3F0199-02, Submits Change in Analysis of Record for SBLOCA & 10CFR50.46 Notification.Attachment to Ltr Includes Description of How Limitations & Conditions Outlined in RELAP5 SER Are Addressed
| ML20206P710 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/07/1999 |
| From: | Roderick D FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F0199-02, 3F199-2, NUDOCS 9901130134 | |
| Download: ML20206P710 (11) | |
Text
v y
Flrrida Power l
E?.2T,"
l 8r. = :: a o -
January 7,1999 3F0199-02 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Change in Analysis of Record for Small Break Loss of Coolant Accident and 10CFR50.46 Notification
Reference:
Letter from NRC to Framatome Technologies Incorporated," Acceptance for Referencing of Topical Report BAW-10192-P, BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants,"
dated February 18,1997
Dear Sir:
This submittal is made in accordance with Title 10 Code of Federal Regulations Part 50.46 (10CFR50.46), Section (a)(3), which requires that the NRC be notified of a change to an acceptable evaluation model (EM) that results in a change in peak clad temperature (PCT) greater than 50 F.
Florida Power Corporation (FPC) is replacing the CRAFT 2 EM with RELAP5/ MOD 2-B&W (RELAP5) EM as the code of record for the Crystal River Unit 3 (CR-3)
Small Break Loss-of-Coolant Accident (SBLOCA) analysis. The transition from CRAFT 2 to RELAP5 involves a change (decrease) in PCT in excess of 50 F. The calculation establishing
//
RELAP5 as the analytical code of record for SBLOCA was implemented at CR-3 on December
//
10, 1998.' 10CFR50.46 Section (a)(3)(ii) requires that this condition be reported to the NRC within 30 days of the change.
In the referenced letter, the NRC issued a generic Safety Evaluation Report for the application of,
RELAP5-based EM as described in topical report BAW-10192-P-A. The attachment to this letter includes a description of how the limitations and conditions outlined in the RELAP5 Safety Evaluation Report are addressed. The conclusions reached by the analysis are as follows:
1.
RELAPS is acceptable for use in CR-3 specific SBLOCA applications.
- 2. Compared to CRAFT 2, the use of RELAPS results in reduced PCTs for SBLOCAs for the existing CR-3 plant configuration.
h.,yj i
9901130134 990107 l
PDR ADOCK 05000302 P
PDR M
CRYSTAL RWER ENERGY COMPLEX: 16760 W. Power Line Street
q l
U.S. Nuclear Regulatory Commission 3F0199-02 Page 2 of 2 The.attachinent to this letter also provides information that demonstrates continued compliance with all requirements of 10CFR50.46. Therefore, no further actions are required by FPC. In
. addition, no NRC action is requested.
If you have any questions regarding this submittal, please contact Mr. Sid Powell, Manager, Nuclear Licensing at (352) 563-4883.
- incerely, 1
Daniel L. Roderick Director Nuclear Engineering and Projects DLR/pei Attachment xc:
Regional Administrator, Region II Senior Resident Inspector j
NRR Project Manager l
l
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 l
l i
i 1
ATTACHMENT i
1 1
l I
i a
s 5
U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 1 of 8 l.
Introduction A.
Purpose and Overview The purpose of this attachment is to provide a summary of the technical details for CR-3's transition from a CRAFT 2 based evaluation model (EM) to RELAP5/ MOD 2-B&W (RELAP5) based EM for Small Break Loss-of-Coolant Accident (SBLOCA) analyses.
The primary emphasis of this attachment is to demonstrate that the CR-3 specific application of RELAP5 for SBLOCA analyses meets the limitations and conditions of the Safety Evaluation Report (SER) for the RELAP5 EM. A comparison of the CRAFT 2 and RELAP5 peak clad temperatures (PCTs) and other Title 10 Code of Federal Regulations Section 50.46 (10CFR50.46) criteria are also provided for the limiting SBLOCA scenario (Cold Leg Pump j
Discharge break). The final section summarizes the conclusions from the analytical effort that support the change of EM. A list of References is given at the end of this attachment.
B.
Background
The function of the Emergency Core Cooling System (ECCS) is to protect the core in the event of a Loss-of-Coolant Accident (LOCA).10CFR50.46 requires that the evaluation of ECCS l
performance for a commercial nuclear power plant must meet the following criteria-1.
The calculated PCTs are less than 2200 F.
2.
The maximum calculated local cladding oxidation is less than 17%
3.
The maximum calculated core-wide oxidation does not exceed 1% of the fuel cladding.
4.
The cladding remains amenable to cooling.
5.
Long-term cooling must be established and maintained after the LOCA.
The first four criteria are demonstrated by analytical methodology, while the last criterion is demonstrated by the combination of ECCS performance, equipment availability, and operational practices.
The distinction between the Small Break and Large Break Loss-of-Coolant Accident (LBLOCA) is the size of the break. In the CRAFT 2-based EM, the LBLOCAs are those breaks which are 2
greater than or equal to 0.5 ft areas, while SBLOCAs are those breaks which are smaller than 0.5 2
ft and the rate of the break floiv exceeds the makeup capability. In the RELAP5-based EM, the l
maximum SBLOCA break size is 0.75 ft. The RELAP5 EM uses a LBLOCA transition break 2
2 2
methodology between break sizes of 0.75 ft and 2.0 ft and a pure LBLOCA methodology above 2.0 112 Spectra of break sizes are then analyzed to determine the most limiting break size. This submittal focuses on the new SBLOCA analyses using the RELAP5-based EM. FPC is currently planning to transition from CRAFT 2 to RELAP5 for LBLOCA later in 1999. During the interim, the CRAFT 2-based EM will be retained as the analysis of record for breaks larger than 2
0.75 ft,
l
T-U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 2 of 8 The existing analyses were performed to demonstrate compliance with 10CFR50.46. These analyses were developed and performed by Babcock and Wilcox (B&W - now Framatome Technologies, Incorporated - FTI). The overall SBLOCA analysis methodology consists of the following process:
- Thermal-hydraulic system analysis - CRAFT 2 (Reference 4)
. A more precise core mixture height calculation during the quiescent portion of the SBLOCA event, if the thermal-hydraulic analysis predicts core uncovering -
FOAM 2 (Reference 5)
- Ilot pin heat-up calculation, if the core mixture height calculation predicts uncovering -THETA-1B (Reference 6) l l
The SBLOCA thermal-hydraulic methodology is currently based on the CRAFT 2 code, as described in Reference 3.
Using CRAFT 2, the limiting break size as identified in Section 2
14.2.2.5.7.2 of the CR-3 Final Safety Analysis Report (FSAR) is a 0.125 ft break at the bottom of the cold leg reactor coolant pump discharge piping. This case resulted in significant core i
uncovery, and the maximum PCT predicted for this case using the CRAFT 2-based approach was l
1859 F. This result met the first criterion of 10CFR50.46, and the remaining criteria pertinent to l
analytical methodology (local and core-wide oxidation, hydrogen generation, coolable core I
geometry) were also met.
The existing CRAFT 2-based analyses represent conservative and valid licensing results based on the boundary conditions analyzed. However, these analyses are being replaced with new analyses that use the RELAP5 EM, which utilizes improved calculational methods. The RELAPS EM, described in References 1 and 2, was devised to provide flexibility for addressing new fuel designs, and to accommodate component performance issues (e.g., reduced ECCS flows, steam generator tube plugging, etc.). The RELAP5 EM also includes requirements that j
are generally more restrictive than those used to license the CRAFT 2-based EM. The transition to the improved RELAP5 calculational methods and meeting the prescribed requirements for the L
use of the new EM provide assurance that CR-3's SBLOCA analyses remain valid and demonstrate continued compliance with the 10CFR50.46 criteria.
CR-3 has used the RELAPS methodology in a limited application to support an Improved Technical Specification (ITS) change when Emergency Diesel Generator loading and Emergency Feedwater challenges were recognized (Reference 7). The approval of the ITS change, as well as the limited use of RELAP5, is documented in Reference 8.
t C.
Full Application of RELAP5 for SBLOCA The purpose of this attachment is to describe CR-3's transition from a CRAFT 2-based SBLOCA methodology to a RELAP5 methodology. As noted above, the latter has already been used in a limited application, and it has been determined to be beneficial for CR-3 to fully transition to this f
more advanced methodology. It is noted that the approved use of RELAPS is subject to eleven i
~
U.S. Nuclear Regulatory Commission Attaciunent 3F0199-02 Page 3 of 8 conditions described in the SER for BAW-10192-P. Therefore, the following section addresses the limitations and conditions associated with the use of the RELAP5-based EM for SBLOCA applications.
H.
Safety Evaluation Report Limitations and Conditions j
The NRC Safety Evaluation Report (SER) on BAW-10192-P (Reference 1) contained eleven conditions related to the use of the RELAP5 EM. These conditions are classified by applicability.
The first category consists of conditions that apply to LBLOCA analyses but do not apply to SBLOCA analysis. The second category encompasses conditions that apply to both SBLOCA and LBLOCA analysis applications. The categories and their constituent conditions are listed (by the numerical scheme used in the SER) below.
A.
Conditions Not Applicable to SBLOCA Analyses These conditions specifically pertain to the RELAP5 LBLOCA EM analyses reviewed and discussed in Section 3 of the BAW-10192-P SER (Reference 1). These conditions were characterized in the SER as pertaining to LBLOCA analysis only. Therefore, these conditions do not apply to the SBLOCA analyses and are not addressed in this submittal. The condition number and the applicable SER section [in brackets] are shown below:
(3) The limiting linear heat rate (LHR) for LOCA limits is determined by the power level and the product of the axial and radial peaking factors... FTI must revalidate the acceptability of the evaluation model peaking methods if: (1) significant changes are found in the core elevation at which the minimum core LOCA margin is predicted or (2) the core maneuvering analyses radial and axial peaks that approach the LOCA LHR limits differ appreciably from those used to demonstrate Appendix K compliance [3.17].
2 2
(4) The mechanistic ECCS bypass model is acceptable for cold leg transition (0.75 ft to 2.0 ft )
and hot leg break calculations. The non-mechanistic ECCS bypass model must be used in 2
the large cold leg break (22.0 ft ) methodology since the demonstration calculations and sensitivities were run with this model [3.9].
(6) LOCA limits for three pump operation must be established for each class of plants by application of the methodology described in this report. An acceptable approach is to demonstrate that three-pump operation is bounded by four pump LHR limits [3.16].
(7) The limiting ECCS configuration, including minimum versus maximum ECCS, must be determined for each plant or class of plants using this methodology [3.18].
(11) B&W-designed plants have internal reactor vessel vent valves (RVVVs) that provide a path
(
for core steam venting directly to the cold legs. The BWNT LOCA evaluation model credits the RVVV steam flow with the loop steam venting for LBLOCA analyses.. This l
demonstration should be performed at least once for each plant type (raised loop and lowered loop) and be judged applicable for all LBLOCA break sizes [3.1].
1 U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 4 of 8 l
IL -
Conditions applicable to LBLOCA and SBLOCA Analyses The following conditions pertain to the RELAP5 LBLOCA and SBLOCA EM analyses reviewed and discussed in Section 5 of the BAW-10192-P SER (Reference 1) and BAW-10164 (Reference 4). These conditions are common to the methods used to analyze both LBLOCAs and SBLOCAs. Therefore these conditions are applicable to this submittal. Brief summaries of the evaluations of these criteria are included below:
(1) The LOCA methodology should include any NRC restrictions placed on the individual codes used in the evaluation model (EM). In this case, the specific code is RELAP5, which is described by topical report BAW-10164 (Reference 4).
Response: The aspects of the EM that warranted conditions include:
. Film boiling correlations i
The condition pertains to the use of unreviewed film boiling heat transfer correlations included in the code as an option. The Chen-Sundaram-Ozkaynak (CSO) core heat transfer correlation, which has not yet been reviewed, was not used. CR-3 verified that the analysis used the Condie-Bengston IV correlation, which has been reviewed (BAW-10164) for use in EM applications.
. Use of B&W auxiliary feedwater model for Once-Through Steam Generators This condition pertains to the use of an unreviewed modeling option. The B&W auxiliary feedwater model for OTSGs was reviewed in Revision 3 of BAW-10164 and subsequently received approval. Therefore, this model was used in the analysis.
. Pre-rupture cladding swell modeling for pre-rupture strain This condition pertains to the effect of fuel pre-rupture clad swelling on flow distribution.
The maximum pre-rupture cladding strain is less than 20% of the rupture strain, therefore, there are no significant flow effects.
. Decay heat standard This condition pertains to the use of a standard other than te ANS 1971 standard prescribed by Appendix K. The decay heat standard used, ANS 193 is consistent with l
the ANS 1971 standard, and is augmented by a 1.2 multiplier on the decay power I
consistent with the requirements of Appendix K.
l U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 5 of 8
+ Use of static properties for input to critical flow models (Extended IIenry-Fauske and Moody)
This condition pertains to the use of static properties in the look-up tables containing the Extended 11enry-Fauske and Moody correlation data in the code. A verification of the use of static properties as input to critical flow tables was performed.
- Loop seal clearing This condition pertains to the use of the interphase drag model implemented in REl.AP5, which tends to overpredict the clearing of a loop seal by available steam flow. Unlike some other designs, B&W plants include Reactor Vessel Vent Valves (RVVVs) that provide a direct core steam venting path to the cold leg break locations, precluding the formation of, and obviating the need for clearing, a loop seal.
- Non-condensible gas impact on safety injection flow This condition pertains to the potential for non-condensible gases to impede ECCS performance. This is generally a concem for recirculating (U-tube) steam generator designs, which may rely upon reflux cooling, wherein the steam generators may be relied upon as a heat sink during the accident. Ilowever, the analysis showed that the amount of non-condensibles generated would be small for B&W reactors, and would not impede ECCS performance. Moreover, because of the RVVV feature inherent in B&W design, such gases could be vented out the cold leg break, and no significant concentrations of gases would degrade the condensation heat removal of the OTSGs.
. CriticalI-feat Flux correlation This condition pertains to the use of a critical heat flux correlation appropriate to the fuel type considered. This analysis used the BWC correlation, which is appropriate for the Mark B-9 and B-10 fuel assemblies.
- Use of Wallis and Upper Plenum Test Facility (UPTF) parameters at the tube bundle and SG plenum inlet This condition pertains to the use of counter-current flow limiting models for U-tube steam generator simulations. It is not applicable to OTSG designs.
- Limitations of the BWUMV CIIF correlation This condition pertains to the use of the BWUMV CIIF correlation, which is limited to pressures above 1300 psia. The RELAP5 code logic does not use this correlation below l
1300 psia.
1
^
U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 6 of 8 l
l
. Core form losses due to ruptured fuel l
l This condition pertains to the impact of ruptured fuel on core flow. This condition largely applies to large break LOCA EM calculations, but form loss coefficients due to ruptured fuel are not excluded from any LOCA analyses.
. User specified parameters l
This condition pertains to the use of code input parameters defined in BAW-10164.
Verification that the code input specified by BAW-10164 was performed.
In summary, all NRC conditions on the use of the RELAP5 code were satisfied by demonstrating compliance with a condition, using appropriate modeling, or indicating that a particular condition did not apply to the B&W 177 fuel assembly or 205 fuel assembly designs. CR-3 has a B&W 177 fuel assembly core.
(2) The guidelines, code options, and prescribed input specified in Tables 9-1 and 9-2 in both Volume I and Volume 11 of BAW-10192P should be used in LBLOCA and SBLOCA evaluation model applications, respectively.
i Response: The prescribed guidelines, options, and input specified in these tables have l
been updated and are adhered to, as required by Reference 1.
(5) Time-in-life LOCA limits must be determined with, or shown to be bounded by, a specific l
application of the NRC-approved evaluation model.
Response: Time-in-life (TIL) calculations for SBLOCA applications are not required unless the fuel pin heat-up is sufficient to cause cladding rupture. To maximize the likelihood of cladding rupture, conservative assumptions were imposed to ensure the maximum cladding hoop stress, PCT, and normalized heating ramp rate limit. Further, parametric analyses are performed to ensure that the time of maximum PCT coincides with the predicted time of rupture. This method ensures that the calculated PCT will bound any PCT predicted by a consistent TIL analysis with appropriate TIL pin parameters.
(8) For the small break model, the hot channel radial peaking factor to be used should correspond to that of the hottest rod in the core, and not to the radial peaking factor of the 12 hottest bundles.
Response: Each pin in the hot bundle (12 assemblies) is peaked to the hot pin radial value.
(9) The constant discharge coefficient model (discharge coefficient = 1.0) referred to as the "fligh or Low Break Voiding Normalized Value," should be used for all small break analyses. The model which changes the discharge coefficient as a function of void fraction, i.e., the " Intermediate Break Voiding Normalized Value," should not be used unless 2 l
l
U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 7 of 8 transient is analyzed with both discharge models and the intermediate void method produces the more conservative result.
Response: The constant discharge coefficient is used, and verified to be used, for each SBLOCA analysis.
(10) For a specific application of the FTI small break LOCA methodology, the break size which yields the local maximum PCT must be identified. In light of the different possible behaviors of the local maximum, FTl should justify its choice of break sizes in each application to assure that either there is no local maximum or the size yielding the maximum local PCT has been found. Break sizes down to 0.01 ft should be considered.
2 Response: The SBLOCA break spectrum (down to 0.01 ft2) is performed to determine the local maximum PCT. The break sizes analyzed are chosen to ensure that the local peak has been appropriately defined.
III.
Summary of Key Results The use of the RELAP5 methodology for this SBLOCA analysis demonstrated compliance with the analytical criteria of 10CFR50.46. A brief comparison of the limiting Cold Leg Pump Discharge CRAFT 2 and RELAP5 break size and maximum PCT results is shown below. The fuel clad oxidation-related criteria are also met using either CRAFT 2 (utilizing the FOAM 2 and THETA codes) or RELAP5 based methods:
CRAFT 2 RELAP5 Limiting Break Size 0.125 ft 0.120 ft 2
2 Maximum PCT (versus 2200 F limit) 1859 F 1583 F Local Oxidation (versus 17% limit):
2.6%
0.646 %
Whole Core Oxidation:(versus 1% limit) less than 1%
less than 1%
The comparison above shows that the change in PCT is greater than 50 F, and is therefore significant. However, the analysis also demonstrates that using the RELAP5 methodology continues to meet the 10CFR50.46 criteria that pertain to analytical methodology. The core geometry also remains amenable to cooling, since resultant fuel deformations were found to maintain coolable configurations. Finally, long-term core cooling is assured through demonstrating that the core is quenched, and that pumped injection is available.
IV.
Conclusion In conclusion, FPC has demonstrated that the RELAP5 methodology as outlined in BAW-10192-PA is appropriate for use in the CR-3 Small Break LOCA analysis. This submittal addresses the limitations and conditions imposed by the associated SER. The RELAP5-based methodology was previously used for CR-3 in an approved, but limited, application. This
U.S. Nuclear Regulatory Commission Attachment 3F0199-02 Page 8 of 8 submittal is notification that the methodology will be used for the SBLOCA analysis of record for GR-3 and to meet 10CFR50.46 reporting requirements.
VL References
- 1. Letter from J. E. Lyons (NRC) to J. H. Taylor (FTI), " Acceptance for Referencing of Topical Report BAW-10192-P, "BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants." February 18,1997.
- 2. Framatome Technologies, Inc. Topical Report, BAW-10192-P Revision 0, "BWNT LOCA Evaluation Model for OTSG Plants." Febmary 1994.
- 3. Florida Power Corp. Calculation, F-98-0008 Revision 0, "SBLOCA Analysis for CRAFT 2 to RELAPS Transition," (FTI Document 86-5001942-00). August 1998.
- 4. Babcock and Wilcox Topical Report, BAW-10164-P Revision 2, "RELAP5 - An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis."
February 1994.
. 5. Babcock and Wilcox Topical Report, BAW-10155A, " FOAM 2 - Computer Code to Calculate Core Level Swell and Mass Flow Rate During SBLOCA." October 1987.
- 6. Babcock and Wilcox Topical Report, BAW-10095A Revision 1," THETA-1B - A Computer Code for Nuclear Reactor Core Thermal Analysis." April 1978.
- 7. Letter from J. P. Cowan (FPC) to USNRC, " Technical Specification Change Request Notice 210." June 14,1997.
- 8. Letter from L. Raghavan (USNAC) to R. A. Anderson (FPC)," Crystal River Unit 3 -
Issuance of Amendment RE: Small Break Loss of Coolant Accident Mitigation (TAC No.
M98991)." January 24,1998.
l l
l l
,