 20, 2016

ML16007A180 OFFICE NRR/DORL/LPLlll-1/PM N RR/DORL/LPLI 11-1 /LA NRR/DSS/SRXB/BC NAME ADietrich MHenderson EOesterle DATE 1/07/2016 1/12/2016 1/13/2016 OFFICE NRR/DORL/LPLI 11-1 /BC NRR/DORL/LPLlll-1/PM NAME DWrona ADietrich DATE 1/19/2016 1/20/2016