ML20137P779
| ML20137P779 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/01/1997 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9704100005 | |
| Download: ML20137P779 (8) | |
Text
_ _ _
Ctitief Support Departmeest l
.A
=spr 10 CFR 50.90
- PECO NUCLEAR ecco tee <ov comoeer i
965 Chesterbrook Boulevard
- A Unit of PECO Energy wayne. PA 19087-6691
.j April 1,1997 Docket No. 50-278 License No. DPR-56 U.S. Nuclear Regulatory Commission
' Attn: Document Control Desk Washington, DC 20555 i
Subject:
Peach Bottom Atomic Power Station, Unit 3 Non-Proprietary Response to Request for Additional information 4
Regarding License Change Application ECR 96-02609 t
I i
Reference:
1.
Letter from R. M. Butrovich (GE Nuclear Energy) to H. J..
Diamond (PECO Energy Company), " Peach Bottom 3, Cycle 11 SLMCPR Licensing Clarification, Revision 2," dated March 16,1997
.i
Dear Sir:
- In our "05000277/LER-1996-009, :on 961001,declared High Pressure Coolant Injection (HPCI) Sys Inoperable Due to Bearing Misalignment. Realigned Bearing Housing,Performed [[procedure" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Satisfactorily & Declared HPCI Sys Operable|letter dated March 12,1997]], PECO Energy provided a response to your Request for Additional Information (RAl) regarding License Change Application (LCA) ECR 96-02609. This LCA revises Technical Specification (TS) Section 2.0, " Safety Limits," and was submitted to the U. S. Nuclear Regulatory Commission on October 30,1996. The Attached Reference 1 letter is a non-proprietary version of the response provided by General Electric.
j i
If you have any questions, please do not hesitate to contact us.
F Verytrulyyours, kdfd60 M'Cthi YL-j 4
Wunugef.
G. A.; Hunger, Jr.,
f)
Director Licensing -
/
Enclosure:
Attachment.
00 %:
H. J. Miller, Administrator, Region I, USNRC
.)
. L. Schmidt, USNRC Senior Resident inspector, PBAPS W
R. R. Janati, Commonwealth of Pennsylvania E S E %,h N 9704100005 970401 7'
- ADOCK 05000278 {s
-P PDR W
t i
r ATTACHMENT Letter from R. M. Butrovich (GE Nuclear Energy) to H. J. Diamond (PECO Energy Company), " Peach Bottom 3, Cycle 11 SLMCPR Licensing Clarification, Revision 2,"
dated March 16,1997 1
I
'Q l
. )
GENuclearEnergy
?
' Richard M Burrovich Nude Iael-Awicas '
het Piotect Mar:ayr GeneralEleanc Depoy Castle Hnne Road P O Boa 180. M:C A33
%ngton. NC 284314780 910 G15 6?G6 DaComm 8*232426G fx 910 615 5G84 March 16,1997 cc: T. R. Loomis RMB: 97-029, R2 J. L. Embley C. L. Ileck Mr. II. J. Diamond Fuels & Services Division PECO NUCLEAR 965 Chesterbrook Houlevard Wayne, PA 19.087-5691
SUBJECT:
Peach Bottom 3, Cycle 11 SLMCPR Licensing Clarification, Revision 2
REFERENCES:
[l } NEDC-32505P, R-Factor Calculation Methodfor GE11, GE12 and GE13 Fuel, November 1995.
(2} GeneralElectricStandardApplicationforReactorFuel,NEDE-240l1-P-A-Il-US, August 1996.
(3} Licensing Topical Report, General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.
t
\\4} Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations, NEDC-1 3260lP, December 1996.
The following information is being provided in response to NRC requests for additional information in connection with their review of the Peach Bottom 3, Cycle i1 Safety Limit MCPR (SLMCPR) calculation. In our phone conversation with the NRC staff on Thursday, January 30,1997 additional information was requested in four areas. The purposes of this letter are: (1) state the request, (2) document the responses given d.: ring the phone conversation, and (3) provide additional relevant information.
I
/
. Mr. H. J. Diamond ~
2 3/16/97 1.
. Provide a Reference for the R-factor Methodology The R-fdctor calculation methodology (Reference [1]) is listed as Reference 3-13 in Chapter 3 of Reference [2]. Revision 13 of GESTAR also has this same reference.
2.
Discuss the Sensitivity of the Calculated SLMCPR to Reactor Pressure Figure 4-4 of GETAB (Reference [3]) shows the effect on critical power of varying the system pressure. For a constant channel inlet subcooling of about 20 Btu /lb. the critical power varies inversely with pressure. This is consistent with previous experience in the range of 800 to 1400 psia. As indicated in Section 4.4 of Reference
[3], the features of these data are typical of all the boiling transition data (both water and Freon) that has been used to establish GETAB.
i An example of how the calculated SLMCPR value changes with system pressure is shown for a particular core containing gel I fuel in Table 1. The core was evaluated at the nominal system pressure of 1055 psia and also at arbitrary pressure increases of 100 and 200 psia. For the elevated pressure cases the absolute value of the core inlet enthalpy was increased so that the amount of subcooling at the core inlet would be constant. As expected for this scenario the core MCPR decreases as pressure is increased corresponding to the decrease in critical power suggested by the trend in
~
Figure 4-4 of Reference [3].
i Mr. II. J. Diamond 3
3/16/97 ne i 1
l i
I i
1 i
l 4
l
'3.
- Discuss Quarter Core versus Full Core Evaluations of the SLMCPR
'.Mr. H. J. Diamond 4
3/16/97
~
The currently approved procedure is modeled after the P1 process computer model. In this model the entire power distribution and TIP instrument uncertainty is applied to the TIP signal, and the change in nodal power is directly transferred to the four bundles surrounding the TIP string.
L i
Mr. II. J. Diamond 5
3/16/97 e
4.
Discuss the Calculational Process for Evaluating CPRs and the St,MCPR The calculation process is described in Section IV-3-6 in Appendix IV of Reference
[3]; A flow chart for the process is provided in Figure IV-4 of that reference. The reactor and core models used to evaluate the initial and perturbed bundle MCPR values are equivalent to those used by the process computer. A discussion of how these models are applied is given in Section IV-3-5 of Reference [3].
. The transient change in the MCPR for the transient event that yields the highest calculated reduction in MCPR is added together with uncertainties associated with that change to the SLMCPR value to define the operating limit MCPR (OLMCPR). Additional details of this process are provided in Section 6.3 of Reference [3].
The SLMCPR value can be thought of as the lowest allowed value of all MCPRs in the core such that the most limiting point of the most limiting single failure transient will still assure that 99.9% of the fuel rods will no.t experience boiling transition.
Mr. II. J. Diamond 6
3/16/97
)
The initial MCPR distribution and all the perturbations to that distribution are all analyzed using reactor and core models equivalent to those used by the P1 plant -
j process computer. This process is no different than what is described in Reference [3].
The 3D BWR simulator (PANACEA)is used as the evaluation engine. The plant process computer is emulated by choosing the appropriate input options for the 3 D j
BWR simulator. This is the currently approved approach.
i i
Very Truly yours,
. M. Butrovic i d