ML20063F297
| ML20063F297 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/01/1994 |
| From: | Denton R BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M87690, TAC-M88143, NUDOCS 9402140232 | |
| Download: ML20063F297 (7) | |
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- 1 BALTIMORE GAS AND ELECTRIC 1650 CALVERT CUFFS PARKWAY. LUSBY, MARYLAND 20657-4702 ROBERT E DENTON Vict PRE $lDE,47
[Chfy3{y},}994 NUCLE AR ENERGY
{4:0)760-44%5 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 Response to NRC Request for Additional Information Regarding Heatup/Cooldown Curves and Variable Low Temperature Overpressure Protection, Unit 1, and An Extension of the Heatup/Cooldown Curves, Unit 2, for the Calvert Cliffs Nuclear Power Plant (TAC Nos. M87690: M88143)
REFERENCES:
(a)
Ixtter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr R. E. Denton (BG&E), dated January 5,1994, Request for Additional Information Regarding Heatup/Cooldown Curves and Variable Low Temperature Overpressure Protection, Unit 1, and An Extension of the Heatup/Cookfown Curves, Unit 2, for the Calvert Cliffs Nuclear Power Plant (TAC Nos. M87690; M88143)
(b)
Letter from Mr. R. E. Denton (BG&E) to NRC Document Control Desk, dated September 3,1993, Unit 1 License Amendment Request; Variable Low Temperature Overpressure Protection v
(c)
Letter from Mr. R. E. Denton (BG&E) to NRC Document Control Desk, dated November 1,1993, Unit 2 License Amendment Request; Extension of Unit 2 Heatup and Cooldown Curves This letter provides Baltimore Gas and Electric Company's response to your request for additional information (RAl) (Reference a). The RAI pertains to two of our requests regarding the heatup and cooldown curves and variable low temperature overpressure protection for Unit 1 (Reference b), an:1 the extension of the heatup and coolf-m curves for Unit 2 (Reference c). The specific information you requested and our responses are u 'ined in Attachment (1).
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9402140232 940201 DR ADOCK 05000317 l
9
., Document Control Desk i
Februa:y 1,1994 Page 2 -
l Should you have any further questions regarding this matter, we will be pleased to discuss them with
.you.
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- I Very truly yours,
.j I
RED /DJM/dlm Attachment I
cc:
D. A. Brune, Esquire l
J. E. Silberg, Esquire i
R. A. Capra, NRC-D. G. Mcdonald, Jr., NRC T. T. Martin, NRC P. R. Wilson, NRC R. I. McLean, DNR J. H. Walter, PSC i
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ATTACIIMENT (1)
Response to NRC Request for Additional Informatica
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Calvert Cliffs Units 1 and 2 Revision of Pressure-Temperature Curves UNIT 1 Question 1.
Describe the ' method by which the 2.61x1019 n!cm2 value was detennined, and the operating timefor which it has been estimated.
Resp <mse 1
2.61x1019 n/cm2 is the fluence at which the limiting Unit 1 beltline material is presently predicted to reach the Pressurized Thermal Shock (PTS) screening criteria, as defined in 10 CFR 50.61.
It should be noted, however, that plant-specific surveillance data '
(Reference 1) indicates that this Unit 1 beltline material can accommodate a higher fluence than used here.
The calculations were performed as prescribed in 10 CFR 50.61. The maximum flyence corresponds to the weld 2-203 A,B,C screening criterion temperature limit of 270 F as follows:
.2-203 A,B,C Copper 0.21 w/o Reference 6, Att ATable 4-1 Nickel 0.88 w/o Reference 6. Att ATable 4-1 Chemistry Factor (CF) 210 Regulatory Guide 1.99 Revision 2
-50,[F Initial RTuor(IRTuor)
Reference 6, Att A Table 4-1 Margin (M)
+56 Reference 6, Att ATable 4-3 Screening Criterion (ART) 270,F 10 CFR 50.61 ART = IRTNor + dRTuor + M Regulatory Guide 1.99 Revision 2 270 = -50
+d dRTuor = 264,RTuor + 56 F
dRTuor = CF
- f**(0.28-0.10* log (f))
Regulatory Guide 1.99 Revision 2 f=2.61E+19 n/cm l
l From 10 CFR 50.61, "f means the best estimate neutron fluence in units of.1019 n/cm2 1
(E > 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the.
location where the materialin question receives the highest fluence for the period of senice in question." Note that this methodology is a function of material properties only.
Baltimore Gas and Electric Company presently projects the above fluence value to be achieved at the limiting beltline material in year 2004 assuming continuation of current core design. Reference (2) documents the basis for this projection.
i n/cm2 he peak azimuthal and axial value on the inside surface of the Question 2.
Is 2.61x1019 t
pressure vessel or the value at the location of the welds?
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Response
2.61x1019 n/cm2 is the peak value at the location of weld 2-203 A,B,C. See response to Question 1. An administrative change will be'made to the proposed figures transmitted in Reference (5). The words ". ATTHE INNER SURFACE OF THE REACTOR VESSEL".
should have been left out of the title. This change deletes these words. ' The' peak fluence locatien is defined in the Bases. It is not necessary to include this information in the title of the figures.
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ATTACIIMENT (1)
Resp (mse to NRC Request for AdditionalInformation Calvert Cliffs Units I and 2 Revision of Pressure-Temperature Curves Question 3.
Assuming thepeak azimuthallocation does not correspond to the azimuthallocation of.
the last surveillance capsule, how was the extrapolation made from the surveillance capsule measured value to the value at the location of the weld, or was an all-analytical method used?
Response
Baltimore Gas and Electric Company's fluence methodologies are described ' in Reference (2). For Calvert Cliffs' surveillance capsule methodology in general, a numerical model is constructed via the DOT-IV transport code and benchmarked to surveillance capsule measured activities and Pool Critical Assembly results. This modelis then employed i
to generate fast neutron fluxes at the interior of the reactor vessel. These fluxes are numerically integrated over the reactor operating history to yield azimuthally and axially dependent fluences at the clad-base-metalinterface.
Question 4, in view of: the current staff recommendation for using cross sections based on the ENDFIB-VI data, have you estimated (past and future) fluence to the pressure vessel using ENDFIB47 based cross sections?
Response
The ENDF/B-VI cross section library has not yet had an independent technical review i
completed and is not appropriate for general use. Therefore, data from it were not used.
l Question S.
Was low leakage taken into account and, ifyes, how?
Response
The pressure-temperature curves are based on the limiting fluence value for the limiting beltline material. That value does not change as fuel management practices change.
Fluence projections for Unit I consider a low-leakage core for Cycle 10 and low-fluence cores with guide-tube flux suppressors for Cycles 11 and beyond when estimating the 2
operating time at which 2.61x1019 n/cm will be reached. The methodology is described in Reference (2).
The introductory paragraph of your request for additional information contains two additional items which require clarification.
a.
A variable setpoint will be usedfor the POR Vs during shutdown cooling.
Response
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The variable trip setpoint for the power +perated relief valves (PORVs) will be used during forced circulation (reactor coolant pumps in operation) and natural circulation. While on shutdown cooling (SDC), a single trip setpoint for the PORVs -
will be maintained. His is necessaiy because the Tcold resistance temperature 2
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1 ATTACHMENT (1)
Response to NRC Request for Additional Information Calvert Cliffs Units 1 and 2 Revision of Pressure-Temperature Curves detectors are not in the Dow stream while on SDC and may not adequately monitor Teactor vessel temperature.
b.
The submittal does notprovide any information as to... how the new setpoints relate to the copper and nickel content of the limiting welds.
Itesponse The Copper and Nickel content of the limiting weld are used to determine the 1/4T and 3/4T Adjusted Reference Temperatures (ARTS) using the methods-of Regulatory Guide 1.99 Revision 2 as follows:
f= 1.56E+ 19 n/cm(-0.24*x)
Regulatory Guide 1.99 Revision 2 1/4 T: f=fsurt cxp ART =-50 + 209.8*1.56"(0.28-0.10* log (1.56)) + 56 = 241.4, tory Guide 1 ART =lRTuor + dRTuor + M Regula F
3/4 T: f=fsort cxp -0.24*x)
Regulatory Guide 1.99 Revision 2 i
f=5.53E+18 n/cm(-
ART =IRTsor + dRTuor + M Regulatogy Guide 1.99 Revision 2 ART =-50 + 209.8*0.553**(0.28-0.10* log (0.553)) + 56 = 181.0 F The 1/4T and 3/4T ARTS are used in the Appendix G methodology, which was submitted in Reference (3) and approved in Reference (4).
As_ stated in Reference (5), the variable trip setpoint for the PORVs is set such that peak transient pressure will not exceed the Appendix G limits. This approach is consistent with the guidance in Standard Review Plan 5.2.2.
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I ATTACIIMENT (1)
Resp (mse to NRC Request for Additional Information Calvert Cliffs Units 1 and 2 Revision of Pressure-Temperature Curves UNIT 2 Question 1.
Ilow was thejluence value of1.9x1019nicm2 detennined?
Resrumse The, Unit 2 pressure-temperature curves are based on 1/4T and 3/4T ARTS of 171 F and 125 F, respectively. The curves are bounding for any 1/4T and 3/4T ARTS up to these values. 1.92x1019 n/cm2 is tige lowest fluence which causes the 1/4T or 3/4T ART of any beltline material to reach 171 F or 125 F, respectively.
The first component to reach these values is plate D-8906-1 at the 3/4T location..Using Regulatory Guide 1.99, Revision 2, the fluence at the clad-base-metal interface is calculated as follows:
Plate D-8906-1 Copper 0.15 w/o Reference 6, Att ATable 4-2 Nickel 0.56 w/o Reference 6, Att A Table 4-2 Chemistry Factor (CF) 108 Regulatory Guide 1.99 Revision 2 Initial RTsor(IRTuor)
+ 10, F Reference 6, Att A Table 4-2 Margin
+34,F Reference 6, Att A Table 4-3 3/4 T: ART =IRTuor + dRTuor + M Regulatory Guide 1.99 Revision 2 G=4.06E+18 n/cm2 (0.28-0.10* log (G)) + 34 125 = + 10 + 108* G *
- 0 T: f0= G *exp(+0.24*x)
Regulatory Guide 1.99 Revision 2 IU=1.92E+19 n/cm2 1.92x1019 2
n/cm is the limiting fluence at the clad-base-metalinterface of plate D-8906-1.
Question 2.
Same as Unit 1, Question 4.
Response
The ENDF/B-VI cross section library has not yet had an independent technical review completed and is not appropriate for general use. Therefore, data from it were not used.
Question 3.
Same as Unit 1, Question S
Response
ne pressure-temperature cunes are based on the limiting fluence value for the limiting beltline material. That value does not change as f el management practices change.
Fluence projections for Unit 2 consider low-leakage cores for Cycles 8 and beyond when estimating the operating time at which 1.92x1019 n/cm2 will be reached. The methodology is
~ described in Attachment (C) of the enclosure contained in Reference (6), and approved in Reference (7).
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A'ITACHMENT (1)
Response to NRC Request for Additional Information
' Calvert Cliffs Units 1 and 2 Revision of Pressure-Temperature Curves REFERENCES (1) letter from Mr. R. E. Denton to NRC Document Control Desk, dated November 29,1993,.
Request for Approval to use Plant-Specific Data (2)
Letter from Mr. R. E. Denton to NRC Document Control Desk, dated June 22,1993, -
Analysjs of Calvert Cliffs Unit No.1 Reactor Vessel Surveillance Capsule Withdrawn from.
the 97 Location (3)
Letter from Mr. G. C. Creel to NRC Document Control Desk, dated October 22,1990, Technical Specification Change - Low Temperature Overpressure Protection i
(4)
Letter from _Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton (BG&E), dated December 18,1990, Issuance of Amendment for Calvert Cliffs Nuclear Power Plant Unit 2 (5)
Letter from Mr. R. E. Denton (BG&E) to NRC Document Control Desk,- dated September 3,1993, License Amendment Request; Variable Low Temperature Overpressure Protection (6)
Letter from Mr. G. C. Creel to NRC Document Control Desk, dated December 13, 1991, Response to the 1991 Pressurized Thermal Shock Rule (7)
Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton (BG&E), dated.
May 24,1993, Response to the 1991 Pressurized Thermal Shock (PTS) Rule,10 CFR 50.61, Calvert Cliffs Nuclear Power Plant, Unit No. 2 (TAC No. M82505) t l
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