ML14099A450
| ML14099A450 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/06/2014 |
| From: | Thomas Wengert Plant Licensing Branch III |
| To: | Weber L Indiana Michigan Power Co |
| Tom Wengert, NRR/DORL 415-4037 | |
| References | |
| TAC MF2916 | |
| Download: ML14099A450 (12) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106 May 6, 2014
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNIT 1 -REQUEST FOR ADDITIONAL INFORMATION ON THE APPLICATION FOR AMENDMENT TO RESTORE NORMAL REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE CONSISTENT WITH PREVIOUSLY LICENSED CONDITIONS (TAC NO.
MF2916)
Dear Mr. Weber:
By letter dated October 8, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13283A121), as supplemented by letter dated April 29, 2014 (ADAMS Accession No. ML14121A422), Indiana Michigan Power Company (I&M) submitted an application for a license amendment to restore the normal reactor coolant system operating pressure and temperature consistent with previously licensed conditions for the Donald C. Cook Nuclear Plant, Unit 1.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject submittal and determined that additional information is needed to complete the review, as described in the enclosed request for additional information (RAI). The NRC staff discussed the RAI in draft form with your staff on March 26, 2014.
As discussed with Ms. Helen Etheridge of your staff on May 2, 2014, I&M will respond to this RAI within 30 days of the date of this letter, with the exception of SCVB RAI-3(c), SCVB RAI-9(a), and SCVB RAI-9(b). As agreed, I&M will identify the response dates for these items in its 30-day RAI response letter.
Please feel free to contact me at (301) 415-4037 if you need any further clarification of the questions in the enclosure.
Docket No. 50-315
Enclosure:
Request for Additional Information cc: Distribution via ListServ Sincerely, Thomas J. Wengert, Sen1or Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
REQUEST FOR ADDITIONAL INFORMATION RESTORATION OF NORMAL OPERATING PRESSURE AND NORMAL OPERATING TEMPERATURE DONALD C. COOK NUCLEAR PLANT UNIT 1 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 By letter dated October 8, 2013 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML13283A121 ), as supplemented by letter dated April 29, 2014 (ADAMS Accession No. ML14121A422), Indiana Michigan Power Company (I&M, the licensee) submitted for the U.S. Nuclear Regulatory Commission (NRC) review and approval a license amendment request (LAR) to restore normal reactor coolant system (RCS) operating pressure and temperature for the Donald C. Cook Nuclear Plant (GNP), Unit 1, consistent with previous licensed conditions. To complete its review, the NRC staff requests the following additional information.
Containment and Ventilation Systems Branch (SCVBl SCVB RAI-1 Reference 1, Enclosure 6, Section 5.4.1.5: it is noted that the proposed vessel/core inlet temperature of 514.6 degrees Fahrenheit (°F) is greater than the analysis of record (AOR) vessel/core inlet temperature of 506.6°F and the proposed RCS pressure 2317 per square inch absolute (psia) (includes uncertainty) is greater than the AOR RCS pressure of 2100 psia.
(a)
Please justify why the most limiting Joss-of-coolant accident (LOCA) short term Mass and energy (M&E) release and containment response for the proposed minimum vessel/core inlet temperature of 514.6°F and pressure of 2317 psia (includes uncertainty) is bounded by the LOCA short term M&E release and containment response in the AOR vessel/core inlet temperature of 506.6°F and RCS pressure of 2100 psia.
(b)
Please explain why subtracting (instead of adding) uncertainty from the realistic value would give a conservative input value of 514.6°F for the vessel/core inlet temperature and would give conservative results for the M&E release.
SCVB RAI-2 Reference 1, Enclosure 6, Section 5.4.2.2:
(a)
Describe the most limiting LOCA break in the AOR from the containment response standpoint. Explain why the most limiting break in the AOR would also be the most limiting break for the proposed RCS normal operating pressure/normal operating temperature (NOP/NOT) conditions.
Enclosure (b)
The UFSAR Section 14.3.4.1.3.1.3 input assumption 7 states that the air recirculation fan is effective 132 seconds after the high-1 containment pressure bistable signal is actuated. Please explain the basis for changing this time to 300 seconds.
(c)
Please explain the basis for the containment spray actuation time of 315 seconds in the proposed evaluation. What is the containment spray actuation time in the AOR? In case the AOR spray actuation time is different from 315 seconds, please justify that the change is less conservative.
SCVB RAI-3 Reference 1, Enclosure 6, Section 5.4.2.4 states, in part:
For the containment integrity analysis, this was completed by evaluating the effects of increased delay times for CTS actuation and containment air recirculation fan actuation on the LOCA containment integrity analysis.
(a)
Explain the method of evaluation described in the above statement.
(b)
If the subject evaluation/analysis methodology is different from the currently used methodology, provide justification.
(c)
As per the UFSAR Section 14.3.4.1.3.1.3. the AOR input assumption for the initial containment air temperature is 56°F in the upper and 60"F in the lower volume. Since it is not stated that these initial conditions were revised, the NRC staff assumes that the proposed analysis used the same initial temperatures. Please provide justification that these initial temperatures are conservative and would maximize the peak containment pressure and temperature transients compared to the transient obtained by using the initial technical specification (TS) 3.6.5 maximum value of 1 OOoF in the upper volume and 120"F in the lower volume.
SCVB RAI-4 Reference 1, Enclosure 6, Section 5.4.2.5:
(a)
The AOR vessel/core inlet temperature shown in the table is 552.5°F. Please explain why this is different from the AOR vessel/core inlet temperature of 506.6°F stated in Section 5.4.1.5.
(b)
Please explain why the RCS pressure of 2100 psia in the AOR and 2250 psia (2317 psia including uncertainty) in the proposed change in NOP/NOT are not included for comparison as key input parameters.
(c)
Please justify why the AOR is bounding for both M&E release and containment response even though the proposed RCS pressure of 2317 psia (uncertainty included) is significantly greater than the AOR RCS pressure of 2100 psia.
(d)
Aside from the key parameters stated in the table, what are other input parameters for long term M&E analysis that differ in the AOR and the proposed analysis? Please provide these values.
(e)
Aside from the key parameters stated in the table, what are other input parameters, for long term containment gas temperature response for equipment environmental qualification that differ in the AOR and the proposed analysis? Please provide these values.
(f)
Please explain why the NOPINOT core stored energy value is less than the AOR core stored energy value, even though the average proposed RCS temperature of 571QF is greater than the average RCS AOR temperature of 556°F.
SCVB RAI-5 Reference 1, Enclosure 6, Section 5.4.2.6 states, in part:
The evaluation of the long-term LOCA M&E and peak containment pressure is predicated upon the continued application of the operability assessment supporting NSAL-11-5 (Reference 3), in conjunction with the AOR.
(a)
Is the AOR based on corrected M&E release, resolving the issues identified in NSAL-11-5 (Reference 2)?
(b)
Describe the operability assessment supporting NSAL-11-5 (Reference 2), including the impact of the corrected M&E release (due to the issues identified in NSAL-11-5), on the containment response A OR.
SCVB RAI-6 Reference 1, Enclosure 6, Section 5.5.1.4:
Explain the basis for selecting the break sizes 1.4 tf, 0.865 ft2, 0.857 ft2, 0.834 ft2, 0.808 ft2, and 1.0 ft2 and their corresponding power levels at which the M&E release analysis is performed.
SCVB RAI-7 Reference 1, Enclosure 6, Section 5.5.1.5 states, in part*.
All of the analyzed breaks conservatively assumed dry saturated steam releases (no entrainment) except the full [double ended rupture] DER at 0 percent initial power. As a result, the small DER with dry saturated steam release was analyzed at 0 percent power, represented by a 1.0 ft2 break (smaller than the area of a single integral flow restrictor) from the faulted-loop SG and a 1.0 ft2 break for the reverse-flow blowdown from the intact-loop SGs.
Please explain why entrainment was assumed in the full DER at zero percent initial power, and explain why, as a result, the small OER with dry saturated steam release was analyzed at zero-percent power.
SCVB RAI-8 Reference 1, Enclosure 6, Table 5.5.1-1:
Explain what is meant by the title of the fourth column "Rod Motion (sec)". Describe its method of calculation, and how it affects the M&E release.
SCVB RAI-9 Reference 1, Enclosure 6, Section 5.5.2.2:
(a)
The input assumption in the seventh bullet states the containment upper volume initial temperature as srF. By a sensitivity analysis, please justify that this initial condition is conservative and would maximize the peak containment temperature transient compared to the transient obtained by using the initial TS maximum value of 120°F.
(b)
By sensitivity analysis, please justify how using initial containment upper volume temperature of 5r'F would maximize the peak containment pressure transient compared to the transient using the initial TS maximum value of 120"F.
SCVB RAI-10 Please describe the impact of the changes in M&E release during LOCA and main steam line break (MSLB) accident to the following containment analyses:
(a)
Sump water temperature response (b)
Net Positive Suction Head (NPSH) analysis for Emergency Core Cooling System (ECCS) and the containment heat removal pumps that draw suction from the containment sump in the post-accident recirculation mode.
SCVB RAI-11 Considering the changes (due to change in NOP/NOT) in the M&E release during LOCA and MSLB accident, please provide the following information for NRC staff review. Refer to SECY 0014 for more information.
(a)
The value of the required NPSH (i.e., NPSHR) for the ECCS and the containment heat removal pumps that draw suction from the containment sump in the post-accident recirculation mode.
(b)
The basis for the NPSHR, including the standard on which it is based. As an example, the NPSHR for ECCS and containment heat removal system pumps is commonly based on the Hydraulic Institute standard, to which the NPSHR is equal to the available NPSH determined in a factory test at the pump design flow with a three-percent drop in the total dynamic head.
(c)
The uncertainty in the factory tested value of NPSHR based on the actual site conditions.
(d)
The minimum NPSH available in the proposed analysis at each pump inlet based on maximizing the sump water temperature along with maximizing the suction strainer head loss based on Generic Safety Issue 191 resolution and maximizing piping head loss.
(e)
The minimum NPSH margin for each pump and its percentage, based on NPSHR with uncertainty added.
SCVS RAI-12 NUREG~OBOO, Standard Review Plan 6.2.1.5 describes the minimum containment pressure analysis for the ECCS performance capability. Regulatory Guide (RG) 1.157, Section 3.12.1 provides guidance for calculating the containment pressure response used for evaluating cooling effectiveness during the post-blowdown phase of a LOCA. The RG states that the containment pressure should be calculated by including the effects of containment heat sinks and operation of all pressure-reducing equipment assumed to be available. Using the above guidance please describe the impact of the changes in M&E input on the minimum containment pressure analyses for ECCS performance during a LOCA and MSLB accident.
Radiation Protection and Consequence Branch {ARCBl ARCS RAI-1 On page 1 of Enclosure 8 to the October 8, 2013, application, Section 2. 0 lists the proposed revised parameters for the proposed amendment. Please provide additional information regarding the effect these revisions will have on each of the radiological design basis accident analyses.
(a) For the revised analyses, were any changes made to the methodologies that are in the current analyses of record?
(b) What are the revised calculated dose values for the Exclusion Area Boundary, Low Population Zone, and Control Room?
ARCS RAI-2 In Section 3.0 of Enclosure 8 to the October 8, 2013, application, it states that the proposed revised parameters will affect the previous maximum and minimum RCS liquid mass values. It also states that these values are used in some of the dose consequence analyses.
(a) What are the revised values for the maximum and minimum RCS liquid mass?
(b) Which dose consequence analyses are being referred to in the above statement and how does that affect the resulting dose values for those analyses?
Electrical Engineering Branch fEEEB)
Background
The licensee proposes to implement a return to RCS NOP/NOT conditions for CNP Unit 1 by increasing the current operating nominal full-power pressurizer pressure from 2100 psia to 2250 psia and increasing the current operating nominal full-power T avg from 556"F to 571"F.
Implementation of the program is proposed to occur prior to GNP Unit 1 Cycle 26 startup (October 2014).
The licensee states that the proposed change also revises the Containment Air Recirculation/Hydrogen Skimmer (CEQ) fan start time from "108 seconds< CEQ fan delay<
132 seconds" to "270 seconds < CEQ fan delay < 300 seconds" and revise containment spray system (CTS) actuation time delay from 115 seconds to 315 seconds.
The CTS design at GNP includes a time delay relay in the CTS pump start circuitry that is presently used to properly sequence the pump onto the emergency diesel generator (EDG) bus and prevent overloading of the diesel. To offset the adverse effects of the proposed increase in full power average RCS temperature on best estimate loss-of-coolant accident peak cladding temperature (BELOCA-PCT), the setting of this time delay relay is increased to further delay CTS actuation following a Hi-Hi containment pressure signal. The new time delay setting continues to support proper EDG bus loading, but also results in a higher containment pressure during large break loss of coolant accident RCS blowdown, which limits the rate of RCS mass release to containment and improves BELOCA-PCT results.
In order for the NRC staff to verify that there are no adverse effects on the EDGs and no environmental qualification changes of electrical equipment due to the proposed changes, please provide the following information:
EEEB RAI-1 Provide the loading sequence for each EDG at GNP Unit 1. In your response, please describe the changes that have been made to the EDG loading sequence and any changes in motor loads as a consequence of higher containment pressure.
EEEB RAI-2 Please provide a summary of your analysis and tests performed to show that the EDGs will continue to perform within the voltage and frequency requirements during the load sequencing and steady state operation after implementing the delayed CTS pump start. Please confirm whether the proposed delay in CTS pump start can result in overlapping of loads during EDG load sequencing if the containment Hi-Hi signal is generated after the permissive from the sequencer timer during LOCAs (for any break sizes).
EEEB RAI-3 Updated Final Safety Analysis Report Section 8.1.2 states:
The engineered safety feature (ESF) Loads are sequenced onto the Reserve Auxiliary Transformers (RATs), under accident conditions, using the same timing relays and sequence as used for the EDG sequencing. The RATs supply the reserve auxiliary power for both units. Under certain plant conditions and grid loading conditions, and with proper precautions and limitations, it is possible for either transformer No. 4 or transformer No. 5 to feed the entire plant auxiliary load.
Please provide a summary of your analysis to show that the RATs can provide the required voltage to ESF loads without actuating protective devices and meet the accident analysis assumptions.
EEEB RAI-4 to AEP-NRC-2013-79, "UFSAR Section 6.3.2 Pages Marked To Show Proposed Changes, Containment Spray Systems, System Design," paragraph A states:
Delaying actuation of the Containment Spray Pumps for a period of time following a containment pressure Hi-Hi signal, for example, has the effect of allowing a higher containment backpressure during the initial phases of a loss of coolant accident, which provides a benefit to the Peak Clad Temperature analysis.
Please explain the impact of additional loading on the RATs and EDGs and potential changes in the fuel oil consumption as a result of the changes in backpressure that ECCS loads could be subjected to during loss of coolant accidents.
EEEB RAI-5 Please provide a discussion of the impact (e.g., changes to environmental qualification parameters such as temperature and pressure) on Environmental Qualification (EQ) of equipment affected by the proposed change. To demonstrate that the equipment remained qualified for service in the revised pressure and temperature environment, please provide a comparison of the EO overlays with containment temperature and pressure profiles including how margins are being maintained for qualified equipment in accordance with Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50.49.
Nuclear Performance and Code Review Branch ($NP8)
SNPB RAI-1 Table 2.1-1 of WCAP-17762-NP contains nuclear steam supply system (NSSS) design parameters which, as described in Section 2.1.2, "are used as the basis for the design transients and for the systems, structures, components, accidents and fuel analyses and evaluations." Each of the eight cases listed in Table 2.1-1 has a different steam outlet pressure
"8" listed, ranging from 618 to 851 psia.
The description of the NSSS design transient evaluations in Section 3.1, however, states that, consistent with the measurement uncertainty recapture uprate license amendment issued by the NRC on December 20, 2002 (ADAMS Accession No. ML023470126), "the full power steam pressure will continue to be limited to a minimum of 679 psia (administratively limited to 690 psia for conservatism)." These limits are higher than the steam outlet pressures listed in Table 2.1-1 for Cases 1, 2, and 6.
Please clarify how the limits from Section 3.1 listed above interact with the design parameters from Table 2.1-1 for the design transient evaluations.
SNPB RAI-2 Section 5.1.1 of WCAP-17762-NP describes the best-estimate large break loss-of-coolant accident (BE LBLOCA) analysis for the DC Cook Unit 1 return to NOP/NOT program. This analysis estimates the impact of fuel thermal conductivity degradation (TCD) by using a method found acceptable by NRC staff in a letter dated March 7, 2013 (ADAMS Accession No. ML13077A137). In this method, margin to the 10 CFR 50.46(b)(1) peak cladding temperature limit of 2200°F is recaptured by adjusting input parameters, including accounting for peaking factor burndown in the fuel.
Table 1 of the previous TCD evaluation for DC Cook Unit 1, from March 19, 2012 (ADAMS Accession No. ML12088A104), included burnup-dependent limits for both heat flux hot channel factor, F0, and the enthalpy rise hot channel factor, F *H* In WCAP-17762-NP, Table 5.1.1-1 presents the Fa and FaH burndown as a function of rod burnup used in the LBLOCA analysis.
The AOR (available in ADAMS at Accession No. ML080090268) accounts for neither TCD nor peaking factor burndown, and provides single limits for F0 and F ll.H*
If peaking factor burndown is required to recapture margin to the PCT limit of 2200°F for the return to NOP/NOT analysis, how will this be addressed in the CNP Unit 1 Core Operating Limits Report and/or technical specifications?
SNPB RAI-3 Sections 6.2.3 and 6.2.4 of WCAP-17762-NP discuss the evaluation of fuel performance at the CNP Unit 1 return to NOP/NOT conditions. Section 6.2.3 provides the acceptance criteria for the fuel, and Section 6.2.4 provides the results of the evaluations with respect to these criteria.
(a) It is stated in Section 6.2.4 that u[n]o explicit PAD calculations were used to evaluate the fuel rod design criteria at NOP/NOT conditions." Discussion in several of the evaluations notes that the effects of returning to NOP/NOT at GNP Unit 1 will be offset by available margin.
Was a quantitative margin assessment performed?
(b) It is also noted in Section 6.2.4 that "[t[he PAD code with USNRC-approved models for in-reactor behavior is used to calculate the fuel rod performance over its irradiation history."
The NRC-approved current version of the PAD code, PAD 4.0, does not include approved method for modeling TCD.
In evaluating the fuel acceptance criteria for the return to NOP/NOT conditions, was consideration given to the effects of TCD on fuel performance? If a quantitative margin assessment was performed per SNPB-RAI-3(a), did this margin assessment include the effects ofTCD?
REFERENCES
- 1.
Letter from l&M to NRC dated October 8, 2013, "Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315 License Amendment Request Regarding Restoration of Normal Reactor Coolant System Operating Pressure and Temperature Consistent With Previously Licensed Conditions" (ADAMS Accession No. ML t 3283A 121 ).
- 2.
NSAL-11-05, "Westinghouse LOCA Mass and Energy Release Calculation Issues,"
July 26, 2011 (ADAMS Accession No. ML13239A479).
Please feel free to contact me at (301) 415-4037 if you need any further clarification of the questions in the enclosure.
Docket No. 50-315
Enclosure:
Sincerely, IRA!
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Request for Additional Information cc: Distribution via ListServ DISTRIBUTION.
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NAME TVVengert MHenderson UShoop DATE 05/6/14 04/30/14 03/05/14 OFFICE DSS/SCVB/BC DE/EEEB/BC LPL3-1/BC NAME RDennig (JBettfe for)
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DATE 03/11/14 03/07/14 05/6/14 OFFICIAL RECORD COPY
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JDean 02/18/14 LPL3-1/PM TVVengert 05/6/14