ML20044G578
| ML20044G578 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/07/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044B514 | List: |
| References | |
| GL-88-11, NUDOCS 9306030339 | |
| Download: ML20044G578 (4) | |
Text
%
DiCLOSURE SAFETI EVAIUATIOff BY 'IHE OFFICE OF NUCLEAR REACIOR RD3UIATION REIATED 10 AMDIEMDTP 70 FACIIIIY OPERATDC LICDGE
- 10. DPR-53
[
BALTIMDRE GAS MiD EIECTRIC CCEPANY CALVERT CLIFFS IUCIIAR PCEER FIR 7T. Ul7IT 1 DOCKE'T 10. 50-317
1.0 INTRODUCTION
By letter dated February 6,1992, the Baltircru Gas aM Electric C2rparry (the licensee) requested permimion to revise the pressure /tecperature (P/T) limits in the Calvert Cliffs Nuclear Power Plant, Lhit 1 Technical Specifications, Section 3.4.
The prtposed P/J limits ser2 requested for a re.rtron fluence applicable up to 3.25E19 rVcn at the inner surface of the reactor v m l.
This correspan:is to about 22 effective full power years (EFPY). The proposed P/T limits were developed using Regulatory Guide (RG) 1.99, Revision 2.
Generic Istter 88-11, "HRC Positicn en Radiation Eobrittleoent of Reactor Vessel Materials and Its Effect on Plant Operations," rm --ads RG 1.99, Rev.
2, be used in calculating P/T limits, unless the use of different rethcds can bejustified.
7b evaluate the P/T limits, the staff uses the follcwiry !GC regulations ard guidance: Apperdices G ard H of 10 CTR Part 50; the ASIM Standards and the ASME Cbde, which are referenced in Appendices G and H; 10 CTR 50.36(c)(2);
i RG 1.99, Rev. 2: Standard Revie4 Plan (SRP) Section 5.3.2; ard Generic Intter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CPR 50.36 to provide Tbchnical Specifications for the operation of the plant. In particular,10 CFR 50.36(c)(2) requires that limiting corr 11tions of operation be included in the Technical Specifications. Tha P/T limits are anorg the limitirq canditions of operation in the Technical Specifications for all cotrercial nuclear plants in the U.S.
Apperdices G ard H of 10 CER Part 50 describe specific requi-uts for fracture totghness ard reactor I
vessel raterial surveillarce that must be considered in setting P/T limits.
An acceptable rethod for uac.d.zwtirq the P/T limits is described in SRP l
Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness ard testirq requirunents for reactor vessel raterials in accordance with the ASME Cbde ard, in particular, that the beltline materials in the surveillarx:e capsules be tested in accordance with AppeMix H of 10 CFR Part 50.
Appendix H, in turn, refers to ASIM Stardards. These tests define the extent of vessel
~
l 9306030339 920507 PDR ADOCK 05000317 P
b 2
e2rittlemnt at the tim of capsule withdrawal in tem of the irrrease in refererce terperature. Apperrlix G also requires the li nsee to predict the effects of neutron irradiation on vessel e 2 rittlenent by calculating the adjusted referen tenpcrature (ART) and Charpy upper shelf energy (USE).
Generic letter 88-11 requested that li nsees and pemittees use the raethods in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor vessel materials. This guide defines the ART as the sum of unirradiated refererce tenperature, the increase in reference temperature resultirg frun neutron irradiation, ard a nargin to ao:: cunt for uncertainties in the prediction method.
2.0 EVAIIJATICN The staff evaluated the effect of neutron irradiation embrittlemnt on eacti beltline material in the Calvert Cliffs 1 reactor vessel. The amount of irradiation erbrittlemnt was calculated in aa:ctdance with RG 1.99, Rev. 2.
The staff detemined that the raterial with the highest ART at 22 EFPY was the intermdiate shell weld 2-203A, B, ard C with 0.21% o:5per (Cu), 0.87% nickel (Ni), ard an initial Rr dt of -50*F.
n For the limiting beltline mterial, the inte=diate shell weld, the staff calculated the ART to be 251.8'F at 1/4T (T = reactor vessel beltline thickness) ard 392.3*F for 3/4T at 22 EFP{. The staff used a neutron fluence of 1.94E19 TVcn at 1/4T ard 6.88E18 rVcn at 3/4T. The ARP was determned by Section 1 of RG 1.99, Rev. 2, because only one surveillance capsule has been renoved frun the reactor vessel.
The licensee used the rethod in RG 1.99, Pet. 2, to calculate an ART of 253.7'F at 1/4T ard 193.8 at 3/4T for the same limitirq weld, 2-203A. The licensee's ARIs are rcre conservative than the staff's ARTS, therefore, they are acceptable. Substituting the licensee's ARTS into egaations in SRP 5.3.2, l
the staff verified that the prrmed P/T limits for heatup, cooldown, ard hydrotest meet the beltline raaterial requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, Apperdix G of 10 CFR Part 50 also im P/T limits based an the reference tenperature for the reactor vessel closure flange materials.Section IV.A.2 of A;perdix G states that when the pressure exoecds 20% of the preservice system hydrostatic test pressure, the tenperature of the closure flarge regions highly stressed by the bolt preload l
rust exoecd the nil-ductility transition referen tsperature of the material in those regicos by at least 120'F for romal operation ard by 90*F for hydrostatic pressure tests ard leak tests.
Based an the flarge reference taperature of -10*F, the staff has determned that the prr-ed P/T limits satisfy Section IV.A.2 of Appendix G.
Section IV.A.1 of Apperdix G requires that the predicted 01arpy USE at erd of life be above 50 ft-lb. The limitirq unirradiated USE is that of lower shell ocurse plate D-7207-1 which is 77 ft-lb. Usity Figure 2 of RG 1.99, Pei. 2, 1
Iq s
3 the staff detemined that the USE at end of life would be 53.2 ft-lb. B is is greater than 50 ft-lb ard, therefore, is ao ptable.
?O CD! CLUSICN The staff concitxles that the proposed P/T limits for the reactor coolant and criticality are valid for a syste:n for heatup, cooldawn, leak p,the inner surface of the reactor vessel neutron fluence up to 3.25E19 TVc:n at (about 22 EFPY) because the pW limits conform to the requirements of Apperdix G of 10 GR Part 50. The su M P/T limits also satisfies Generic Istter 88-11 because the licensee used the rethod in R31.99, Rev. 2 to calculate the adjusted refererce te::perature. Hence, the proposed P/T limits may be in.urprated into the Calvert Cliffs 1 Tecimical Specifications.
4.0 Ytm<tBCl3 1.
Regulatory Guide 1.99, Radiaticn Embrittlement of Reactor Vxe:a1 Materials, Revision 2, May 1988 2.
NURD3-0800, Standard Review Plan, Section 5.3.2:
W ture Limits 3.
J. S. Perrin et al, " Final Report on Calvert Cliffs Lhit 1 Ibclear Plant Reactor Pressure Vessel Surveillance Prepcuu: Capsule 263," C+> J+.r 15, 1980 4.
February 6,1992, letter fIun G. C. Creel (BME) to USNRC Wwnt Cbntrol Desk, subject: Calvert Cliffs Ibclear Power Plant, Unit 1; License Areidgrut. Request: TNmture Overpressure Protectico (LTOP).
h
+
l
[
4 i
SALP INPUT FACILITY NAME:
Calvert Cliffs Unit 1
SUMMARY
OF REVIEW ACTIVITIES The staff reviews the licensee's pressure-temperature limits in the Calvert Cliffs Unit 1 Technical Specifications based on Generic Letter 88-11. Generic Letter 88-11 recommends the licensee to use Regulatory Guide 1.99, Rev.
2, to calculate the nil-doctility reference temperature, RT which is a parameter in establishing the pressure-temperature 11m, its. The staff calculates the RT, of g
the limiting beltline material, and compares it to the licensee's RT,. Based on the limiting RT ;, the staff verifies the licensee's g
g pressure-temperature limits using Standard Review Plan 5.3.2. The staff also verifies that the upper shelf energy of all reactor vessel beltline materials complies with Appendix G to 10 CFR 50.
NARRATIVE DISCUSSION OF LICENSEE PERFORMANCE-FUNCTIONAL AREA i
SAFETY ASSESSMENT /OUALITY VERIFICATION The licensee's calculation of RT follows the method in RG 1.99, g,
Rev.
2.
The proposed pressure-temperature limits are within the limits of SRP 5.3.2 and satisfy Appendix G to 10 CFR 50.
Quality control in preparing the pressure-temperature limits is evident.
Implementation of Generic Letter 88-11 is timely and ef fective.
l Licensee's performance is good.
r AUTHOR:
John Tsao, EMCB/DET (301)504-2702 i