ML110450537
| ML110450537 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/11/2011 |
| From: | Trepanier T Calvert Cliffs 3 Nuclear Project, Constellation Energy Nuclear Group, EDF Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 11-013 | |
| Download: ML110450537 (35) | |
Text
Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 CENG a joint venture of E
Cnsergnation~
CALVERT CLIFFS NUCLEAR POWER PLANT NRC 11-013 February 11, 2011 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Supplement to License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel
REFERENCE:
(a) Letter from Mr. T. E. Trepanier (CCNPP) to Document Control Desk (NRC), dated November 23, 2009, License Amendment Request:
Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel During review of information provided to support Calvert Cliffs transition from Westinghouse nuclear fuel to AREVA nuclear fuel, the Nuclear Regulatory Commission staff has requested supplemental information be provided. The supplemental information is provided in Attachment (1). In addition, to resolve outstanding issues with some of the methodologies used in the evaluation of the transition to AREVA nuclear fuel, we propose the adoption of certain license conditions in Appendix C of Renewed License Nos. DPR-53 and DPR-69 for Calvert Cliffs Units 1 and 2, respectively. The proposed license conditions are contained in Attachment (2). The information in this response does not change the No Significant Hazards Determination previously provided in Reference (a).
Attachment (1) contains information that is proprietary to AREVA, therefore, it is accompanied by an affidavit signed by AREVA, owner of the information (Attachment (3). The affidavit sets forth, with specificity, the considerations listed in 10 CFR 2.390(b)(4). Accordingly, it is requested that the information that is proprietary to AREVA be withheld from public disclosure. The non-proprietary version of Attachment (1) is included as Attachment (4).
t41o
Document Control Desk February 11, 2011 Page 2 Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.
Very truly your Thomas E. Trepan6 Plant General Manager STATE OF MARYLAND TO WIT:
COUNTY OF CALVERT I, Thomas E. Trepanier, being duly sworn, state that I am Plant General Manager - Calvert Cliffs Nuclear Power Plant, LLC (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants.
Such information has been reviewed in accordance with compan acticen believe it to be reliable.
Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of
- o*( .e.A-r ,this dayof Fe~bjucL ,2011.
'jb TNESS y Hand and Notarial Seal:
' -. ,yT Public Wendy L. Hunter My Commission Expires: 0*4left county, NOTARY Mlarylane/
PUBLIC / 0 /-V My Convinision Expires 1/N/O14 Date TET/PSF/bjd
Attachment:
(1) Proprietary Supplemental Information (2) License Conditions (3) AREVA Proprietary Affidavit (4) Non-Proprietary Supplemental Information
Document Control Desk February 11, 2011 Page 3 cc: Without Attachment (1)
D. V. Pickett, NRC Resident Inspector, NRC W. M. Dean, NRC S. Gray, DNR
ATTACHMENT (2)
LICENSE CONDITIONS
-j Calvert Cliffs Nuclear Power Plant, LLC February 11, 2011
ATTACHMENT (2)
LICENSE CONDITIONS Draft License Conditions for Unit 1 Amendment No. Additional Conditions Implementation Date 297 For the Asymmetric Steam Generator This amendment is effective Transient analysis performed in immediately and shall be accordance with the methodology of implemented within 60 days of Technical Specification 5.6.5.b.8, the completion of the Unit 1 2012 methodology shall be revised to capture refueling outage.
the asymmetric core inlet temperature distribution and application of local peaking augmentation factors. The revised methodology shall be applied to Calvert Cliffs Unit 1 core reload designs starting with Cycle 20.
For the Seized Rotor Event analysis performed in accordance with the methodology of Technical Specification 5.6.5.b.8, the methodology shall be revised to capture the asymmetric core inlet flow distribution. The revised methodology shall be applied to Calvert Cliffs Unit I core reload designs starting with Cycle 20.
For the Control Element Assembly Ejection analysis performed in accordance with the methodology of Technical Specification 5.6.5.b.l 1, the cycle-specific hot zero power peak average radial fuel enthalpy is calculated based on a modified power dependent insertion limit with Control Element Assembly Bank 3 assumed to be fully inserted (only in the analysis, not in actual plant operations). This revised methodology shall be applied to Calvert Cliffs Unit 1 core reload designs starting with Cycle 20.
The Small Break Loss of Coolant accident performed in accordance with the methodology of Technical Specification 5.6.5.b.9 shall be analyzed using a break spectrum with augmented detail related to break size. This revised methodology shall be applied to Calvert Cliffs Unit 1 core reload designs starting with Cycle 20.
Core Operating Limits Report Figures 3.1.6, 3.2.3, and 3.2.5 shall not be I
ATTACHMENT (2)
LICENSE CONDITIONS Amendment No. Additional Conditions Implementation Date changed without prior NRC review and approval until an NRC-accepted generic, or Calvert Cliffs-specific, basis is developed for analyzing the Control Element Assembly Rod Bank Withdrawal Event, the Control Element Assembly Drop, and the Control Element Assembly Ejection (power level-sensitive transients) at full power conditions only.
Approval of the use of S-RELAP5 (Technical Specification 5.6.5.b.8) is restricted to only those safety analyses that confirm acceptable transient performance relative to the specified acceptable fuel design limits. Prior transient specific NRC approval is required to analyze transient performance relative to reactor coolant pressure boundary integrity until NRC approval is obtained for a generic or Calvert Cliffs-specific basis for the use of the methodology in Technical Specification 5.6.5.b.8 to demonstrate reactor coolant pressure boundary integrity.
For the RODEX2-based fuel thermal-mechanical design analysis performed in accordance with the methodology of Technical Specification 5.6.5.b.3, Calvert Cliffs Unit 1 core reload designs (starting with Cycle 20) shall satisfy the following criteria:
- a. Predicted rod internal pressure shall remain below the steady state system pressure.
- b. The linear heat generation rate fuel centerline melting safety limit shall remain below 21.0 KW/ft.
For the Control Element Assembly Ejection analysis, Calvert Cliffs Unit 1 core reloads (starting with Cycle 20) shall satisfy the following criteria:
- a. Predicted peak radial average fuel enthalpy when calculated in accordance with the methodology of Technical Specification 5.6.5.b. 11 shall remain below 200 cal/g.
2
ATTACHMENT (2)
LICENSE CONDITIONS Amendment No. Additional Conditions Implementation Date
- b. For the purpose of evaluating radiological consequences, should the S-RELAP5 hot spot model predict fuel temperature above incipient centerline melt conditions when calculated in accordance with the methodology of Technical Specification 5.6.5.b.8, a conservative radiological source term (in accordance with Regulatory Guide 1.183, Revision 0) shall be applied to the portion of fuel beyond incipient melt conditions (and combined with existing gap source term), and cladding failure shall be presumed.
The approval of the emergency core cooling system evaluation performed in accordance with the methodology of Technical Specification 5.6.5.b.7 shall be valid only for Calvert Cliffs Unit 1, Cycle
- 20. To remove this condition, Calvert Cliffs shall obtain NRC approval of the analysis of once- and twice-burned fuel for core designs following Unit 1 Cycle 20.
3
ATTACHMENT (2)
LICENSE CONDITIONS Draft License Conditions for Unit 2 Amendment No. Additional Conditions Implementation Date 273 For the Asymmetric Steam Generator This amendment is effective Transient analysis performed in immediately and shall be accordance with the methodology of implemented within 60 days of Technical Specification 5.6.5.b.8, the issuance.
methodology shall be revised to capture the asymmetric core inlet temperature distribution and application of local peaking augmentation factors. The revised methodology shall be applied to Calvert Cliffs Unit 2 core reload designs starting with Cycle 19.
For the Seized Rotor Event analysis performed in accordance with the methodology of Technical Specification 5.6.5.b.8, the methodology shall be revised to capture the asymmetric core inlet flow distribution. The revised methodology shall be applied to Calvert Cliffs Unit 2 core reload designs starting with Cycle 19.
For the Control Element Assembly Ejection analysis performed in accordance with the methodology of Technical Specification 5.6.5.b.1 1, the cycle-specific hot zero power peak average radial fuel enthalpy is calculated based on a modified power dependent insertion limit with Control Element Assembly Bank 3 assumed to be fully inserted (only in the analysis, not in actual plant operations). This revised methodology shall be applied to Calvert Cliffs Unit 2 core reload designs starting with Cycle 19.
The Small Break Loss of Coolant accident performed in accordance with the methodology of Technical Specification 5.6.5.b.9 shall be analyzed using a break spectrum with augmented detail related to break size. This revised methodology shall be applied to Calvert Cliffs Unit 2 core reload designs starting with Cycle 19.
Core Operating Limits Report Figures 3.1.6, 3.2.3, and 3.2.5 shall not be 4
ATTACHMENT (2)
LICENSE CONDITIONS Amendment No. Additional Conditions Implementation Date changed without prior NRC review and approval until an NRC-accepted generic, or Calvert Cliffs-specific, basis is developed for analyzing the Control Element Assembly Rod Bank Withdrawal Event, the Control Element Assembly Drop, and the Control Element Assembly Ejection (power level-sensitive transients) at full power conditions only.
Approval of the use of S-RELAP5 (Technical Specification 5.6.5.b.8) is restricted to only those safety analyses that confirm acceptable transient performance relative to the specified acceptable fuel design limits. Prior transient specific NRC approval is required to analyze transient performance relative to reactor coolant pressure boundary integrity until NRC approval is obtained for a generic or Calvert Cliffs-specific basis for the use of the methodology in Technical Specification 5.6.5.b.8 to demonstrate reactor coolant pressure boundary integrity.
For the RODEX2-based fuel thermal-mechanical design analysis performed in accordance with the methodology of.
Technical Specification 5.6.5.b.3, Calvert Cliffs Unit 2 core reload designs (starting with Cycle 19) shall satisfy the following criteria:
- a. Predicted rod internal pressure shall remain below the steady state system pressure.
- b. The linear heat generation rate fuel centerline melting safety limit shall remain below 21.0 KW/ft.
For the Control Element Assembly Ejection analysis, Calvert Cliffs Unit 2 core reloads (starting with Cycle 19) shall satisfy the following criteria:
- a. Predicted peak radial average fuel enthalpy when calculated in accordance with the methodology of Technical Specification 5.6.5.b. 1I shall remain below 200 cal/g.
5
ATTACHMENT (2)
LICENSE CONDITIONS Amendment No. Additional Conditions Implementation Date
- b. For the purpose of evaluating radiological consequences, should the S-RELAP5 hot spot model predict fuel temperature above incipient centerline melt conditions when calculated in accordance with the methodology of Technical Specification 5.6.5.b.8, a conservative radiological source term (in accordance with Regulatory Guide 1.183, Revision 0) shall be applied to the portion of fuel beyond incipient melt conditions (and combined with existing gap source term), and cladding failure shall be presumed.
The approval of the emergency core cooling system evaluation performed in accordance with the methodology of Technical Specification 5.6.5.b.7 shall be valid only for Calvert Cliffs Unit 2, Cycle
- 19. To remove this condition, Calvert Cliffs shall obtain NRC approval of the analysis of once- and twice-burned fuel for core designs following Unit 2 Cycle 19.
6
ATTACHMENT (3)
AREVA PROPRIETARY AFFIDAVIT Calvert Cliffs Nuclear Power Plant, LLC February 11, 2011
AFFIDAVIT COMMONWEALTH OF VIRGINIA )
) ss.
CITY OF LYNCHBURG )
- 1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.
- 2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.
- 3. I am familiar with the AREVA NP information contained in the attachment to a letter from T.E. Trepanier (Calvert Cliffs Nuclear Power Plant) to Document Control Desk (NRC) entitled "Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel,'! numbered NRC 11-013 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.
- 4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in
accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."
- 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:
(a) The information reveals details of AREVA NP's research and development plans and programs or their results.
(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.
The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.
- 7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
- 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.
SUBSCRIBED before me this ____
day of 'C
,2011.
Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/14 Reg. # 7079129 SHERRY L. MCFAOEN Notary Public ,
Commonwealth of Virginia Mrc7079129 14 My Commission Expires Oct 31, 2'014 r
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Calvert Cliffs Nuclear Power Plant, LLC February 11, 2011
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Supplemental Information 1 Small Break LOCA Break Size Increment Study To verify that the most limiting SB LOCA peak cladding temperature was determined based on a 0.01 ft2 break size increment the following sensitivity study was performed.
The cladding temperature for the published limiting case (break size of 0.09 ft2) turned over nearly coincidently with the. safety injection tank (SIT) discharge. The peak cladding temperature occurred
[ I after the start of the SIT discharge. However, SIT discharge did not occur for the next smaller break size (0.08 ft2). Given this information it could be postulated that the limiting peak cladding temperature could occur at a break size between 0.08 ft2 and 0.09 f.
To further investigate the issue, AREVA performed additional SB LOCA studies, specifically, at 0.087 ft2, 0.088 W, and 0.089 ft2. These break areas represent an increment in break diameter of 0.02 inches. Note the SIT injection rate was multiplied by 2 to demonstrate the timing dependence of SIT injection on the hot spot cladding temperature using a single y-axis scale.
The results of those cases are shown below in Figures 1-1 through 1-4. They produced lower peak cladding temperatures than the previously identified limiting case. The results are reasonable, with the time between peak cladding temperature occurrence and SIT discharge monotonically decreasing as break size increased - until, for the limiting case, the SIT discharge occurred almost simultaneously with peak cladding temperature. From the study, it can be seen that the cases performed encompassed the possible timing dependencies between peak cladding temperature and SIT injection. The scope of cases cover the phenomena where the escalation of cladding temperature was terminated by [
I to the limiting case where the [ ] occurred essentially at the same time. For break sizes less than 0.09 ft2 (4.1 inches), the model was biased to ensure
[ ]. For break sizes larger than 0.09 f, [
] with an associated reduction in peak cladding temperature. Thus, the supporting calculations confirm that within a change in break diameter of 0.02 inches, the 0.09 ft2 break remains the limiting case.
1
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 0.087 ft2 Break 2000 1800 1600
~1400 1200 LL I--
1000
. 800
-- Hot Spot Cladding Temperature E600 ----.. SIT Discharge Rate x2 400 200 0
1000 1500 ID:49235 24Jan2011 21:15:03 087.dmx Time (sec)
Figure 1-1, Peak Cladding Temperature for a 0.087 ft2 Break 2
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 0.088 ft2 Break 2000 1800 1600 if g 1400 o 1200 U-I--
0 1000
- 800 Hot Spot Cladding Temperature 600 --- a SIT Discharge Rate x2 1-400 200 0
1000 1500 ID:49235 24Jan2011 21:15:03 087.dmx Time (sec)
Figure 1-2, Peak Cladding Temperature for a 0.088 ft2 Break 3
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 0.089 ft2 Break 2000 1800 1600 g 1400 o 1200 FL I,-
1000 800 G)
M. 600 -* Hot Spot Cladding Temperature E, ------ SIT Discharge Rate x2.
400 200 0
1000 1500 11:49235 24Jan2011 21:15:03 087.dmx Time (sec)
Figure 1-3, Peak Cladding Temperature for a 0.089 ft2 Break 4
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 0.090 ft2 Break 2000 1800 1600
_ 1400 1200 1000 800 CO
) 600 Hot Spot Cladding Temperature E
01)
U SIT Discharge Rate x2 1-400 200 0
1000 1500 0:49235 24Jan2011 21:15:03 087.dmx Time (sec)
Figure 1-4, Peak Cladding Temperature for a 0.090 ft2 Break 5
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Supplemental Information 2 The figures below are provided in connection with discussions regarding the Response to Question 33 provided in Reference (1).
6
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION PCT - Heat Structure 207-38 PCT = 1682.2 "F, at Time = 33.12 s, on 4% Gad Rod 2000 1500 CL 1000 0.
a_
500 F 00 100 200 300 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 7
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 8
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 9
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Heat Transfer Coefficient - Heat Structure 207-38 60 I1 I F I I 50
-4
-230 (D..
0 0
C-I.-
10 0 L 0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 10
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Core Outlet Mass Flow 50 40 30 20 U) 10 E
.0 0
U)
V)
-10
-20
-30
-40
-50 0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 1I
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION 12
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Containment Pressure 60 Containment 50 40 CL 1030 a.
20 10 0
0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 13
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Upper Plenum Pressure 500 I I I I I 400 300 al CL Cl) 200 100 L
0 0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 14
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Core Inlet Mass Flow 100 50 U)ý E
0 0 U)
U)
-50
-100 0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 15
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Reactor Power 5e+08 0
a- 3e+08 Toa Reco oe RacorPo
--- erfrm isio Ir Oe+00 0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2O11 07:40:44 RSTPLT 16
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Break Flow 10 Vessel Side Pump Side Total Cl)
E 5
co 0
UL 0
0 50 100 150 200 250 300 350 400 Time (s)
ID:25522 25Jan2011 07:40:44 RSTPLT 17
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Supplemental Information 3 In connection with continued discussions of the Response to Question 12 provided in Reference (2), the following information is provided.
The mixture level shown in Figure 3-2 locates an approximate elevation where there is a large void discontinuity. It is not used in the code for any system calculations. The nodal average void fraction along with the non-equilibrium and non-homogeneous fluid conditions are used in the actual clad thermal calculation.
The S-RELAP5 determines where the mixture level is in vertically-oriented volumes. The mixture level within that volume is determined from the void fractions above and below the detected mixture level.
For the period of interest, this translates to [ ]. Note the volumes in this region are
[ ] tall. This is a calculated mixture level that is based on [
1; i.e., even though the lower volumes have begun to accumulate more liquid (void fraction decreasing), it has essentially no effect on the volume in which the calculated mixture level resides.
The mixture level estimated above agrees well with the code-calculated mixture level shown in Figure 3-2. This is not an actual mixture height as one would derive from a code with a true drift flux or level swell model, but a reasonably accurate approximation from which the calculation can be understood.
Direct interpretation of the mixture height from the void fraction distribution gives the same approximate result within one core volume.
Figure 3-2 also shows that the core inlet flow slightly exceeded [
], evidenced by the increase in the core liquid level (decrease in the core entrance voiding shown in Figure 3-1) and also the increase in the downcomer collapsed liquid level. So, there is little or no change in the core mixture level during this time and the level stays relatively constant with the temperature increasing slightly until just before the mixture advances one S-RELAP node, at around I[.
Figure 3-3 shows that during this period [ J, the total high pressure safety injection flow rate and the break flow rate crossed, with the high pressure safety injection exceeding break flow after about [ ]. This is consistent with the core and downcomer level response shown in Figure 3-2.
The adjustment of mixture level established sufficient cooling to slow the hot spot temperature rise as shown in the cladding thermal response shown in Figure 5-13 in the calculation file. The SIT injection then increased the supply of cooling water, lowered the system pressure, and established a clear path to the end of the temperature excursion.
18
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Figure 3-1, Void Fractions - Two-Phase Mixture Level Figure 3-2, Two-Phase Mixture and Collapsed Liquid Levels in the Core 19
ATTACHMENT (4)
NON-PROPRIETARY SUPPLEMENTAL INFORMATION Figure 3-3, Break, Total High Pressure Safety Injection, and Core Inlet Flow Rates References
- 1. Letter from Mr. G. H. Gellrich (CCNPP) to Document Control Desk (NRC), dated December 30, 2010, Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel
- 2. Letter from Mr. T. E. Trepanier (CCNPP) to Document Control Desk (NRC), dated January 14, 2011, Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel 20