ML20216E650
| ML20216E650 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/23/1987 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20216E656 | List: |
| References | |
| TAC-60075, NUDOCS 8706300731 | |
| Download: ML20216E650 (29) | |
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~,t NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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ALARAMA POWER COMPANY DOCKET NO. 50-348 i
JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSEE Amendment No. 71 License No. NPF-2 4
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Alabama Power Company (the
' licensee) datbd October 25,19A5, as supplemented September 29, 1986, complies'with~the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the l
Commission';
C.
There is reasonable assurance: (i) that the activities authorized by 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 Comission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-2 is hereby amended to read as follows:
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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
71
, are hereby incorporated in the license. _ The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of receipt of this amendment.
FOR THE NilCLEAR REGULATORY COMMISSION g
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Elinor G. Adensam, Director Project Directorate 11-1 Divisibn' of Reactor Projects I/II Office of Nuclear Reactor Regulation Attachnent:
Changes to the Technical Specifications Date of Issuance: June 23, 1987
ATTACHMENT TO LICENSE AMENDMENT NO. 71 FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages Insert Pages 3/4 4-29 3/4 4-29 3/4 4-3b 3/4 4-30 B3/4 4-6 B3/4 4-6 B3/4 4-7 B3/4 4-7 B3/4 4-9 B3/4 4-9 B3/4 4-10 B3/4 4-10 B3/4 4-10A B3/4 4-14 i
33/4 4-14 I
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MATERIAL PROPERTY BASIS CONTROLLING MATE RI AL LOWER SHELL (PLATE NO.86919 2)
COPPE R CONTENT
- 0.14 WT%
NICKEL CONTENT
- 0.56 WT%
5' F INITIAL MT NOT t,'47, N4' F 1
- 3/4T,1215" F CURVES APPLICA8LE FOR HE ATUP RATES UP TO SO' F/HR FOR THE SERVICE
, PERIOD UP TO 16 EFPY 2500' i i,iiiiiiiiii s
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FOR THE SERVICE PERIOD UP TO 16 EFPY 250 l
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INDICATED TEMPERATURE (O F)
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FIGURE 3.4-2 FARLEY UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 16 EFPY FARLEY - UNIT 1 3/4 4-29 AMENDMENT NO. EP>> 71 36'uumi A rWM u4' sin' du 11 h
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4 MATERI AL PROPERTY BA51$
CONTROLLING MATE RI AL : LOWER SHELL (PL ATE NO. 56919 2) 0.14 WT%
COPPE R CONTENT NICKEL CONTENT 016 WT%
RT AFTER 16 E FPY 1/4T,146.4 F NDT 3/4T,1215' F 4
CURVES APPLICABLE FOR COOLDOWN UP TO 100* F/HR FOR THE SE RVICE PERIOD UP TO 16 EFPY 2500 J
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FIGURE 3.4-3 FARLEY UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 16 EFPY FARLEY - UNIT 1 3/4 4-30 AMENDMENT NO.58, 71
_ _ = _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ - _ - - _ _ _ _ _ - _ _ _ _ _ _ _.
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. REACTOR COOLANT SYSTEM BASES
--=
=================--------,=-
Reducing T to less than 500*F prevents the release of activity should a steam generakor tube rupture since the saturation pressure of the primary av coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4,.10 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G as required per 10 CFR Part 50 Appendix G.
l
- 1) The reactor coolant-temperature and 'pr' essure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.
a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.
b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of nonductile f ailure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2) These limit lines shall be calculated periodically using methods provided below.
- 3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.
FARLEY-UNIT 1 B 3/4 4-6 AMENDMENT NO. 4. 71
REACTOR COOLANT SYSTEM BASES 4)
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 j
difference between the pressurizer and the spray fluid is greater than 320*F.
i 5)
System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vesse1~ Code,Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with ASTM E185-82 and in accordance with additional reactor vessel requirements. These properties are then evaluated in 1
accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel C(de and the calculation nethods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."
Heatup and cooldown 1.imit curves are calculated using the most limiting value of the -nil-ductility reference temperature, RTndt, at the end of 16 effective full power years (EFPY) of service life'.' 'The 16 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the_ limiting unirradiated material.
The selection of such a limiting RTndt assures that all components in the Reactor Coolant System will be. operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RTndt" 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 caus7 an increase in the RTndt. Therefore, an adjusted reference temperature, based upon the fluence and copper and nickel content of the material in questivo, can be predicted using Figure B 3/4.4-1, Figure B 3/4.4-2 and the recommeedations of Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted aojustments for this shif t in RTndt at the end of 16 EFPY.
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FIGURE B 3/4.4-1 FAST NEUTRON FLUENCI (E>l MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE (EFPY)
FARLEY - UNIT 1 B 3/4 4-10 AMENDMENT NO. N.71
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FARLEY - UNIT 1 8 3/4 4-10 A AMEN 0 MENT NO.71
REACTOR COOLANT' SYSTEM BASES r.....===================------------
--=
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the intide to the outside and the pressure limit must at all times be based on analysis of the most critical. criterion.
Finally, the 10 CFR Part 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel *lange must be considered.
l This Rule states that the minimum metal temperature of the closure flange regions be at least 120'F higher than the limiting RTndt for these regions when the pressure exceeds 20 percent of the;preservice hydrastatic test pressure (621 psig for Farley Unit'1).
In addition, the new 10 CFR Part 50 Rule states that-a plant specific fracture evaluation may be performed to justify less limiting requirements.
As a result, such a fracture analysis was performed for Farley Unit 2 1 These Farley Unit 2 fracture analysis results are applicable to Farley Unit 1 since the pertinent parameters are identical for both plants.
Based upon this fracture analysis, the 16 EFPY heatup and cooldown curves are impacted by the new 10 CFR Part 50 Rule as shown on, Figures 3.4-2 and 3.4-3.
Although'the pressurize'r operates in tempfrature ranges above those for which there.is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two RHR relief valves or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR 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 j
]
The inservice inspection and testing programs for ASME Code Class 1, 2 and 3
]
components ensure that the structural integrity and operational readiness of these components will 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 written relief has been granted by the
' Commission pursuant to 10CFR Part 50.55a(g)(6)(i).
3/4.4.12 REACTOR VESSEL HEAD VENTS The 0PERABILITY of the Reactor Head Vent System ensures that adequate core cooling can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel.
This system is in accordance with 10CFR50.44(c)(3)(iii).
FARLEY-UNIT 1 B 3/4 4-14 AMENDMENT NO. 47, 71
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SAFETY EVALUATION BY THE OFFICE OF NilCLFAR REACTOR REGULATION RELATFD TO AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. NPF-?
ALABAMA POWER COMPANY JOSEPH M. FARLEY NtlCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-348 INTRODUCTION In a letter from R. P. Mcdonald to S. A. Varga dated October P.5,1985, the Alabama Power Company (the licensee) requested changes to the Joseph M. Farley Nuclear. Plant, linit 1 (Farley-1) Heatup/Cooldown Curves and supporting bases.
The staff reviewed the licensee's submittal and det' ermined that the effect of neutron irradiation on all beltline mat'e'rlais had not been assessed by the licensee. The staff's concerns are documented in a letter from E. A. Reeves to R. P. Mcdonald dated June 16, 1986.
In response to this letter, the licensee provided a revised set of heatup and cooldown curves, which were to be appli-cable for 16 effective full power years (EFPY). The revised curves and their bases were submitted in a letter from R. P. Mcdonald to L. S. Rubenstein dated September 29, 1986. The curves and bases are to be contained in Figures 3.4-2 and 3.4-3 and Bases Section 3/4.4.10 of the Farley-1 Technical Specifications.
DISCllSSION Heatup/Cooldown curves must be calculated in accordance with the requirements of Appendix G of 10 CFR 50, which became effective on July 26, 1983.
Appendix G of 10 CFR 50 requires that the reactor vessel beltline and closure flange region materials meet the safety margins of Aopendix G of the ASME Code Section III. To calculate pressure-temperature limit: iri accordance with these requirements, the effect of neutron irradiation, boltup, pressure and thermal stresses on the limiting reactor vessel beltline and closure flange region materials must be estimated. The effect of neutron irradiation on the Farley-1 beltline materials is documented in Westinghouse Report WCAP-10934, Rev. 2 dated June 1986, which is in Attachment 2 to the licensee's submittal dated September 29, 1986. The effect of boltup, pressure and thermal stresses on the reactor vessel closure flange region are documented in letters from R. P. Mcdonald to S. A. Varga dated June 18, 1984, and October 25, 1985.
~
a
EVAL llAT10N The metbods recorrnended by the NRC staff for calculating the effect of neutron irradiation darage are documented in Regulatory Guide 1.99, " Effects of Residual Elenents on Predicted Radiation Damage to Reactor Vessel Materials."
The relationships docurented in Regu7atory Guide 1.99 were empirically derived from materials that wre irradiated in commercial nuclear reactor surveillance capsules. The est per1.inent erMrical relationships are contained in Revision 2 to the guide.
Revision 2 has been reviewed by the NPC staff and published for comment. The licensee has us.ed the methods recommended in Regulatory Guide 1.09, Revision 2, to calculate the effect of neutron irradiation on the Farley-1 beltline materials.
Neutron irradiation damage is measured by an increase in a material's reference tempe rature. The value of the reference temperature that results from neutron irradiation damage is called the m-+erial's adjusted reference temperature, ART. The limiting ART was used to calculate pressure-temperature limits for the Farley-1 beltline materials. These limits were calculated in accordance with the reouirements in Appendix G of the ASME Code Section III. The NRC staff has evaluated these 1% its using the calculation methods recommended in Standard Review Frlan (SRP) 5.L2, " Pressure-Temperature Limits."
The stresses in the closure flenge region resulting from pressure, thermal effects, and boltup were calcMated by the licensee using finite element analysis. The closure head ar;d vessel flange geometry used in the finite element analysis was modelled for a typical 4-loop reactor vessel. The Farley-1 plant is a 3-loop rea ctor vessel. The geometry of the closure flange region in the Farley-1 reactor vessel is slightly different than that of the typical 4-loop reactor vessel.
To account for these differences, the licensee used the computation method of Reference 1 to perform a stress analysis of the Farley-1 vessel based on the finite element analysis and an analytical comparison of critical dimensions of the two types of vessels. Their analysis indicates that the typical 4-loop reactor vessel and the Farley-1 reactor vessel have essentially equivalent pressure and boltup stresses at the critical closure flange region. Hence, the stresses from boltup and pressure used for the typical 4-loop plant were used in the fracture mechanics evaluation for Farley-1. The stresses at the critical closure flange region resulting from thermal conditions during heatup or cooldown of the Farley-1 vessel were detennined by the computation method in Reference 1 to be significantly less than those calculated for the typical 4-loop plant.
Fracture mechanics evaluations at three discontinuity locations in the closure flange region were performed in accordance with the methodology in Appendix A of ASPE Code Section XI. In this analysis the licensee used all the safety l
factors reouired by Appendix G of the ASME Code, except for the Code recommended flaw size, to determine the closure flange location that would be considered the critical location. The location with the highest stress l
intensity factor after applying safety margins was con 3idered the critical l
closure flange location. The critical location was determined to be the outside surface at the discontinuity between the flange and upper shell of the reactor vessel.
l l
9
-3 4
The postulated flaw size recorrnended by Appendix G of the ASME Code was used for evaluating the beltline recion, but was not used in evaluating the closure flange region. The postulated ? law size recommended by Appendix G has a depth of i the section thickness (i T) and a length of li times the section thickness. The section thickness at the critical flange location for Farley-1 is 0.125 inches. Appendix G of the ASME Code indicates that smaller defect sizes may be used on an individual case basis, if a smaller size of maximum postulated defect can be assured. The postulated defect used in the licensee's analysis was a 0.625 inch deep by 3.75 inches long surface flaw. The licensee's justification for using a smaller flaw size in evaluating the closure flange region than that used in evaluating the beltline region is that the volunetric examination of the closure flange location will assure detection of the critical size flaw.
Volumetric examination of the reactor vessel flange-to-upper shell weld and specified adjacent base material is accomplished by two ultrasonic scan routines. Coverage from the flange side of the weld involves use of angled longitudinal waves from the flange seal surface. Beam angles are selected based on their ability..to provide coverage of the weld and specified adjacent base material to the extent practical and provide near normal incidence to the plane of the weld. Refracted beam angles in the range 0 to 16 are typically used for these examinations. Examinations from the shell side of the weld involve O', 45, and 60* refracted angle beam coverage from the vessel inside diameter surface. Angle beam scanning is performed in two directions, parallel to the weld and perpendicular to the weld from the shell side. Access for the j
shell side examination is limited to the Ten Year Inservice Inspection outage when the core barrel is removed from the reactor vessel.
The licensee indicates that the fact that postulated flaws are surface related is significant from a detection probability point of view.
Incipient cracks starting at right angles to a given surface (OD or ID) provide favorable conditions.for detection via ASME Code specified 45 shear wave ultrasonic examinations from the opposite surface.
Circumferential flaws are oriented j
favorably for detection during axial scanning. Axial flaws are oriented favorably for detection during circumferential scans. Circumferentially oriented flaws in the vessel flange weld region also provide favorable conditions for detection during ultrasonic examinations from the flange ceal surface.
Additional justifications for pemitting smaller postulated flaws in the closure flange region than the size postulated for the beltline region are described in the staff's regulatory analysis of public comments which is in to the staff's report SECY-83-80, "10 CFR Part 50-General Revision of Appendices G and H, Fracture Toughness and Reactor Vessel Material Surveillance Reouirements," February 25, 1983.
As previously reported, the licensee's fracture mechanics evaluation was perfomed in accordance with the methodology in Appendix A of ASME Code Section XI.
In this method, the stress intensity factors at the crack tip are calculated by linearizina the stress around the postulated flaw.
The linearized stress is divided into membrane and bending stresses. The Appendix A method of linearizing stress' resulted in negative membrane stresses
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-a.
when considering boltup, pressure and thermal condition during heat-up. The licensee considered the negative membrane stresses equal to zero when deter-mining the stress intensity factor resulting from thermal conditions during heat-up. The NRC staff considers this acceptable, since it conservatively represents the stress condition resulting from heat-up.
The licensee used the negative value of membrane stress when determining the stress intensity factor resulting from boltup and pressure conditions. The neoative membrane stress will result in a reduction in the calculated stress intensity factor, since the stress intensity factor is the sum of a positive bending stress and a negative membrane stress. A negative value of membrane stress does not represent the real membrane stress resulting from boltup and pressure conditions. However, the non-conservatism resulting from a negative valued membrane stress will be offset by a high value for the bending stress that results from the linearizing method. Several methods of calculating stress intensity factors for a stress distribution similar to that in the closure flange region were evaluated in Reference 2.
The Appendix A method of linearizing the stress around the postulated flaw produced conservative stress intensity factors when compared to those calculated using a finito element analysis method, an A'SME Code Section UI. Appendix G method recommended for nonlinear stress ' distributions, and a poly-nomial method (Reference 3). This
~
comparison indicates that the Appendix A method of linearizing stress will result in an acceptable fracture mechanics analysis for evaluating flaws in the closure flange region of the reactor vessel.
Using the stress intensity factors calculated in accordance with Appendix A of the ASME Code Section XI and the safety margins of Appendix G of the ASME Code with a postulated flaw of 0.625 inch deep by 3.75 inches long, the licensee proposed pressure-temperature limits for the closure flange region materials.
The pressure-temperature limits for the closure flange reg'on materials were j
combined with the limits for the beltline region to develop the Farley-1 Heatup/Cooldown Curves.
SAFETY
SUMMARY
Based on the method documented in Regulatory Guide 1.99, Revision 2, for evaluating the effect of neutron irradiation on reactor vessel beltline materials, and the method of calculating pressure-temperature limits in SRP 3.6.2, the licensee's proposed Heatup/Cooldown Curves for 16 EFPY meet the safety margins of Appendix G of the ASME Code, j
Based on the licensee's finite element analysis, the fracture mechanics analysis perfonned in accordance with Appendix A of Section XI of the ASME Code, and the licensee's and NRC staff's justification for considering smaller postulated flaw i
sizes based on SECY-83-80, the licensee's proposed pressure-temperature limits for the closure flange region meet the safety margins of Appendix G of the ASME Code.
Based on the above two conclusions, the proposed Heatup/Cooldown Curves that are I
contained in the licensee's letter dated September 29, 1986, meet the safety margins of Appendix G,10 CFR 50 for 16 EFPY and are acceptable Farley-1 Technical Specification.
ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no sigr.ificant increase in individual or cumulative occupational radiation exposure. The Commission has l
previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded, based on the consideration discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
References 1.
" Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components (Pressurized, Water Cooled Systems), U.S. Department of Commerce, December 1, 1958 and February 27, 1959, pp. 58, 59, 60, Addendum No. 1.
2.
Bloom, J.M., and Van Der Sluys, W.A., " Determination of Stress Intensity Factors for Gradient Stress Fields," Journal of Pressure Vessel Technology, Vol. 99, August 1977.
3.
Buchalet, C.B., and Bamford, W.H., " Stress Intensity Factor Solutions for Continuous Surface Flaws in Reactor Pressure Vessels,"
. Mechanics of Crack Growth, ASTM STP 590, American Society for Testing and Materials, 1976.
Principal Contributor:
B. Elliot l
Dated: June 23', 1987
-[*
June 23, 1987 0
Docket No. 50-348 EB.4 Erimes u Gray file i
Docket file J. Partlow B. Elliot NRC PDR T.Barnhart(4)
Local POR W. Jones Mr. R. P. Mcdonald-PAD 2 Reading E. Butcher Senior Vice President S. Varga ACRS(10) i Alabama Power Company P. Anderson C. Miles Post Office Box 2641 E. Reeves (2)
L. Tremper, LFMB Birmingham, Alabama 35291-0400 OGC-Bethesda ACRS (10)
L. Harmon
-C. tulesrGRA
Dear Mr. Mcdonald:
E. Jordan T. Itemper, LF"!L
SUBJECT:
TECHNICAL SPECIFICATION AMENDMENT REGARDING HEATUP AND C00LDOWN LIMITATIONS, JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 (TAC N0. 60075)
The Comission has issued the enclosed Amendment No.
to Facility Operating License NPF-2 for the Joseph M. Farley Nuclear Plant, Unit 1 i
The amendment. consists of changes to the Technical Specifications, in response to your application transmitted by letter dated October 25, 1985, as supple-mented September 29, 1986..
The amendment revises Technical Specification Figures 3.4-2 and 3.4-3, heatup and cooldown limitations based on results of analysis of Capsule "U" Reactor Vessel Material Radiation Surveillance Program. Analysis results are included in the Westinghouse Report WCAP-10934, Revision 2.
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next regular bi-weekly Federal Register notice.
Sincerely, Edward A. Reeves, Project Manager Project Directorate 11-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Amendment No. 71 to NPF-2 3.
Safety Evaluation cc: w/ enclosures:
See next page
- See P ev ous Concurrenc 11-1 j'Id PM P
II-1*
OGC PAnderson ER es:bg EAdensam BVogler*
l 6//787 6/ /87 6/p/87 6/g/87
/
a Mr. ' R. P. Mcdonald Alabama Power Company
.loseph M. Farley' Nuclear Plant.
i cc:
' Mr. W. O. Whitt D. Biard MacGuineas. Esquire Executive Vice President Volpe, Roskey and Lyons Alabana Power Company 918_16th Street, N.V.
Post Office Box 2641 Washington, DC 20006 Rirmingham, Alabama 35291-0400 Charles R. Lowman Mr. Louis B. Long, General Manager Alabama Electric Corporation i
Southern Company Services, Inc.
Post Office Rox 550 Post Office Box 2625 Andalusia, Alabama 36420 Birmingham, Alabama 35202 Chairman Regional Administrator, Pegion II Houston County Commission U.S. Nuclear Regulatory Comission Dothan, Alabama 36301 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303 Ernest L. Blake, ilr., Esquire Shaw, Pittman, Potts 'a'n'd Trowbridge Claude Earl Fox, M.D.
2300 N Street, N.W.
~
State Health Officer Washington, DC 20037 State Department of Public Health State Office Building Montgomery, Alabama 36130 Robert A. Buettner, Esquire Balch, Bingham, Baker, Hawthorne, Mr. J. D. Woodard Williams and Ward General Manager - Nuclear Plant Post Office Box 306 Post Office Box 470 Birmingham, Alabama 35201 Ashford, Alabama 36312 Resident Inspector U.S. Nuclear Regulatory Comission Post Office Box 24 - Route 2 Columbia, Alabama 36319 i
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ALARAMA POWER COMPANY DOCKET NO. 50-348 l
JOSEPH M. FARLEY NUCLEAR PLANT, ifNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. NPF-2 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Alabama Power Company (the
' licensee) datb' October 25, 19.85., as supplemented September 29, 1986, d
complies"with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endanaerina 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 license 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-2 is hereby amended to read as follows:
i
I
)
,N*
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(P.) Technical Specifications i
l The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
71 are hereby incorporated in the license. The licensee J
shall operate the facility in accordance with the Technical Specifications.
3.-
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of receipt of this amendment.
FOR THE NUCLEAR REGULATORY' COMMISSION l
i 1
j a d e, dis "
\\
Elinor G. Adensam, Director Project Director' ate 11-1 i
Divi's' ion'of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
i Changes to the Technical Specifications Date of-Issuance: June 23, 1987 l
i
\\
ATTACHMENT TO LICENSE AMENDMENT NO. 71 FACILITY OPERATING LICENSE N0. NPF-2 DOCKET NO. 50-348-Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages Insert Pages j
3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 B3/4 4-6 B3/4 4-6 B3/4 4-7 83/4 4-7 83/4 4-9 B3/4 4-9 B3/4 4-10 B3/4 4-10 83/4 4-10A j
B3/4 4-14 1
13/4 4-14 i
i
- '* i i
MATERI AL PROPERTY BA$lS CONTROLLING M ATE RI AL LOWER SHELL (PL ATE NO.86919-2)
COPPE R CONTENT 0.14 WT%
NICKEL CONTENT 0.56 WT%
l 5' F INITIAL RT NDT 0 F RT AFTE R 16 E FPY 1/4T,146.4 NDT 3/4T,121.5' F CURVES APPLICABLE FOR HEATUP RATES UP TO SO' F/HR FOR THE SERVICE
- PERIOD UP TO 16 E FPY l I f i 1 1 1 I I I i1 A
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FIGURE 3.4-2 FARLEY UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 16 EFPY FARLEY - UNIT'1 3/4 4-29 AMENDMENT N0. EP>, 71
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,e, MATERI AL PROPERTY BA$t$
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RT AFTER 16 E FPY 1/4T,144.4 F NDT 3/47,121,5' F CURVES APPLICABLE FOR COOLDOWN UP TO 100' F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY 2500 2
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c.
REACTOR COOLANT SYSTEM BASES
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=================--------------- -
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Reducing T to less than 500*F prevents the release of activity should a steam generakor tube rupture since the saturation pressure of the primary av coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific
. activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by.the data obtained.
3/4.4.10 PRESSURE / TEMPERATURE LIMITS 1
The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure vessel Code, Section 111, Appendix G as required per 10 CFR Part 50 Appendix G.
- 1) The reactor coolant temperature and'pfessure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.
a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation.
b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of nonductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2) These limit lines shall be calculated periodically using methods provided below.
- 3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.
FARLEY-UNIT 1 B 3/4 4-6 AMENDMENT NO. 76. 71
e.
REACTOR COOLANT SYSTEM BASES
-=
- 4) 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 320 F.
- 5) 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 reactor vessel are determined in accordance with ASTM E185-82 and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Sumer Addenda to Section III of the ASME' Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."
Heatup and cooldown lini,t curves are calculated using the most limiting value of the nil-ductili,ty re.ference temperature, RTndt, at the end of 16 effective
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full power years (EFPY) of service life. The 16 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material.
The selection of such a limiting RTndt assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requi rements.
The reactor vessel materials have been tested to determine their initial RTndt; 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 Therefore, an adjusted reference temperature, based upon the in the RT ndt.
fluence and copper and nickel content of the material in question, can be predicted using Figure B 3/4.4-1, Figure B 3/4.4-2 and the recommendations of Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTndt at the end of 16 EFPY.
FARLEY-UNIT 1 B 3/4 4-7 AMENDMENT NO. M. 71
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ilEACTOR COOL ANT SYSTEM,
,g BASES
==______===_______
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
l Finally, the 10 CFR Part 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered.
This Rule states that the minimum metal temperature of the closure flange regions be at least 120*F higher than the limiting RTndt for_ these regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Farley Unit 1).
In addition, the new 10 CFR Part 50 Rule states that a plant specific fracture evaluation may be performed to justify less limiting requirements. As a result, such a fracture analysis was performed for Farley Unit 2.
These Farley Unit 2 fracture analysis results are applicable to Farley Unit 1 since the pertinent parameters are identical for both plaats.
Based upon this fracture analysis, the 16 EFPY heatup and cooldown curves are impacted by the new 10 CFR Part 50 Rule as shown on, Figures 3.4-2 and 3.4-3.
Although 'the pressurizer operates in teri$e'rature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two RHR relief valves or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR 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 for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will 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 written relief has been granted by the Commission pursuant to 10CFR Part 50.55a(g)(6)(1).
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 gases in the reactor vessel.
This system is h accordance with 10CFR50.44(c)(3)(iii).
FARLEY-UNIT 1 B 3/4 4-14 AMENDMENT NO. f/, 71
-