ML17138A213
| ML17138A213 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/18/2017 |
| From: | Wink R Ameren Missouri, Union Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LDCN 16-0011, ULNRC-063 72 | |
| Download: ML17138A213 (16) | |
Text
Ameren MISSOURI Callaway Plant May 1$. 2017 ULNRC-063 72 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.90 Ladies and Gentlemen:
DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT I UNION ELECTRIC CO.
RENEWED FACILITY OPERATING LICENSE NPF-30 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION CONCERNING REVISION OF TS 5.6.5, CORE OPERATING LIMITS REPORT (COLR), TO ALLOW THE USE OF THE PARAGON AND NEXUS CORE DESIGN METHODS (LDCN 16-0011)
By letier dated October 11, 2016 (ADAMS Accession No. ML16286A553). Union Electric Company (dba Ameren Missouri) submitted a license amendment request (LAR) for the Callaway Plant. The proposed amendment would modify Technical Specification (TS) 5.6.5.
CORE OPERATING LIMITS REPORT (COLR). to reference and allow use of the NRC approved methodologies described in WCAP-1 6045-P-A. Qualification of the Two Dimensional Transport Code PARAGON, WCAP-I6045-P-A. Addendum 1-A. Qualification of the NEXUS Nuclear Data Methodology. and WCAP-10965-P-A. Addendum 2-A.
Qualification of the New Pin Power Recovery Methodology. for the Callawa Plant.
During its review of the submitted report. the NRC staff determined that requests for additional information (PAls) were needed to complete its review. The NRC transmitted PAls to Ameren Missouri in electronic form on April 25. 2017. and requested that the responses be provided by May 25, 2017. Accordingly. the PAl responses are provided in an enclosure to this letter.
Specifically, Enclosure I contains the Ameren Missouri responses to the PAl questions.
Enclosures 2 and 3 contain a drafi revised TS markup and revised, clean typed pages of TS 5.6.5. respectively, for information. A supplement to Ameren Missouris LAR, providing a revision to the changes proposed for TS 5.6.5, will be submitted by June 15. 2017 (to allow time for review by Callaway Plants Onsite Review Committee).
P.O. Box 620 Fulton, MO 6525!
AmerenMissouri.com
ULNRC-06372 May 1$ 2017 Page 2 This letter does not contain new commitments. If there are any questions. please contact Mr. Torn Elwood at 314-225-1905.
I declare under penalty of peiury that the foregoing is true and correct.
Sincerely,
/ 1er C. Wink
(
Manager, Regulatory Affairs Executed on: 5// // 7
Enclosure:
1.
Responses to Requests for Additional Information 2.
Draft Technical Specification Page Markups 3.
Draft Retyped Technical Specification Pages
ULNRC-06372 May 18.2017 Page 3 cc:
Mr. Kriss M. Kennedy Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington. TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steeclman, MO 65077 Mr. L. John KIos Project Manager, Callaway Plant Oftice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 081-14 Washington. DC 20555-0001
ULNRC-063 72 May 18. 2017 Page 4 Index and send hardcopy to QA File A160.0761 Hardcopy:
Certuec Corporation 6100 Western Place. Suite 1050 fort Worth, IX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)
Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:
F. M. Dia I. F. Herrmann
- 3. L. Cox L. H. Kanuckel R. C. Wink T. B. Elwood B. D. Richardson J. W. Knaup Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. Jay Silberg (Pillsbury Winthrop Shaw Pittman LLP)
Missouri Public Service Commission Mr. Steve Feeler (DNR)
Mr. Robert Stout (DNR)
Enclosure I to ULNRC-06372 Page 1 of 2 ENCLOSURE 1 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION Request (RAI 1)
NRC Generic Letter (GL) 88-16. Removal of Cycle-Specific Parameter Limits from Technical Specifications: allows licensees to control certain reactor physics parameters on a cycLe-by-cycle basis by moving them to a separate report. provided they specify the calculation methodology. The discussion in the GL references terms indicating that approval of licensee control of these parameters is contingent on selection of a single. specific methodology (e.g., the specified methodology and an NRC-approved methodology, etc.).
The proposed amendment would preserve both the current PHOENIX-P/ANC nuclear design methodology and the new PARAGON-NEXUS/ANC nuclear design methodology in the licensees TS. By leaving both the current and proposed methodologies in the TS. the proposed amendment leaves open the possibility that either methodology could be used to calculate the core operating limits addressed in the COLR. effectively creating a new.
ambiguous methodology. The NRC staff has therefore been unable to determine that the proposed amendment is consistent with GL 88-16 guidance. Please propose a revision to TS 5.6.5 thaI provides cor a single. unambiguous method to determine how the core operating limits discussed in the submittal will be calculated.
Response to RAI I In response to this request. WCAP-11 596-P-A, Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores. is being removed from the list of analytical methods used to determine the core operating limits, as contained in TS 5.6.5.b.
Enclosures 2 and 3 contain a draft revised TS markup and clean typed pages of TS 5.6.5.
respectively, reflecting removal of WCAP-l 1596-P-A. The markup and clean typed, revised pages for TS 5.6.5 will be formally provided in a supplement to Ameren Missouris license amendment request. to be submitted by June 15, 2017.
Enclosure I to ULNRC-06372 Page2 of2 Request (RAI 2)
(Clarification) Addendum 1-A to WCAP-1 6045-P-A brings about a change to the way boron letdown curves are calculated and input into the overall nuclear design method. Because of this. in previous reviews of PARAGON-NEXUS/ANC applications, the NRC staff found it necessary to verify that no changes were made to the analysis methods for post-loss of coolant accident (LOCA) subcriticality and boric acid precipitation behavior. Please verify that no changes are made to these analysis methods as part of this LAR.
Response to RAI 2 The use of WCAP-16045-P-A. Addendum 1-A. for Callaway does not affect the inputs or method(s) for ensuring core subcriticality. for both short and long-term post-LOCA conditions. thereby precluding the potential for return to power following a large break LOCA. Since neither the post-LOCA boron source concentration nor heat generation are impacted by the use of Addendum 1 -A. the current emergency operating procedure timing for boric acid precipitation and the action time for switching to simultaneous injection wilt continue to remain valid. Core design specific parameters that are verified each cycle to be conservative with respect to the LOCA inputs and refueling boron concentration will continue to be calculated using NRC approved methods.
ENCLOSURE 2 DRAFT MARKUP OF TECHNICAL SPECIFICATION PAGES
Page2of4 Reporting Requirements 1c Cr 5.6
/. /-.;s,)
5.6 Reporhng Requirements 5.6.5 CORE OPERATING UMITS REPORT tCOLR a.
Core operating limits shah 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.
Moderator Temperature Coefficient limits in Specification 3 1 3, 2.
Shutdown Bank Insertion Limit for Specification 3 1 5, 3.
Control Bank Insertion Limits lot Specification 3 1 6, 4.
Axial Flux Difference Limits for Specification 3 2 3, 5.
Heat Flux Hot Channel Factor, F0(Z), F0RTF, K(Z). W(Z) and FQ Penalty Factors for Specification 3 2 1, 6.
Nuclear Enthalpy Rise Hot Channel Factor F, FH RTP and Power Factor Multiplier, PF,,H. limits for Specification 3 2 2, 7.
Shutdown Margin Limits for Specifications 3 11 3 1 4, 3 1 5, 3 1 6, and 3 1 8, 8.
Reactor Core Safety Limits Figure for Specification 2 1.1, 9.
Overtemperature T and Overpower L\\T Setpoint Parameters for Specification 3 3 1 and 10.
Reactor Coolant System Pressure and Temperature DNB Limits for Specification 3.4 1.
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.
2.
WCAP-10216-P-A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL AND FQ SURVEILLANCE TECHNICAL SPECIFICATION.
3.
WCAP-10266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE.
(continued)
CALLAWAY PLANT 5,0-24 Amendment No. 215 to ULNRC06372
/\\
IL
C / /
Reporting Requirements 5.6 5.6 Reporting Requirements 4.
WCAP-12610-P-A, VANTAGE ÷ FUEL ASSEMBLY REFERENCE CORE REPORT 5.
WCAP-11397-P-A, REVISED THERMAL DESIGN PROCEDURE.
6.
WCAP-14565-P-A, VIPRE-Ol MODELING AND QUALIFICATION FOR PRESSURIZED WATER REACTOR NON-LOCA THERMAL HYDRAUUC SAFETY ANALYSIS.
7.
WCAP-10851-P-A, IMPROVED FUEL PERFORMANCE MODELS FOR WESTINGHOUSE FUEL ROD DESIGN AND SAFETY EVALUATIONS.
8.
WCAP-1 5063-P-A, WESTINGHOUSE IMPROVED PERFORMANCE ANALYSIS AND DESIGN MODEL (PAD 4.0).
9.
WCAP-8745-P-A, DESIGN BASES FOR THE THERMAL OVERPOWER DT AND THERMAL OVERTEMPERATURE DT TRIP FUNCTIONS 10.
WCAP-10965-P-A, ANC: A WESTINGHOUSE ADVANCED NODAL COMPUTER CODE.
t-we-AP159G-rA, QUAUFtCATION OF THE P1 IO[NIX-P/ANC NUCLEAR DEGIGN SYSTEM ron PnE:SSURIZ[D WATER 12.
WCAP-1 3524-P-A, APOLLO: A ONE DIMENSIONAL NEUTRON DIFFUSION THEORY PROGRAM.
13 WCAP-14565-P-A Addendum 2-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for
._ g PWR Low Pressure Applications.
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance far each reload cycle to the NRC.
(continued)
CALLAWAY PLANT 5.0-25 Amendment No. 216 to ULNRCO6372 Page 4 of 4 LDCN 16-0011 Tech Spec S.S.5.b Markup INSERT A
- 11. WCAP-10965-P-A Addendum 2-A, Qualification of the New Pin Power Recovery Methodology.
INSERT B
- 14. WCAP-16045-P-A, Qualification of the TwoDimensional Transport Code PARAGON.
- 15. WCAP-6045-P-A Addendum 1-A, Qualifcation of the NEXUS Nuclear Data Methodology.
ENCLOSURE 3 DRAFT RETYPED TECHNICAL SPECIFICATION PAGES to ULNRCO6372 Page 2 of 6 Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CER 50.4.
5.6.1 Not Used.
5.6.2 Annual Radiological Environmental Operatinci Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be subm iffed by May 1 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environm ental monitoring program for the reporting period.
The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections lV.B.2, IV.B.3, and IV.C.
The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in a format similar to the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3 Radioactive Effluent Release Reoort The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submiffed prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a sum mary 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 lV.B.1.
5.6.4 Not used.
(continued)
CALLAWAY PLANT 5.0-23 Amendment No. 215 to ULNRCO6372 Page 3 of6 Reporting Requirements 5.6 5.6 Reporting Requirements 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 docu mented in the COLR for the following:
1.
Moderator Temperature Coefficient limits in Specification 3.1.3, 2.
Shutdown Bank Insertion Limit for Specification 31.5, 3.
Control Bank Insertion Limits for Specification 3.1.6, 4.
Axial Flux Difference Limits for Specification 3.2.3, 5.
Heat Flux Hot Channel Factor, EQ(Z), FQRTP, K(Z), W(Z) and EQ Penalty Factors for Specification 3.2.1, 6.
Nuclear Enthalpy Rise Hot Channel Factor F, FH RTP and Power Factor Multiplier, PFH, limits for Specification 3.2.2, 7.
Shutdown Margin Limits for Specifications 3.1.1, 3.1.4, 3.1.5, 3.1.6, and 3.1.8, 8.
Reactor Core Safety Limits Figure for Specification 2.1.1, 9.
Overtemperature.\\T and Overpower L\\T Setpoint Parameters for Specification 3.3.1, and 10, Reactor Coolant System Pressure and Temperature DNB Limits for Specification 3.4.1.
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.
2.
WCAP-1 0216-P-A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL AND EQ SURVEILLANCE TECHNICAL SPECIFICATION.
3.
WCAP-1 0266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE.
(continued)
CALLAWAY PLANT 5.0-24 Amendment No. 215 to ULNRCO6372 Page 4 of 6 Reporting Requirements 5.6 5.6 Reporting Requirements 4.
WCAP-1 261 0-P-A, VANTAGE + FUEL ASSEMBLY REFERENCE CORE REPORT.
5.
WCAP-11397-P-A, REVISED THERMAL DESIGN PROCEDURE.
6.
WCAP-14565-P-A, VIPRE-Ol MODELING AND QUALIFICATION FOR PRESSURIZED WATER REACTOR NON-LOCA THERMAL-HYDRAULIC SAFETY ANALYSIS.
7.
WCAP-10851-P-A, IMPROVED FUEL PERFORMANCE MODELS FOR WESTINGHOUSE FUEL ROD DESIGN AND SAFETY EVALUATIONS.
8.
WCAP-15063-P-A, WESTINGHOUSE IMPROVED PERFORMANCE ANALYSIS AND DESIGN MODEL (PAD 4.0).
9.
WCAP-8745-P-A, DESIGN BASES FOR THE THERMAL OVERPOWER DT AND THERMAL OVERTEMPERATURE DI TRIP FUNCTIONS.
10.
WCAP-1 0965-P-A, ANC: A WESTINGHOUSE ADVANCED NODAL COMPUTER CODE.
11.
WCAP-1 0965-P-A Addendum 2-A, Qualification of the New Pin Power Recovery Methodology.
12.
WCAP-1 3524-P-A, APOLLO: A ONE DIMENSIONAL NEUTRON DIFFUSION THEORY PROGRAM.
13.
WCAP-1 4565-P-A Addendum 2-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications.
14.
WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON.
15.
WCAP-16045-P-AAddendum 1-A, Qualification of the NEXUS Nuclear Data Methodology.
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
(continued)
CALLAWAY PLANT 5.0-25 Amendment No. #
to ULNRCO6372 Page 5 of 6 Reporting Requirements 5.6 5.6 Reporting Requirements U.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT tPTLR) a.
RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing and PORV lift setting as well as heatup and coold own rates shall be established and documented in the PTLR for the following:
1.
Specification 3.4.3, RCS Pressure and Temperature (PIT) Limits, and 2.
Specification 3.4.12, Cold Overpressure Mitigation System (COMS).
b.
The analytical methods used to determine the RCS pressure and temperature and COMS PORV limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-NP-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.
c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 Not used.
5.6.8 PAM Report When a report is required by Condition B or F of LCO 3.33, Post Accident Monitoring (PAM) Instrumentation, a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.9 Not used.
(continued)
CALLAWAY PLANT 5.0-26 Amendment No. #
to ULNRCO6372 Page 6 of 6 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program. The report shall include:
a.
The scope of inspections performed on each SG; b.
Degradation mechanisms found; c.
Nondestructive examination techniques utilized for each degradation mechanism; U.
Location, orientation (if linear), and measured sizes (if available) of service induced indications; e.
Number of tubes plugged during the inspection outage for each degradation mechanism; 1.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator; and g.
The results of condition monitoring, including the results of tube pulls and in-situ testing.
CALLAWAY PLANT 5.0-27 Amendment No. 215