ML121220306
| ML121220306 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/30/2012 |
| From: | John Stanley Constellation Energy Nuclear Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML121220306 (11) | |
Text
Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 CENG.
a joint venture of CALVERT CLIFFS NUCLEAR POWER PLANT April 30, 2012 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317 10 CFR 50.46 30-day Report and Annual Report for Changes to the Emergency Core Cooling System Performance Analysis
REFERENCES:
(a) Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP),
dated February 18, 2011, Amendment re: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (TAC Nos. ME2831 and ME2832)
(b) Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC),
dated March 31, 2011, 10 CFR 50.46 30-day Report for Changes to the Emergency Core Cooling System Performance Analysis (c) Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC),
dated January 19, 2012, 10 CFR 50.46 30-day Report for Changes to the Emergency Core Cooling System Performance Analysis This letter is submitted pursuant to 10 CFR 50.46(a)(3)(ii) to provide notification of a significant change to the peak cladding temperature analysis result for the large break loss-of-coolant accident (LB LOCA) and the small break loss-of-coolant accident (SB LOCA) analyses. Because the effect on the peak cladding temperature of the changes is greater than 50'F from the temperature calculated for the limiting transient using the last acceptable model, the analysis changes qualify as significant as defined in 10 CFR 50.46(a)(3)(i) and, consequently, are provided in Attachment (1).
The analyses for the LB LOCA and SB LOCA Emergency Core Cooling System performance have been re-performed for Unit 1, Cycle 21. The analyses were performed using the latest Nuclear Regulatory Commission accepted versions of the AREVA evaluation models for Calvert Cliffs pressurized water reactors [Reference (a)]. The new analyses explicitly model the AREVA fuel introduced into Unit 1, Cycle 21 in the spring of 2012. Reference (a) documents the Nuclear Regulatory Commission's approval VI-
Document Control Desk April 30, 2012 Page 2 of the use of AREVA methods for reactor core designs that include both AREVA and Westinghouse fuel assemblies and reactor core designs that include only AREVA fuel assemblies for both Calvert Cliffs Units 1 and 2.
The results of the new LB LOCA and SB LOCA analyses conform to the Emergency Core Cooling System acceptance criteria of 10 CFR 50.46(b) and are discussed in Reference (a). With the implementation of the approved license amendment on Unit 1, the new LB LOCA and SB LOCA analyses constitute the new licensing basis for Unit I on April 6, 2012 when Unit 1 entered Mode 4.
Attachment (1) contains the results of the change in the peak cladding temperature based on the change in vendor methodology used for the analyses for Unit 1.
The analyses presented in Reference (a) that are now being implemented on Unit I were performed for both Units 1 and 2. A 10 CFR 50.46 30-day letter (Reference b) was previously submitted to the Nuclear Regulatory Commission to document a > 50'F change in peak clad temperature for Unit 2 due to the change in analyses.
In addition, AREVA has reported changes to or errors in the acceptable evaluation models for calendar year 2011. These changes are required to be reported within 30 days of implementation of the new licensing basis for Unit 1 because the magnitude of the peak cladding temperature change for the SB LOCA analysis exceeds 50'F. This annual report is contained in Attachment (2). These changes were previously reported for Unit 2 in Reference (c).
Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.
Very truly yours,
>~ for James J. Stanley Manager-Engineering Services JJS/PSF/bjd Attachments: (1) 10 CFR 50.46- 30 Day Report (2) 10 CFR 50.46 - Annual Report cc: N. S. Morgan, NRC Resident Inspector, NRC W. M. Dean, NRC S. Gray, DNR
ATTACHMENT (1) 10 CFR 50.46 - 30 DAY REPORT Calvert Cliffs Nuclear Power Plant, LLC April 30, 2012
ATTACHMENT (1) 10 CFR 50.46 - 30 DAY REPORT INTRODUCTION This letter is submitted pursuant to 10 CFR 50.46(a)(3)(ii) to provide notification of a significant change to the peak cladding temperature analysis result for the Unit I large break loss-of-coolant accident (LB LOCA) and the Unit 1 small break loss-of-coolant accident (SB LOCA) analyses. Because the effect on the peak cladding temperature of the changes is greater than 50'F from the temperature calculated for the limiting transients using the last acceptable model, the analyses changes qualify as significant as defined in 10 CFR 50.46(a)(3)(i) and, consequently, are provided below.
Calvert Cliffs request to amend the Unit I renewed operating licenses to transition from Westinghouse to AREVA-designed fuel was approved by the Nuclear Regulatory Commission (NRC) in Reference 1. As part of that transition, the Emergency Core Cooling System (ECCS) performance for both the LB LOCA and the SB LOCA were re-analyzed. The analyses were performed with the latest NRC accepted versions of EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Rectors," and EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based." The analyses comply with the limitations and constraints imposed by Reference 1.
Note that these new LB LOCA and SB LOCA analyses are not assessments (i.e., they do not provide an estimate of the effect of the changes on the limiting ECCS analysis). Rather, they are complete re-analyses that use acceptable evaluation models that are applicable to Calvert Cliffs Unit I and Unit 2. A summary of the new analysis and their compliance with 10 CFR 50.46 is provided below.
The analyses described in Reference (1) that were used to amend the licenses were performed prior to calendar year 2011. A 10 CFR 50.46 30-day letter (Reference 2) was submitted to the NRC to document a > 50'F change in peak clad temperature (PCT) for Unit 2 due to this change in vendor methodologies.
REFERENCE ANALYSES LB LOCA The Unit 1 LB LOCA ECCS performance analysis was performed with the AREVA evaluation method approved in Reference 1. The analysis included the implementation of the M5 cladding material properties into the LB LOCA methodology.
The analysis resulted in an absolute change in PCT from the prior Westinghouse analysis of record of greater than 50'F. A comparison of PCT results is provided in Table 1. Note, the change in PCT is due to a switch in fuel vendors and is not the result of a calculation or methodology error.
Table 1, AREVA versus Westinghouse LB LOCA PCT Analysis Results Item PCT, -F New Analysis of Record (includes AREVA fuel) 1,670 Old Analysis of Record (Westinghouse fuel) 2,057 Table 2 provides the results of the new LB LOCA analysis demonstrating conformance with the acceptance criteria of 10 CFR 50.46(b).
1
ATTACHMENT (1) 10 CFR 50.46 - 30 DAY REPORT Table 2, LB LOCA versus Acceptance Criteria Parameter Criterion Result Peak Cladding Temperature, 'F 2200 1670 Maximum Cladding Oxidation, % <17 0.907 Maximum Core-Wide Cladding Oxidation, % <1 0.011 Coolable Geometry Yes Yes SB LOCA AREVA performed a complete SB LOCA analysis for Calvert Cliffs. The analysis was performed with the AREVA evaluation method approved in Reference 1.
The analysis resulted in an absolute change in PCT from the prior Westinghouse analysis of record of greater than 50'F. A comparison of PCT results is provided in Table 3. Note, the change in PCT is due to a switch in fuel vendors and is not the result of a calculation or methodology error.
Table 3, AREVA versus Westinghouse SB LOCA PCT Analysis Results Item PCT, 'F AREVA Fuel Analysis of Record 1,626 Westinghouse Fuel Analysis of Record 1,855 Table 4 provides the results of the new SB LOCA analysis demonstrating conformance with the acceptance criteria of 10 CFR 50.46(b).
Table 4, SB LOCA versus Acceptance Criteria Parameter Criterion Result Peak Cladding Temperature, 'F 2200 1626 Maximum Cladding Oxidation, % <17 1.769 Maximum Core-Wide Cladding Oxidation, % <1 0.015 Coolable Geometry Yes Yes
SUMMARY
The new LB LOCA and SB LOCA analyses constitute new licensing basis analyses (analyses-of-record) for Calvert Cliffs Unit 1. They are used as the reference analyses to evaluate the impact on peak cladding temperature of changes to or errors in the AREVA LB LOCA and SB LOCA evaluation models and their application to Calvert Cliffs.
REFERENCES
- 1. Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP), dated February 18, 2011, Amendment re: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel
- 2. Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC), dated March 31, 2011, 10 CFR 50.46 30-day Report for Changes to the Emergency Core Cooling System Performance Analysis 2
ATTACHMENT (2) 10 CFR 50.46 - ANNUAL REPORT Calvert Cliffs Nuclear Power Plant, LLC April 30, 2012
FAB12-174 Page A-3 Attachment to FAB12-174 Calvert Cliffs Unit 1 and Unit 2 NUCLEAR PLANTs - 10CFR50.46 ANNUAL REPORT -
EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL CHANGES In accordance with the annual reporting requirements of 10CFR50.46 (a)(3)(ii), the following is a summary of the limiting design basis accident (loss-of-coolant accident) analysis results established using the current Calvert Cliffs ECCS evaluation model.
Table 1 Calvert Cliffs Realistic Large Break LOCA PCT PCT Section Licensing Basis PCT (September 2009) 1670 OF RLBLOCA & S-RELAP5 - FIJ Multiplier
& Underpredicting Liquid Entrained to 0 OF 1 SG Tubes RLBLOCA Upper Plenum Modeling 0 OF 2 S-RELAP5 Sleicher-Rouse correlation +8 OF 3 Updated Licensing'Basis PCT 1678 OF Net Change +8 OF
FAB12-174 Page A-4 Table 2 Calvert Cliffs Small Break LOCA PCT PCT Attachment Licensing Basis PCT (September 2009) 1626 OF S-RELAP5 Sleicher-Rouse correlation +69 OF 3 Updated Licensing Basis PCT 1695 OF Net Change +69 OF A detailed discussion of the ECCS evaluation model change outlined above is attached to this memorandum.
This submittal is based on information provided by AREVA, Inc. in Letter Numbers FAB10-484, FAB11-226, and FAB12-39.
FAB12-174 Page A-5 Section 1 STEAM GENERATOR TUBE LIQUID ENTRAINMENT MODELING ERROR
Background
The AREVA Realistic Large Break Loss-of-Coolant Accident (RLBLOCA) analysis methodology used in the Calvert Cliffs large break LOCA analysis of record uses a bias on interphase friction at the steam generator tube sheet entrance to establish liquid entrainment into the steam generator tubes during a large primary system pipe break.
The bias is established by comparing calculated results from the S-RELAP5 computer code with established test data (i.e., Upper Plenum Test Facility (UPTF) Tests 10 and 29). The UPTF test facility contains a full scale, four loop pressurized water reactor test assembly complete with the hardware required to represent geometry specific phenomena occurring during a large or small break LOCA. The S-RELAP5 input parameter controlling steam generator liquid entrainment is interphase friction. The range of interphase friction spans several orders of magnitude between the flow regimes occurring in the hot leg, hot leg riser, steam generator inlet plenum and steam generator tube sheet. As a quantitative determination of the uncertainty in interphase friction is not feasible, a bias is applied to establish a conservatively bounding value.
The magnitude of the bias is determined by adjusting the S-RELAP5 RLBLOCA multiplier "FIJ" until the S-RELAP5 code conservatively over-predicts the entrainment observed in UPTF Tests 10 and 29. Upon review, the original FIJ multiplier of 1.75 used in the Calvert Cliffs analysis does not meet the bias criteria described above and under-predicts steam generator liquid entrainment.
Results A re-evaluation of steam generator liquid entrainment using the S-RELAP5 code and the criteria discussed above established the appropriate value for the "FIJ" multiplier to be 5.0 for the Calvert Cliffs configuration. AREVA has updated the input parameter accordingly. Based on sensitivity studies, AREVA has bounded the effect of the steam generator liquid entrainment under-prediction on the Calvert Cliffs emergency core cooling system evaluation model results.
There is no need to adjust the interphase friction models used in the analysis of SBLOCAs, since the flow phenomena in SBLOCA is completely different from that in LBLOCA. Models and benchmarks for SBLOCA applications are completely independent of those used for LBLOCA analyses. No errors have been noted by AREVA in the simulation of counter-current flow occurring in the progression of a SBLOCA.
A penalty of 01F has been applied to the RLBLOCA limiting PCT to address the effect of the error on the Calvert Cliffs analysis.
FAB12-174 Page A-6 Section 2 Liquid fallback from the Upper Plenum to the Hot Channel Back-ground The impact of liquid and vapor flow spikes from the upper plenum (UP) into the hot channel (HC) and surrounding six assembly regions of the core and a non-physical flow pattern in the upper plenum was evaluated for the RLBLOCA methodology. Even though CCFL modeling was applied at the HC exit junction, it will not be activated due to the negative spikes in steam velocities (from upper plenum to HC).
Results The current RLBLOCA reactor vessel modeling was traced back to the EMF-2103 sample problem for a 3-loop VV plant. This W 3-loop plant has a geometry feature in the upper plenum know as "flow mixers or standpipes". Due to this geometry feature, the upper plenum was broken into two sections, one to an open hole region and one to a flow mixer region. The modeling in the sample problem blocked the cross flow between radial junctions in the first level of upper plenum and this was carried through in plants without flow mixers as a methodology conservatism.
The UP nodalization for these plant cases was revised to make it consistent with the geometry. In addition, in all plant cases, a high reverse loss coefficient is applied to the HC and central core to UP junctions at the beginning of the core reflooding phase.
Cases were rerun that had liquid down flow and potentially affect the AOR PCT limit.
The PCT impact for Calvert Cliffs Units 1 and 2 is 0 F.
FAB12-174 Page A-7 Section 3 S-RELAP5 Code Programming of Sleicher-Rouse correlation
Background
Sleicher-Rouse is one of the correlations used to define the heat transfer between the fuel and coolant. This correlation is applicable to both Large and Small Break analyses performed with the S-RELAP5 computer code.
During development of a BWR LOCA methodology based on S-RELAP5, the behavior of the Sleicher-Rouse correlation relative to other single-phase vapor heat transfer correlations was reviewed and it was questioned whether Sleicher-Rouse correlation was correct. It was also discovered that another industry code uses the Sleicher-Rouse correlation, but the form for the correlation is different than that used in S-RELAP5 implementation of Sleicher-Rouse. The concern is related to the form of the equation for calculating the exponent of the temperature ratio correction term.
The S-RELAP5 form is n = -logio(Tw/Tg)1 4 + 0.3 The alternate form used in another industry code is n = -[Iogio(Tw/Tg)]11 4 + 0.3 The alternate form appears to be more consistent with other heat transfer correlations and expected physical trends.
Results Preliminary assessments of the potential impact of using the alternate Sleicher-Rouse correlation form were performed. A development version of S-RELAP5 was prepared with the alternate Sleicher-Rouse form and several code validation and plant sample problems were repeated. Additional analyses were performed using a different heat transfer correlation for single-phase vapor heat transfer. The assessments included analyses for both RLBLOCA and SBLOCA.
The PCT impact for Calvert Cliffs Units 1 and 2 RLBLOCA is +8 F. The PCT impact for Calvert Cliffs Units 1 and 2 SBLOCA is +69 F.