ML20064A438
| ML20064A438 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/27/1990 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20064A437 | List: |
| References | |
| NUDOCS 9009040214 | |
| Download: ML20064A438 (9) | |
Text
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4 2500
,ii,,,iiiii LEAK TEST LIMIT
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2250 :::: RTgg AFTER 14 EFPY 1
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l 1/4T: 152'F
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3/4T: 124*F J
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_ 1750
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0 UNACCEPTABLE I
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k OPERATION J
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_{1500
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$ 1250
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E IHEATUP RATES UP TO 60*F/HR.
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h1000
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O5
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ACCEPTABLE OPERATION ::
2
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750
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j CRITICALITY LIMIT BASED ON 250 INSERVICE HYDROSTATIC TEST. :
TEMPERATURE (279'F) FOR THE --
SERVICE PERIOD UP TO 14 EFPY ;;
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O 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG. F)
Figure 3.4-2 Farley Unit 2 Reactor Coolant System Heatup Limitations l
Applicable for the First 14 EFPY.
I rarley-Unit 2 3/4 4-29 Amendment No.
9009040214 900827 PDR ADOCK 05000364 P
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j 2500 i
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2250 :::: RTNTDAFTER 14 EFPY
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1/4T: 152*F
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j 2000 Z
3/4T: 124'F l
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_ 1750 f
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OPERATION r
w 1500
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g 1250
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-Q 1000 f
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ACCEPTABLE OPERATION ::
5 E
- COOLDOWN RATES
$7 750
- oF/HR.
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-s 500 40
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60 l
100 250 1
0 0
50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG. F)
Figure 3.4-3 Farley Unit 2 Reactor Cooling System Cooldown Limitations Applicable for the First 14 EFPY, i
Farley-Unit 2 3/4 4-30 Amendment No.
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The pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200'F/hr respectively.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than i
320'F.
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System preservice hydrotests and'in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.
The fracture toughness properties of the ferritic materials in the i
reactor vessel are determined in accordance with ASTM E185-82, and in
_l accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of'the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in VCAP-7924-A, " Basis for-Heatup and Cooldown Limit Curves, April 1975."
Heatup and cooldown limit curves are calculated using the most.'imiting at de end M value of the nil-ductility reference temperature, RT 14effectivefullpoweryears(EFPY)ofservicelife.dt,he14EFPY.servicelife_l n
T at the 1/4T location in-periodischosensuchthatthelimitingRT,i,thelimitingunirradiated the core region is greater'than the RT o
material. TheselectionofsuchalimlikngRT assures that all components in the heactor Coolant System vill be,, operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RTnde; the results of these tests are shown in Table B 3/4.4-1.
Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation-can cause an increase in the RT,,g.
Therefore, an adjusted reference temperature, based upon the fluence and the nickel and copper content of the material in question, can be predicted using VCAP-12471 and the re'-
commendations of Regulatory Guide-1.99, Revision 2, " Radiation Embrittle-ment of Reactor Vessel Materials." 'The heatup and cooldown limit curves of l
Figures 3.4-2 and 3.4-3 include-predicted adjustments for this shift in RT,,,
at the end of 14 EFPY.
FARLEY-UNIT 2 B 3/4 4-7 AMENEMNT NO.
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FARLEY-UNIT 2 B 3/4 4-10 AMENDMENT NO.
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i The use of the composite curve'is necessary to set conservative heatup i
limitations because it is possible f(r conditions to exist such that over=the course;of the heatup ramp the contro'. ling condition.svitches from the inside i
to the outside and the pressure limit must at all times be based on analysis
'I of the most critical criterion.
Finally, the 10CFR Part 50, Appendi(-G Rule which addresses the metal temperature of'the closure head-flenge and vessel flange must be considered.-
This. Rule states.that the minimum oetal temperature of the. closure flange.
regions be at least 120'F higher trian the limiting RT
. h hae agkw whenthepressureexceeds20percentofthe-preservic,e,liydrostatictest pressure (621 psig for Farley Upst 2)..In addition,- the new 10CFR Part 50, Rule states that a plant specific fracture evaluation may be performed to justify less limiting requirements. Based upon such a fracture analysis for u
Farley Unit 2, the 14 EFPY heatup and cooldown curves are impacted by the.
l-nev 10CFR Part 50 Rule as shown on Figures 3.4-2 and L.4-3.-
Although the pressurizer operates in temperature ranges above those.for which~
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there is reason for-concern of non-ductible failure, operating limits'are-provided to assure compatibility of operation vith the fatigue analysis j'
performed in accordance with the ASME Code requirements.
The OPERABILITY of two RHR' relief valves or:an RCS vent opening of greater 4
than or equal to 2.85 square inches ensures that the'RCS vill be protected from pressure transients which could exceed the. limits of Appendix;G to 10CFR Part 50 when'one or more of the RCS cold legs are:less than or equal to 310'F.
Either RHR relief valve:has adequate relieving capability to-protect-the RCS from overpressurization when the transient is limited to either (1).
the start of an idle RCP with the secondary water temperature of the: steam:
generator less than or equal to 50'F above the RCS cold leg temperatures!or (2) the start of 3 charging pumps and their: injection into a water solid RCS.
3/4.4.11 STRUCTURAL INTEGRITY The inservice inspection and testing programs fori ASME Code Class 1,: 2 and 3 components ensure that the structural integrity and operational readiness of these components vill be maintained at an-acceptable level throughout the.
life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as-required by 10CFR Part 50.55a(g) except where specific vritten relief has been granted by l
the Commission pursuant to 10CFR Part 50.55a(g)(6)(i).
3/4.4.12 REACTOR VESSEL HEAD VENTS The OPERABILITY of the Reactor Head Vent' System ensures that adequate core cooling can be maintained in the event of the accumulation'of non-condensable l-gases in the reactor vessel.
This system is in accordance with 10CFR50.44(c)(3)(iii).
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FARLEY-UNIT 2 B 3/4 4-14 AMENDMENT NO.
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i Significant Hazards Evaluation i
' Pursuant to 10 CFR 50.92 i
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Significant Basards Evaluati c Put:suant to 10 CFR 50.92 for the Proposed Change 3 to FNP unit 2' hchnical' Specifications for the Purpose of Incorporating Heatup and Cooldown curves Applicable for 14 EPPY Proposed Change Replace the existing heatup/cooldown curves found in Technical Specifications Section 3/4.4.10 with the new heatup/cooldown curves applicable.through 14 ETPY.
As_a result of incorporating the new heatup/cooldown curves,'the following changes to the Technical Specifications will also be required:
- 1. Figures 3.4-2 and 3.4-3 of the Technical Specifications will be revised to incorporate the proposed heatup/cooldown curves for.the first 14 ETPY.
In additius, Bases Section B 3/4.4.10 will be revised to reflect that the proposed heatup/cooldown curves are applicable for 14 EFPY as opposed to the current 8 ETPY.
- 2. Bases figure B 3/4.4-1, titled " Fast Neutron Fluence" will be deleted since it is no longer used by Regulatory Guide 1.99, Revision 2.
Background
The ability of the large steel vessel containing the reactor core and its-primary coolant to resist fracture constitutes an important factor in ensuring plant safety. The beltline region of the reactor. pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA533 Grade B Class 1 (the base material of the Farley' Unit 2 reactor pressure vessel beltline) are well documented in literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
A method for performing analyses to guard against brittle fracture in reactor f
7 pressure vessels has been presented in " Protection Against Non-ductile Failure,"
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Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The i
method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RT I'
ndt l
The proposed amendment would modify a portion of the Farley Unit 2 Technical l
Specifications, Section 3/4.4.10, titled " Pressure / Temperature Limits" by incorporating heatup and cooldown curves based on the analysis of capsule X.
The methodology used to perform the analysis of Capsule X was updated to Regulatory Guide 1.99 Revision 2 as required by Generic Letter 88-11 dated July 12, 1988. The Capsule X analysis is documented in WCAP-12471, titled i
" Analysis of Capsule X from the Alabama Power Company, Joseph M. Farley Unit 2~
Reactor Surveillance Program," which was submitted to the NRC by Alabama Power L
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Page 2 Company letter dated April 12, 1990.
Revisions have also been made to the Farley Unit 2 Technical Specifications Bases Section 3/4.4.10 to reflect that the curves are valid for the first 14 ETPY of operation.
In addition, Figure B 3/4.4-1, titled " Fast Neutron Fluence as a Function of Pull Power _ Service Life" has been deleted from the Technical Specifications since it is not used by Regulatory Guide 1.99 Revision 2.
Analysis Alabama Power Company has reviewed the requirements of 10 CFR 50.92 as they.
relate to the proposed change to the heatup/cooldown curves and considers this change not to involve a significant hazards consideration.
In support of this conclusion, the following analysis is provided:
(1) The proposed change will not significantly increase the probability or consequences of an accident previously evaluated.
Neither the probability nor the consequence of a previously evaluated accident is increased due to the updated pressure-temperature operating limits. The adjusted reference temperature of the limiting beltline material was used to correct the beltline pressure-temperature curves to account for irradiation effects.
Thus, the operating limits are adjusted to incorporate the initial fracture toughness conservatism present when the reactor vessel was-new. The adjusted reference temperature calculations were performed utilizing the guidance contained in Regulatory Guide 1.99, Revision 2.
The updated curves provide assurance that brittle fracture of the reactor vessel is prevented; therefore, the consequences of a previously evaluated accident are not significantly increased as a result of this change.
(2) The proposed change will not create the possibility of a new or different kind of accident frcm any accident previously evaluated.
l The updated pressure-temperature. operating limits will not create the possibility of a new or different kind of accident than previously evaluated. The revised operating limits are merely an update of the old limits by taking into account the efeects of irradiation embrittlement, uti'.izing criteria defined in Regulatory Guide 1.99, l
Revision 2.
The updated pressure-temperature curves are conservatively adjusted to account for the effects of irradiation on the limiting reactor vessel material. No physical changes to the plant are being made, therefore, no new modes of operation are provided.
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Page 3 (3) We. proposed change does not involve a significant reduction in a-margin of safety.
%e method for-performing analyses to guard against brittle fracture in reactor pressure vessels as presented-in " Protection Against-Non-ductile Failure," Appendi'. G to Section III of the ASME Boiler and Pressure Vessel Code has been used. his method utilizes the fracture mechanics concepts and is based on the reference nil-
%ese methods;have been used to set ductility temperature (RT the operating limits for Nir). ley Unit 2 Land take into account the-E effect of. irradiation on the reactor vessel. materials while maintaining a required margin of. safety. %erefore, the proposed change does not-involve a significant reduction in a margin of safety.
Conclusion Based on the preceding analysis it-is concluded that operation of the Joseph M. Farley Nuclear Plant Unit'2 in accordance with the proposed change does not result in a significant increase ~in the probability or f
consequences of an accident previously evaluated, create the possibility.
of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
Therefore, Alabama Power Company has determined 'that the ' proposed change meets the requirements of 10 CFR 50.92(c) and does not involve a significant hazards consideration.
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