ML20034G942
| ML20034G942 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/03/1993 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034G943 | List: |
| References | |
| NUDOCS 9303120137 | |
| Download: ML20034G942 (6) | |
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'o UNITED STATES g
E NUCLEAR REGULATORY COMMISSION o
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- E W ASHINGTON, D. C. 20555
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l COMMONWEALTH ED 20N COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2
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AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.123 License No. DPR-19 j
1 1.
The Nuclear Regulatory Commission (the Commission) has found that.
1 A.
The application for amendment by the Commonwealth Edison Company l
(the licensee) dated September 14, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as i
amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; i
B.
The facility will operate in conf:;rcity with the application, the provisions of the Act and the rules and regulations of the Commission; I
C.
There is reasonable assuranc: (i) th:t '.he activities authorized i
- j this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-19 is bereby amended to read as follows:
9303120137 930303 VDR ADOCK 05000237 P
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. (2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.123, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of the date of its issuance to ba 1
implemented within 30 days.
l FOR THE NUCLEAR REGULATORY COMMISSION kl4 f.
OV ames E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 3, 1993
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l ATTACHMENT TO LICENSE AMENDMENT NO. 123 i
FACILITY OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 l
Revise the Appendix A Technical Specifications by removing the pages ident1fied below and inserting the attached paga,
The revised pages are I
identified by the captioned amendment number and contain marginal lines
)
indicating the area of change.
REMOVE INSERT 3/4.6-23 3/4.6-23 B 3/4.6-26 B 3/4.6-26 B 3/4.6-26a B 3/4.6-26a i
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9 DRESDEN 11 DPR-19 Amendment No. 123 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) 1600_
5 A - SSEM MfDROTEST UWT CURVE A E
wrrH FUEL N VESSEL 5 B - NON-NUREAR HC.ATUP/
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1400 :-
VAL'o To 16 rJPY E C - NUCLEAR CDRE CRITCAL)
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Uu!T. VAU TD 16 EFPf 5
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1200 E-
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'1 BELTLINE:
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12 82*F j
14 87*F O
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iiiiiiiii, iiiiu s ii iiiiiiiiil i:::: 111. iiiiii;ii, iiiiiiisi, iiiiiiiii 0
50 100 150 200 250 300 350 l
MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
FIGURE 3.6.1.
3/4.6-23
DRESDEN 11 DPR-19 Amendment No. 123 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
B.
Pressurization Temperature - The reactor vessel is a primary barrier against the release of fission products to the environs.
In order to i
provide assurance that this barrier is maintained at a high degree of integrity, pressure-temperature limits have been established for the operating conditions to which the reactor vessel can be subjected.
Figure 3.6.1 presents the pressure-temperature curves for those opera-t ting conditions; Inservice Hydrostatic Testing (Curve A), Non-Nuclear Heatup/Cooldown (Curve B), and Core Critical Operation (Cur ve C).
These curves have been established to be in conformance with Appendix G l
to 10 CFR 50 and Regulatory Guide 1.99, Revision 2, and take into account the change in reference nil-ductility transition temperature (RT,33) as a result of neutron embrittlement. The adjusted reference temperature (ART) of the limiting vessel material is used to account i
for irradiation effects.
l Three vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non-beltline region (other than the closure flange region); and 3) the closure flange region. The beltline region is defined as th M region of the reactor vessel that directly surrounds the effective. tight of the reactor core (between the bottom and the top of active fuel), and i
is subject to an RT, adjustment to Ecount for irradiation embrittle-ment. Thenon-beltl3ine and closure flange regions receive insufficient fluence to necessitate an RT adjustment. These regions contain 37 components which include; th,e reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do not directly surround the reactor core. Although the closure flange region is a non-beltline region, it (the closure flange region) is treated separately for the development of the pressure-temperature curves to address 10 CFR 50 Appendix G requirements.
In evaluating the adequacy of the steel which comprises the reactor vessel, it is necessary that the following be established:
- 1) the RT,31 for all vessel and adjoining materials; 2) the relationship between RT,3, and integrated neutron flux (fluence, at energies greater than one Hev); and 3) the fluence at the location of a postulated flaw.
Boltun Temperature The initial RT of the main closure flanges, the shell and head materials conn,ec, ting to these flanges, the connecting welds and the 3
vertical electroslag welds which terminate immediately below the l
vessel flange have an RT,33 of 40*F.
Therefore, the minimum allowable boltap temperature is established as 100*F (RT
+ 60*F) 33 which includes a 60*F conservatism required by the origi,nal ASME Code of construction.
B 3/4.6-26 l
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DRESDEN 11 DPR-19 Amendment No. 123 Curve A - Hydrotestina As indicated in Curve A of Figure 3.6.1 for system hydrotesting, the minimum metal temperature of the n actor vessel shell is 100*F 1
for reactor pressures less than 312 psig.
This 100*F minimum boltup temperature is based on an RT of 40*F for the electroslag weld immediately below the vessel fl,ange and a 60*F conservatism required by the original ASME Code of construction.
At reactor pressures greater than 312 psig the minimum vessel metal temperature is established as 130*F. The 130*F minimum temperature l
is based on a closure flange region RT of 40*F and a 90*F con-I servatism required by 10 CFR 50 Append,ix G for pressure in excess of 20% of the preservice hydrostatic test pressure (1563 psig).
At approximately 650 psig the effects of pressurization are more limiting than the boltup stresses at the closure flange region, hence a family of non-linear curves intersect the 130*F vertical line. Belt-line as well as non-beltline curves have been provided to allow separate monitoring of the two regions.
Beltline curves as a function of vessel exposure for 12,14, and 16 effective full power years (EFPY) are presented to allow the use of the appropriate curve up to 16 EFPY of operation.
Curve B - Non-Nuclear Heatup/Cooldown Curve B of Figure 3.6.1 applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).
The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress.
l As indicated by the vertical 100*F line, the boltup stresses at the closure flange region are most limiting for reactor pressures below approximately 110 psig.
For reactor pressures greater than approximately 110 psig, pressurization and thermal stresses become more limiting than the boltup stresses, which is reflected by the non-linear portion of Curve B.
The non-linear portion of the curve is dependent on non-beltline and beltline regions, with the beltline region temperature limits having been adjusted to account for vessel irradiation (up to a vessel exposure of 16 EFPY).
The
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non-beltline region is limiting between approximately 110 psig and 830 psig. Above approximately 830 psig, the beltline region becomes limiting.
Curve C - Core Critical Operation Curve C, the core criticcl operation curve shown in Figure 3.6.1, is generated in accordance with 10 CFR 50 Appendix G which requires core critical pressure-temperature limits to be 40*F above any Curve A or B limits.
Since Curve B is more limiting, Curve C is Curve B plus 40*F.
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B 3/4.6-26a
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